ML19168A028

(Sps), Units 1 and 2 - Supplement to Subsequent License Renewal Application Change Notice 3

ML19168A028
Person / Time
Site:	Surry
Issue date:	06/10/2019
From:	Mark D. Sartain Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation
References
19-248
Download: ML19168A028 (90)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 June 10, 2019

• United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY

• SURRY POWER STATION (SPS) UNITS 1 AND 2 Serial No.:

NRA/DEA:

Docket Nos.:

License Nos.:

SUPPLEMENT TO SUBSEQUENT LICENSE RENEWAL APPLICATION CHANGE NOTICE 3 10 CFR 50 10 CFR 51 10 CFR 54 19-248 R2' 50-280/281 DPR-32/37 By letter dated October 15, 2018 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML18291A842), Virginia Electric and Power Company (Dominion Energy Virginia) submitted an application for the subsequent license renewal of Renewed Facility Operating License Nos. DPR-32 and DPR-37 for the Surry Power Station.

The purpose of this letter is to update the SLRA to incorporate changes discussed with NRC staff during recent NRC onsite audits, document MRP 227-A, Revision 1 updates and update other editorial items. provides descriptions of ten topics that require a SLRA supplement and

• identifies each affected SLRA section and/or table. Enclosure 2 includes mark-ups of each affected SLRA section and/or table being supplemented, as described in. It should be noted that changes to two commitments (Items #16 and #34) are provided in Table A4.0-1.

Serial No.: 19-248 Docket Nos.: 50-280/281 SLRA Supplement - Change Notice 3 Page 2 of 6 If there are any questions regarding this submittal or if additional information is needed, please contact Mr. Paul Aitken at (804) 273-2818.

Sincerely, Mark D. Sartain Vice President - Nuclear Engineering and Fleet Support COMMONWEAL TH OF VIRGINIA COUNTY OF HENRICO DIANE E. AITKEN NOTARY PUBLIC REG. #7763114 COMMONWEALTH OF VIRGINIA MVCOMMISSION EXPIRES MARCH 31, 2022 The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, 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 in 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 JQ___ day of ~

• , 2019.

My Commission Expires:~ 3lJ 2.Dt:z_

Commitments made in this letter:

The Licensee Commitments identified in Table A4.0-1 of Appendix A, Final Safety Analysis Report Supplement, are proposed to support approval of the subsequent renewed operating licenses and may change during the NRC review period.

Enclosures:

Topics that Require a SLRA Supplement -

SLRA Mark-ups - Change Notice 3

cc:

(w/o Enclosures except *)

Serial No.: 19-248 Docket Nos.: 50-280/281 SLRA Supplement - Change Notice 3 Page 3 of 6 U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector Surry Power Station Mr. Emmanuel Sayoc

• NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop O 11 F1 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. Tam Tran*

NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop O 11 F1 11555 Rockville Pike Rockville, Maryland 20852-2738 Ms. Karen Cotton NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. G. Edward Miller NRC Senior Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-9E3 11555 Rockville Pike Rockville, Maryland 20852-2738

State Health Commissioner Virginia Department of Health James Madison Building - y!h Floor 109 Governor Street Room 730 Richmond, Virginia 23219 Mr. David K. Paylor, 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 Mr. Justin Williams, Director Division of Land Protection and Revitalization 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. Craig R. Nicol, Regional Director Virginia Department of Environmental Quality Tidewater Regional Office 5636 Southern Blvd Virginia Beach, VA 23462 Serial No.: 19-248 Docket Nos.: 50-280/281 SLRA Supplement - Change Notice 3 Page 4 of 6

Serial No.: 19-248 Docket Nos.: 50-280/281 SLRA Supplement - Change Notice 3 Ms. Jewel Bronaugh, 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. Robert W. Duncan, Director Virginia Department of Game and Inland Fisheries P.O. Box 90778 Henrico, VA 23228 Mr. Allen Knapp, Director Virginia Department of Health Office of Environmental Health Services 1 09 Governor St, 5th Floor Richmond, VA 23129 Ms. Julie Lagan, Director Virginia Department of Historic Resources State Historic Preservation Office 2801 Kensington Ave Richmond, VA 23221 Mr. Steven G. Bowman, Commissioner Virginia Marine Resources Commission 2600 Washington Ave Newport News, VA 23607 Dr. Mary Fabrizio, Professor Virginia Institute of Marine Science School of Marine Science 7509 Roper Rd, Nunnally Hall 135 Gloucester Point, VA 23062 Ms. Angel Deem, Director Virginia Department of Transportation Environmental Division 1401 East Broad St Richmond, VA 23219 Mr. Stephen Moret, President Virginia Economic Development Partnership 901 East Byrd St Richmond, VA 23219 Page 5 of 6

Serial No.: 19-248 Docket Nos.: 50-280/281 SLRA Supplement - Change Notice 3 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 Mr. Jeff Caldwell, Director Virginia Department of Emergency Management 10501 Trade Rd Richmond, VA 23236 Mr. Bruce Sterling, Chief Regional Coordinator Virginia Department of Emergency Management 1070 University Blvd Portsmouth, VA 23703 Mr. Sanford B. Wanner, Administrator Surry County 45 School Street Surry, VA 23883 Page 6 of 6 Serial No.: 19-248 Docket Nos.: 50-280/281 TOPICS THAT REQUIRE A SLRA SUPPLEMENT Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

Change Notice 3 SPS SLRA The following ten topics require the SLRA to be supplemented:

1.

Components Supports and NSSS Fatigue TLAA AMR Lines

2.

Aging Management of Neutron Shield Tank Serial No.: 19-248 Page 2 of 10

3.

Fire Protection and Domestic Water Tank Foundation Aging Management

4.

Pressure-Temperature Limits (Section 4.2.5): TLAA Evaluation Editorial Correction

5.

10 CFR Part 50, Appendix J program (82.1.32): UFSAR Supplement Revision

6.

Fire Water System program (82.1.16): Enhancement Clarified, Operating Experience Update, Program Description Revision, and UFSAR Supplement Revisions

7.

Masonry Walls program (82.1.33): Enhancement Revision

8.

Structures Monitoring program (82.1.34): Program Description Revision, UFSAR Supplement Revision, and Enhancements Revised/Added

9.

Appendix C - MRP-227-A Gap Analysis for PWR Vessel Internals Aging Management: GAP Analysis Tables Clarification /Revision

10. Further Evaluation of Aging Management Associated With Increase in Porosity and Permeability Due to Leaching of Calcium Hydroxide and Carbonation in Inaccessible Concrete Areas

L Change Notice 3 SPS SLRA The following ten topics require the SLRA to be supplemented:

Serial No.: 19-248 Page 3 of 10

1. Component Supports and NSSS Supports Fatigue TLAA AMR Lines A fatigue TLAA was inadvertently identified for structural members associated with Component Supports and NSSS Supports in SLRA Section 3.5.2.2.2.5; and Tables 3.5.1, 3.5.2-36 and 3.5.2-38. Further review has confirmed that there are no TLAAs associated with Component Supports and NSSS Supports.

Based on the above, the SLRA is supplemented as shown in Enclosure 2, to clarify that there are no fatigue TLAAs associated with Component Supports and NSSS Supports in the following:

SLRA Section SLRA Table 3.5.2.2.2.5 3.5.1 3.5.2-36 3.5.2-38

2. Aging Management of Neutron Shield Tank The External Swiaces Monitoring of Mechanical Components program was assigned to manage loss of material for the external surfaces of the neutron shield tank. However, it was deemed more appropriate for the Structures Monitoring program to manage this aging effect. In addition, SLRA Section 2.4.1.38 (NSSS Supports) is updated to clarify that the neutron shield tank is evaluated with the reactor coolant system.

Based on the above, the SLRA is supplemented as shown in Enclosure 2, in the following:

SLRA Section SLRA Table 2.4.1.38 3.1.2-3 3.1.2.1.3 3.5.1

3. Fire Protection and Domestic Water Tank Foundation Aging Management Operating experience identified degradation of the oiled-sand cushion beneath the fire protection/domestic water tank foundations.

Loss of material and loss of form of the earthfill material (oiled-sand cushion) beneath the fire protection/domestic water tank foundations will be managed by the Structures Monitoring program (B2.1.34).

Change Notice 3 SPS SLRA Serial No.: 19-248 Page 4 of 10 Based on the above, the SLRA is supplemented as shown in Enclosure 2, to incorporate_ management of the oiled-sand cushion beneath the fire protection/domestic water tank foundations in the following:

SLRA Sections SLRA Tables 2.4.1.27 2.4.1-27 3.5.2.1.27 3.5.1-058 3.5.2-27

4. Pressure-Temperature Limits (Section 4.2.5): TLAA Evaluation Editorial Correction An editorial correction is made to the fifth paragraph of the TLAA Evaluation in Section 4.2.5 to specify the nozzle forging materials are documented in Tables 4.2.4-1, 4.2.4-3, 4.2.4-5, and 4.2.4-7.

Based on the above, Section 4.2.5 is supplemented, as shown in Enclosure 2, to specify the nozzle forging materials are documented in Tables 4.2.4-1, 4.2.4-3, 4.2.4-5, and 4.2.4-7.

5. 10 CFR Part 50, Appendix J program (82.1.32): UFSAR Supplement Revision.

The 10 CFR Part 50, Appendix J program UFSAR Supplement is revised to include "subject to the requirements of 10 CFR Part 54" to be consistent with the NUREG-2191 Table X1-01, "FSAR Supplement Summaries for GALL-SLR Chapter XI Aging Management Programs."

SLRA Section A1.32 is supplemented, as shown in Enclosure 2, to include the UFSAR Supplement revision described above.

6. Fire Water System program (82.1.16): Enhancement Clarified, Operating Experience Update, Program Description Revision, and UFSAR Supplement Revisions The Fire Water System program revision includes the following items:

Exception #1 is deleted with the commitment to permanently remove the exterior insulation of fire protection/domestic water tanks (a.k.a. fire water storage tanks).

Enhancement #7 is revised as follows:

Prior to the subsequent period of extended operation, the insulation on the exterior surfaces of the fire water storage tanks (FWSTs) will be permanently

Change Notice 3 SPS SLRA Serial No.: 19-248 Page 5 of 10 removed. Wall thickness measurements will be performed on external tank areas exhibiting unexpected degradation. RefurbishmenUrecoating will be performed consistent with the severity of the degradation identified and commensurate with the potential for loss of intended function. Inspections of external tank surfaces will be on a refueling cycle frequency.

Enhancement #11 is revised to require visual inspection and wall thickness examination of the Unit 1 hydrogen seal oil deluge sprinkler piping that does not allow drainage as part of drainage reconfiguration. Wall thickness examination of the Unit 1 main transformer deluge sprinkler piping that does not allow drainage will also be performed as part of the drainage reconfiguration.

Piping with unexpected degradation will be replaced.

Operating experience #6 and #7 for the FWSTs internal inspections results are revised to include August 2018 internal visual inspection results and March 2019 bottom thickness measurements.

Operating experience #10 for the fire protection system flow test results is revised to present motor driven fire pump and diesel driven fire pump discharge pressure data from the 2014 through 2019 flow tests.

The Fire Water System program UFSAR Supplement is revised as follows:

A cracking aging effect that was not incorporated with SLRA Change Notice #2 is incorporated.

The following qualification commitment is incorporated and also included in the AMP program description:

The training and qualification of individuals involved in coating/lining. inspections of non-cementitious coatings/linings are conducted in accordance with ASTM International Standards endorsed in RG 1.54 including guidance from the NRC staff associated with.a particular standard.

Fire Hydrant flushing procedures were revised to ensure the following information is captured in a condition report when a fire hydrant barrel does not drain in 60 minutes:

Where soil conditions or other factors are such that a hydrant barrel does not drain within 60 minutes, or where groundwater level is above that of the hydrant drain, the hydrant drain shall be plugged and the water in the barrel shall be pumped out.

Dry barrel hydrants that will be subject to freezing weather and have plugged drains shall be identified clearly as needing pumping after operation.

SLRA Section B2.1.16, Section A 1.16 and Table A4.0-1, Item 16 are supplemented, as shown in Enclosure 2, with enhancement revisions, operating experience revisions and UFSAR Supplement revisions.

Change Notice 3 SPS SLRA

7. Masonry Walls program (82.1.33): Enhancement Revision Serial No.: 19-248 Page 6 of 10 The Masonry Walls program is revised to apply Enhancement #2 to Element 6, Acceptance Criteria.

SLRA Section 82.1.33 is supplemented, as shown in Enclosure 2, with the enhancement #2 clarification.

8. Structures Monitoring program (82.1.34): Program Description Revision, UFSAR Supplement Revision, and Enhancements Revised/Added The Structures Monitoring program is revised to include the following new enhancements:

Procedures will be enhanced to specify that evaluation of neutron shield tank findings consider its structural support function for the reactor pressure vessel.

Procedures will be enhanced to also include LOCAs as events that require evaluation for potentially degraded structures by Civil/Mechanical Design Engineering.

Procedures will be enhanced to include aging management of erosion for the oiled-sand cushion in the fire protection and domestic water storage tank foundations.

The Structures Monitoring program description and UFSAR Supplement are revised to include an aging effect of loss of material and loss of form for aging management of the earthfill material (oiled-sand cushion) beneath the fire protection/domestic water tank foundations.

SLRA Section B2.1.34, Section A1.34 and Table A4.0-1, Item 34 are supplemented, as shown in Enclosure 2, to include the program description and UFSAR Supplement revisions and enhancement revisions/additions described above.

Change Notice 3 SPS SLRA Serial No.: 19-248 Page 7 of 10

9. Appendix C -

MRP-227-A Gap Analysis for PWR Vessel Internals Aging Management: GAP Analysis Tables Clarification/Revision The following SLRA Appendix C components are supplemented, as shown in Enclosure 2, to incorporate the revisions/clarifications described below. Unless noted otherwise, reference Westinghouse letter AMLR-17-35-P Revision 1 for additional details.

SLRA Appendix C Components Changes or Clarifications Component App. C Description of Change Table (page#)

CRGT C3.3-3 Note #6 was used to reference the Areva evaluation for Flexures (page C-29) the Unit 1 (AREVA) flexures to identify Safety and Economic Consequences, FMECA Groups, and Risk Categorization. The SLR Inspection Category remains "N" (No Additional Measures).

r Fuel Alignment C3.3-3 Referred to Note #7 for further explanation of Pins (Upper)

(page C-31) degradation mechanisms for upper fuel alignment pins.

Wear-related surface degradation is considered the leading degradation mechanism.

Core Barrel:

C3.3-3 Corrected an editorial error to indicate the SLR Core Barrel (page C-36)

Inspection Category is "X" instead of "E".

Flange (Surface)

Core Barrel:

C3.3-3 Corrected an editorial error to indicate the SLR Core Barrel (page C-36)

Inspection Category is "N" instead of "E".

Outlet Nozzle Core Barrel:

C3.3-3 Corrected an editorial error that originated in Change Lower Axial (page C-36)

Notice #2. The lower flange weld was supposed to be Weld removed, and Note #5 added, in Change Notice #2.

Lower Flange Instead, the lower axial weld was removed and Note #5 Weld was added. This error has been corrected in Change Notice #3.

Fuel Alignment C3.3-3 Referred to Note #7 for further explanation of Pins (Lower)

(page C-37) degradation mechanisms for lower fuel alignment pins.

Wear-related surface degradation is considered the leading degradation mechanism.

CRGT C3.3-3 Note #6 is added to Table C3.3-3.

Flexures (page C-41)

Change Notice 3 SPS SLRA Fuel Alignment Pins CRGT Sheaths and C-Tubes Alignment and Interfacing Components:

Clevis insert bolts, Clevis insert dowels; Clevis bearing Stellite wear surface. Radial Support Keys, Stellite wear surfaces.

CRGT Guide Cards CRGT Sheaths and C-Tubes CRGT Sheaths and C-Tubes Core Barrel:

Middle Axial Weld Lower Axial Weld Upper Girth Weld Core Barrel:

Lower Flange Weld & Upper Axial Weld Lower Support Forging &

Lower Support C3.3-3 (page C-41)

C4.3-1 (page C-54)

C4.3-1 (page C-56)

\\-

C4.3-1 (page C-58)

C4.3-1 (page C-58)

C4.3-2 (page C-59)

C4.3-2 (page C-59)

C4.3-2 (page C-60)

C4.3-2 (page C-60)

Note #7 is added to Table C3.3-3.

Serial No.: 19-248 Page 8 of 10 Expansion link to the CRGT continuous section sheaths and C-tubes has been added along with Note #18 for associated Primary component CRGT guide plates (cards).

The addition of Note #18 resulted in the previous Note

1. 18 being renumbered to Note #19.

Revised Note #2 to remove the following sentence:

"Interim Guidance issued in PWROG Letter OG 18-76 amends the requirements regarding baseline examinations".

Note #18 is added to Table C4.3-1.

Included a new line to identify an Expansion item for "Control Rod Guide Tube Continuous section sheaths and C-tubes" For the core barrel assembly, changed Note #9 to Note

1. 8 for the middle axial weld (MAW) and lower axial weld (LAW), and for the upper girth weld (UGW). The reason for the change is the removal of Note #8, and the renumbering of Note #9.

For the core barrel assembly, change Note #9 to Note

1. 8 for the lower flange weld (LFW), and for the upper axial weld (UAW). The reason for the change is the removal of Note #8, and the renumbering of Note #9.

An error was corrected that originated in Change Notice

1. 2. The correct examination method for the lower support forging and for the lower support column bodies

Change Notice 3 SPS SLRA Column Bodies (cast)

Core Barrel:

MAW, LAW, UGW, LFW, UAW C4.3-2 (page C-61)

Serial No.: 19-248 Page 9 of 10 (cast) is VT-3 as indicated in MRP-227, Revision 1, but had been incorrectly changed to EVT-1 in CN#2. Note 5 was added to the Lower Support Column Bodies (cast).

Removed Note #8 and renumbered Note #9 to be Note

1. 8.

SLRA Appendix C Gap Analysis changes resulted in aging evaluation changes for the

• following components:

Control rod guide tube (CRGT) continuous section sheaths and C-Tubes. An expansion link to the CRGT continuous section sheaths and C-tubes has been added for the associated Primary component CRGT guide plates (cards).

Core barrel outlet nozzle welds are eliminated as an expansion component in MRP-227, Revision 1.

Based on the above, the SLRA is supplemented as shown in Enclosure 2, to revise aging evaluation for the CRGT continuous section sheaths and C-tubes and the core barrel outlet nozzle welds in the following:

SLRA Tables 2.3.1-2 3.1.1-053a

3. f 1-059a 3.1.2-2

Change Notice 3 SPS SLRA Serial No.: 19-248 Page 10 of 10

10. Further Evaluation of Aging Management Associated With Increase in Porosity and Permeability Due to Leaching of Calcium Hydroxide and Carbonation in Inaccessible Concrete Areas Further evaluation of aging management associated with increase in porosity and I

permeability due to leaching of calcium hydroxide and carbonation in inaccessible concrete areas is clarified in SLRA Sections 3.5.2.2.1.9, 3.5.2.2.2.1 (4 ),

and 3.5.2.2.2.3(3). The further evaluations are clarified to indicate that there is evidence of leaching in the accessible areas exposed to a water-flowing environment. Evaluation has determined that the observed leaching in accessible areas did not adversely impact the structural integrity or resulted in a loss of intended function of the inaccessible concrete areas. This applies to Containment and concrete structures within the scope of subsequent license renewal.

Based on the above, the SLRA is supplemented as shown in Enclosure 2, to clarify the further evaluations in the following:

SLRA Section 3.5.2.2.1.9 3.5.2.2.2.1(4) 3.5.2.2.2.3(3)

SLRA MARK-UPS CHANGE NOTICE 3 Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2 Serial No.: 19-248 Docket Nos.: 50-280/281

Change Notice 2 SPS SLRA SLRA Section Table 2.3.1-2 2.4.1.27 2.4.1.38 Table 2.4.1-27 3.1.2.1.3 Table 3.1.1 Table 3.1.2-2 Table 3.1.2-3 3.5.2.1.27 3.5.2.2.1.9 3.5.2.2.2.1(4) 3.5.2.2.2.3(3) 3.5.2.2.2.5 Table 3.5.1 Table 3.5.2-27 Table 3.5.2-36 Table 3.5.2-38 4.2.5 App~n.cliiAzv..,:/1;;:i'..i.:;t,i:.

A1.16 A1.32 A1.34 Table A4.0-1 SLRA As-Submitted Pages*

2-58, 2-59 2-289 2-303 2-332 3-16 3-56, 3-62 3-90, 3-91 3-105 3-717 3-738 3-741 3-744 3-746 3-759, 3-760, 3-764 3-835 3-848, 3-849 3-854 4-71 i*:

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A-14 A-27 A-28, A-29 A-71 to A-73, A-92 Serial No.: 19-248 Docket Nos.: 50-280/281 Page 2 of 74 Topic#

9 3

2 3

2 9

9 2

3 10 10 10 1

1, 2, 3 3

1 1

4

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'i' 6

5 8

6, 8

Change Notice 2 SPS SLRA SLRA Section B2.1.16 B2.1.33 B2.1.34 Table C3.3-3 Table C4.3-1 Table C4.3-2 SLRA As-Submitted Pages*

B-108toB-119 B-220 to B-222 B-223 to B-229 C-29, C-31, C-36, C-37, C-41 C-54, C-56, C-58 C-59, C-60, C-61 6

Q 8

9 9

9 Serial No.: 19-248 Docket Nos.: 50-280/281 Pag_e 3 of 74 Topic#

• SLRA As-Submitted page numbers may not correspond to the page numbers in Enclosure 2.

Surry Power Station Units 1 and 2 Application for Subsequent License Renewal Change Notice 3 Mechanical Systems Table 2.3.1-2 Reactor Vessel Internals Subcomponent Alignment and interfacing (clevis insert bolt)

Alignment and interfacing (clevis insert dowel)

Alignment and interfacing (clevis insert wear surface)

Alignment and interfacing (internals hold-down spring)

Alignment and interfacing (radial support key wear surface)

Alignment and interfacing (thermal sleeve)

Alignment and interfacing (upper core plate alignment pin wear surface)

Alignment and interfacing (upper core plate alignment pin)

Baffle former (baffle edge bolt)

Baffle former (baffle former bolt)

Baffle former (baffle plate)

Baffle former (corner bolt)

Bottom mounted instrumentation (column body)

Bottom mounted instrumentation (flux thimble tube)

Control rod guide tube (continuous section sheath and C-tube}

Control rod guide tube (guide plate)

Control rod guide tube (guide tube support pin nut) (Unit 1 only)

See Table 2.1.5-1 for definitions of intended functions.

