ML24129A200

Update to Subsequent License Renewal Application, Supplement 2

ML24129A200
Person / Time
Site:	Summer
Issue date:	05/06/2024
From:	James Holloway Dominion Energy South Carolina
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
24-157
Download: ML24129A200 (1)

Text

Dominion Energy South Carolina, Inc.

5000 Dominion Boulevard. Glen Allen, VA 23060 Dominior.Energy.com May 6, 2024 United States Nuclear Regulatory Commission Attention: Document Control Desk 11555 Rockville Pike Rockville, MD 20852 DOMINION ENERGY SOUTH CAROLINA, INC.

VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 Dominion

~

Energy:

Serial No.:

NRA/SS:

Docket No.:

License No.:

10 CFR 50 10 CFR 51 10 CFR 54 24-157 RO 50-395 NPF-12 UPDATE TO SUBSEQUENT LICENSE RENEWAL APPLICATION (SLRA)

SUPPLEMENT 2 By letter dated August 17, 2023 [Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML23233A179], Dominion Energy South Carolina, Inc. (Dominion, Dominion Energy South Carolina, or DESC) submitted an application for the subsequent license renewal of Renewed Facility Operating License No. NPF-12 for Virgil C. Summer Nuclear Station (VCSNS) Unit 1.

Beginning on November 6, 2023, the NRC staff conducted an Aging Management Audit as part of their review of the VCSNS SLRA. As a result of discussions with the NRG staff throughout the audit, additional information necessary for the NRG staff to complete their technical review was identified. Some of the associated updates to the SLRA were provided in the Supplement 1 letter, dated April 1, 2024 [ADAMS Accession No. ML24095A207], and additional updates to the SLRA are provided in this Supplement 2 letter. provides a description of each of the topics that require the SLRA to be supplemented and identifies the affected SLRA section and/or table for Supplement 2. includes mark-ups of affected SLRA sections and/or tables for Supplement 2, as described in Enclosure 1.

To aid the staff in assessing changes, Enclosure 2 shows new text as underlined and deleted text as lined-through. Please note that change bars are shown for new text but are not shown for deleted text.

Serial No.: 24-157 Docket No.: 50-395 Page 2 of 6 If there are any questions regarding this submittal or if additional information is needed, please contact Mr. Keith Miller at (804) 273-2569.

Sincerely, James E. Holloway Vice President - Nuclear Engineering and Fleet Support COMMONWEAL TH OF VIRGINIA

)

)

COUNTY OF HENRICO

)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by James E. Holloway, who is Vice President - Nuclear Engineering and Fleet Support of Dominion Energy South Carolina, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that Company, and that the statements in the document are true to the best of his knowledge an J ucnc,.

Acknowledged before me this '9th day of Mo\\/

, 2024.

My Commission Expires: J(Xhuetr~ '3 l1 2.02..B Commitments made in this letter:

KATHRYN HILL BARRET NOTARY PUBLIC COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES JANUARY 31. 2028 Notary Public The Licensee Commitments identified in Table A4.0-1 of Appendix A, FSAR 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 - Supplement 2

cc:

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 Virgil C. Summer Nuclear Station Ms. Lauren Gibson NRC Branch Chief U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 11 E11 11555 Rockville Pike Rockville, Maryland 20852-2738 Ms. Kim Conway NRC Project Manager U.S. Nuclear Regulatory Commission Two White Flint North Mail Stop A 1 OM 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. Ted Smith NRC Branch Chief U. S. Nuclear Regulatory Commission Two White Flint North Mail Stop B 72M 11555 Rockville Pike Rockville, Maryland 20852-2738 Ms. Marieliz Johnson NRC Project Manager U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 7 E 1 11555 Rockville Pike Rockville, Maryland 20852-2738 Serial No.: 24-157 Docket No.: 50-395 Page 3 of 6