Page2-58 Intended Function(s)

Structural Support Structural Support Structural Support Structural Support Structural Support Structural Support Structural Support Structural Support Structural Support Structural Support Flow Distribution, Structural Support Structural Support Structural Support Structural Support Structural Support Structural Support Structural Support Serial No.: 19-248 Page 4 of 74

Surry Power Station Units 1 and 2 Application for Subsequent License Renewal Change Notice 3 Mechanical Systems Table 2.3.1-2 Reactor Vessel Internals Subcomponent Control rod guide tube (guide tube support pin)

(Unit 1 only)

Control rod guide tube (lower flange)

Core barrel (barrel former bolt)

Core barrel (core barrel flange)

GeFe eaFFel feeFe eaFFel el:ltlet ASi!:i!:le)

Core barrel (lower axial weld)

Core barrel (lower flange weld)

Core barrel (lower girth weld)

Core barrel (upper axial weld)

Core barrel (upper flange weld)

Core barrel (upper girth weld)

Lower internals (fuel alignment pin)

Lower internals (lower core plate)

Lower support (column body)

Lower support (column bolt)

Lower support (lower support forging)

No additional measures components Thermal shield (flexure)

Upper internals (fuel alignment pin)

Upper internals (upper core plate)

See Table 2.1. 5-1 for definitions of intended functions.

Page2-59 Intended Function(s)

Structural Support Structural Support Structural Support Flow Distribution, Structural Support StFl:letl:lFal Sl:l1313eFt Structural Support Structural Support Structural Support Structural Support Structural Support Structural Support Structural Support Flow Distribution, Structural Support Structural Support Structural Support Structural Support Flow Distribution, Structural Support Structural Support Structural Support Structural Support Serial No.: 19-248 Page 5 of 74

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Change Notice 3 Structures 2.4.1.27 Fire Protection and Domestic Water Tank Foundation

System Description

The Fire Protection/Domestic Water Tank Foundations are supported on well-tamped sand and gravel with an oiled-sand cushion between the tank and the backfill. To contain this material under the tanks, reinforced concrete ring walls, whose tops are approximately at grade, were constructed just outside the perimeter of the tank. The Fire Protection/Domestic Water Tank Foundations are located adjacent to the Fire Pump House, west of the Intake Canal.

System Evaluation Boundary The evaluation boundary for the Fire Protection/Domestic Water Tank Foundations includes the reinforced concrete ring walls constructed just outside the perimeter of the tank and the oiled-sand cushion under the tank.

The Fire Protection/Domestic Water Tank is evaluated in the fire protection system.

System Intended Functions The Fire Protection/Domestic Water Tank Foundations are relied upon for compliance with regulations for Fire Protection (10 CFR 50.48). Therefore, the Fire Protection/Domestic Water Tank Foundations are within the scope of license renewal in accordance with the criteria of 10 CFR 54.4(a)(3).

UFSAR References Additional details of the Fire Protection and Domestic Water Tank Foundation can be found in the UFSAR, Section 9.10.2.2.1.

Subsequent License Renewal Boundary Drawings The subsequent license renewal boundary drawing for the Fire Protection and Domestic Water Tank Foundation is listed below:

11448-SLRY-1 H Components Subject to Aging Management Review The component types subject to aging management review are indicated in Table 2.4.1-27, Fire Protection and Domestic Water Tank Foundation.

The aging management review results for these component types are indicated in Table 3.5.2-27, Containments, Structures and Component Supports - Fire Protection and Domestic Water Tank Foundation - Aging Management Evaluation.

Page2-289 Serial No.: 19-248 Page 6 of 74

Change Notice 3 System Evaluation Boundary Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Structures The evaluation boundary for the NSSS Supports includes all supports for Nuclear Steam Supply System components. The evaluation boundary for each nuclear steam supply system support lies between the integral attachment on piping and equipment being supported and its Containment concrete supporting structure.

Specifically:

- Pins, bolting, and other removable hardware that are part of the connection to the NSSS equipment integral attachment have been evaluated with the nuclear steam supply system equipment supports.

- Spring supports, sliding surfaces, stainless steel elements, steel elements.

- Exposed portions of the embedded components (i.e. end portion of threaded anchor and nut) and grout are evaluated with the nuclear steam supply system equipment supports.

- Concrete supporting structures (including the embedded portion of threaded anchor) are evaluated with the Containment.

- Integral attachments for the nuclear steam supply system piping and equipment are evaluated for aging management with the specific nuclear steam supply system equipment.

- The neutron shield tank is evaluated with the reactor coolant system.

- Snubbers are active components and not subject to aging management.

System Intended Functions Portions of the NSSS Supports perform the following safety-related function: The NSSS Supports provide structural support for safety-related SSCs. Therefore, the NSSS Supports are within the scope of license renewal in accordance with the criteria of 10 CFR 54.4(a)(1 ).

UFSAR References Additional details of the NSSS Supports can be found in the UFSAR, Section 15.6.2, Figure 15.6-1,

Figure 15.6-2, Figure 15.6-3, Figure 15.6-4, Table 15.6-1, and Table 15.6-2.

Subsequent License Renewal Boundary Drawings There are no subsequent license renewal boundary drawing for the NSSS Supports.

Components Subject to Aging Management Review The component types subject to aging management review are indicated in Table 2.4.1-38, NSSS Supports.

The aging management review results for these component types are indicated in Table 3.5.2-38, Containments, Structures and Component Supports - NSSS Supports - Aging Management Evaluation.

Page2-303 Serial No.: 19-248 Page 7 of 74

Surry Power Station Units 1 and 2 Application for Subsequent License Renewal Structures Table 2.4.1-27 Fire Protec tion and Domestic Water Tank Foundation Structural Membe r Intended Function(s)

Concrete element Structural Support Oiled-sand cushion Structural Suggort nent types are indicated in Table 3.5.2-27, Containments, rts - Fire Protection and Domestic Water Tank Foundation -

The AMR results for these compo Structures and Component Suppa Aging Management Evaluation.

See Table 2.1.5-1 for definitions of intended functions.

Page2-332 Serial No.: 19-248 Page 8 of 74

Change Notice 3 Aging Effects Requiring Management Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Aging Management Review The following aging effects, associated with the reactor coolant system, require management:

• Cracking

• Cumulative fatigue damage

• Long-term loss of material

• Loss of coating or lining integrity

• Loss of fracture toughness

• Loss of material

• Loss of preload

• Reduction of heat transfer

• Reduced thermal insulation resistance Aging Management Programs The following aging management programs manage the aging effects for the reactor coolant system component types:

• ASME Code Class 1 Small-Bore Piping (82.1.22)

• ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD (82.1.1)

• Bolting Integrity (82.1.9)

• Boric Acid Corrosion (82.1.4)

• Closed Treated Water Systems (82.1.12)

• External Surfaces Monitoring of Mechanical Components (82.1.23)

• Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (82.1.25)

• Internal Coatings/Linings For In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (82.1.28)

• Lubricating Oil Analysis (82.1.26)

• One-Time Inspection (82.1.20)

• Selective Leaching (82.1.21)

• Structures Monitoring (82.1.34)

• Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) (82.1.6)

• Water Chemistry (82.1.2)

Page 3-16 Serial No.: 19-248 Page 9 of 74

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 Component Number 3.1.1-053a Stainless steel, nickel alloy Westinghouse reactor internal Primary components exposed to reactor coolant, neutron flux 3.1.1-053b Stainless steel Westinghouse reactor internal Expansion components exposed to reactor coolant and neutron flux 3.1.1-053c Stainless steel, nickel alloy Westinghouse reactor internal Existing Programs components exposed to reactor coolant, neutron flux 3.1.1-054 Stainless steel bottom mounted instrument system flux thimble tubes (with or without chrome plating) exposed to reactor coolant and neutron flux Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Aging Effect/Mechanism Cracking due to SCC, irradiation-ass isled sec, fatigue Cracking due to sec, irradiation-assisted sec, fatigue Cracking due to SCC, irradiation-assisted sec, fatigue Loss of material due to wear Aging Management Program AMP XI.M16A, PWR Vessel Internals, and AMP XI.M2, Water Chemistry (for sec mechanisms only)

AMP XI.M1 6A, PWR Vessel Internals, and AMP XI.M2,

• Water Chemistry (for sec mechanisms only)

AMP XI. M1 6A, PWR Vessel Internals, and AMP XI.M2, Water Chemistry (for SCC mechanisms only)

AMP XI.M37, Flux Thimble Tube Inspection Page 3-56 Further Evaluation Recommended Yes (SRP-SLR Section 3.1.2.2.9)

Yes (SRP-SLR Section 3.1.2.2.9)

Yes (SRP-SLR Section 3.1.2.2.9)

No Discussion Consistent with NUREG-2191 with exceptions.

Exceptions apply to the NUREG-2191 recommendations for Water Chemistry (82.1.2) program implementation.

The 1,QntrQI rQg gyiQ!l Ml!l (I.QDlinyQy~ ~!ll.liQn sheath ang C-tybe), core barrel (lower flange weld) and core barrel (upper girth weld) align to this item but are listed as Expansion components in the Appendix C Gap Analysis.

See further evaluation in Section 3.1.2.2.9.

Consistent with NUREG-2191 with exceptions.

Exceptions apply to the NUREG-2191 recommendations for Water Chemistry (82.1.2) program implementation.

The brackets, clamps, terminal blocks and conduit straps located on the periphery align to this item, and are expected to be elevated to a Primary Inspection component, as described in the Appendix C Gap Analysis. The upper internals (upper core plate) aligns to this item, and is listed as a Primary Inspection component in the Appendix C Gap Analysis. See further evaluation in Section 3.1.2.2.9.

Consistent with NUREG-2191 with exceptions.

Exceptions apply to the NUREG-2191 recommendations for Water Chemistry (82.1.2) program implementation.

Clevis insert bolts and dowels align to this item, but are listed as Primary Inspection components in the Appendix C Gap Analysis. See further evaluation in Section 3.1.2.2.9.

Not applicable. Loss of material due to wear is addressed in row 3.1.1-028. The associated NUREG-2192 aging item is not used.

Change Notice 3 Serial No.: 19-248 Page 10 of 74

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 Component Number 3.1.1-059a Stainless steel (SS, including CASS, PH SS or martensitic SS) or nickel alloy Westinghouse reactor internal Primary components exposed to reactor coolant and neutron flux Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Aging Effect/Mechanism Loss of fracture toughness due to neutron irradiation embrittlement and for CASS, martensitic SS, and PH SS due to thermal aging embrittlement; changes in dimensions due to void swelling,

distortion ; loss of preload due to thermal and irradiation-enhanced stress relaxation, creep; loss of material due to wear Aging Management Program AMP XI. M1 6A, PWR Vessel Internals Page 3-62 Further Evaluation Recommended Yes (SRP-SLR Section 3.1.2.2.9)

Discussion Consistent with NUREG-2191. The control rod guide tube (continuouli liection shesith s;1nd C-M1e) core barrel (lower flange and lower axial welds) and the lower support (column body) also align to this item, but are listed as Expansion components in the Appendix C Gap Analysis. The lower internals (fuel alignment pin) also aligns to this item, but is listed as an Existing component in the Appendix C Gap Analysis. See further evaluation in Section 3.1.2.2.9.

Change Notice 3 Serial No.: 19-248 Page 11 of74

Table 3.1.2-2 Subcomponent Bottom mounted instrumentation (flux thimble tube)

Control rod guide lube <continuous section sheath and C-Miel Control rod guide tube (guide plate)

Control rod guide tube (guide tube support pin nut)

(Unit 1 only)

Control rod guide tube (guide tube support pin) (Unit 1 only)

Reactor Vessel, Internals, and Reactor Coolant System - Reactor Vessel Internals - Aging Management Evaluation Intended Material Environment Aging Effect Requiring Aging Management Programs NUREG-21 91 Table 1 Function(s)

Management Item Item ss Nickel alloy (E) Reactor coolant Cracking PWR Vessel Internals (B2.1.7)

IV.B2.RP-355 3.1.1-053c and neutron flux Water Chemistry (82.1.2)

IV.82.RP-355 3.1.1-053c Loss of fracture toughness; PWR Vessel Internals (82.1.7)

IV.B2.R-424 3.1.1-119 changes in dimensions; loss of preload; loss of material Loss of material Flux Thimble Tube Inspection (82.1.24)

IV.B2.RP-356 3.1.1-028

.S.S.

Stainless

<El Reactor coolant Cracking PWR Vessel Internals (82.1.7)

IV.82.RP-29§ J 1.1-053a

.s1e.e.l and neutron flux Water Chemistr:y (82.1.2)

IV.82 RP-29§ 3.1.1-Q53a Loss of material PWR Vessel Internals <B2.1.7)

IV.B2.RP-296 3.1.1-Q59a ss Cast (E) Reactor coolant Cracking PWR Vessel Internals (B2.1.7)

IV.82.RP-298 3.1.1-053a austenitic

>250°c (>482°F)

Water Chemistry (82.1.2)

IV.B2.RP-298 3.1.1-053a stainless and neutron flux steel Loss of fracture toughness PWR Vessel Internals (82.1.7)

IV.82.RP-297 3.1.1-059a Loss of material PWR Vessel Internals (B2.1. 7)

IV.B2.RP-296 3.1.1-059a Stainless (E) Reactor coolant Cracking PWR Vessel Internals (82.1.7)

IV.B2.RP-298 3.1.1-053a steel and neutron flux Water Chemistry (B2.1.2)

IV.B2.RP-298 3.1.1-053a Loss of material PWR Vessel Internals (82.1. 7)

IV.B2.RP-296 3.1.1-059a ss Nickel alloy (E) Reactor coolant Cracking PWR Vessel Internals (82.1.7)

IV.82.RP-355 3.1.1-053c and neutron flux Loss of fracture toughness; PWR Vessel Internals (B2.1.7)

IV>B2.RP-287 3.1.1-059b loss of preload ss Nickel alloy (E) Reactor coolant Cracking PWR Vessel Internals (B2.1. 7)

IV.B2.RP-355 3.1.1-053c and neutron flux Water Chemistry (82.1.2)

IV.B2.RP-355 3.1.1-053c Loss of fracture toughness; PWR Vessel Internals (82.1.7)

IV.B2.RP-287 3.1.1-059b loss of preload Loss of material; loss of PWR Vessel Internals (82.1.7)

IV.B2.RP-285 3.1.1-059c preload Notes C

D E, 3 E, 2 c

Q c

C D

C A

C D

A C.10 C.10, 11 A.10 8.10 C,10 C.10 Surry Power Station. Units 1 and 2 Application for Subsequent License Renewal Page 3-90 Change Notice 3 Serial No. : 19-248 Page 12 of 74

Table 3.1.2-2 Subcomponent Control rod guide tube (lower flange)

Core barrel (barrel former bolt)

Core barrel (core barrel flange)

GeFe eaFFel (seFe eaFFel ettllel

~

Core barrel (lower axial weld)

Core barrel (lower flange weld)

Reactor Vessel, Internals, and Reactor Coolant System - Reactor Vessel Internals - Aging Management Evaluation Intended Material Environment Aging Effect Requiring Aging Management Programs NUREG-2191 Table 1 Function(s)

Management Item Item ss Cast (E) Reactor coolant Cracking PWR Vessel Internals (B2.1.7)

IV.B2.RP-298 3.1.1-053a austenitic

>2so*c (>482°F)

Water Chemistry (B2.1.2)

IV.B2.RP-298 3.1.1-053a stainless and neutron flux steel Loss of fracture toughness PWR Vessel Internals (B2.1.7)

IV.B2.RP-297 3.1.1 -059a Stainless (E) Reactor coolant Cracking PWR Vessel Internals (B2.1.7)

IV.B2.RP-298 3.1.1-053a steel and neutron flux Water Chemistry (B2.1.2)

IV.B2.RP-298 3.1.1-053a Loss of fracture toughness PWR Vessel Internals (B2.1. 7)

IV.B2.RP-297 3.1.1-059a ss Stainless (E) Reactor coolant Cracking PWR Vessel Internals (B2.1.7)

IV.B2.RP-273 3.1.1-053b steel and neutron flux Water Chemistry (B2.1.2)

IV.B2.RP-273 3.1.1-053b Loss of fracture toughness; PWR Vessel Internals (B2.1.7)

IV.B2.RP-274 3.1.1-059b changes in dimensions; loss of preload Loss of material PWR Vessel Internals (B2.1.7)

IV.B2.RP-345 3.1.1-059c FD;SS Stainless (E) Reactor coolant Cracking PWR Vessel Internals (B2.1.7)

IV.B2.RP-280 3.1.1-053a steel and neutron flux Water Chemistry (B2.1.2)

IV.B2.RP-280 3.1.1-053a Loss of material PWR Vessel Internals (B2.1.7)

IV.B2.RP-345 3.1.1-059c ss Stai Riess (I;) ReasleF seelaRI GmsldR§ PWR l,lessel IRISFRals (82. ~.7) llJ.82.RP 278 a.u l:leae sleel aRel R9ttlF9R flttl(

WaleF Gl'leFRiSlf'/ (82. ~.2)

IV82.RP 278 a.u l:leae bass ef FRaleFial PWR Vessel IR!eFRals (82.~.7) lll82.RP 2QQe a.u l:leQe ss Stainless (E) Reactor coolant Changes in dimensions PWR Vessel Internals (B2.1.7)

IV.B2.RP-270 3.1.1-059a steel and neutron flux Cracking PWR Vessel Internals (B2.1.7)

IV.B2.RP-387a 3.1.1-053b Water Chemistry (B2.1.2)

IV.B2.RP-387a 3.1.1-053b Loss of fracture toughness PWR Vessel Internals (B2.1. 7)

IV.B2.RP-388a 3.1.1-059b ss Stainless (E) Reactor coolant Changes in dimensions PWR Vessel Internals (B2.1. 7)

IV.B2.RP-270 3.1.1-059a steel and neutron flux Cracking PWR Vessel Internals (B2.1.7)

IV.B2.RP-280 3.1.1-053a Water Chemistry (B2.1.2)

IV.B2.RP-280 3.1.1-053a Loss of fracture toughness PWR Vessel Internals (B2.1.7)

IV.B2.RP-297 3.1.1-059a Notes C, 8 D, 8 C, 8 A

B A

A B

A A

A B

A A;--B 8,-B G

C A

B A

C A

B C

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Page 3-91 Change Notice 3 Serial No.: 19-248 Page13of74

Table 3.1.2-3 Reactor Vessel, Internals, and Reactor Coolant System - Reactor Coolant - Aging Management Evaluation Component Intended Material Type Function(s)

Pressurizer PB Steel with (upper head and stainless cladding) steel cladding Pump casing PB Cast (reactor coolant) austenitic stainless steel Tank (neutron PB;SS Steel shield)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Environment (E) Air - indoor uncontrolled (E) Air with borated water leakage (I) Reactor coolant (E) Air - indoor uncontrolled (I) Reactor coolant

>250°C (>482°F)

(E) Air - indoor uncontrolled (E) Air with borated water leakage (I) Closed-cycle cooling water (E) Concrete Aging Effect Requiring Aging Management Programs Management Loss of material External Surfaces Monitoring of Mechanical Components (B2.1.23)

Loss of material Boric Acid Corrosion (B2.1.4)

Cracking ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD (B2.1.1)

Water Chemistry (B2.1.2)

Cumulative fatigue damage TLAA Loss of material Water Chemistry (B2.1.2)

Cracking External Surfaces Monitoring of Mechanical Components (B2.1.23)

Loss of material External Surfaces Monitoring of Mechanical Components (B2.1.23)

Cracking ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD (B2.1.1)

Water Chemistry (B2.1.2)

Cumulative fatigue damage TLAA Loss of fracture toughness Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) (B2.1.6)

TLAA Loss of material Water Chemistry (B2.1.2)

Loss of material E':JcteFAal S1:1rfaees MeAiteriA!J ef Meel'laAieal C8FAJ38A9A!S (82.1.2d)

Structure~ Moni!Qring (B2.1 ~4)

Loss of material Boric Acid Corrosion (B2.1.4)

Loss of material Closed Treated Water Systems (B2.1.12)

None None Page 3-105 NUREG-2191 Table 1 Notes Item Item IV.C2.R-431 3.1.1-124 C

IV.C2.R-17 3.1.1-049 A

IV.C2.R-58 3.1.1-040 A

IV.C2.R-25 3.1.1-042 A

IV.C2.R-25 3.1.1-042 B

IV.C2.R-223 3.1.1-009 A

IV.C2.RP-23 3.1.1-088 B

V.A.EP-103c 3.2.1-007 A

IV.C2.R-452b 3.1.1-136 A

IV.C2.R-09 3.1.1-033 A

IV.C2.R-09 3.1.1-033 B

IV.C2.R-223 3.1.1-009 A

IV.C2.R-52 3.1.1-050 A

IV.C2.R-52 3.1. 1-050 E, 1 IV.C2.RP-23 3.1.1-088 B

IV.C2.l=I 4a1 d.1.1 124 A

IIIA~ TP-~02 l~.1-Q77 6

IV.C2.R-17 3.1.1-049 A

IV.C2.RP-221 3.1.1-089 B

IV.E.RP-353 3.1.1-105 A

Change Notice 3 Serial No.: 19-248 Page 14 of 74

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Change Notice 3 Aging Management Review 3.5.2.1.27 Fire Protection and Domestic Water Tank Foundation Materials The materials of construction for the fire protection and domestic water tank foundation structural members are:

• Concrete

• Earthfill (rip-rap, stone, soil)

Environment The fire protection and domestic water tank foundation structural members are exposed to the following environments:

• Air - outdoor

• Groundwater

• Soil

• Water - flowing Aging Effects Requiring Management The following aging effects, associated with the fire protection and domestic water tank foundation structural members, require management:

• Cracking

• Cracking and distortion

• Increase in porosity and permeability

• Loss of bond

• Loss of material, loss of form

• Loss of material (spalling, scaling)

• Loss of material (spalling, scaling) and cracking

• Loss of strength Aging Management Programs The following aging management programs manage the aging effects for the fire protection and domestic water tank foundation structural members:

• Structures Monitoring (82.1.34)

Page3-71 7 Serial No.: 19-248 Page 15 of 74

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Change Notice 3 Aging Management Review 3.5.2.2.1.9 Increase in Porosity and Permeability Due to Leaching of Calcium Hydroxide and Carbonation Increase in porosity and permeability due to leaching of calcium hydroxide and carbonation could occur in inaccessible areas of concrete elements of PWR and BWR concrete and steel containments. Further evaluation is recommended if leaching is observed in accessible areas that impact intended functions. Acceptance criteria are described in BTP RLSB-1 (Appendix A.1 of this SRP-SLR).

[3.5.1-014] - UFSAR Section 15.3.1 discusses concrete mix designs. Reinforced concrete structures at SPS were designed, constructed, and inspected in accordance with ACI and ASTM standards, which provide for a good quality, dense, well-cured, and low permeability concrete. The mix proportions were established in accordance with ACl-301, "Specifications for Structural Concrete for Buildings." Procedural controls ensured quality throughout the batching, mixing, and placement processes. The ASME Section XI, Subsection IWL program (B2.1.30) and the Structures Monitoring program (B2.1.34) identify and manage any cracks in the containment concrete. Crack control was achieved through proper sizing, spacing, and distribution of reinforcing steel in accordance with ACl-318-63, "Building Code Requirements for Reinforced Concrete." The Structures Monitoring program (B2.1.34) and the ASME Section XI, Subsection IWL program (B2.1.30) inspect for evidence of leaching of calcium hydroxide and carbonation in accessible, and normally inaccessible structural components when scheduled maintenance work and planned plant modifications permit access. The Structures Monitoring program (B2.1.34) and the ASME Section XI, Subsection IWL program (B2.1.30) require that evaluation of inspection results includes consideration of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to inaccessible areas. Although plant operating experience has identified evidence of leaching of calcium hydroxide and carbonation, it has been determined that the observed leaching did not adversely impact the structural integrity or result in a loss of intended function of the conta inment structures.Plant operating experienoe has not identified any aging effeots related to inorease in porosity and permeability due to leashing of oaloium hydroxide and oarbonation. The Struotures Monitoring program (82.1.34) and the /\\SME Scotian XI, Subseotion l\\o.'L program (82.1.30) confirm the absenoe of aging effeots related to leashing of oaloium hydroxide and oarbonation.