Mr. G. Edward Miller NRC Senior Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop O-9E3 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. Gregory Lindamood Santee Cooper - Nuclear Coordinator Virgil C. Summer Nuclear Station P. 0. Box 88 Jenkinsville, SC 29065 Stephen R. Pelcher Santee Cooper - Deputy General Counsel One Riverwood Drive Moncks Corner, SC 29461 Elizabeth Johnson Director, Historical Services, D-SHPO South Carolina Department of Archives and History 8301 Parklane Road Columbia, SC 29233 Johnathan Leader State Archaeologist South Carolina Institute of Archaeology and Anthropology 1321 Pendleton St, 1st Floor, Suite 16 Columbia, SC 29208 Fran Marshall Director, Office of Environmental and Public Health Serial No.: 24-157 Docket No.: 50-395 Page 4 of 6 South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Myra Reese Director, Office of Environmental Affairs South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201

Serial No.: 24-157 Docket No.: 50-395 Nate Haber Director Bureau of Water - Water Quality Division South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Lorianne Riggin Director of Environmental Programs South Carolina Department of Natural Resources 1000 Assembly Street Columbia, SC 29201-3117 Duane Parrish Director, South Carolina Parks, Recreation, and Tourism 1205 Pendleton St. Ste 248 Columbia, SC 29201 Florence Belser Chair, Public Service Commission of South Carolina 101 Executive Center Drive #100 Columbia, SC 29210 Christy Hall Secretary of Transportation South Carolina Department of Transportation 955 Park Street Columbia, SC 29201 Kim Stenson Director, South Carolina Emergency Management Division 2779 Fish Hatchery Rd West Columbia, SC 29172 Rhonda Thompson Chief, Bureau of Air Quality South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Renee Shealy Chief, Bureau of Environmental Health Services South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Page 5 of 6

Serial No.: 24-157 Docket No.: 50-395 Henry Porter Chief, Bureau of Land and Waste Management South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Jennifer Hughes Chief, Bureau of Water South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Chris Stout Chief, Office of Ocean & Coastal Resource Management South Carolina Department of Health and Environmental Control 1362 McMillan Avenue North Charleston, SC 29405 Page 6 of 6

Supplement 2 VCS SLRA TOPICS THAT REQUIRE A SLRA SUPPLEMENT Dominion Energy South Carolina, Inc.

(Dominion Energy South Carolina, or DESC)

Virgil C. Summer Nuclear Station Unit 1 Serial No.: 24-157

Supplement 2 VCSSLRA The following seven topics require the SLRA to be supplemented:

Serial No.: 24-157 Page 2 of6

1.

10 CFR 54.4(a)(3) - Regulated Events: Station Blackout Corrections

2.

Fire Service System Enhancement: Engine Driven Pump Jacket Water Heat Exchanger Core and Strainer Elements

3.

Primary Shield Wall Evaluation: Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation

4.

Aging Management of Electrical and Instrumentation and Controls

5.

Fatigue Management: High Energy Line Break Identification Correction

6.

Fire Protection: Provide Revised and Additional Commitments

7.

Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems: Completion of Commitment

Supplement 2 VCS SLRA The following seven topics require the SLRA to be supplemented:

Serial No.: 24-157 Page 3 of 6

1. 10 CFR 54.4(a)(3) - Regulated Events: Station Blackout Corrections Transmission conductors are not included in the 115 kV Parr offsite power recovery path to the Engineered Safety Features (ESF) transformer XTF-4. The second bullet in SLRA Section 2.1.3.4 will be revised to remove the term "transmission conductors."

Additionally, section 2.5.1.3 will be clarified to state that the 115kV offsite power recovery path contains the switchyard bus, and the 230kV offsite power recovery path contains the switchyard bus and overhead transmission conductors.

Based on the above, the SLRA is supplemented, as shown in Enclosure 2, to incorporate corrections of electrical component descriptions, as shown in the following:

SLRA Section 2.1.3.4 (Station Blackout) 2.5.1.3 (High Voltage Insulators)

2. Fire Service System Enhancement: Engine Driven Pump Jacket Water Heat Exchanger Core and Strainer Elements A diesel fire pump engine jacket water heat exchanger tube leak required the replacement of a heat exchanger core in January 2024. To address future aging of the diesel driven fire pump engine jacket water heat exchanger, Dominion will replace the heat exchanger core periodically. Review of the available electronic maintenance records from 1997 to present indicated that the tube bundle had been in service the entire time. A 20-year replacement frequency has been chosen to provide reasonable assurance that the heat exchanger will continue to perform its intended function. The 20-year replacement frequency is also consistent with the treatment of a similar heat exchanger core at North Anna Power Station. The periodic replacement is documented via a new subsequent license renewal commitment: 'The diesel driven fire pump engine jacket water heat exchanger core will be replaced at least once every 20 years." With this change, the diesel fire pump jacket water heat exchanger core is not subject to aging management review per 10 CFR 54.21 (a)(1 )(ii).