Therefore, aging effects due to leaching of calcium hydroxide and carbonation are not applicable, and a plant-specific aging management program for inaccessible areas to manage the effects of increase in porosity and permeability due to leaching of calcium hydroxide and carbonation is not required.

Page 3-738 Serial No.: 19-248 Page 16 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Aging Management Review

[3.5.1-047] - Leaching - UFSAR Section 15.3.1 discusses concrete mix designs. Reinforced concrete structures at SPS were designed, constructed, and inspected in accordance with ACI and ASTM standards, which provide for a good quality, dense, well-cured, and low permeability concrete. The mix proportions were established in accordance with ACl-301, "Specifications for Structural Concrete for Buildings." Procedural controls ensured quality throughout the batching, mixing, and placement processes. The Structures Monitoring program (B2.1.34) identifies and manages any cracks in the concrete structures. Crack control was achieved through proper sizing,

spacing, and distribution of reinforcing steel in accordance with ACl-318-63, "Building Code Requirements for Reinforced Concrete." Additionally, the Structures Monitoring program (B2.1.34) inspects for evidence of leaching of calcium hydroxide and carbonation in accessible, and normally inaccessible structural components when scheduled maintenance work and planned plant modifications permit access. The Structures Monitoring program (B2.1.34) requires that evaluation of inspection results includes consideration of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in,

degradation to inaccessible areas. Although plant operating experience has identified evidence of leaching of calcium hydroxide and carbonation, it has been determined that the observed leaching did not adversely impact the structural integrity or result in a loss of intended function of the associated concrete structures.Plant operating experience has not identified any aging effects related to increase in porosity and permeability due to leashing of caloium hydroxide and carbonation. Therefore, a plant-specific aging management program for inaccessible areas to manage the effects of increase in porosity and permeability due to leaching of calcium hydroxide and carbonation is not required.

3.5.2.2.2.2 Reduction of Strength and Modulus Due to Elevated Temperature

[Reduction of strength and modulus of concrete due to elevated temperatures could occur in PWR and BWR Group 1-5 concrete structures. For any concrete elements that exceed specified temperature limits, further evaluations are recommended. Appendix A of American Concrete Institute (AC!) 349-85 specifies the concrete temperature limits for normal operation or any other long-term period. The temperatures shall not exceed 66°C (150°F) except for local areas, which are allowed to have increased temperatures not to exceed 93°C (200°F). Further evaluation is recommended of a plant-specific program if any portion of the safety-related and other concrete structures exceeds specified temperature limits [i.e., general area temperature greater than 66°C (150°F) and local area temperature greater than 93°C (200°F)]. Higher temperatures may be allowed if tests and/or calculations are provided to evaluate the reduction in strength and modulus of elasticity and these reductions are applied to the design calculations. The acceptance criteria are described in BTP RLSB-1 (Appendix A.1 of this SRP-SLR).

Page 3-741 Serial No.: 19-248 Page 17 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Aging Management Review

[3.5.1-051] - UFSAR Section 15.3.1 discusses concrete mix designs. Reinforced concrete structures at SPS were designed, constructed, and inspected in accordance with ACI and ASTM standards, which provide for a good quality, dense, well-cured, and low permeability concrete. The mix proportions were established in accordance with ACl-301, "Specifications for Structural Concrete for Buildings." Procedural controls ensured quality throughout the batching, mixing, and placement processes. The Structures Monitoring program (82.1.34), which includes Group 6 structures, identifies and manages any cracks in the concrete structures. Crack control was achieved through proper sizing, spacing, and distribution of reinforcing steel in accordance with ACl-318-63, "Building Code Requirements for Reinforced Concrete." Additionally, the Structures Monitoring program (82.1.34) inspects for evidence of leaching of calcium hydroxide and carbonation in accessible, and normally inaccessible structural components when scheduled maintenance work and planned plant modifications permit access. The Structures Monitoring program (82.1.34) requires that evaluation of inspection results includes consideration of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in. degradation to inaccessible areas. Although plant operating experience has identified evidence of leaching of calcium hydroxide and carbonation, it has been determined that the observed leaching did not adversely impact the structural integrity or result in a loss of intended function of the associated concrete structures. Plant operating experience has not identified any aging effects related to increase in porosity and permeability due to leaching of calcium hydrmcide and carbonation. Therefore, a plant-specific aging management program for inaccessible areas to manage the effects of increase in porosity and permeability due to leaching of calcium hydroxide and carbonation is not required.

3.5.2.2.2.4 Cracking Due to Stress Corrosion Cracking, and Loss of Material Due to Pitting and Crevice Corrosion Cracking due to SSC and loss of material due to pitting and crevice corrosion could occur in: (a)

Group 7 and 8 SS tank liners exposed to standing water; and (b) SS and aluminum alloy support members; welds; bolted connections; or support anchorage to building structure exposed to air or condensation (see SRP SLR Sections 3.2.2.2.2, 3.2.2.2.4, 3.2.2.2.8, and 3.2.2.2. 10 for background information).

For Group 7 and 8 SS tank liners exposed to standing water, further evaluation is recommended of plant-specific programs to manage these aging effects. The acceptance criteria are described in BTP RLSB 1 (Appendix A.1 of this SRP SLR).

Page 3-744 Serial No.: 19-248 Page 18 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Aging Management Review

[3.5.1-100] - Plant-specific OE has identified pitting or crevice corrosion or cracking for stainless steel piping components exposed to air or condensation (see Further Evaluation 3.4.2.2.2). The Structures Monitoring program (B2.1.34) will manage the aging of stainless steel and aluminum alloy components to ensure that these components continue to perform their intended functions during the subsequent period of extended operation.

3.5.2.2.2.5 Cumulative Fatigue Damage Due to Fatigue Evaluations involving time-dependent fatigue, cyclical loading, or cyclical displacement of component support members, anchor bolts, and welds for Groups 81.1, B1.2, and B1.3 component supports are TLAAs as defined in 10 CFR 54. 3 only if a CLB fatigue analysis exists.

TLAAs are required to be evaluated in accordance with 10 CFR 54.21(c). The evaluation of this TLAA is addressed in Section 4. 3, "Metal Fatigue Analysis," and/or Section 4. 7, "Other Plant Specific Time-Limited Aging Analyses," of this SRP-SLR. For plant-specific cumulative usage factor calculations, the method used is appropriately defined and discussed in the applicable TLAAs.

[3.5.1 -053] - The evaluation of fatigue for component support members, anchor bolts, and *.velds for Group B1.1 oomponents is addressed as a TL.AA in SL.RA Scotian 4.3.2, ASME Code, Scotian Ill, Class 1 P:atigue Analyses. The evaluation of fatigue for component support members, anchor bolts, and welds for Group B1.2 components are addressed as TL.Ms in SLR/\\ Section 4.3.3,

/\\ll>JSI B31.1 Allmvable Stress Analyses.There are no TLAAs associated with component support members, anchor bolts, and welds for Groups B1.1 and B1.2 component supports. Group B1.3 component supports are associated with BWRs: therefore, not applicable.

3.5.2.2.2.6 Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation Reduction of strength, loss of mechanical properties, and cracking due to irradiation could occur in PWR and BWR Group 4 concrete structures that are exposed to high levels of neutron and gamma radiation. These structures include the reactor (primary/biological) shield wall, the sacrificial shield wall, and the reactor vessel support/pedestal structure. Data related to the effects and significance of neutron and gamma radiation on concrete mechanical and physical properties is limited, especially for conditions (dose, temperature, etc.) representative of light water reactor (LWR) plants. However, based on literature review of existing research, radiation fluence limits of 1 x 1019 neutronslcm2 neutron radiation and 1 x 108 Gy (1 x 1010 rad) gamma dose are considered conservative radiation exposure levels beyond which concrete material properties may begin to degrade markedly (Ref 17, 18, 19).

Page 3-746 Serial No.: 19-248 Page 19 of 74

Table 3.5.1 Summary of Aging Management Programs for Containments, Structures and Component Supports Evaluated in Chapters II and Ill of the GALL-SLR Report Item Component Number 3.5.1-052 Groups 7, 8 - steel components: tank liner 3.5.1-053 Support members; welds; bolted connections; support anchorage to building structure 3.5.1-054 All groups except 6:

concrete (accessible areas):

all 3.5.1-055 Building concrete at locations of expansion and grouted anchors; grout pads for support base plates 3.5.1-056 Concrete: exterior above-and below-grade; foundation ; interior slab Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Aging Effect/Mechanism Cracking due to SCC; Loss of material due to pitting and crevice corrosion Cumulative fatigue damage due to cyclic loading (Only if CLB fatigue analysis exists)

Cracking due to expansion from reaction with aggregates Reduction in concrete anchor capacity due to local concrete degradation/ service-induced cracking or other concrete aging mechanisms Loss of material due to abrasion; cavitation Aging Management Program Plant-specific aging management program TLAA, SRP-SLR Section 4.3 Metal Fatigue, and/or Section 4.7 Other Plant-Specific Time-Limited Aging Analyses AMP XI.S6, Structures Monitoring AMP XI.S6, Structures Monitoring AMP XI.S7, Inspection of Water-Control Structures Associated with Nuclear Power Plants or the FERC/US Army Corp of Engineers dam inspections and maintenance programs.

Page 3-759 Further Evaluation Recommended Yes (SRP-SLR Section 3.5.2.2.2.4)

Yes (SRP-SLR Section 3.5.2.2.2.5)

No No No Discussion Not applicable. See further evaluation in Section 3.5.2.2.2.4.

GeAsisleAI wi!R ~H::J~EeG ~~ g~. Gl:lFl'Hllali>,<e fali§l:le ElaA'la§e ef 13elliA§ aAEl steel eleA'leAls is a TLM.!:c!Qt sir212licable. Th§r§ sir§ OQ TLAAs aiisQQisited with sur212ort m§m!;lers anQhQr QQlts sing welQll fQ[ QQ[O(;!Qnenl sur212orts. See further evaluation in Section 3.5.2.2.2.5.

Consistent with NUREG-2191.

Consistent with NUREG-2191.

Consistent with NUREG-2191. Loss of material of concrete elements exposed to water-flowing is managed by the Inspection of Water-Control Structures Associated with Nuclear Power Plants (B2.1.35) program.

Change Notice 3 Serial No.: 19-248 Page 20 of 74

Table 3.5.1 Summary of Aging Management Programs for Containments, Structures and Component Supports Evaluated in Chapters II and Ill of the GALL-SLR Report Item Component Number 3.5.1-057 Constant and variable load spring hangers; guides; stops 3.5.1-058 Earthen water-control structures: dams; embankments; reservoirs; channels; canals and ponds 3.5.1 -059 Group 6: concrete (accessible areas): all 3.5.1-060 Group 6: concrete (accessible areas): exterior above-and below-grade; foundation Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Aging Effect/Mechanism Loss of mechanical function due to corrosion, distortion, dirt or debris accumulation, overload, wear Loss of material; loss of form due to erosion, settlement, sedimentation, frost action, waves, currents, surface runoff, seepage Cracking; loss of bond ; and loss of material (spalling, scaling) due to corrosion of embedded steel Loss of material (spalling, scaling) and cracking due to freeze-thaw Aging Management Program AMP XI.S3, ASME Section XI, Subsection IWF AMP XI.S7, Inspection of Water-Control Structures Associated with Nuclear Power Plants or the FERG/US Army Corp of Engineers dam inspections and maintenance programs.

AMP XI.S7, Inspection of Water-Control Structures Associated with Nuclear Power Plants or the FERG/US Army Corp of Engineers dam inspections and maintenance programs.

AMP XI.S7, Inspection of Water-Control Structures Associated with Nuclear Power Plants or the FERG/US Army Corp of Engineers dam inspections and maintenance programs.

Page 3-760 Further Evaluation Recommended No No No No Discussion Consistent with NUREG-2191.

Consistent with NUREG-2191. Loss of material; loss of form of earthen dike and embankment is managed by the Inspection of Water-Control Structures Associated with Nuclear Power Plants (82.1.35) program. A different siging msinsigement tirQgrsim (Str1JQ\\1Jre§ MQnitQring (82.1.34)) i§ i;;redited fQr msinsiging IQss Qf msiterisil s1nd IQ§§ Qf fQrm Qflhe Qiled-§sind Q!.l~hiQn sutitiQrting the Fire PrQtectiQn/Domestic Wsiter S!Qrsige tanks Consistent with NUREG-2 191. Cracking; loss of bond; and loss of material (spalling, scaling) of group 6 concrete elements (accessible areas) is managed by the Inspection of Water-Control Structures Associated with Nuclear Power Plants (82.1.35) program.

Consistent with NUREG-2191. Loss of material (spalling, scaling) and cracking due to freeze-thaw of group 6 concrete elements (accessible areas) is managed by the Inspection of Water-Control Structures Associated with Nuclear Power Plants (82.1.35) program.

Change Notice 3 Serial No.: 19-248 Page 21 of 74

Table 3.5.1 Summary of Aging Management Programs for Containments, Structures and Component Supports Evaluated in Chapters II and Ill of the GALL-SLR Report Item Component Number 3.5.1-076 Sliding surfaces: radial beam seats in BWR drywell 3.5.1-077 Steel components: all structural steel 3.5.1-078 Stainless steel fuel pool liner 3.5.1-079 Steel components: piles 3.5.1-080 Structural bolting 3.5.1-081 Structural bolting 3.5.1-082 Structural bolting Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Aging Effect/Mechanism Loss of mechanical function due to corrosion, distortion, dirt or debris accumulation,

overload, wear Loss of material due to corrosion Cracking due to sec; Loss of material due to pitting and crevice corrosion Loss of material due to corrosion Loss of material due to general, pitting, crevice corrosion Loss of material due to general, pitting,

crevice corrosion Loss of material due to general, pitting,

crevice corrosion Aging Management Program AMP XI.S6, Structures Monitoring AMP XI.S6, Structures Monitoring AMP XI. M2, Water Chemistry, and monitoring of the spent fuel pool water level and leakage from the leak chase channels.

AMP XI.S6, Structures Monitoring AMP XI.S6, Structures Monitoring AMP XI.S3, ASME Section XI, Subsection IWF AMP XI.S6, Structures Monitoring Page 3-764 Further Evaluation Recommended No No No No No No No Discussion Not applicable - BWR only.

Consistent with NUREG-2191. For those components credited with a fire barrier function, the Fire Protection Program (B2.1.15) is used in conjunction with the Structures Monitoring Program (B2.1.34) to manage loss of material. In li!QQi!iQn !Q QQn!2inm!1nl~ Str!.!Ql!.!r!1~ 2nd QQmQQn!1nt Su1212or:t~* the n!1utrQn ~hi!11Q t2nk which is

!1V21u2jed with the re2Q!Qr cool2n! ~~§!!1m i§ 21igned tQ

!hi~ it!1m.

Consistent with NUREG-2191 with exceptions.

Exceptions apply to the NUREG-2191 recommendations for Water Chemistry (B2.1.2) program implementation.

Monitoring of the spent fuel pool water level and leakage from the leak chase channels is performed by the Structures Monitoring (B2.1.34) program.

Consistent with NUREG-2191.

Consistent with NUREG-2191.

Consistent with NUREG-2191.

Consistent with NUREG-2191.

Change Notice 3 Serial No.: 19-248 Page 22 of 74

Table 3.5.2-27 Containments, Structures and Component Supports - Fire Protection and Domestic Water Tank Foundation -

Aging Management Evaluation Structural Intended Material Environment Aging Effect Requiring Aging Management Programs NUREG-2191 Table 1 Notes Member Function(s)

Management Item Item Concrete element ss Concrete (E) Air - outdoor Cracking Structures Monitoring (B2.1.34)

II I.A3.TP-204 3.5.1-043 E, 1, 2

II I.A3.TP-25 3.5.1-054 A, 2 Cracking ; Joss of bond; and Structures Monitoring (B2.1.34)

III.A3.TP-26 3.5.1-066 A, 2 loss of material (spalling, scaling)

Loss of material (spalling,

Structures Monitoring (B2.1.34)

IIJ.A3.TP-23 3.5.1-064 A, 2 scaling) and cracking (E) Groundwater Cracking ; loss of bond; and Structures Monitoring (B2.1.34)

IIJ.A3.TP-212 3.5.1-065 A, 2 loss of material (spalling, 111.A3.TP-27 3.5.1-065 A, 2 scaling)

Increase in porosity and Structures Monitoring (B2.1.34)

III.A3.TP-29 3.5.1-067 A, 2 permeability; cracking ; loss of material (spalling, scaling)

(E) Soil Cracking Structures Monitoring (B2.1.34)

III.A3.TP-204 3.5.1-043 E, 1, 2

Cracking and distortion Structures Monitoring (B2.1.34)

III.A3.TP-30 3.5.1-044 A, 2 Cracking; loss of bond; and Structures Monitoring (B2.1.34)

III.A3.TP-212 3.5.1-065 A, 2 loss of material (spalling,

III.A3.TP-27 3.5.1-065 A, 2 scaling)

Increase in porosity and Structures Monitoring (B2.1.34)

III.A3.TP-29 3.5.1-067 A, 2 permeability; cracking; loss of material (spalling, scaling)

(E) Water - flowing Increase in porosity and Structures Monitoring (B2.1.34)

III.A3.TP-24 33.5.1-067 A, 2 permeability; loss of strength Qil~Q-IH!nQ

~

.E.fil1hfill_

(El Air - Q!.l\\QQQr LQ~~ Qf m2t~ri2I* IQ~~ Qf fQrm Str!.JQ!!.lr~~ MQnitQring (B2.1.J4) 111.A§.T-22 J:i.1-Q58 Ll

~

!.r.iJ2:.r.ru2.

~!Qn~ ~Qil)

Table 3.5.2-27 Plant-Specific Notes:

1.

The plant-specific aging management program used to manage the applicable aging effect(s) for this component type, material, and environment combination is the Structures Monitoring (B2.1.34) program.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Page 3-835 Change Notice 3 Serial No.: 19-248 Page 23 of 74

2.

Concrete element includes the ring wall.

3.

Structures Monitoring (82.1.34} program instead of Inspection of Water-Control Structures Associates with Nuclear Power Plants (82.. 35 program will manage loss of materials: loss of form of the oiled-sand cushion supporting the Fire-Protection/Domestic Water Storage tanks.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Page 3-836 Change Notice 3 Serial No.: 19-248 Page 24 of 74

Table 3.5.2-36 Containments, Structures and Component Supports - Component Supports - Aging Management Evaluation Structural Intended Material Member Function(s)

Aluminum EN;SS Aluminum elements Bolting ss Steel Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Environment (E) Air (E) Air - indoor uncontrolled (E) Air with borated water leakage Aging Effect Requiring Aging Management Programs Management Loss of material; cracking Structures Monitoring (B2.1.34)

Gl:lA'll:llali\\*e faligl:le eaA'!age

+tAA (ORiy if GLB faligl:le aAalysis

~

Loss of material Structures Monitoring (B2.1.34)

ASME Section XI, Subsection IWF (B2.1.31)

Structures Monitoring (B2.1.34)

Loss of preload ASME Section XI, Subsection IWF (B2.1.31)

Structures Monitoring (B2.1.34)

Loss of material Boric Acid Corrosion (B2.1.4)

Page 3-849 NUREG-2191 Table 1 Notes Item Item III.B2.T-37b 3.5.1-1 00 A, 2 III.B3.T-37b 3.5.1 -100 A, 2 III.B4.T-37b 3.5.1-100 A, 2 III.B5.T-37b 3.5.1-100 A, 2 111.81.2.T 2e a.e.1 Qea A

III.B2.TP-248 3.5.1-080 A

III.B3.TP-248 3.5.1-080 A

III.B4.TP-248 3.5.1-080 A

III.B5.TP-248 3.5.1-080 A

III.B1.2.TP-226 3.5.1-081 A

III.B1.2.T-24 3.5.1-091 A

III.B2.TP-43 3.5.1-092 A

III.B3.TP-43 3.5.1-092 A

III.B4.TP-43 3.5.1-092 A

III.B5.TP-43 3.5.1-092 A

III.B1.2.TP-229 3.5.1-087 A

III.B2.TP-261 3.5.1-088 A

III.B3.TP-261 3.5.1-088 A

III.B4.TP-261 3.5.1-088 A

III.B5.TP-261 3.5.1-088 A

III.B1.2.T-25 3.5.1-089 A

III.B2.T-25 3.5.1-089 A

III.B3.T-25 3.5.1-089 A

III.B4.T-25 3.5.1-089 A

III.B5.T-25 3.5.1-089 A

III.B5.T-25 3.5.1-089 A

Change Notice 3 Serial No.: 19-248 Page 25 of 74

Table 3.5.2-36 Containments, Structures and Component Supports - Component Supports - Aging Management Evaluation Structural Intended Material Member Function(s)

Grout ss Grout Sliding surfaces ss Lubrite Spring support ss Steel Stainless steel ss Stainless elements steel Steel elements EN;SS Steel Vibration isolation ss Non-metallic elements (e.g., rubber)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Environment (E) Air - indoor uncontrolled (E) Air - indoor uncontrolled (E) Air - indoor uncontrolled (E) Air with borated water leakage (E) Air (E) Air - indoor uncontrolled (E) Air with borated water leakage (E) Air - indoor uncontrolled Aging Effect Requiring Aging Management Programs Management Reduction in concrete Structures Monitoring (B2.1.34) anchor capacity Loss of mechanical function Structures Monitoring (B2.1.34)

Loss of mechanical function ASME Section XI, Subsection IWF (B2.1.31)

Loss of material Boric Acid Corrosion (B2.1.4)

Loss of material; cracking ASME Section XI, Subsection IWF (B2.1.31)

G1,1A't1,1lali*t1e fali§1,1e elaA'ta§e

+tAA (ORiy if GLB fali§1,1e aAalysis 9*i&ls}

Loss of material ASME Section XI, Subsection IWF (B2.1.31)

Structures Monitoring (B2.1.34)

Loss of material Boric Acid Corrosion (B2.1.4)

Reduction or loss of isolation Structures Monitoring (B2.1.34) function Page 3-850 NUREG-2191 Table 1 Notes Item Item III.B1.2.TP-42 3.5.1-055 A

III.B2.TP-42 3.5.1-055 A

III.B3.TP-42 3.5.1-055 A

III.B4.TP-42 3.5.1-055 A

III.B5.TP-42 3.5.1-055 A

III.B2.TP-46 3.5.1-074 A

III.B1.2.T-28 3.5.1-057 A

III.B2.T-25 3.5.1-089 A

III.B1.2.T-36b 3.5.1-099 A

III.B1.2.T 26 a.6.1 gi;a A,4 III.B1.2.T-24 3.5.1-091 A, 1 III.B2.TP-43 3.5.1-092 A, 1 III.B3.TP-43 3.5.1-092 A, 1 III.B4.TP-43 3.5.1-092 A, 1 III.B5.TP-43 3.5.1-092 A, 1 III.B1.2.T-25 3.5.1-089 A, 1 III.B2.T-25 3.5.1-089 A, 1 III.B3.T-25 3.5.1-089 A, 1 111.B4.T-25 3.5.1-089 A, 1 III.B5.T-25 3.5.1-089 A, 1 III.B4.TP-44 3.5.1-094 A

Change Notice 3 Serial No.: 19-248 Page 26 of 74

Table 3.5.2-38 Containments, Structures and Component Supports - NSSS Supports - Aging Management Evaluation Structural Intended Material Environment Aging Effect Requiring Aging Management Programs NUREG-2191 Table 1 Notes Member Function(s)

Management Item Item Bolting ss High-strength (E) Air Cracking ASME Section XI, Subsection IWF (82.1.31) 111.81.1.TP-41 3.5.1-068 A

steel Stainless (E)Air Loss of material; cracking ASME Section XI, Subsection IWF (82.1.31) 111.81.1.T-36b 3.5.1-099 A

steel (E) Air with borated None None 111.81.1.TP-4 3.5.1-098 A

water leakage Steel (E) Air - indoor Loss of material ASME Section XI, Subsection IWF (82.1.31) 111.81.1.TP-226 3.5.1-081 A

uncontrolled Loss of preload ASME Section XI, Subsection IWF (82.1.31) 111.81.1.TP-229 3.5.1-087 A

(E) Air with borated Loss of material Boric Acid Corrosion (82.1.4) 111.81.1.T-25 3.5.1-089 A

water leakage Grout ss Grout (E) Air - indoor Reduction in concrete Structures Monitoring (82.1.34) 111.81.1.TP-42 3.5.1-055 A

uncontrolled anchor capacity Sliding surfaces ss Lubrite (E) Air - indoor Loss of mechanical function ASME Section XI, Subsection IWF (82.1.31) 111.81.1.TP-45 3.5.1-075 A

uncontrolled Stainless steel ss Stain less (E) Air Loss of material; cracking ASME Section XI, Subsection IWF (82.1.31)

III.B1.1.T-36b 3.5.1-099 A, 2 elements steel (E) Air with borated None None 111.81.1.TP-4 3.5.1-098 A, 2 water leakage Steel elements ss Steel (E) Air - indoor Gt1A'lt1lati"'e fali§t1e aaA'la§e

+!:AA III.B1.1.T 26 a.e.1 oea A;-4 uncontrolled (Only if GLB fati§t1e analysis

~

Loss of material ASME Section XI, Subsection IWF (82.1.31) 111.81.1.T-24 3.5.1-091 A, 1 Loss of mechanical function ASME Section XI, Subsection IWF (82.1.31) 111.81.1.T-28 3.5.1-057 A, 1 (E) Air with borated Loss of material Boric Acid Corrosion (82.1.4) 111.81.1.T-25 3.5.1-089 A, 1 water leakage Table 3.5.2-38 Plant-Specific Notes:

1.

Steel elements include support members, spring supports, bearing plates, base plates, and connections, including maraging steel.

2.

Stainless steel elements include support members.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Page 3-855 Change Notice 3 Serial No.: 19-248 Page 27 of 74

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Change Notice 3 Time-Limited Aging Analyses 4.2.5 PRESSURE-TEMPERATURE LIMITS TLAA

Description:

10 CFR 50 Appendix G requires that the RV be maintained within established pressure-temperature (P-T) limits, including heatup and cooldown operations. These limits specify the maximum allowable pressure as a function of reactor coolant temperature. As the RV is exposed to increased neutron irradiation, its fracture toughness is reduced. The P-T limits must account for the anticipated RV fluence.

The current P-T limits are based upon fluence projections for 60 years of plant operation. Because they were based upon a fluence assumption of 60 years of operation, the P-T limits analyses meet the definition of 10 CFR 54.3(a) (Reference 1. 7-2) and have been identified as TLAAs.

TLAA Evaluation:

Heatup and cooldown limit curves are calculated using the most limiting value of RT NOT corresponding to the limiting material in the beltline region of the RV. The most limiting RT NOT of the material in the core region (beltline) of the RV is determined by using the unirradiated RV material fracture toughness properties and estimating the irradiation induced shift (LlRT NOT)

• RT NOT increases as the material is exposed to fast neutron irradiation; therefore, to find the most limiting core region (beltline) RT NOT at any time, LlRT NOT due to the neutron radiation exposure associated with that time must be added to the original unirradiated RT NOT Using the ART values, P-T limit curves are determined in accordance with the requirements of 10 CFR Part 50, Appendix G, as augmented by ASME Code,Section XI, Appendix G.

The P-T limits for 48 EFPY (currently maintained in the Technical Specifications for Units 1 and 2) are based on the K1a methodology and the latest fluence data.

According to NUREG-2192, Section 4.2.2.1.4, the P-T limits for the subsequent period of extended operation need not be submitted as part of the SLRA since the P-T limits are required to be updated through the 10 CFR 50.90 licensing process when necessary for P-T limits that are located in the Technical Specifications. The current licensing basis will ensure that the P-T limits for the subsequent period of extended operation will be updated prior to exceeding the EFPY for which they remain valid.

Nozzle materials were evaluated in WCAP-18242-NP at 48 EFPY and 68 EFPY; the nozzle forging materials evaluated are documented in Tables 4.2.4-1, 4.2.4-3, 4.2.2 54.2.4-5, and e4.2.4-7. All nozzle materials were assigned the fluence values at the postulated 1 /4T flaw location for each specific nozzle in Table 4.2.1-1 and Table 4.2.1-2. Thus, Unit 1 Inlet Nozzle 1 and Unit 2 Inlet Nozzle 1 and Outlet Nozzle 3 have neutron fluence values greater than 1.0 x 10 17 n/cm2 (E > 1.0 MeV) at 68 EFPY. In order to fully assess the Units 1 and 2 P-T limit curves applicability to 68 EFPY, a nozzle corner fracture mechanics analysis was completed for all nozzle materials.

Page4-71 Serial No. : 19-248 Page 28 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix A - UFSAR Supplement monitoring of air moisture content and contaminants such that specified limits are maintained, and performance of opportunistic inspections of components for indications of loss of material.

This program is based on the Surry response to NRC GL 88-14, "Instrument Air Supply Problems;"

and utilizes guidance and standards provided in EPRI TR 108147 "Compressor and Instrument Air System Maintenance Guide: Revision to NP-7079," and ANSI/ISA-S7.3-1975, "Quality Standard for Instrument Air." The Compressed Air Monitoring program activities implement the moisture content and contaminant criteria of ANSI/ISA-S7.3-1975 (incorporated into ISA-S7.0.01-1996).

Program activities include air quality checks at various locations to ensure that dew point, particulates, and hydrocarbons are maintained within the specified limits. Opportunistic inspections of the internal surfaces of select compressed air system components for signs of loss of material will be performed.

A1.15 FIRE PROTECTION The Fire Protection program is an existing condition and performance monitoring program comprised of functional tests and visual inspections. The program manages:

• loss of material for fire-rated doors, fire damper housings, the halon systems, RCP oil collection system, steel seismic gap covers and the low-pressure carbon dioxide systems

• loss of material (spalling) or cracking for concrete structures, including fire barrier walls, ceilings, and floors

• hardening, shrinkage, and loss of strength for elastomer fire barrier penetration seals and seismic gap elastomers

• loss of material, change in material properties, cracking/delamination, and separation for non-elastomer fire barrier penetration seals, fire stops, fire wraps, and coatings cracking/delamination, and separation

• loss of material and cracking for aluminum seismic gap covers This program includes fire barrier inspections. The fire barrier inspection program requires periodic visual inspection of fire barrier penetration seals, fire barrier walls, ceilings, and floors, fire damper housings, and periodic visual inspection and functional tests of fire-rated doors to demonstrate that their operability is maintained. The program also includes periodic inspections and functional tests of the halon systems and low-pressure carbon dioxide systems.

A1.16 FIRE WATER SYSTEM The Fire Water System program is an existing condition monitoring program that manages cracking. loss of material, flow blockage due to fouling, and loss of coating integrity for in-scope water-based fire protection systems. This program manages aging effects by conducting periodic PageA-13 Serial No.: 19-248 Page 29 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix A - UFSAR Supplement visual inspections, flow testing, and flushes consistent with provisions of the 2011 Edition of National Fire Protection Association (NFPA) 25. Testing of sprinklers that have been in place for 50 years is performed consistent with NFPA 25, 2011 Edition. With exception of two locations, portions of the water-based fire protection system that have been wetted but are normally dry have been confirmed to drain and are not subjected to augmented testing and inspections.

The water-based fire protection system is normally maintained at required operating pressure and is monitored such that loss of system pressure is detected and corrective actions initiated. Piping wall thickness measurements are conducted when visual inspections detect surface irregularities indicative of unexpected levels of degradation. When the presence of organic or inorganic material sufficient to obstruct piping or sprinklers is detected, the material is removed and the source is detected and corrected. Non-code inspections and tests follow site procedures that include inspection parameters for items such as lighting, distance offset, presence of protective coatings, and cleaning processes that ensure an adequate examination.

The training and qualification of individuals involved in coating/lining inspections of non-cementitious coatings/linings are conducted in accordance with ASTM International Standards endorsed in RG 1.54 including guidance from the staff associated with a particular standard.

A1.17 OUTDOOR AND LARGE ATMOSPHERIC METALLIC STORAGE TANKS The Outdoor and Large Atmospheric Metallic Storage Tanks program is an existing condition monitoring program that manages the effects of loss of material and cracking on the outside and inside surfaces of aboveground metallic tanks constructed on concrete or soil. This program is a condition monitoring program that manages aging effects associated with outdoor tanks with internal pressures approximating atmospheric pressure including the refueling water storage tanks (RWSTs), refueling water chemical addition tanks (CATs), emergency condensate storage tanks (ECSTs), and the emergency condensate makeup tanks (ECMTs). This program also manages aging of the fire protection/domestic water storage tanks (FWSTs) bottom surfaces exposed to soil.

The program includes preventive measures to mitigate corrosion by protecting the external surfaces of steel components per standard industry practice. The RWSTs are insulated and rest on a concrete foundation covered with an oil sand cushion. Caulking is used at the concrete-component interface of the RWSTs. The ECSTs and ECMTs are internally coated and protected by concrete missile barriers. Weep holes, located around the circumference of the ECSTs where the concrete missile shield meets the concrete foundation, allow drainage of leakage or condensation to the outside perimeter of the ECSTs. The weep holes will be inspected for water leakage once each refueling cycle. The CATs are skirt supported and insulated with sprayed-on rigid polyurethane foam.

The program manages loss of material on tank internal bare metal surfaces by conducting visual inspections. Surface exams of external tank surfaces are conducted to detect cracking on the PageA-14 Serial No.: 19-248 Page 30 of 74 l

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Change Notice 3 Appendix A - UFSAR Supplement A1.31 ASME SECTION XI, SUBSECTION IWF The ASME Section XI, Subsection IWF program is an existing condition monitoring program that manages loss of material, cracking, loss of preload, and loss of mechanical function for supports of Class 1, 2, and 3 components. There are no Class MC supports at SPS. This program consists of periodic visual examination of piping and component supports for signs of degradation, evaluation, and corrective actions. This program recommends additional inspections beyond the inspections required by the 1 O CFR Part 50.55a ASME Section XI, Subsection IWF program. This includes a one-time inspection within five years prior to entering the subsequent period of extended operation of an additional 5% of the sample populations for Class 1, 2, and 3 piping supports. The additional supports will be selected from the remaining population of IWF piping supports and will include components that are most susceptible to age-related degradation. For high-strength bolting with an actual yield strength equal to or greater than 150 ksi in sizes greater than one inch nominal diameter, volumetric examination comparable to that of ASME Code,Section XI, Table IWB-2500-1, Examination Category B-G-1 are performed to detect cracking in addition to the VT-3 examination. If a component support does not exceed the acceptance standards of IWF-3400, but is electively repaired to as-new condition, the sample is increased or modified to include another support that is representative of the remaining population of supports that were not repaired.

A1.32 10 CFR PART 50, APPENDIX J The 10 CFR Part 50, Appendix J program is an existing performance monitoring program that manages cracking, loss of leak tightness, loss of material, loss of preload and loss of sealing.

Leakage rates through the Containment pressure boundary are monitored, including the Containment liner, associated welds, penetrations, isolation valves, fittings, and other access openings to detect degradation of the Containment pressure boundary. Corrective actions are taken if leakage rates exceed acceptance criteria. Leakage rate testing is performed in accordance with the regulations and guidance provided in 10 CFR Part 50 Appendix J, Option B; Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program ;" afl4-NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J~" and subject to the requirements of 1 O CFR Part 54.

PageA-27 Serial No.: 19-248 Page 31 of 74

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Change Notice 3 Appendix A - UFSAR Supplement A1.33 MASONRY WALLS The Masonry Walls program is an existing condition monitoring program that is implemented as part of the Structures Monitoring program (A 1.34) and manages loss of material, cracking, and loss of material (spalling and scaling) that could impact the intended function of the masonry walls.

The Masonry Walls program consists of inspections, consistent with Inspection and Enforcement Bulletin (IEB) 80-11 and plant-specific monitoring proposed by Information Notice (IN) 87-67, for managing shrinkage, separation, gaps, loss of material and cracking of masonry walls such that the evaluation basis is not invalidated and intended functions are maintained. The inspections of the masonry walls within the scope of subsequent license renewal are conducted by qualified personnel at a frequency not to exceed five years.

A1.34 STRUCTURES MONITORING The Structures Monitoring program is an existing condition monitoring program that monitors the condition of structures and structural supports that are within the scope of subsequent license renewal to manage the following aging effects:

• Cracking

• Cracking and distortion

• Cracking, loss of material

• Cracking, loss of bond, and loss of material (spalling, scaling)

• Increase in porosity and permeability, cracking, loss of material (spalling, scaling)

• Loss of material

• Loss of material, loss of form

• Loss of material, change in material properties

• Loss of material (spalling, scaling) and cracking

• Loss of mechanical function

• Loss of preload

• Loss of sealing

• Reduction in concrete anchor capacity

• Reduction of foundation strength and cracking

• Reduction or loss of isolation function This program consists of periodic visual inspection and monitoring the condition of concrete and steel structures, structural components, component supports, and structural commodities to ensure that aging degradation (such as those described in ACI 349.3R, ACI 201.1 R, and other documents)

PageA-28 Serial No.: 19-248 Page 32 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix A - UFSAR Supplement will be detected, the extent of degradation determined and evaluated, and corrective actions taken prior to loss of intended functions. Inspections also include seismic joint fillers, elastomeric materials; and steel edge supports and steel bracings associated with masonry walls, and periodic evaluation of groundwater chemistry and opportunistic inspections for the condition of below grade concrete. Quantitative results (measurements) and qualitative information from periodic inspections are trended with photographs and surveys for the type, severity, extent, and progression of degradation. The acceptance criteria are derived from applicable consensus codes and standards.

For concrete structures, the program includes personnel qualifications and quantitative acceptance criteria of ACI 349.3R-02, "Evaluation of Existing Nuclear Safety-Related Concrete Structures." The inspection of structural components, including masonry walls and water-control structures, are performed at intervals not to exceed five years, except for wooden poles, which are inspected on a 10-year frequency.

Qualified inspectors identify changes that could be indicative of Alkali-Silica Reaction (ASR). If indications of ASR development are identified, the evaluation considers the potential for ASR development in concrete that is within the scope of the Structures Monitoring program (A 1.34), the ASME Section XI, Subsection IWL program (A 1.30), or the Inspection of Water-Control Structures Associated With Nuclear Power Plants program (A 1.35).

ASME Code,Section XI, visual examinations (VT-1) are conducted to detect cracking of stainless steel and aluminum components.

PageA-29 Serial No.: 19-248 Page 33 of 74

Table A4.0-1 Subsequent License Renewal Commitments Program Commitment The Fire Water System program is an existing condition monitoring program that will be enhanced as follows:

1. PFeeeeh::1Fes iAs13eetisA §1:1isaAee will 13e Fe*rises ts Feei1:1iFe Fe13laeeFAeAt sf aAy s13FiAkleF tl:iat sl=lsws aAy sf tl=le fellswiA§ :

leak,;1§e, esFFssisA, 131:lysieal saFAa§e, lsasiA§, 13aiAtiA§ 1:1Aless 13aiAtes 13y tl=le s13FiAkleF FAaA1:1faet1:1FeF, SF iASSFFeet sFieAtatisA. S13FiAkleFs at tl=le fellswiA§ lseatisAs will 13e asses ts tl=le test see13e: +Re ~aswaste Faeility, A1:11EiliaFy BeileF, MaiAteAaAee B1:1ilsiA§, GeAseAsate Pslisl=liA§ B1:1ilsiA§, ba1:1ASF'.f B1:1ilsiA§, aAs Mael=liAe S1=le13 B1:1ilsiA§. (Comgleted Change Notice 1)

2. Prior to 50 years in service, sprinkler heads will be submitted for field-service testing by a recognized testing laboratory consistent with NFPA 25, 2011 Edition, Section 5.3.1. Additional representative samples will be field-service tested every 10 years thereafter to ensure signs of aging are detected in a timely manner. For wet pipe sprinkler systems, a one-time test of sprinklers that have been exposed to water including the sample size, sample selection criteria, and minimum time in service of tested sprinklers will be performed. 81 !ils!Qh unit, s! ~s1m12l!il of J0.{Q ors! msiximum of t!;ln ~grinkl!;lr~ with no mor!;l than four sgrinklers ger structure shall be tested. Testing is based on a minimum time in service of fiftl£ l£ears and severitll of ogerating conditions for each gogulation. (Revised Change Notice 2)

Fire Water

3. Procedures will be revised to specify:

16 System program

a. Standpipe and system flow tests for hose stations at the hydraulically most limiting locations for each zone of the system on a five year interval to demonstrate the capability to provide the design pressure at required flow.

b. Acceptance criteria for wet pipe main drain tests. Flowing pressures from test to test will be monitored to determine if there is a 10% reduction in full flow pressure when compared to previously performed tests. The Corrective Action Program will determine the cause and necessary corrective action.

c. If a flow test or a main drain test does not meet acceptance criteria due to current or projected degradation additional tests are conducted. The number of increased tests is determined in accordance with the corrective action process; however, there are no fewer than two additional tests for each test that did not meet acceptance criteria. The additional inspections are completed within the interval in which the original test was conducted. If subsequent tests do not meet acceptance criteria, an extent of condition and extent of cause analysis is conducted to determine the further extent of tests. The additional tests include at least one test at the other unit with the same material, environment, and aging effect combination.

d. Main drains for the standpipes associated with hose stations within the scope of subsequent license renewal will also be added to main drain testing procedures.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix A - UFSAR Supplement PageA-71 AMP Implementation Program will be implemented and inspections or tests begin 5 years before the subsequent period of extended operation.

Inspections or tests that are to be completed prior to the subsequent 82.1.16 period of extended operation are completed 6 months prior to the subsequent period of extended operation or no later than the last refueling outage prior to the subsequent period of extended operation.

Change Notice 3 Serial No.: 19-248 Page 34 of 74

Table A4.0-1 Subsequent License Renewal Commitments Program Commitment

4. Procedures will be revised to gerform internal visual insgections of sgrinkler and deluge system giging to identit'. internal corrosion, foreign material, and obstructions to flow. Follow-ug volumetric examinations will be gerformed if internal visual insgections detect age-related degradation in excess of what would be exgected accounting for design, grevious insgection exgerience, and insgection interval. If organic Qr foreign material, or internal flow blockage that could result in failure of system function is identified, then an obstruction investigation will be gerformed within the Corrective Action Program that includes removal of the material, an extent of condition determination, review for increased insgections, extent of follow-ug examinations, and a flush in accordance with NFPA 25, 2011 Edition, Annex D.5, Flushing Procedures. The internal visual insgections will consist of the following : (Relocated from Commitment 10 and corrected - Change Notice 2}

a. Wet gige sgrinkler systems - 50% of the wet gige sgrinkler systems in scoge for subsequent license renewal will have visual internal insgections of giging by removing a hydraulically remote sgrinkler, gerformed eve[Y five years, consistent with NFPA 25, 2011 Edition, Section 14.2. During the next five-year insgection geriod, the alternate systems greviously not insgected shall be insgected.

b. Pre-action sgrinkler systems - gre-action sgrinkler systems in scoge for subsequent license renewal will have visual internal insgections of giging by removing a hydraulically remote nozzle, gerformed eve[Y five years, consistent with NFPA 25, 2011 Edition. Section 14.2.

C.

Deluge systems - deluge systems in scoge for subsequent license renewal will have visual internal insgections of giging by removing a hydraulically remote nozzle, gerformed eve[Y five years, consistent with NFPA 25, 2011 Edition, Section 16 Fire Water 14.2.