This heat exchanger core replacement will also be described in the fire service system boundary description.

Supplement 2 VCSSLRA Serial No.: 24-157 Page 4 of 6 Additionally, flow blockage aging effect will be added to the aging management review

{AMR) rows for the fire service system strainer elements in Table 3.3.2-13.

Based on the above, the SLRA is supplemented as shown in Enclosure 2, to identify the heat exchanger core replacement frequency and to add flow blockage aging effect, as shown in the following:

SLRA Section SLRA Table 2.3.3.13 3.3.2-13 A4.0-1 #50

3. Primary Shield Wall Evaluation: Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation SLRA Section 3.5.2.2.2.6 states the impact of gamma heating on the primary shield wall (PSW) has been evaluated and it was concluded that the maximum PSW concrete temperature would be less than 128°F. Subsequent review has determined that the maximum PSW concrete temperature needs to be updated to be less than 145°F instead of 128°F.

Based on the above, the SLRA is supplemented as shown in Enclosure 2, to correct the maximum PSW concrete temperature, as shown in the following:

SLRA Section 3.5.2.2.2.6

4. Aging Management of Electrical and Instrumentation and Controls A table documenting the aging management evaluation that references NUREG-2191, Table 3.6.1, item 3.6.1-001 for electrical equipment subject to 10 CFR 50.49 Environmental Qualification requirements was not included in SLRA Section 3.6.

Based on the above, the SLRA is supplemented as shown in Enclosure 2 to add new table 3.6.2-4, as shown in the following:

SLRA Section SLRA Table 3.6.2 3.6.2-4

Supplement 2 VCS SLRA Serial No.: 24-157 Page 5 of 6

5. Fatigue Management: High Energy Line Break Identification Correction SLRA Section 4.3.5 describes the identification and evaluation of ASME Section Ill, Class 1 high energy line break locations as a time limited aging analyses (TLAA). FSAR Section 3.6.2.1.2 indicates thatASME Section Ill, Class 2 and 3 and specific non-nuclear safety class high energy line break locations also have a time limiting aspect.

Based on the above, the SLRA is supplemented, as shown in Enclosure 2, to identify and assess the high energy line break locations for ASME Section Ill, Class 2 and 3 piping and non-nuclear safety class piping, as shown in the following:

SLRA Section 4.3.5 A3.3.5

6. Fire Protection: Provide Revised and Additional Commitments The following revisions and additional enhancements are being included in the Fire Protection program (GALL-SLR Section XI.M26):

An enhancement is being added to describe aging management for elastomeric components.

An enhancement is being revised to identify aging effects requiring aging management for susceptible non-elastomeric fire protection components.

An enhancement is being added to identify aging effects, aging management, and trending of inspection results for CO2 fire protection system components.

Two existing enhancements are being revised to be consistent with the above changes.

Based on the above, the SLRA is supplemented, as shown in Enclosure 2, to revise aging management program commitments, as shown in the following:

SLRA Section SLRA Table A1.15 A4.0-1 #15 B2.1.15

Supplement 2 VCSSLRA Serial No.: 24-157 Page 6 of 6

7. Inspection of Overhead Heavy Load and Light Load (Related to Refueling)

Handling Systems: Completion of Commitment The enhancement of the Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems aging management program (GALL-SLR Section XI.M23) has been implemented. The crane inspection procedures have been revised to (1) specify visual inspections for the effects of general corrosion, deformation, cracking, and wear on the rails, bridges, structural members, and structural components of the system and to (2) provide guidance for monitoring the structural connections for looseness, missing or loose bolts and/or nuts, loss of material due to general corrosion, cracking, loss of preload and other conditions indicative of loss of preload.

Based on the above, revisions are being made in the following areas of the SLRA, as shown in Enclosure 2, to indicate completion of the commitment:

SLRA Section SLRA Table 82.1.13 A4.0-1 #13 82-1

Supplement 2 VCSSLRA SLRA MARK-UPS SUPPLEMENT 2 Dominion Energy South Carolina, Inc.