System program

5. Procedures will be revised to perform system flow testing at flows representative of those expected during a fire. A flow resistance factor (C-factor) will be calculated to compare and trend the friction loss characteristics to the resu lts from previous flow tests. (Renumbered Change Notice 2)

6. PFeseElufes faf hyElfaAl flushiA§ y,*ill ee fe>w<iseEI ta feeiuife fully a13eAiA§ the hyElfaAl aAEI fully flawiA§ the hyElfaAl faf Re less thaA aAe miAute aAEI uAtil fafei§A malefial has sleafeEI. IA aEIElitieA. J3feseElufes will ee fe>w<iseEI te eeseFYe ElfaiAiA§ ef the hyElfaAl eaffel aAEI alsa feeiuife the eaffel ee 13um13eEI Elfy sheulEI it Rel ElfaiA withiA eQ ffliAutes. 1-lyElrnAts eutsiee the 13rntesteEI arna that am withiA the sso13e of sueseeiueAt liseAse FeAewal will ee aEIEleEI to the flush sso13e.(Comgleted Change Notice 1 and renumbered Change Notice 2

7. +he Fife VIJatef Systeffl J3fO§faffl will ee fe*.iiseEI to 13efi0Elisally iAs13est the iAsulateEI eidefiof suffases of the fife watef taAks eA a ~ Q yeaf ffeeiueAsy ElufiA§ the sueseeiueAl 13efiaEI ef e13efatieA. IAsulaliaA is feme..,eEI le J3f9>w<iee a ffliAiffluffl iAs13estioA 13a13ulatiaA af ~a aAe seiuafe foot saffl13les. +he saffl13les will ee ElistfieuteEI iA sush a way that iAs13estiaAs assuf aA the taAk Elaffle. Aeaf the laAk eaUaffl. al 13aiAts whern stfusturnl su1313aFts. 13i13e. af iAstfuffleAt A9i!:i!:les 13eAetfate the iAsulatieA aAel whefe watef saulel sallest. IA aElelitiaA. iAs13estiaA lasatiaAs will ee eases aA the lil~elihaael af saFFasiaA uAeleF iAsulatiaA aoourriA§. Prior to the subsequent geriod of extended ogeration, the insulation on the exterior surfaces of the fire water storage tanks (FWSTs} will be germanently removed. Wall thickness measurements will be gerformed on external tank ares;1§ exhibiting un!;lxgected degradation. Refurbishment/recoating will be gerformed consistent with the severi!y of the degradation identified and commensurate with the gotential for loss of intended function. lnsgections of external tank surfaces will be on a refueling cycle frequency. (Renumbered Change Notice 2 and revised Change Notice 3}

Surry Power Station. Units 1 and 2 Application for Subsequent License Renewal Appendix A - UFSAR Supplement PageA-72 AMP Implementation Program will be implemented and inspections or tests begin 5 years before the subsequent period of extended operation. Inspections or tests that are to be completed prior to the subsequent 82.1.16 period of extended operation are completed 6 months prior to the subsequent period of extended operation or no later than the last refueling outage prior to the subsequent period of extended operation. Change Notice 3 Serial No.: 19-248 Page 35 of 74

Table A4.0-1 Subsequent License Renewal Commitments Program Commitment

8. PFeeeeh::1Fes feF FAaiAliAe stFaiAeF fh,1sl:liA§ will ee Fe't'ised te Feei1:1iFe fl1:1sl:liA§ 1:1Atil eleaF wateF is eeseFYed afteF eael:l e13eratieA eF flew test IA additieA te fl1:1sl:liA§ afteF e13eratieA, tl:le ~adwaste Faeility FAaiAliAe straiAeF will Feei1:1iFe aA iAs13eetien e't'el)'

fi*,*e yeaFs feF daFAa§ed and eeFFeded 13aFts.(Comi;ileted - Change Notice 1 and renumbered Change Notice 2}}

9. A procedure will be created to provide a Turbine Building oil deluge systems spray nozzle air flow test to ensure that patterns are not impeded by plugged nozzles, to ensure that nozzles are correctly positioned, and to ensure that obstructions do not prevent discharge patterns from wetting surfaces to be protected. (Renumbered Change Notice 2) rn. PFeeed1:1Fes will ee FeYised te 13effeFFA iAteFAal 1.<<is1:1al iAs13eetieAs ef s13FiAkleF aAd del1:1§e systeFA 13i13iA§ te identify iAteFAal eeFFesien, feFei§n FAateFial, aAd eestF1:1etiens te flew. Fellew 1:113 \\lel1:1FAetFie eMaFAinatieAs will ee 13effeFFAed if iAteFAal 't'is1:1al iAs13eetiens detest a§e Felated de§FadatieA iA eiEeess ef wl:lat we1:1ld ee eM13eeted aeee1:1AtiA§ feF desi§A, we*.<<ie1:1s iAs13eetieA eM13eFieAee, aAd iAs13eetieA iAtePw'al. If eF§aAie eF ferni§A FAatmial, eF iAteFAal flew eleeka§e tl:lat ee1:1ld rns1:1lt iA fail1:1rn ef systeFA f1:1Aetien is identified, tl:len an eestF1:1etien in't'esti§atieA will ee 13effeFFAed witl=lin tl:le GeFFeeti*.<<e Aetien PFe§FaFA tl:lat iAel1:1des FeFAe*,al ef tl:le FAateFial, aA eMteAt ef eenditien deteFFAinatien, Fe*.<<iew feF ineFeased ins13eetiens, eMtent ef fellew 1:113 elEaFAiAatiens, and a fl1:1sl:l in aeeeFdaAee witl=I ~IFPA 2e, 2011 Editien, AAAei1 g_e, Fl1:1sl:liA§ PFeeed1:1Fes. +l=le inteFAal 1Jis1:1al ins13eetieAs will eeAsist ef tl=le fellewiA§: (Old Enhancement 9 was relocated to Enhancement 4 Change Notice 2)

a. !Alet 13i13e Sl3FiAl~leF systeFAS e0% ef tl=le wet 13i13e s13FiAl~leF systeFAs iA see13e feF s1:1eseei1:1eAt lieeAse FeAewal will l:la>.<<e

't'is1:1al iAteFAal iAs13eetieAs ef 13i13iA§ ey FeFAe*,in§ a l:lydFa1:1lieally FeFAete s13FinkleF, 13effeFFAed e't'ei:y fi*.<<e yeaFs, seAsisteAt witl:l ~IFP/\\ 2e, 2011 Editien, Seetien 14.2. (;)1:1Fin§ tl=le neMt fi'4*e yeaF iAs13estieA 13mied, tl:le alteFAate systeFAs 13rn\\*ie1:1sly 16 Fire Water Aet iAs13eeted sl:lall ee iAs13ested. System program

b. PFe aetieA Sl3FiAkleF systeFAS 13Fe aetieA s13FiAkleF systeFAs iA see13e feF s1:10seei1:1eAt lieeAse Fenewal will Ra\\le \\lis1:1al iAteFAal ins13estiens ef 13i13in§ ey FeFAe\\lin§ a 1:lydra1:1lieally FeFAete Aezzle, 13effeFFAed e>.<<el)' fi1Je yeaFs, sensistent witl=I

~JFP.O. 2e, 2011 Editien, Sestien 14.2. C. (;)el1:1§e systeFAs del1:1§e systeFAs in see13e feF s1:10SeE11:1ent lieense FeAewal will Ra\\le 1,is1:1al inteFnal ins13eetiens ef 13i13in§ ey FeFAe*,in§ a l=lydra1:1lieally FeFAete Aezzle, 13effeFFAed e*,el)' fi*,e yeaFS, eensistent witl=I ~IFP.o. 2e, 2011 Editien, SeetieA ~

10. Procedure will be revised to provide inspection guidance related to lighting, distance and offset for non-ASME Code inspections. The procedure will specify adequate lighting be verified at the inspection location to detect degradation.

Lighting may be permanently installed, temporary, or portable (e.g., flashlight), as appropriate. For accessible surface inspections, inspecting from a distance of two to four feet (or less) will be appropriate. For distant surface inspections, viewing aids such as binoculars may be used. For viewing angles which may prevent adequate inspection, a viewing aid such as an inspection mirror or boroscope should be used.

11. The Unit 1 hydrogen seal oil system deluge sprinkler pipe and Unit 1 station main transformer '1A' deluge sprinkler piping will be reconfigured to allow drainage.A§ i;iact Qf th~ drainag~ r~QQofigurs;itiQn vi§ys;il in§i;i~Q1iQn§ and ws;ill thiQkn~§§ measurements will be i;ierformed on the Unit 1 hJ'.drogen seal oil SJ'.Stem deluge si;irinkler i;iii;ie that does not drain. In addition, wall thiQkness examination of the Unit 1 main transformer deluge sgrinkler giging that does not allow drainage will also be i;ierformed as i;iart of the drainage reconfiguration. Pii;iing with unexi;iected degradation will be rei;ilaced. (Revised Change Notice 3)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix A - UFSAR Supplement PageA-73 AMP Implementation Program will be implemented and inspections or tests begin 5 years before the subsequent period of extended operation. Inspections or tests that are to be completed prior to the subsequent B2.1.16 period of extended operation are completed 6 months prior to the subsequent period of extended operation or no later than the last refueling outage prior to the subsequent period of extended operation. Change Notice 3 Serial No. : 19-248 Page 36 of 74

Table A4.0-1 Subsequent License Renewal Commitments Program Commitment

12. The grogram will be revised to reguire insgections and tests be gerformed by gersonnel gualified in accordance with site grocedures and grograms for the sgecified task. (Added Change Notice 2}

13. Procedures will be revised to reguire when degraded coatings are detected by internal coating insgections, accegtance criteria and corrective action recommendations consistent with the Internal Coatings/Linings for ln-Scoge Piging, Piging Comgonen!~. Heat Exchangers and Tanks (82.1.28} grog ram are followed in lieu of NFPA 25 section 9.2.7 (1 }, (2}, and (4}.

When interior gitting or general corrosion (beyond minor surface rust) is detected, tank wall thickness measurements are conducted as stated in NFPA 25 Section 9.2.7(3} in vicinity of the loss of material. Vacuum box testing as stated in NFPA 25 Section 9.2.7(5} is conducted when gitting, cracks, or loss of material is detected in the immediate vicinity of welds. (Added Change Notice 2}

14. Procedures will be revised to address recurring internal corrosion with the use of Low Frequency Electromagnetic Technique (LFET) or a similar technique on 100 feet of piping during each refueling cycle to detect changes in the pipe wall thickness. LFET screening or a similar technique will also be performed on accessible interior fire water storage tank Fire Water 16 System program bottoms during periodic inspections. The procedure will specify thinned areas found during the LFET screening be followed up with pipe wall thickness examinations to ensure aging effects are managed and wall thickness is within acceptable limits. In addition to the pipe wall thickness examination, the performance of opportunistic visual inspections of the fire protection system will be required whenever the fire water system is opened for maintenance.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix A - UFSAR Supplement PageA-74 AMP Implementation Program will be implemented and inspections or tests begin 5 years before the subsequent period of extended operation. Inspections or tests that are to be completed prior to the subsequent B2.1.16 period of extended operation are completed 6 months prior to the subsequent period of extended operation or no later than the last refueling outage prior to the subsequent period of extended operation. Change Notice 3 Serial No.: 19-248 Page 37 of 74

Table A4.0-1 Subsequent License Renewal Commitments Program Commitment The Structures Monitoring program is an existing condition monitoring program that will be enhanced as follows:

1. Procedures will be revised to include inspection of the following structures that are within the scope of subsequent license renewal: decontamination building, radwaste facility, health physics yard office building, laundry facility, and machine shop.

lnsgections for the added structures will be gerformed under the enhanced grogram in order to establish guantitative baseline insgection data grior to the subseguent geriod of extended oi;ieration. (Revised Change Notice 1)

2. Procedures will be revised to add the oiled-sand cushion to the insgection of the fire grotection/domestic water tank foundation.(Added Change Notice 3)

3. Procedures will be revised to include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting. For structural bolting consisting of ASTM A325, ASTM A490, ASTM F1852 and/or ASTM F2280 bolts, the preventive actions for storage, lubricant selection, and bolting and coating material selection discussed in Section 2 of the Research Council for Structural Connections publication, "Specification for Structural Joints Using High-Strength Bolts," will be used.

4. The checklil;;t for l;;tructural and suggort steel will be revised to indicsite: "Are anll connection memt;!ers loose, misl;;ing or damaged (bolts, rivets, nuts, etc.)?". (Added Change Notice 2)

5. Prases1:1res will 13e revises ta req1:1ire at least f.i*,e years af mE13erieAse (ar /\\GI iAs13estar sertif.isatiaA) fer saAsrete iAs13estors Structures 34 Monitoring ta 13e saAsisteAt witl=l AGI J49.J~ GG~. Procedures will be revised to reguire at least five )lears of exgerience (or ACI program insgector certification) for concrete insgectors to be consistent with ACI 349.3R-002. Procedures will be revised to eliminate ogtions for insgector gualifications that are not consistent with ACI 349.3R-002.(Revised Change Notice 2)

6. Proses1:1res,,,*ill ee re*,ises ta iAs13est,..,,aaseA 1301,.rer 13ales aA a rn year freq1:1eAsy. Procedures will be revised to sgecifll that wooden gole insgections will be gerformed eve[ll ten )lears bl,'. an outside firm that grovides wooden gole insgection services that are consistent with standard indust[ll gractice. Visual examinations mall be augmented with soundings or other technigues aggrogriate for the t)li;ie, condition, and treatment of the wooden goles, including borings to determine the location and extent of decall and excavation to determine the extent of decal£ at the groundline. (Revised Change Notice 2)

7. Procedures will be revised to sgecifll that evaluation of insgection results includes consideration of the accegtabilitll of inaccessible areas when conditions exist in accessible areas that could indicate the gresence of, or result in, degradation to such inaccessible areas. (Added Change Notice 2)

8. Procedures will be enhanced to sgecifll VT-1 insgections to identifll cracking on stainless steel and aluminum comgonents.

A minimum of 25 insgections will be gerformed eve[ll ten )lears during the subseguent geriod of extended ogeration from each of the stainless steel and aluminum comgonent gogulations assigned to the Structures Monitoring grogram. If the comgonent il:! measured in linear feet, at least one foot will be insgected to gualifll as an insgection. For other comgonents, at least 20% of the surface area will be insgected to gualifll as an insgection. The selection of comgonents for insgection will consider the severit)l of the environment. For examgle, comgonents gotentialll£ exgosed to halides and moisture would be insgected, since those environmental factors can facilitate stress corrosion cracking. (Added Change Notice 2) Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix A - UFSAR Supplement PageA-95 AMP Implementation Program enhancements for SLR will be B2.1.34 implemented 6 months prior to the subsequent period of extended operation. Change Notice 3 Serial No. : 19-248 Page 38 of 74

Table A4.0-1 Subsequent License Renewal Commitments Program Commitment

9. Procedures will be enhanced to sr;iecib£ for the samr;iling-based insr;iections to detect cracking in stainless steel and aluminum comr;ionents, additional insr;iections will be conducted if one of the insr;iections does not meet accer;itance criteria due to current or r;irojected degradation unless the cause of the aging effect for each ar;ir;ilicable material and environment is corrected b~ rer;iair or rer;ilacement for all comr;ionents constructed of the same material and exr;iosed to the same environment. No fewer than five additional insr;iections for each insr;iection that did not meet accer;itance criteria or 20 r;iercent of each ar;ir;ilicable material, environment, and aging effect combination will be insr;iected, whichever is less.

Additional insr;iections will be comr;ileted within the 10-~ear insr;iection interval in which the original insr;iection was conducted. The resr;ionsible engineer will initiate condition rer;iorts to generate work orders to r;ierform the additional insr;iections. The resr;ionsible engineer will evaluate the insr;iection results, and if the subseguent insr;iections do not meet acc~r;itance criteria, an extent of condition and extent of cause anal~sis will be conducted. The resr;ionsible engineer will then determine the further extent of insr;iections. Additional samr;iles will be insr;iected for an~ recurring degradation to ensure corrective actions ar;ir;iror;iriatel~ address the associated causes. The additional insr;iections will include insr;iections of comr;ionents with the same material, environment, and aging effect combination at both Unit 1 and Unit 2. If an~ r;irojected insr;iection results will not meet accer;itance criteria r;irior to the next scheduled insr;iection, insr;iection freguencies will be adjusted as determined b~ the Corrective Action Program. (Added Change Notice 2}

10. Procedures will be enhanced to sr;iecib£ that evaluation of neutron shield tank findings consider its structural sur;ir;iort function for the reactor r;iressure vessel. (Added Change Notice 3}

11. Procedures will be enhanced to also include LOCAs as events that reguire evaluation for r;iotentiall~ degraded structures b~

Civil/Mechanical Design Engineering. (Added Change Notice 3) The Inspection of Water Control Structures Associated with Nuclear Power Plants program is an existing condition monitoring program that will be enhanced as follows: Inspection of

1. Procedures will be revised to provide guidance for specification of bolting material, lubricants and sealants, and installation Water Control Structures torque or tension to prevent degradation and assure structural bolting integrity.

35 Associated with

2. Procedures will be revised to specify the preventive actions for storage discussed in Section 2 of Research Council for Nuclear Power Structural Connections publication "Specification for Structural Joints Using ASTM A325 or A490 Bolts" for ASTM A325, Plants program ASTM F1852, ASTM F2280, and/or ASTM A490 structural bolts.

3. Procedures will be revised for concrete inspection to require at least five years of experience (or ACI inspector certification) to be consistent with ACI 349.3R-2002.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix A - UFSAR Supplement PageA-96 AMP Implementation Program enhancements for SLR will be implemented B2.1.35 6 months prior to the subsequent period of extended operation. Change Notice 3 Serial No.: 19-248 Page 39 of 74

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Change Notice 3 Appendix B - Aging Management Programs 82.1.16 Fire Water System Program Description The Fire Water System program is an existing condition monitoring program that manages loss of material, flow blockage, cracking and loss of coating integrity for in-scope water-based fire protection systems. This program manages aging effects by conducting periodic visual inspections, flow testing, and flushes. Testing and inspections are conducted on a refueling outage interval as allowed by NUREG-2191, Section XI.M27, Table XI.M27~1. "Fire Water System Inspection and Testing Recommendations". There are no nozzle strainers, glass bulb sprinklers, fire pump suction strainers, or foam water sprinkler systems within the scope of subsequent license renewal. The Fire Water System program will include testing a representative sample of the sprinklers prior to fifty years in service with additional representative samples tested at 10-year intervals. Sprinkler testing will be performed consistent with the 2011 Edition of NFPA 25, "Standard For The Inspection, Testing and Maintenance of Water-Based Fire Protection Systems," Section 5.3.1. The fifty year in-service date for sprinklers is October 26, 2021. Portions of water-based fire protection system components that have been wetted, but are normally dry, such as dry-pipe or preaction sprinkler system piping and valves, were designed and installed with a configuration and pitch to allow draining. With the exception of two locations, Engineering walkdowns confirmed the as-built configuration that allows draining and does not allow water to collect. Corrective actions have been initiated for the two locations to verify a flow blockage condition does not exist and to restore the two locations to original configuration requirements that allow draining and do not allow water to collect. After corrective actions, portions of the water-based fire protection system that have been wetted, but are normally dry, will not be subjected to augmented testing and inspections beyond those required by NUREG-2191, AMP XI.M27, Table XI. M27-1. The water-based fire protection system is normally maintained at required operating pressure and is monitored such that loss of system pressure is detected and corrective actions initiated. A low pressure condition is alarmed in the Main Control Room by the auto start of the electric motor driven fire pump, followed by the start of the diesel-driven fire pump if the low pressure condition continues to exist. The status of the fire pumps is indicated in the Main Control Room and at the fire pump control panels in the pump house. Both fire pumps may be manually started from the control room. Piping wall thickness measurements are conducted when visual inspections detect surface irregularities indicative of unexpected levels of degradation. When the presence of organic or inorganic material sufficient to obstruct piping or sprinklers is detected, the material is removed and the source is detected and corrected. Page B-108 Serial No.: 19-248 Page 40 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs Inspections and tests are performed by personnel qualified in accordance with procedures and programs to perform the specified task. Non-code inspections and tests follow procedures that include inspection parameters for items such as lighting, distance, offset, presence of protective coatings, and cleaning processes that ensure an adequate examination. If a flow test (i.e., NFPA 25, 2011 Edition, Section 6.3.1) or a main drain test (i.e., NFPA 25, 2011 Edition, Section 13.2.5) does not meet the acceptance criteria due to current or projected degradation, additional tests are conducted. The number of increased tests is determined in accordance with the site's corrective action process; however, there are no fewer than two additional tests for each test that did not meet the acceptance criteria. The additional inspections are completed within the interval (i.e., five years or annual/refueling) in which the original test was conducted. If subsequent tests do not meet the acceptance criteria, an extent of condition and extent of cause analysis is conducted to determine the further extent of tests required. The additional tests will include at least one test at the other unit on site with the same material, environment, and aging effect combination. In addition to piping replacement, actions will be taken to address instances of recurring corrosion due to microbiological induced corrosion. Low Frequency Electromagnetic Technique (LFET) or similar scanning technique will be used for screening 100 feet of accessible piping during each refueling cycle to detect changes in the wall thickness of the pipe. Thinned areas found during the LFET scan are followed up with pipe wall thickness examinations to ensure aging effects are managed and that wall thickness is within acceptable limits. In addition to the pipe wall thickness examination, opportunistic visual inspections of the fire protection system will be performed whenever the fire water system is opened for maintenance. Aging of the external surfaces of buried and underground fire main piping is managed by the Buried and Underground Piping and Tanks program (B2.1.27). Loss of material and cracking of the internal surfaces of cement lined buried and underground fire main piping are managed by the Internal Coatings/Linings For In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program (B2.1.28). Aging of the fire water storage tank bottom surfaces exposed to oil soil are managed by the Outdoor and Large Atmospheric Metallic Storage Tanks program (B2.1.17). When degraded coatings are detected during internal inspections of the fire water storage tanks, acceptance criteria, and corrective action recommendations of the Internal Coatings/Linings For In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program (B2.1.28) are followed. The training and qualification of individuals involved in coating/lining inspections of non-cementitious coatings/linings are conducted in accordance with ASTM International Standards endorsed in RG 1.54 including guidance from the staff associated with a particular standard. Page B-109 Serial No. : 19-248 Page 41 of 74

Change Notice 3 NUREG-2191 Consistency Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs The Fire Water System program is an existing program that, following enhancement, will be consistent, with exception, to NUREG-2191, Section XI M27, Fire Water System. Exception Summary The following program element(s) are affected: Detection of Aging Effects (Element 4)

1.

(Deleted exception for fire water storage tanks insulated external surface inspections - Change Notice 3) The fire water storage tanlm are insulated oarbon steel tanlm looated in an outdoor environment. ~JU REG 2191, AMP XI.M27, Table XI.M27 1 and note 10 recommends the insulated mcternal surfaoes of fire water storage tan Im be inspeoted for signs of degradation on a refueling outage interval for signs of degradation. This would require insulation removal each refueling cycle. Therefore, inspections of the external carbon steel surfaces of the fire water storage tanks will be performed on a 10 year frequency during the subsequent period of operation. Justification for Exception: The line item in NU REG 2191, Section XI.M27, Table XI.M27 1, for water storage tank external surfaces recommends the inspection guidance of NFPA, 2011 Edition, Section 9.2.5.5, which requires inspeotion of insulated tank surfaoes. ~ff PA, 2011 Edition, Scotian 9.2.5.5, does not provide specifio inspection guidance for oorrosion of metallio surfaoes under insulation in an outdoor air environment. NU REG 2191, Section XI.M29, Outdoor and Large Atmospheric Metallic Storage Tanks, element 4, provides inspeotion guidanoe for oorrosion under insulation for insulated oarbon steel tanl<s looated in an outdoor environment. NUREG 2191, Section XI.M29, Table XI.M29 1, recommends a 10 year frequency for corrosion under insulation during the subsequent period of operation.

2.

NUREG-2191, Table XI.M27-1, note 10 recommends main drain tests at each water-based system riser to determine if there is a change in the condition of the water piping and control valves on an annual or refueling outage interval. Surry Power Station will perform the main drain tests on twenty percent of the standpipes and risers every refueling cycle. Justification for Exception As indicated by NUREG-2191 Table XI.M27-1, note 10, access for some inspections is feasible only during refueling outages which are scheduled every eighteen months. Main drain tests on twenty percent of the standpipes and risers every eighteen months provide adequate information to determine the condition of the fire water piping is maintained consistent with the design basis. Page B-11 0 Serial No. : 19-248 Page 42 of 74

Change Notice 3 Enhancements Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs Prior to the subsequent period of extended operation, the following enhancement(s) will be implemented in the following program element(s): Parameters Monitored or Inspected (Element 3), Detection of Aging Effects (Element 4), Acceptance Criteria (Element 6), and Corrective Actions (Element 7)

1.

(Sprinkler inspections - Completed Change Notice 1)

2.

Prior to 50 years in service, sprinkler heads will be submitted for field-service testing by a recognized testing laboratory consistent with NFPA 25, 2011 Edition, Section 5.3.1. Additional representative samples will be field-service tested every 10 years thereafter to ensure signs of aging are detected in a timely manner. For wet pipe sprinkler systems, a one-time test of sprinklers that have been exposed to water including the sample size, sample selection criteria, and minimum time in service of tested sprinklers will be performed. At each unit, a sample of 3% or a maximum of ten sprinklers with no more than four sprinklers per structure shall be tested. Testing is based on a minimum time in service of fifty years and severity of operating conditions for each population. (Revised - Change Notice 2)

3.

Procedures will be revised to specify:

a. Standpipe and system flow tests for hose stations at the hydraulically most limiting locations for each zone of the system on a five year interval to demonstrate the capability to provide the design pressure at required flow.

b. Acceptance criteria for wet pipe main drain tests. Flowing pressures from test to test will be monitored to determine if there is a 10% reduction in full flow pressure when compared to previously performed tests. The Corrective Action Program will determine the cause and necessary corrective action.

c. If a flow test or a main drain test does not meet acceptance criteria due to current or projected degradation additional tests are conducted. The number of increased tests is determined in accordance with the corrective action process; however, there are no fewer than two additional tests for each test that did not meet acceptance criteria. The additional inspections are completed within the interval in which the original test was conducted. If subsequent tests do not meet acceptance criteria, an extent of condition and extent of cause analysis is conducted to determine the further extent of tests. The additional tests include at least one test at the other unit with the same material, environment, and aging effect combination.

d. Main drains for the standpipes associated with hose stations within the scope of subsequent license renewal will also be added to main drain testing procedures.

Page B-111 Serial No.: 19-248 Page 43 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B -Aging Management Programs

4.

Procedures will be revised to perform internal visual inspections of sprinkler and deluge system piping to identify internal corrosion, foreign material, and obstructions to flow. Follow-up volumetric examinations will be performed if internal visual inspections detect an unexpected level of degradation due to corrosion product deposition. If organic or foreign material, or internal flow blockage that could result in failure of system function is identified, then an obstruction investigation will be performed within the Corrective Action Program that includes removal of the material, an extent of condition determination, review for increased inspections, extent of follow-up examinations, and a flush in accordance with NFPA 25, 2011 Edition, Annex 0.5, Flushing Procedures. The internal visual inspections will consist of the following : (Relocated from Enhancement 1 O and Corrected - Change Notice 2)

a. Wet pipe sprinkler systems - 50% of the wet pipe sprinkler systems in scope for subsequent license renewal will have visual internal inspections of piping by removing a hydraulically remote sprinkler, performed every five years, consistent with NFPA 25, 2011 Edition, Section 14.2. During the next five-year inspection period, the alternate systems previously not inspected shall be inspected.

b. Pre-action sprinkler systems - pre-action sprinkler systems in scope for subsequent license renewal will have visual internal inspections of piping by removing a hydraulically remote nozzle, performed every five years, consistent with NFPA 25, 2011 Edition, Section 14.2.

c. Deluge systems - deluge systems in scope for subsequent license renewal will have visual internal inspections of piping by removing a hydraulically remote nozzle, performed every five years, consistent with NFPA 25, 2011 Edition, Section 14.2.

Parameters Monitored or Inspected (Element 3), Detection of Aging Effects (Element 4), and Monitoring and Trending (Element 5)

5.

Procedures will be revised to perform system flow testing at flows representative of those expected during a fire. A flow resistance factor (C-factor) will be calculated to compare and trend the friction loss characteristics to the results from previous flow tests.-(Renumbered :. Change Notice 2) Parameters Monitored or Inspected (Element 3) and Detection of Aging Effects (Element 4)

6.

(Hydrant flushing Completed :_Change Notice 1 and renumbered - Change Notice 2)

7.

Prior to the subsequent period of extended operation, the insulation on the exterior surfaces of the fire water storage tanks (FWSTs) will be permanently removed. Wall thickness measurements will be performed on external tank areas exhibiting unexpected degradation. Refurbishment/recoating will be performed consistent with the severity of the degradation identified and commensurate with the potential for loss of intended function. Inspections of external tank surfaces will be on a refueling cycle frequency. The Fire Viator Sy<stem program PageB-112 Serial No.: 19-248 Page 44 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs \\*,ill be revised to periodically inspect the insulated exterior surfaces of the fire water tanks on a 1 O year frequency during the subsequent period of operation. Insulation is removed to provide a minimum inspection population of 25 one square foot samples. The samples will be distributed in such a 1.vay that inspections occur on the tank dome, near the tank bottom, at points 1Nhere structural supports, pipe, or instrument nozzles penetrate the insulation and where,,..ater could collect. In addition, inspection locations will be based on the likelihood of corrosion under insulation occurring.(Renumbered - Change Notice 2 and revised - Change Notice 3)

8.

(Strainer flushing G~ompleted :..Change Notice 1 and renumbered - Change Notice 2)

9.

A procedure will be created to provide a Turbine Building oil deluge systems spray nozzle air flow test to ensure that patterns are not impeded by plugged nozzles, to ensure that nozzles are correctly positioned, and to ensure that obstructions do not prevent discharge patterns from wetting surfaces to be protected. (Renumbered - Change Notice 2) (Old Enhancement #9 was Rrelocated to Enhancement 4 - Change Notice 2) Detection of Aging Effects (Element4)

10. Procedure will be revised to provide inspection guidance related to lighting, distance and offset for non-ASME Code inspections. The procedure will specify adequate lighting be verified at the inspection location to detect degradation. Lighting may be permanently installed,

temporary, or portable (e.g., flashlight), as appropriate. For accessible surface inspections, inspecting from a distance of two to four feet (or less) will be appropriate. For distant surface inspections, viewing aids such as binoculars may be used. For viewing angles which may prevent adequate inspection, a viewing aid such as an inspection mirror or boroscope should be used. 11. The Unit 1 hydrogen seal oil system deluge sprinkler pipe and Unit 1 station main transformer '1A' deluge sprinkler piping will be reconfigured to allow drainage. As part of the drainage reconfiguration, visual inspections and wall thickness measurements will be performed on the Unit 1 hydrogen seal oil system deluge sprinkler pipe that does not drain. In addition, wall thickness examination of the Unit 1 main transformer deluge sprinkler piping that does not allow drainage will also be performed as part of the drainage reconfiguration. Piping with unexpected degradation will be replaced. (Revised - Change Notice 3)

12. The program will be revised to require inspections and tests be performed by personnel qualified in accordance with site procedures and programs for the specified task. (Added Change Notice 2)

13. Procedures will be revised to require when degraded coatings are detected by internal coating inspections, acceptance criteria and corrective action recommendations consistent with the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers and Page B-113 Serial No.: 19-248 Page 45 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs Tanks (B2.1.28) program are followed in lieu of NFPA 25 section 9.2.7 (1), (2), and (4). When interior pitting or general corrosion (beyond minor surface rust) is detected, tank wall thickness measurements are conducted as stated in NFPA 25 Section 9.2.7(3) in vicinity of the loss of material. Vacuum box testing as stated in NFPA 25 Section 9.2.7(5) is conducted when pitting, cracks, or loss of material is detected in the immediate vicinity of welds. (Added Change Notice 2) Detection of Aging Effects (Element4) and Acceptance Criteria (Element 6)

14. Procedures will be revised to address recurring internal corrosion with the use of Low Frequency Electromagnetic Technique (LFET) or a similar technique on 100 feet of piping during each refueling cycle to detect changes in the pipe wall thickness. LFET screening or a similar technique will also be performed on accessible interior fire water storage tank bottoms during periodic inspections. The procedure will specify thinned areas found during the LFET screening be followed up with pipe wall thickness examinations to ensure aging effects are managed and wall thickness is within acceptable limits. In addition to the pipe wall thickness examination, the performance of opportunistic visual inspections of the fire protection system will be required whenever the fire water system is opened for maintenance.

Operating Experience Summary The following examples of operating experience provide objective evidence that the Fire Water System program has been, and will be effective in managing the aging effects for SSCs within the scope of the program so that their intended functions will be maintained consistent with the current licensing basis during the subsequent period of extended operation.

1.

In January 2012, an Engineering walkdown of the fire protection piping header along the north wall of the Unit 2 Turbine Building revealed a potential leak location on the supply line to a hose rack. The flanged connection and straight pipe were removed and replaced.

2.

In January 2012, a section of 2-inch fire protection "drop" piping in the Turbine Building developed a leak. The investigation for extent of condition and determination for the extent of fire protection piping to be inspected and replaced, as necessary, involved inspections of three locations in the Turbine Building and three locations in the Auxiliary Building. Microbiologically induced corrosion (MIC) was evident in many locations, but the extent of corrosion was not as severe in the Auxiliary Building as it was in the Turbine Building. Despite the less severe corrosion in the Auxiliary Building, the three segments of piping that were inspected were replaced. Similarly, one of the three segments of piping in the Turbine Building was replaced. A capital project was proposed for a multi-year process of replacing segments of 2-inch, 4-inch, and 10-inch piping in the Turbine Building. The initial phase that was completed included replacing 200 feet of ten inch piping in the Turbine Building. Additional phases were proposed, and described in the Fire Protection Strategic Plan. See April 2013 and November 2015 operating experience. Page B-114 Serial No.: 19-248 Page 46 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs

3.

In June 2012, during inspection of Auxiliary Building fire protection piping minor sediment was discovered in the supply header to the Unit 1 cable tunnel sprinklers. Debris and MIC nodules were discovered inside a spool piece and accessible four inch piping. The sediment and debris were removed, the visual inspection was performed, and the blind flanges and spool pieces were replaced. The necessary pipe replacement is included in the Fire Protection Strategic Plan.

4.

In March 2013, NRC Information Notice 13-06, "Corrosion in Fire Protection Piping Due to Air and Water Interaction", identified industry operating experience involving the loss of function of fire protection water systems due to the potential for adverse air and water interactions in pre-action and dry-pipe systems. Engineering evaluated the potential for similar adverse conditions and associated degradation in deluge systems at Surry Power Station that are periodically flow tested. Subsequently, in January 2018, a walkdown was performed to confirm that plant design specifications on drainage features for piping downstream of all in-scope pre-action and deluge valves in the fire protection system continued to be in effect. Two locations, one relating to main transformer 1A and one relating to Unit 1 generator hydrogen seal oil system, were identified as having a potential for adverse air and water interactions and entered into the corrective action program.

5.

In April 2013, a section of two 10-inch fire protection system piping in the Turbine Building developed a leak. A walkdown of six locations was performed to determine extent of condition in the Turbine Building and the Auxiliary Building. MIC was evident in four locations, but the extent of corrosion in the Auxiliary Building was not as severe. Replacement of 4-inch and 10-inch fire protection header is a like-for-like replacement. The replacement of the Turbine Fire Protection Header was split into four different phases. One phase was to be accomplished each year. The second phase is planned to replace approximately 400 feet of ten-inch header pipe and 200 feet of two-inch hose station pipe. The necessary pipe replacement is included in the Fire Protection Strategic Plan.

6.

In February 2014, visual and volumetric inspections were performed for Fire Protection/domestic water storage tank 1A to determine the extent of additional degradation that had occurred since similar inspections were completed in December 2008. The most significant degradation was noted on the tank floor. The result of the visual inspection was that coating degradation was continuing, and that some bare metal was evident. Similarly, volumetric examinations found additional thinning for the tank floor. /\\n engineering e*;aluation projeoted that the tanl( floor plate would reaoh minimum aooeptable thiokness prior to the expiration of the Unit 2 renmued operating lioense. Monitoring of the tank floor will oontinue until the tank floor is repaired or replaoed. The neoessary tanl( repair or replaoement is inoluded in the fire Proteotion Strategio Plan. Follow-up visual examinations were performed in August 2018 and follow-up wall thickness examinations were performed in March 2019. Prior wall thickness measurements were confirmed to be attributed to laminations that existed from original steel plate fabrication. An engineering evaluation projected the tank floor plate Page B-115 Serial No.: 19-248 Page 47 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs would maintain acceptable wall thickness thoughout the subsequent period of extended operation. Work orders were generated to refurbish/recoat the FWST interior surfaces prior to the subsequent period of extended operation.

7.

In August 2014, visual and volumetric inspections were performed for Fire Protection/domestic water storage tank 1 B to determine the extent of additional degradation that had occurred since similar inspections were completed in December 2008. The most significant degradation was noted on the tank floor. The result of the visual inspection was that coating degradation was continuing, and that some bare metal was evident. Volumetric mmminations found some thinning of the tank floor. /\\n engineering evaluation projected that the tank floor plate would reach minimum acceptable thicl<ness prior to the expiration of the Unit 2 renewed operating license. Monitoring of the tank floor will continue until the tanl{ floor is repaired or replaced. Follow-up visual examinations were performed in August 2018 and follow-up wall thickness examinations were performed in March 2019. Prior wall thickness measurements were confirmed to be attributed to laminations that existed from original steel plate fabrication. An engineering evaluation projected the tank floor plate would maintain acceptable wall thickness thoughout the subsequent period of extended operation. Work orders were generated to refurbish/recoat the FWST interior surfaces prior to the subsequent period of operation.

8.

In September 2014, a materials analysis was performed on buried cement lined grey cast iron fire main piping that was fractured during flow testing of hose station valves. The fracture was attributed to a latent material defect in the cast iron. The piping was removed and replaced with an equivalent spool piece. Based on the oxidation along the top segment of the crack, the pipe was cracked for a long period of time. High levels of calcium deposits on the fracture (from the cement lining) indicate that the pipe was partially cracked at the top segment before factory installation of the cement liner (manufacturing process). Material analysis of the pipe determined that the microstructure consisted of graphite flakes that were approximately 75% ferrite and 25% pearlite. This resulted in a reduction in the supplied material hardness. Failure of pipe was not preventable through maintenance. The failure was caused by ground settling. During the pipe replacement it was observed that there was vertical misalignment between the replacement pipe and the existing buried pipe, which indicated that the buried side piping was exerting a large bending load at the anchor/foundation. This bending load along with the pre-existing crack and lower hardness value caused the pipe fracture. The balance of the failed pipe was found in good condition with no significant loss of cement lining material, corrosion, cracking, fouling, or reduction of pipe interior diameter.

9.

In November 2015, an effectiveness review of the Fire Protection Program aging management activity (AMA) (UFSAR Section 18.2.7) was performed. The AMA was evaluated against the performance criteria identified in NEI 14-12 for the Detection of Aging Effects, Corrective Actions, and Operating Experience program elements. A comprehensive fire water system assessment recommended a large scale piping replacement of turbine building and auxiliary Page B-1 16 Serial No.: 19-248 Page 48 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs building piping. The large scale piping replacement project to be performed over multiple refueling outages was identified as a measure to address degradation in carbon steel system piping and to ensure that system intended functions were maintained. Completed and closed phases of this effort have included replacement of approximately 400 feet of 4 inch piping and 200 feet of 2 inch piping in 2014 and approximately 567 feet of 4 inch piping and 303 feet of 2 inch piping in 2015. An additional phase replacing approximately 175 feet of 4 inch piping and 100 feet of 2 inch piping has been completed and is awaiting final testing. Work documents for additional phases are planned and issued for work extending into 2019.

10. In April 2016, March 2019, results from fire protection system 2500 gpm flow tests with the motor driven fire pump in April 2016, July 2013, and April 2010 from 2014 through 2019 consistently showed that thesatisfactory system pressure is higher than the required value for the corresponding flow rate. In 2016, the result indicated that the measured pressure exceeded the required pressure by fourteen psi. In 2013, the measured pressure was thirteen psi higher than required. The result in 2010 measured a pressure that v,as 19 psi higher than required. The trend from these results does not indicate significant degradation over the s+*five-year interval, particularly considering the two most recent measurements. Results from fire protection system 2500 gpm flow tests with the diesel driven fire pump from 2014 through 2019 also consistently showed satisfactory system pressure for the corresponding flow rate.

There is confidence that continued implementation of flow monitoring for the fire protection system using the three year interval required by the Technical Requirements Manual will effectively manage aging prior to a loss of intended function. 11. 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.

12. In November 2017, as part of oversight reviews of the Fire Protection Program AMA (UFSAR Section 18.2.7), an inconsistency was identified in the performance interval for system integrity demonstration by main drain testing. The test interval had been extended from quarterly to each 18 months but the extended interval had not been incorporated into program documents.

An Engineering Assignment to review operating experience to trended performance data to 2011 has been completed with no significant degrading trends observed. The new interval is Page B-11 7 Serial No. : 19-248 Page 49 of 74 l

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs consistent with the test interval of NFPA 25 (2011 Edition) Table 13.1.1.2 modified by NUREG-2191, Section XI.M27, Table XI.M27-1, Note 10.

13. In January 2018 an aging management program effectiveness review was performed for the Fire Protection Program AMA (UFSAR Section 18.2.7). Information from the summary of that effectiveness review is provided below:

The Fire Protection Program AMA is meeting or exceeding the requirements of selected NEI 14-12, "Aging Management Program Effectiveness," elements. Key activities of the Fire Protection Program AMA that were reviewed include the inspection of components, the evaluation of inspection results, repairs/replacements, corrective actions, and AMA document updates. Engineering reports from 2006 to 2017 of inspections results were reviewed to confirm inspection frequencies were conducted at appropriate intervals and corrective actions taken consistent with the observed aging degradation. The review also included pertinent issues found in the Corrective Action Program from 2006 through 2017 for age related degradation of fire protection components within the scope of license renewal. In the past, multiple fire water piping leaks had been identified in the Unit 1 and Unit 2 Turbine Buildings. As a result, a five phase large scale fire protection piping replacement project has been underway since 2015 to replace Turbine Building header piping and hose station piping as well as the Unit 1 and Unit 2 Auxiliary Building Hose station piping. Two of the Turbine Building phases are complete and two are waiting on testing. Phase five includes the remaining scope in the turbine building and the entire scope in the Auxiliary Building and is planned to start in 2018. Once complete, a large majority of the above ground fire protection piping in the plant will have been replaced, including areas where reoccurring leaks were previously identified. The fire water/domestic water storage tanks are managed by the Tank Inspection Activities AMA (UFSAR Section 18.1.3); but, are also discussed here for overall fire protection performance considerations. The fire water/domestic water storage tanks were found to have failing internal coatings and loss of material on the tank floors. Estimates for projected useable tank lifetime and evaluations for additional monitoring were performed. Recommendations are being prepared for repair or replacement project considerations. Multiple operating issues, and obsolescence of the diesel driven fire pump resulted in a design change that replaced the diesel driven fire pump and associated control panel. The new diesel driven fire pump has exhibited substantially improved performance compared to the original fire pump. Activities to implement NFPA 25, 1998 Edition, Section 2-3.1.1 (1998 edition), testing of sprinklers that have been in service for fifty years have been initiated to prove continued functionality. The Unit 1 and Unit 2 turbine building sprinklers have been sampled and will be tested by 2021, when fifty years of service is reached. Page B-118 Serial No.: 19-248 Page 50 of 74

Change Notice 3 Recurring Internal Corrosion (RIC) Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs Recurring internal corrosion, including through-wall failures due to microbiological induced corrosion, has occurred on several occasions. Periodic fire protection system piping flushes, flow testing and piping thickness measurements will be performed to identify pipe degradation prior to loss of system intended function. Periodic visual inspections and tank bottom thickness measurements are performed on the fire water storage tanks. In addition to recent piping replacements in the Turbine Building and the Auxiliary Building to address instances of RIC due to microbiologically-influenced corrosion, Low Frequency Electromagnetic Technique (LFET) or a similar technique on 100 feet of piping during each refueling cycle to detect changes in the pipe wall thickness. LFET screening or a similar technique will also be performed on accessible interior fire water storage tank bottoms during periodic inspections. Thinned areas found during the LFET scan are followed-up with pipe wall thickness examinations to ensure aging effects are managed and that wall thickness is within acceptable limits. In addition to the pipe wall thickness examination, opportunistic visual inspections of the fire protection system will be performed whenever the fire water system is opened for maintenance. The above examples of operating experience provides objective evidence that the Fire Water System program includes activities to perform periodic fire main and hydrant inspections and flushing, sprinkler inspections, functional test, and flow tests to identify loss of material, flow blockage, and loss of coating integrity for in-scope water-based fire protection systems within the scope of subsequent license renewal, and to initiate corrective actions. Occurrences identified under the Fire Water System 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. Appropriate 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 Fire Water System program, following enhancement, will effectively identify aging, and initiate corrective actions, prior to a loss of intended function. Conclusion The continued implementation of the Fire Water System program, following enhancement, will provide 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. Page B-119 Serial No.: 19-248 Page 51 of 74

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Change Notice 3 Appendix B - Aging Management Programs 82.1.33 Masonry Walls Program Description The Masonry Walls program is an existing condition monitoring program that manages loss of material, cracking, and loss of material (spalling and scaling) for masonry walls. The Masonry Walls program is implemented as part of the Structures Monitoring program (B2.1.34). The Masonry Walls program consists of inspections, consistent with IE Bulletin 80-11 (IEB 80-11), "Masonry Wall Design," and plant-specific monitoring proposed by IN 87-67, "Lessons Learned from Regional Inspections of Licensee Actions to IE Bulletin 80-11," for managing shrinkage, separation, gaps, loss of material and cracking of masonry walls such that the evaluation basis is not invalidated and intended functions are maintained. The Masonry Walls program relies on periodic visual inspections, conducted by qualified personnel at a frequency not to exceed five years, to monitor and maintain the condition of masonry walls within the scope of subsequent license renewal so that the established evaluation basis for each masonry wall remains valid during the subsequent period of extended operation. Qualifications for personnel perform ing inspections and evaluations are consistent with ACI 349.3R-02, "Evaluation of Existing Nuclear Safety-Related Concrete Structures". Inspections are performed and results evaluated consistent with applicable industry documents to ensure that a loss of intended function does not occur. Conditions found to impact the intended function of the masonry wall or invalidate its evaluation basis are documented and entered into the Corrective Action Program for evaluation which will result in analysis, repair or replacement. Masonry walls that are considered fire barriers are also managed by the Fire Protection program (B2.1.15). Steel elements of masonry walls are visually inspected by the Structures Monitoring program (B2.1.34). NUREG-2191 Consistency The Masonry Walls program is an existing program that, following enhancement, will be consistent with NUREG-2191, Section XI.S5, Masonry Walls. Exception Summary None Enhancements Detection of Aging Effects (Element 4)

1.

Procedures will be revised to clarify qualifications for personnel performing inspections of masonry walls and concrete to be consistent with ACI 349.3R-02. Page B-224 Serial No.: 19-248 Page 52 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs Monitoring and Trending (Element 5). Acceptance Criteria (Element 6)

2.

Procedures will be revised to explicitly address the trending of inspection results and projection to the next inspection interval. The procedure will be revised to include acceptance criteria for masonry wall inspections that will be used to ensure observed aging effects (cracking, loss of material, or gaps between the structural steel supports and masonry walls) do not invalidate the evaluation basis of the wall or impact its intended function. Operating Experience Summary The following examples of operating experience provide objective evidence that the Masonry Walls program has been, and will be effective in managing the aging effects for SSCs within the scope of the program so that their intended functions will be maintained consistent with the current licensing basis during the subsequent period of extended operation.

1.

In May 2009, during a walkdown of the Unit 1 Normal Switchgear Room a crack was identified around a concrete block wall. Inspection of the opposite side of the wall showed that the same crack existed on the Unit 2 side. The crack was less than 1/8 inch wide on both sides. The crack was repaired by work order. which was completed and accepted on 5/26/2009.

2.

In June 2012, while performing inspections, a 0.050 inch crack was observed in the masonry block and mortar of the Unit 1 'A' Fuel Oil Pump House exterior. The crack width decreased to 0.025 inch between the mortar and the masonry as it progressed along the west wall to the south wall. A work order was issued and the wall crack was repaired.

3.

In May 2015, an approximate 1/2 inch diameter hole was identified in a masonry block wall which is located between Battery Room 1A and the Unit 1 Emergency Switchgear Room. This wall is a fire barrier wall. The hole did not completely penetrate the block wall and may have been created for an anchor bolt that has since been removed. A work order was submitted and the hole in the block wall was repaired. The above examples of operating experience provide objective evidence that the Masonry Walls program includes activities to perform visual inspections to manage loss of material, cracking, and loss of material (spalling and scaling) for masonry walls within the scope of subsequent license renewal, and to initiate corrective actions. Occurrences identified under the Masonry Walls 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 Masonry Walls program, following enhancement, will effectively manage aging prior to a loss of intended function. Page B-225 Serial No.: 19-248 Page 53 of 74

Change Notice 3 Conclusion Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs The continued implementation of the Masonry Walls program, following enhancement, will provide 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. Page B-226 Serial No. : 19-248 Page 54 of 74

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Change Notice 3 Appendix B - Aging Management Programs 82.1.34 Structures Monitoring Program Description The Structures Monitoring program is an existing condition monitoring program that manages aging of the structures and components that are within the scope of subsequent license renewal by managing the following aging effects:

• Cracking
• Cracking and distortion
• Cracking, loss of material
• Cracking, loss of bond, and loss of material (spalling, scaling)
• Increase in porosity and permeability, cracking, loss of material (spalling, scaling)
• Loss of material
• Loss of material, loss of form
• Loss of material (spalling, scaling) and cracking
• Loss of material, change in material properties
• Loss of mechanical function
• Loss of preload
• Loss of sealing
• Reduction in concrete anchor capacity
• Reduction of foundation strength and cracking
• Reduction or loss of isolation function The Structures Monitoring program implements the requirements of 1 O CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," consistent with guidance of U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.160, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," and Nuclear Management and Resources Council 93-01, "Industry Guidelines for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants". The scope of the Structures Monitoring program includes structures and components in the scope of subsequent license renewal. The program relies on periodic visual inspections to monitor and maintain the condition of structures and components within the scope of subsequent license renewal. Inspections are conducted by qualified personnel at a frequency not to exceed five years, except for wooden poles, which will be inspected on a 10-year frequency. The interval between successive recurring inspections may be decreased based on conditions discovered in previous inspections.

Page B-227 Serial No.: 19-248 Page 55 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs Structural monitoring inspections consist primarily of periodic visual examination of accessible structures and components performed by qualified personnel. For concrete and associated components, ACl-349.3R, "Evaluation of Existing Nuclear Safety-Related Concrete Structures," and other applicable industry documents are used as guidance for the inspections, inspector qualifications, and evaluation of inspection results. The inspection program for structural steel is similar to the concrete program and is based on the guidance provided in the AISC Specification for Structural Steel Buildings and Code of Standard Practice. For earthen structures, evaluation of inspection results is performed by a qualified civil/structural engineer. Procedures will include preventive actions to provide reasonable assurance of structural bolting integrity, as discussed in Electric Power Research Institute (EPRI) documents (such as EPRI NP-5067, "Good Bolting Practices, A Reference Manual for Nuclear Power Plant Maintenance Personnel," and TR-104213, "Bolted Joint Maintenance & Application Guide"), American Society for Testing and Materials (ASTM) standards, and AISC specifications, as applicable. In order to evaluate the potential of water to cause degradation of inaccessible below-grade concrete, samples of groundwater will be taken at intervals not to exceed five years. The water chemistry is evaluated, and should the results of water testing indicate potentially harmful levels of substances such as chlorides > 500 ppm, sulfates > 1,500 ppm, or a pH < 5.5, inaccessible areas are assessed for aging when aging degradation exists in accessible areas and opportunistically inspected when excavated. Ground water monitoring has shown the ground water to be non-aggressive, except for one sampling point. In 2007, a sample with a significantly high chloride level was obtained from the Turbine Building sump. Subsequent sample results from this sump have found additional chloride levels above the acceptance limit. An inspection was performed to assess the structure for any degradation that could be attributed to the elevated levels of chloride. The inspection found no evidence of significant degradation. There have been no indications of concrete degradation due to elevated chloride levels anywhere in the plant. Engineering continues quarterly monitoring of the ground water in this sump. For surfaces provided with protective coatings, observation of the condition of the coating is an effective method for identifying the absence of degradation of the underlying material. Therefore, coatings on structures within the scope of the Structures Monitoring program are inspected only as an indication of the condition of the underlying material. ASME Code, Section XI visual examinations (VT-1) or surface examinations will be conducted to detect cracking of stainless steel and aluminum components exposed to aqueous solutions or air environments containing halides. A minimum sample of 25 inspections will be performed from each of the aluminum and stainless steel component populations every ten years. Page B-228 Serial No.: 19-248 Page 56 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs If any sampling-based inspections to detect cracking in aluminum and stainless steel do not meet the acceptance criteria, additional inspections will be conducted, unless the cause of the aging effect for each applicable material and environment is corrected by repair or replacement. There will be no fewer than five additional inspections for each inspection that did not meet acceptance criteria, or 20% of each applicable material, environment, and aging effect combination inspected,

• whichever is less. If any subsequent inspections do not meet acceptance criteria, an extent of condition and extent of cause analysis will be conducted to determine the further extent of inspections required. Additional samples will be inspected for any recurring degradation to ensure corrective actions appropriately address the associated causes. The additional inspections will include inspections of components with the same material, environment, and aging effect combination at both Unit 1 and Unit 2. The additional inspections will be completed within the interval (i.e., 10 year inspection interval) in which the original inspection was conducted. Where practical, the inspections will focus on the bounding or lead components most susceptible to aging because of time in-service, severity of operating conditions, and lowest design margin.

Concrete inspection results are evaluated to identify changes that could be indicative of Alkali-Silica Reaction (ASR) development. If indications of ASR development are identified, the evaluation considers the potential for ASR development in concrete that is within the scope of the ASME Section XI, Subsection IWL program (82.1.30), the Structures Monitoring program (B2.1.34), or the Inspection of Water-Control Structures Associated with Nuclear Power Plants program (82.1.35). In 1988, a research study was performed to evaluate the degradation processes that could affect the reinforced concrete structures. Concrete core samples were secured from the intake canal, Unit 1 Condensate Storage Tank Missile Shield, Unit 2 Safeguards Building and Unit 2 Containment. Based on testing of these samples, the study concluded that there was no evidence of ASR. Evaluation of inspection results includes consideration of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas. Structural sealants, seismic gap joint filler, vibration isolation elements, and other elastomeric materials are monitored for cracking, loss of material, and hardening. These elastomeric elements are acceptable if the observed loss of material, cracking, and hardening will not result in a loss of intended function. Visual inspection of elastomeric elements is supplemented by tactile inspection to detect hardening if the intended function is suspect. Procedures will include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting. For structural bolting consisting of ASTM A325, ASTM A490, ASTM F1852 and/or ASTM F2280 bolts, the preventive actions for storage, lubricant selection, and bolting and coating material selection discussed in Section 2 of the Research Council for Structural Page B-229 Serial No.: 19-248 Page 57 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs Connections publication, "Specification for Structural Joints Using High-Strength Bolts," will be used. Spent fuel pool (SFP) liner leakage through the leak chase channels is monitored. An alarm is provided on the SFP to sound at a level loss of approximately 0.5 feet (UFSAR Section 9.5.3.3). A review of recent leak chase channel monitoring reports shows acceptable leakage rates with no tell-tale drains being completely blocked. The Masonry Walls program (B2.1.33) and the Inspection of Water-Control Structures Associated with Nuclear Power Plants program (B2.1.35) are implemented as part of this program. NUREG-2191 Consistency The Structures Monitoring program is an existing program that, following enhancement, will be consistent with NUREG-2191, Section XI.S6, Structures Monitoring. Exception Summary None Enhancements Prior to the subsequent period of extended operation, the following enhancement(s) will be implemented in the following program element(s): Scope of Program (Element 1)

1.

Procedures will be revised to include inspection of the following structures that are within the scope of subsequent license renewal: decontamination building, radwaste facility, health physics yard office building, laundry facility, and machine shop. Inspections for the added structures will be performed under the enhanced program in order to establish quantitative baseline inspection data prior to the subsequent period of extended operation.

2.

Procedures will be revised to add the oiled-sand cushion to the inspection of the fire protection/domestic water tank foundation. (Added Change Notice 3) Preventive Actions (Element 2)

3.

Procedures will be revised to include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting. For structural bolting consisting of ASTM A325, ASTM A490, ASTM F1852 and/or ASTM F2280 bolts, the preventive actions for storage, lubricant selection, and bolting and coating material selection discussed in Section 2 of the Research Council for Structural Connections publication, "Specification for Structural Joints Using High-Strength Bolts," will be used. Page 8-230 Serial No.: 19-248 Page 58 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs

4.

The checklist for structural and support steel will be revised to indicate: "Are any connection members loose, missing or damaged (bolts, rivets, nuts, etc.)?". Detection of Aging Effects (Element 4)

5.

Procedures will be revised to eliminate options for inspector qualifications that are not consistent with ACI 349.3R-002.

6.

Procedures will be revised to specify that wooden pole inspections will be performed every ten years by an outside firm that provides wooden pole inspection services that are consistent with standard industry practice. Visual examinations may be augmented with soundings or other techniques appropriate for the type, condition, and treatment of the wooden poles, including borings to determine the location and extent of decay and excavation to determine the extent of decay at the groundline.

7.

Procedures will be revised to specify that evaluation of inspection results includes consideration of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas.

8.

Procedures will be enhanced to specify VT-1 inspections to identify cracking on stainless steel and aluminum components. A minimum of 25 inspections will be performed every ten years during the subsequent period of extended operation from each of the stainless steel and aluminum component populations assigned to the Structures Monitoring program. If the component is measured in linear feet, at least one foot will be inspected to qualify as an inspection. For other components, at least 20% of the surface area will be inspected to qualify as an inspection. The selection of components for inspection will consider the severity of the environment. For example, components potentially exposed to halides and moisture would be inspected, since those environmental factors can facilitate stress corrosion cracking. Corrective Actions (Element 7)

9.

Procedures will be enhanced to specify for the sampling-based inspections to detect cracking in stainless steel and aluminum components, additional inspections will be conducted if one of the inspections does not meet acceptance criteria due to current or projected degradation, unless the cause of the aging effect for each applicable material and environment is corrected by repair or replacement for all components constructed of the same material and exposed to the same environment. No fewer than five additional inspections for each inspection that did not meet acceptance criteria or 20 percent of each applicable material, environment, and aging effect combination will be inspected, whichever is less. Additional inspections will be completed within the 10-year inspection interval in which the original inspection was conducted. The responsible engineer will initiate condition reports to generate work orders to perform the additional inspections. The responsible engineer will evaluate the inspection results, and if the subsequent inspections do not meet acceptance criteria, an extent of Page 8-231 Serial No.: 19-248 Page 59 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs condition and extent of cause analysis will be conducted. The responsible engineer will then determine the further extent of inspections. Additional samples will be inspected for any recurring degradation to ensure corrective actions appropriately address the associated causes. The additional inspections will include inspections of components with the same material, environment, and aging effect combination at both Unit 1 and Unit 2. If any projected inspection results will not meet acceptance criteria prior to the next scheduled inspection, inspection frequencies will be adjusted as determined by the Corrective Action Program.

10. Procedures will be enhanced to specify that evaluation of neutron shield tank findings consider its structural support function for the reactor pressure vessel. (Added Change Notice 3)

11. Procedures will be enhanced to also include LOCAs as events that require evaluation for potentially degraded structures by Civil/Mechanical Design Engineering. (Added Change Notice 3)

Operating Experience Summary The following examples of operating experience provide objective evidence that the Structures Monitoring program has been, and will be effective in managing the aging effects for SSCs within the scope of the program so that their intended functions will be maintained consistent with the current licensing basis during the subsequent period of extended operation.

1.

In March 2007, a condition report (CR) was written to document a ground water monitoring sample with a chloride level of 1210 ppm, which exceeded the acceptance limit of <500 ppm. This sample was obtained from the Turbine Building sump. Corporate and site Engineering continue to monitor the quarterly sample results from the Turbine Building sump and have found additional chloride levels above the acceptance limit, as high as 2700 ppm. An inspection of the Turbine Building sump was performed in July 2008 to assess the sump structure for any degradation that could be attributed to the elevated level of chlorides. The inspection found no evidence of significant degradation to the interior concrete. There are no safety-related components in the vicinity of the Turbine Building sump, and there have been no indications of concrete degradation due to elevated chloride levels anywhere in the plant. The source of the chlorides has not been determined. The Turbine Building sump is the deepest dewatering point and closest to the Intake Canal where expected underground leakage from the canal could influence the chloride level. The potential for in-plant sources of chlorides reaching the sump via secondary drains or local ground water was studied and determined to be unlikely. An Engineering evaluation concluded that, while the chloride level has remained high in the Turbine Building sump, the other sumps/piezometer well locations, some of which are located in close proximity to the Turbine Building sump, have been found to be consistently within acceptable levels. Engineering will continue to monitor the chloride levels in the Turbine Building sump on a quarterly basis. The plant procedure has been revised Page B-232 Serial No. : 19-248 Page 60 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs to maintain sampling requirements so that trending may continue but eliminate the comparison to the acceptance criterion for this sampling point.

2.

In May 2011, a spall was found on the inside concrete surface of the bioshield wall of the Unit 2 Containment 'C' steam generator cubicle. The spall was approximately six inches long by six inches wide and 1-1/4 inches deep. The reinforcing steel was not exposed. It was determined that the bioshield wall remained fully functional, but the spalled concrete required repair prior to unit startup to prevent potential degradation of the reinforcing steel. A work order was submitted and the spalled concrete has been repaired.

3.

In December 2011, several embedded anchor bolts for the condenser unit of a Unit 1 Control Room chiller were found to be degraded. The anchor bolts displayed signs of corrosion and material loss. A work order was submitted and the anchor bolts were repaired in December 2011, which consisted of chipping the existing concrete around the anchor bolts until sound metal was reached, performing a weld repair of each anchor bolt, and repairing the concrete slab.

4.

In October 2012, leakage (approximately one gpm) was identified in the bottom portion of the steel to concrete joint (interface between the steel elbow and the concrete pipe) of the Unit 2 'D' 96-inch circulating water line. Corrosion and coating failure on the bottom third of the pipe was observed at this location. The urethane seal around the leading (upstream) edge of the joint was also missing and degraded. A work order was submitted and the Unit 2 'D' 96-inch circulating water line joint has been repaired.

5.

In January 2013, the Service Building roof was leaking, causing water to collect in two locations on the floor of the Service Building hallway. The first location was near the #1 EDG room. The second location was approximately halfway between the doors to the health physics area and the door to the operations annex. A work order was submitted and degraded roof areas were repaired.

6.

In December 2014, a CR was written to document a ground water monitoring sample that showed a chloride level of 610 ppm. The sampling point that exhibited unacceptable chloride levels is located adjacent to the Intake Canal, which draws water from the river. Three months later the same sampling point was found to have chlorides at 676 ppm. These values exceeded the acceptance limit of <500 ppm. The CR evaluation determined that the elevated chloride level was probably due to unusually low rain fall on the James River, temporarily increasing its natural salinity. Results from subsequent monitoring of ground water have been acceptable, and no degradation of concrete due to elevated chloride levels has been identified.

7.

In December 2015, an effectiveness review of the Civil Engineering Structural Inspection Activity (UFSAR Section 18.2.6) was performed. The aging management activity (AMA) was Page B-233 Serial No. : 19-248 Page 61 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs evaluated against the performance criteria identified in NEI 14-12 for the Detection of Aging Effects, Corrective Actions, and Operating Experience program elements. No gaps were identified by the effectiveness review.

8.

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.

9.

In November 2017, as part of oversight review activities, the Civil Engineering Structural Inspection Activity (UFSAR Section 18.2.6) AMA owners confirmed that AMA inspections had been performed and the inspections addressed the required SSCs consistent with the aging management activity commitments required in UFSAR Chapter 18. Security lighting poles were within the scope of license renewal but were not inspected during the Civil Engineering Structural Inspection Activity cycle completed in 2012. The omission of the security lighting poles from the 2012 inspection cycle was entered in the Corrective Action Program. In December 2017, Civil Engineering inspected the light poles and noted no degradation. The License Renewal Application and supporting documentation were reviewed for in-scope structures requiring inspection, and that information was cross-referenced with the implementing procedure to confirm aging management program commitments required by UFSAR Chapter 18 were satisfied. The security lighting poles are identified in the implementing procedure as being within scope of license renewal and will be inspected during subsequent structural inspections.

10. In January 2018, an aging management program effectiveness review was conducted for the Civil Engineering Structural Inspection Activity (UFSAR Section 18.2.6), which include the Structures Monitoring program (82.1.34), Masonry Walls program (82.1.33) and the Inspection of Water-Control Structures Associated with Nuclear Power Plants program (82.1.35). Information from the summary of that effectiveness review is provided below:

The Civil Engineering Structural Inspection Activity is meeting or exceeding the requirements of selected NEI 14-12, "Aging Management Program Effectiveness," elements. Key activities of the AMA that were reviewed included structural inspections for aging management that have been incorporated into the periodic inspections performed for Maintenance Rule compliance. Maintenance Rule inspections, along with trending and evaluation for evidence of Page B-234 Serial No.: 19-248 Page 62 of 74

Change Notice 3 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Appendix B - Aging Management Programs aging effects, ensure the continuing capability of civil engineering structures to meet their intended functions consistent with the current licensing basis. A 10-year review of inspection results and corrective actions did not identify any aging that resulted in a loss of intended function(s). 11. In March 2018, the existing Structures Monitoring program was revised to improve the inspection techniques and to adopt new inspection techniques to manage aging effects associated with ASR degradation of concrete structures and components consistent with industry operating experience IE Notice 2011-20 (IN 2011-20), "Concrete Degradation by Alkali-Silica Reaction," and EPRI Report #3002005389 (2015), "Tools for Early Detection of ASR in Concrete Structures." The above examples of operating experience provide objective evidence that the Structures Monitoring program includes activities to perform volumetric and visual inspections to identify aging effects for structures, structural supports, and structural commodities within the scope of subsequent license renewal, and to initiate corrective actions. Occurrences identified under the Structures Monitoring 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 Structures Monitoring program, following enhancement, will effectively manage aging prior to a loss of intended function. Conclusion The continued implementation of the Structures Monitoring program, following enhancement, will provide 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. Page B-235 Serial No.: 19-248 Page 63 of 74

Table C3.3-3 SLR Expert Panel Review Results Table Assembly Sub-assembly Component Material Screened-in Degradation Mechanisms3 -c e MRP-191, MRP-191, Expert 0 0

l Panel c /

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J 0 2018-022 Upper Control rod guide Bolts (Note 3) 31 6 ss None None sec.

L internals tube assemblies and Fatigue assembly flow downcomers C-tubes (Note 3) 304 ss Wear Wear

Wear, M

Fatigue Guide tube enclosures 304 ss SCC-W, SCC-W SCC-W, L (Note 3) Wear Fatigue Flanges - intermediate 304 ss SCC-W, SCC-W, SCC-W, L (Note 3) Fatigue Fatigue Fatigue CF8 SCC-W, SCC-W, SCC-W, L Fatigue, TE Fatigue, TE Fatigue, TE Flanges - lower 304 ss SCC-W, SCC-W, SCC-W, M Fatigue Fatigue

IASCC, Fatigue, IE CF8 SCC-W, SCC-W, SCC-W, M

Fatigue, Fatigue,

IASCC, TE, IE TE, IE

Fatigue, TE, IE Flexureless inserts (Unit 2 304 ss None None Fatigue L

only) (Note 3) Flexures (Unit 1 only) ~ X-750 sec sec

sec,

~ Fatigue NQN. § Guide plates (cards) 304 ss SCC-W, SCC-W, SCC-W, H

Wear, Wear,

Wear, Fatigue Fatigue Fatigue CF8 SCC-W, SCC-W, H

Wear, Wear,

Fatigue, Fatigue, TE TE, IE Surry Power Station, Units 1 and 2 Page C-29 Application for Subsequent License Renewal Appendix C - MRP-227-A Gap Analysis for PWR Vessel Internals Aging Management Q)

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Table C3.3-3 SLR Expert Panel Review Results Table Assembly Sub-assembly Component Material Screened-in Degradation Mechanismsa MRP-191, MRP-191, Expert Rev. O Rev. 1 Panel c / MRP 2018-022 Upper Mixing devices Mixing devices (Note 3) CF8 SCC-W, SCC-W, SCC-W, internals TE, IE TE, IE

Fatigue, assembly TE, IE (cont.)

Upper core plate and Fuel alignment pins 304 ss Wear

IASCC, fuel alignment pins

Wear, Fatigue, IE

~ Upper core plate 304 ss

Wear, Wear,

IASCC, Fatigue Fatigue, IE

Wear, Fatigue, IE Upper core plate insert 304 ss Wear, IE (Note 3)

Upper core plate insert bolts 316 ss

IASCC, (Note 3)

Fatigue, IE, ISR/IC Upper core plate insert locking 304L SS Fatigue, IE devices and dowel pins (Note 2) (Note 3) Upper core plate insert locking 316 ss Fatigue, IE devices and dowel pins (Note 2) (Note 3) Upper Bolting (Note 3) 316 ss None None None instrumentation conduit and supports 304 ss None None Brackets, clamps, terminal 304 ss None None sec. blks, conduit straps (Note 3) Fatigue CF8 TE

sec, Fatigue, TE, IE Conduit seal assembly: body, 304 ss None None

sec, tubesheets, tubesheet welds Fatigue (Note 3)

Surry Power Station, Units 1 and 2 Page C-31 Application for Subsequent License Renewal Appendix C - MRP-227-A Gap Analysis for PWR Vessel Internals Aging Management Q) Q) < u.!:1 u u ~ C: C: 0 ~ Q) E Q) w a. 0

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Table C3.3-3 SLR Expert Panel Review Results Table Assembly Sub-assembly Component Material Screened-in Degradation Mechanisms3 MRP-191, MRP-191, Expert Rev. O Rev. 1 Panel c / MRP 2018-022 Lower Core barrel Core barrel flange (surface: 304 ss SCC-W, SCC-W, SCC-W, internals upper fla nge weld is included Wear Wear

Wear, assembly below with upper core barrel Fatigue (cont.)

girth welds) Core barrel outlet nozzles 304 ss SCC-W, SCC-W, SCC-W, Fatigue Fatigue

Wear, Fatigue Lower core barrel axial welds 304 ss SCC-W, SCC-W, SCC-W, (includes MAW and LAW)

IASCC, IE IASCC, IE

IASCC,

~ Fatigue, IE, vs Lower core barrel girth welds 304 ss SCC-W, SCC-W, SCC-W, (~ LGW aREI lF\\OJ) IASCC, IE IASCC, IE

IASCC,

~ Fatigue, IE, vs Upper core barrel axial welds 304 ss SCC-W, SCC-W, IE SCC-W, (includes UAWJ IASCC, IE Fatigue Upper core barrel girth welds 304 ss SCC-W, SCC-W, IE SCC-W, (includes UFWJ Note 5 IASCC, IE Fatigue Diffuser plate Diffuser plate (Note 3) 304 ss None None None Flux thimble (tubes) Flux thimble tube plugs Alloy 600 SCC-W, SCC-W, (Note 3) IASCC, IE,

IASCC, vs Fatigue, IE, vs Flux thimbles (tubes)

Alloy 600 SCC-W, SCC-W,

IASCC, IASCC, Wear, IE,

Wear, vs Fatigue, IE, vs Head cooling spray Head cooling spray nozzles 304 ss None None Fatigue nozzles (Note 3)

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Table C3.3-3 SLR Expert Panel Review Results Table Assembly Sub-assembly Component Material Screened-in Degradation Mechanisms3 MRP-191, MRP-191, Expert Rev. O Rev. 1 Panel c / MRP 2018-022 Lower Irradiation specimen Irradiation specimen access 316 ss None internals guides plug (dowel pin) (Note 3) assembly Irradiation specimen access 304 ss IE IE None (cont.) plug (plug) (Note 3) Irradiation specimen access X-750 None plug (spring) (Note 3) Irradiation specimen guide 304 ss Wear, IE Wear, IE SCC-W, (Note 3)

Wear, Fatigue Lower core plate and Fuel alignment pins 304 ss

IASCC, IASCC, fuel alignment pins Wear, IE, Wear, IE, vs vs fr:IQ1w Lower core plate 304 ss SCC-W, SCC-W,

IASCC, IASCC,

IASCC, Wear,

Wear, Wear, Fatigue, IE, Fatigue, IE, Fatigue, IE, vs vs vs Lower support Lower support column bodies CF8 IASCC,TE, IASCC,TE,

IASCC, column assemblies IE, VS IE, VS

Fatigue, TE, IE, VS Lower support column bolts 316 ss

IASCC, IASCC,

Wear, Wear, Fatigue, IE, Fatigue, IE, VS, ISR/IC VS, ISR/IC Lower support column bolt 304L SS

IASCC, locking devices (Note 3)

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a.

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2.

3.

4.

5.

6.

7.

Stress corrosion cracking (SCC) [1A is applicable for sec welds (SCC-W)] Irradiation-assisted stress corrosion cracking (IASCC) Wear Fatigue (FAT) Thermal aging embrittlement (TE) Irradiation embrittlement (IE) Void swelling (VS) Thermal and irradiation-induced stress relaxation or irradiation creep (ISR/IC) P = Primary, E = Expansion, X = Existing, N = No additional measures Degradation mechanism added during Expert Panel review as indicated in L TR-AMLR-17-35 and L TR-AMLR-18-4. Alloy 600 was identified as the material for the support pin nuts at Surry Unit 1. These nuts were replaced as part of the control rod guide tube support pin replacement performed by AREVA. The AREVA evaluation indicates that the aging degradation mechanisms of concern are sec and irradiation-enhanced stress relaxation/irradiation-enhanced creep (ISR/IC). The upper core plate insert locking devices are 304L SS, and the dowel pins are 316 SS. No additional measures. The thermal shield flexure locking devices are 304L SS and the dowel pins are 304 SS. MRP-227, Revision 1, added expansion links from the upper flange weld (UFW) to the lower flange weld (LFW) and to the upper girth weld (UGW). For Unit 1, Babcock & Wilcox replaced the CRGT support pins and flexures with a modified design fabricated from Alloy X-750 during the CRGT replacement. The MRP-191, Revision 2, expert panel considered the Alloy X-750 flexures to be a Category C component. However, AREVA performed an evaluation of the replacement CRGT assemblies, including the replacement flexures. Section 4.1 of the AREVA report (AREVA Licensing Report ANP-357 4, Rev 0, "Surry Unit 1 Modified Replacement CRGT Assembly Reconciliation with MRP-227-A for an 80-Year License", September 2017) listed those Surry Unit 1 replacement CRGT assembly components that were assigned to Categories Band C (i.e., "non-Category A") and did not include the replacement flexures in those categories. Based on this designation as a Category A component, the replacement flexures require no additional measures. Since the flexure design was modified during CRGT replacement, the B&W classification is considered appropriate for the Surry reactor internals program. The fuel alignment pins screen in for multiple mechanisms because of the conservative screening criteria used and the high radiation exposure of the pin locations. However, for the fuel alignment pins in both the upper core plate and the lower core plate, degradation mechanisms other than wear are not expected to impact the function of the pins. either due to the limited amount of degradation anticipated or due to the redundancy of the pins (more than one per fuel assembly). Wear-related surface degradation that has been observed. particularly for pins with Malcomized hardening treatment, is considered the leading degradation mechanism for the fuel alignment pins. Surry Power Station, Units 1 and 2 Page C-41 Change Notice 3 Serial No.: 19-248 Application for Subsequent License Renewal Appendix C - MRP-227-A Gap Analysis for PWR Vessel Internals Aging Management Page 68 of 74

Table C4.3-1 Primary Components Primary Item Effect (mechanism) Expansion Link Examination Method / Frequency (Note 1) (Note 1) Control Rod Guide Loss of material (wear) QQntrQI rQg guige Visual (VT-3) inspections and quantitative Tube Assembly !1.1be QQn!in!JQ!Jli measurements are performed. Per the lleQ!iQn llhei;!lhli requirements ofWCAP-17451-P, the Guide plates ang Q-!1.1beli absence of significant degradation during (cards) <Note 1 BlNef:le the inspections in 2012, confirm that no additional inspection is required prior to the normal ten-year interval8 (Note 2). Control Rod Guide Cracking (SCC, Fatigue) Remaining Enhanced visual (EVT-1) examination to Tube Assembly Irradiation Embrittlement (IE) and accessible determine the presence of crack-like Lower flange welds, Thermal Embrittlement (TE) are CRGT assembly surface flaws in flange welds no later than 2 lower flange refueling outages from the beginning of the LFW applicable aging mechanisms welds first license renewal period and subsequent BM I column exam ination on a ten-year interval.b bodies (Note 3) Core Barrel Cracking (SCC) Upper girth weld Enhanced visual (EVT-1) examination, no Assembly (UGW) later than 2 refueling outages from the Upper flange weld; Lower flange beginning of the first license renewal period and subsequent examination on a ten-year UFW weld (LFW) interval.c Upper axial weld (UAW) Lower support forging. (Note 3) Core Barrel Cracking (SCC, IASCC, Fatigue) Middle and lower Periodic enhanced visual (EVT-1) Assembly Irradiation Embrittlement (IE) is an core barrel axial examination, no later than 2 refueling applicable aging mechanism. welds. outages from the beginning of the first Lower girth weld; license renewal period and subsequent LGW Upper core plate. examinations on a ten-year interval.d (Note 8) Lower support column bodies (cast). (Note 3) Surry Power Station, Units 1 and 2 Page C-54 Application for Subsequent License Renewal Appendix C - MRP-227-A Gap Analysis for PWR Vessel Internals Aging Management Examination Coverage An update provided in MRP 2018-007 indicates wear measurements to be obtained in 37 of the 48 CRGT locations. 100% of outer (accessible) CRGT lower flange weld surfaces and 0.25-inch of the adjacent base metal on the individual periphery CRGT assemblies (Notes 4 and 5). 100% of the accessible weld length of one side of the UFW and 0. 75-inch of adjacent base metal shall be examined. (Notes 6 and 9). 100% of the accessible weld length of the OD of the LGW and 0.75-inch of adjacent base metal shall be examined. (Note 9). Source of Revision/ Addition MRP-227, Rev. 1 added WCAP-17451-P, MRP-2018-007 supplements industry WCAP-17451-P requirements. Rev. 1 added Expansion to remaining CRGT lower flange welds; Rev. 1 removed Expansion to items upper core plate and lower support forging; added 0.25 inch of base metal to examination coverage. MRP-227A initially established examination coverage. MRP-227, Rev 1, removed Expansion to core barrel outlet nozzles, and to lower support column bodies; Rev. 1 added Expansion to UGW, LFW, and UAW, and to lower support forging/casting; reduced coverage to 25%. However MRP 2018-026 increased the required examination coverage. MRP 2018-022 added core barrel axial welds, upper core plate, and lower support column bodies as Expansion items. MRP 2018-026 revised the required examination coverage. Change Notice 3 Serial No.: 19-248 Page 69 of 74

Table C4.3-1 Primary Components Primary Item Effect (mechanism) Expansion Link Examination Method/ Frequency (Note 1) (Note 1) Alignment and Distortion (Loss of load due to None Direct measurement of spring height within Interfacing stress relaxation) (Note 17) three cycles of the beginning of (before or Components after) the first license renewal period. If the first set of measurements is not sufficient to Internals hold down assess remaining life, additional spring spring height measurements will be required.l (Note 17) Alignment and Cracking (SCC), Loss of material None Visual (VT-3) no later than 2 refueling Interfacing (Wear) outages from the beginning of the first Components license renewal period. k Clevis insert bolts Subsequent examinations on a ten-year Clevis insert dowels interval. (Note 4-819.) Alignment and Loss of material (Wear) None Visual inspection for top of CRGTs and/or Interfacing bottom of thermal sleeve guide funnel for Components indications of wear per MRP 2018-01 O (TB-07-02). MRP 2018-027 implements this Thermal sleeves inspection as described in NSAL 18-1. Thermal Shield Cracking (fatigue) or Loss of None Visual (VT-3) no later than 2 refueling Assem bly Material (wear) that results in outages from the beginning of the first thermal shield flexures excessive license renewal period. Subsequent Thermal shield wear, fracture, or full separation exam inations on a ten-year interval. 1 flexures Radial Support Loss of material (Wear) None Visual (VT-3) no later than 2 refueling Keys outages from the beginning of the first Radial support keys license renewal period. Subsequent examinations on a ten-year interval. Stellite wear surfaces (Note 4-819.) Alignment and Loss of material (Wear) None Visual (VT-3) no later than 2 refueling Interfacing outages from the beginning of the first Components license renewal period. Subsequent examinations on a ten-year interval. Clevis bearing Stellite wear surface (Note 4-819) Surry Power Station, Units 1 and 2 Page C-56 Application for Subsequent License Renewal Appendix C - MRP-227-A Gap Analysis for PWR Vessel Internals Aging Management Examination Coverage Measurements should be taken at several points around the circumference of the spring, with a statistically adequate number of measurements at each point to minimize uncertainty. All clevis insert bolts and clevis insert dowels. Wear surfaces for top of CRGT and/or bottom of thermal sleeve guide funnel per MRP 2018-010 (TB-07-02). 100% of thermal shield flexures Wear surfaces and radial support keys Wear surfaces and radial support keys Source of Revision/ Addition MRP-227A. A calculation of required hold down spring height for the 80-year design life confirms that the existing measured spring heights for both units are acceptable, and no further measurements are necessary Clevis insert bolts elevated to the Primary category by MRP 2018-022; the scope is expanded to include clevis insert dowels. Added as a Primary component in MRP 2018-022. MRP-227A Added as a Primary component in MRP 2018-022. Added as a Primary component in MRP 2018-022. Change Notice 3 Serial No.: 19-248 Page 70 of 74

j.

k.

I. Notes: 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 Direct measurements of the hold down spring height were performed for Unit 1 and Unit 2 in 2012. The measurements were obtained at 8 locations around the circumference of the spring. Three measurements were performed at each location. The results indicated an acceptable spring height that confirms the capability of the hold down spring to perform its intended function for 80 years of operation. The clevis insert bolting was inspected for integrity during the 2013 and 2014 outages for Unit 1 and Unit 2, respectively. VT-3 exams were performed using VT-1 acuity. The enhanced visual acuity was used specifically to address industry OE concerns of clevis insert bolt cracking. No issues were identified. VT-3 examinations were performed in 2013 for the six thermal shield flexures in Unit 1. The Unit 2 inspections for the six thermal shield flexures were performed in 2014. All inspection results were satisfactory; there was no evidence of cracking (fatigue) or loss of material (wear). Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3 of MRP-227-A. Examination method updated in MRP-227, Revision 1 based on issuance of WCAP-17451-P for industry use. Interim Guidance issued in PWROG Letter OG-18-76 amends the requirements regarding baseline examinations. The FMECA expert panel determined that Surry would follow MRP-227, Revision 1, for the expansion inspection components of the CRGT lower flange welds, which include the remaining CRGT lower flange welds and the BMI column bodies. The lower support columns (cast) and the upper core plate are added as Expansion links to the lower core barrel girth weld (Primary), and the lower core support forging would be added as an Expansion link to the upper core barrel flange weld (Primary). MRP-227-A Note: A minimum of 75% of the total identified sample population must be examined. Clarification in MRP-227, Revision 1, to state that 0.25 inch of the adjacent base metal must be examined for the CRGT lower flange welds. MRP-227-A Note: A minimum of 75% of the total weld length (examined+ unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined from either the inner or outer diameter for inspection credit. The examination coverage for core barrel welds was redefined in MRP-227, Revision 1. The upper girth weld was moved to an Expansion link from the upper flange weld in MRP-227, Revision 1. MRP 2018-026 revised the examination coverage to require a minimum of 50% of the circumference of either the ID or the OD of the weld being examined MRP-227-A Note: The lower core barrel flange weld may be alternatively designated as the core barrel-to-support plate weld. The lower flange weld was moved to an Expansion link from the upper flange weld in MRP-227, Revision 1. Bracket bolts are not applicable to the Surry design. MRP-227-A Note: A minimum of 75% of the total bolt population (examine+ unexamined), including coverage consistent with the Expansion criteria in Table 5-3 of MRP-227, must be examined for inspection credit. Corner bolts will be added as a Primary component in the next revision of MRP-227. They will be treated the same as baffle-former bolts. MRP-227-A Note: Void swelling effects on the component are managed through management of void swelling on the entire baffle-former assembly. Examination timing and frequency is updated in MRP-227, Revision 1, based in issuance of MRP 2017-009 for industry use. Language clarified/simplified in MRP-227, Revision 1. Sheath or C-tube wear measurement can be considered best practice, but optional until the inspection when the guide card wear just above the sheaths or C-tubes has a ligament that is worn-through or is projected to wear-through before the time of the next inspection. Added as a Primary component in MRP 2018-022. Surry Power Station, Units 1 and 2 Page C-58 Change Notice 3 Serial No.: 19-248 Application for Subsequent License Renewal Appendix C - MRP-227-A Gap Analysis for PWR Vessel Internals Aging Management Page 71 of 74

Table C4.3-2 Expansion Components Expansion Item Effect (mechanism) Primary Link Examination Method/ Frequency (Note 1) Upper Internals Assembly Cracking (Fatigue, wear) Core barrel Visual (VT-3) examination. Upper core plate Lower Girth Re-inspection every 10 years following Weld (Note 2) initial inspection. (Note 3) Control Rod Guide Tube Cracking (SCC, Fatigue) CRGT Lower Enhanced visual (EVT-1) examination to Assembly Irradiation Embrittlement (IE) Flange Welds determine the presence of crack-like and Thermal Embrittlement (TE) (Notes 2 and 6) surface flaws in flange welds. Remaining CRGT lower flange welds (Note 6) are applicable aging Subsequent examination on a ten-year mechanisms. interval. CQDlCQI RQd Guide Iube LQSS Qf Msi!erisil (Wesir) CRGTGuide Per the regiiiremeo!s Qf WCAP-17451-P Assembly Plsi!es (Qsirdsl RevisiQn 2. CQn!iniiQiis seQ!iQo shesiths sing Q-!ubes Bottom Mounted Cracking (Fatigue) including the CRGT Lower Visual (VT-3) examination of BMI column Instrumentation System detection of completely Flange Welds bodies as indicated by difficulty of insertion/ Bottom-mounted fractured column bodies. (Note 2) withdrawal of flux thimbles. instrumentation (BMI) Irradiation Embrittlement (IE) is Re-inspection every 10 years following column bodies an applicable aging mechanism. initial inspection. Flux thimble insertion / withdrawal to be monitored at each inspection interval. Core Barrel Assembly Cracking (SCC, IASCC) Lower core Enhanced visual (EVT-1) examination. Middle axial weld (MAW) Irradiation Embrittlement (IE) is barrel cylinder Re-inspection every 10 years following girth weld and Lower axial weld (LAW) the applicable aging (LGW) initial inspection. mechanism. Core Barrel Assembly Cracking (SCC) Upper core Enhanced visual (EVT-1) examination. Upper girth weld (UGW) barrel flange Re-inspection every 10 years following weld (UFW). initial inspection. Surry Power Station, Units 1 and 2 Page C-59 Application for Subsequent License Renewal Appendix C - MRP-227-A Gap Analysis for PWR Vessel Internals Aging Management Examination Coverage Source of Revision/ Addition 25% of accessible surfaces. MRP-227, Revision 1. (Notes 4 and 5) MRP 2018-022 specified VT-3 examination and 25% coverage. A minimum of75% of the CRGT assembly MRP-227, Revision 1, lower flange weld surfaces and 0.25 inch added the requirement of the adjacent base metal for the flange for 0.25 inch of welds not inspected under the Primary adjacent base metal. link. E2rnminsi!iQn QQversige (;)er !he MRP-227A regiiirements Qf WQAP-17 451-P ReyisiQn 2 100% of BMI column bodies for which MRP-227A difficulty is detected during flux thimble insertion/withdrawal. 100% of the accessible weld length of the MRP 2018-026 OD of the MAW and LAW and 0.75-inch of changed the adjacent base metal shall be examined. examination coverage. (Notes 4 and Q_a) 100% of the accessible weld length of one MRP-227, Revision 1, side of the UGW and 0.75-inch of added the UGW to the adjacent base metal shall be examined Expansion category as (Note ~ ) a link from Primary-Upper flange weld (UFW). MRP 2018-026 changed the examination coverage. Change Notice 3 Serial No.: 19-248 Page 72 of 74

Table C4.3-2 Expansion Components Expansion Item Effect (mechanism) Primary Link Examination Method/ Frequency (Note 1) Core Barrel Assembly Cracking (SCC) Upper core Enhanced visual (EVT-1) examination. Lower flange weld (LFW) barrel flange Re-inspection every 10 years following weld (UFW). initial inspection. Core Barrel Assembly Cracking (SCC, IASCC) Upper core Enhanced visual (EVT-1) examination. Upper axial weld (UAW) Irradiation Embrittlement (IE) is barrel flange Re-inspection every 10 years following an applicable aging mechanism. weld (UFW). initial inspection. Lower Internals Assembly Cracking (SCC) Upper Core Visual (E-V=t4VT-3) examination. Lower support forging Barrel Flange Re-inspection every 10 years following Weld (UFW) initial inspection. (Note 3) (Note 2) Lower Support Assembly Cracking (IASCC) including Lower core Visual (E-V=t4~ ) examination. Lower support column detection of completely barrel girth Re-inspection every 10 years following bodies (cast) fractured column bodies. Weld (Note 2) initial inspection. (Note 3) Irradiation Embrittlement (IE) is an applicable aging mechanism. Core Barrel Assembly Cracking (IASCC, Fatigue) Baffle-former Volumetric (UT) examination Barrel-former bolts (Note 7) Irradiation Embrittlement (IE), bolts Re-inspection every 10 years following void swelling, initial inspection. irradiation-enhanced stress relaxation (ISR) aging mechanisms Surry Power Station, Units 1 and 2 Page C-60 Application for Subsequent License Renewal Appendix C - MRP-227-A Gap Analysis for PWR Vessel Internals Aging Management Examination Coverage Source of Revision/ Addition 100% of the accessible weld length of the MRP-227, Revision 1, OD surface of the LFW and 0.75-inch of added the LFW to the adjacent base metal shall be examined. Expansion category as (Note ea) a link from Primary-Upper flange weld (UFW). MRP 2018-026 changed the examination coverage. 100% of the accessible weld length of one MRP-227, Revision 1, side of the UAW and Y.' of adjacent base added the UAW to the metal shall be examined (Note ~ ) Expansion category as a link from Primary-Upper flange weld (UFW). MRP 2018-026 changed the examination coverage. 25% of the bottom surface. MRP-227, Revision 1, (Notes 4 and 5) added this item to the Expansion category. MRP 2018-022 specified VT-3 examination and 25% coverage. 25% of accessible support column MRP-227, Revision 1, assemblies as visible from above the added this item to the lower core plate. (Note~ 4...aO.Q...2) Expansion category. MRP 2018-022 specified VT-3 examination and 25% coverage. 100% of accessible barrel-former bolts MRP-227A (minimum of 75% of the total population). Accessibility may be limited by presence of thermal shield or neutron pads. (Note4) Change Notice 3 Serial No.: 19-248 Page 73 of 74

Table C4.3-2 Expansion Components Expansion Item Effect (mechanism) Primary Link Examination Method/ Frequency Examination Coverage Source of Revision/ (Note 1) Addition Lower Support Assembly Cracking (IASCC, Fatigue) Baffle-former Volumetric (UT) examination 100% of accessible lower support column MRP-227A Lower support column bolts Irradiation embrittlement (IE), bolts Re-inspection every 10 years following bolts (minimum of 75% of the total and irradiation-enhanced stress initial inspection. population}, or as supported by relaxation (ISR) are applicable plant-specific justification. (Note 4) aging mechanisms Notes: 1 Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3 of MRP-227-A. 2 3 4 5 6 7 The FMECA expert panel determined that Surry would follow MRP-227, Revision 1, for the expansion inspection components of the CRGT lower flange welds, which include the remaining CRGT lower flange welds and the BMI column bodies. The lower support columns (cast) and the upper core plate are added as Expansion links to the lower core barrel girth weld (Primary), and the lower core support forging is added as an Expansion link to the upper core barrel flange weld (Primary). MRP-227-A specifies an EVT-1 examination for the upper core plate, lower support forging, and lower support columns (cast). It is noted that in MRP-227, Revision 1, the inspection technique for these components is changed to a visual (VT-3) examination. MRP-227-A Note: A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions). The examination coverage for the upper core plate, lower support forging, and lower support column bodies was redefined in MRP-227, Revision 1. Remaining CRGT lower flange welds is added as an Expansion component in MRP-227, Revision 1, but as stated in Note 2 above, Surry will inspect in accordance with MRP-227, Revision 1 for this component. MRP 2018-022 was issued on the baffle-former bolt expansion components which specifies that the lower support column bolts remain the first expansion component of the BFB unless a large cluster of BFB indications is discovered during the UT exams. The presence of clustering would trigger expansion of the barrel-former bolts adjacent to the large cluster of BFB indications due to the potential for clustering to result in indications of the barrel bolts. The terms "large cluster and "barrel-former bolts adjacent to the cluster" are defined in M RP 2018-022. The oore barrel outlet nozzle welds are eliminated as an E:>Epansion oomponent in MRP 227. Revision 1. The examination coverage for core barrel welds was redefined in MRP-227, Revision 1. Surry Power Station, Units 1 and 2 Page C-61 Change Notice 3 Serial No.: 19-248 Application for Subsequent License Renewal Appendix C - MRP-227-A Gap Analysis for PWR Vessel Internals Aging Management Page 74 of 74}}