(Dominion Energy South Carolina, or DESC)

Virgil C. Summer Nuclear Station Unit 1 Serial No.: 24-101 Docket No: 50-395

Serial No.24-157 Supplement 2 Page 2 of22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Scoping and Screening Methodology I

Plant electrical system cables, and cable bus are routed from the Engineered Safety Features transformer XTF-4 and Emergency Auxiliary Transformer XTF-31 transformers to the ES 7.2kV Buses XSW-1 DA, and XSW-1 DB.

Offsite power is supplied to the station by the 115 kV Parr ESF Line and 230 kV Bus #3, to the Engineered Safety Features transformer XTF-4 and Emergency Auxiliary Transformer XTF-31 transformers. Restoration of the 7.2kV ES buses XSW-1 DA, XSW-1 DB, XSW-1 EA, and XSW-1 EB through the Engineered Safety Features transformer XTF-4 or Emergency Auxiliary Transformer XTF-31 would by definition terminate an SBO event. The following electrical components are in the scope of license renewal for recovery from Station Blackout:

• 230KV Bus #3 disconnect switches, circuit breakers, associated control components

{including cables), transmission conductors, switchyard bus, high voltage insulators to connect the 230KV Bus #3 Breaker XCB-8892 to Emergency Auxiliary Transformer XTF-31.

• 115KV Parr ESF Line disconnect switches, circuit breakers, associated control components (including cables), voltage regulators, transmission eonductors, switchyard bus, high voltage insulators to connect the 115 kV Parr ESF Line, Circuit Switcher XES4 to the Engineered Safety Features transformer XTF-4.

• The Engineered Safety Features transformer, XTF-4, connection to the 7.2kV bus XSW-1 DX includes insulated cables, cable bus, and circuit breakers.

• The XTF-4 connection via the 7.2kV bus XSW-1 DX to the 7.2kV ES buses XSW-1 DA and XSW-1 DB includes insulated cables, cable bus, and circuit breakers.

• The Emergency Auxiliary Transformer, XTF-31, connection to the 7.2kV ES buses XSW-1DA and XSW-1DB includes insulated cables, cable bus, and circuit breakers.

• ES bus XSW-1 EA is connected to ES bus XSW-1 DA with insulated cables, cable bus, and circuit breakers.

• ES bus XSW-1EB is connected to ES bus XSW-1DB with insulated cables, cable bus, and circuit breakers.

• DG/A {XEG-0001A-DG) is connected to ES bus XSW-1 DA with insulated cables and circuit breakers

• DG/8 (XEG-0001B-DG) is connected to ES bus XSW-1DB with insulated cables and circuit breakers The Alternate AC (MC) source of power from the Parr Hydro Power Station via ESF Transformer XTF5052 to bus XSW-1DX described in FSAR Section 8.1 is not credited for SBO recovery.

Page2-11

Serial No.24-157 Supplement 2 System Evaluation Boundary Page 3 of 22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Mechanical Systems The evaluation boundary for the fire service system components subject to aging management review includes safety-related containment penetration piping and nonsafety-related components that provide support to directly-connected safety-related components, or that retain water in buildings containing safety-related components. The electric and diesel-driven fire pumps and jockey pump are subject to aging management review. The fire pump diesel engine is an active skid-mounted unit. Components attached to the engine and within the skid boundaries are integral parts of the engine. They are tested as a unit and are parts of the active engine component, and are not subject to aging management review. Additionally, the diesel fire pump engine jacket water heat exchanger core will be replaced at least once every 20 years. The diesel fire pump engine exhaust and fuel oil supply components outside of the skid are subject to aging management review. The water supply from the fire pumps, via the yard loop to applicable hydrants, hose reels, and sprinkler systems are subject to aging management review. Water suppression piping within the Auxiliary Service Building is not credited for NFPA-805, but is subject to aging management review to maintain yard loop integrity. The low pressure carbon dioxide tank and supply piping components are subject to aging management review.

Sample piping for the incipient fire (smoke) detectors that monitor the Relay and Upper Cable Spreading Rooms is also subject to aging management review, but does not appear on system drawings. The incipient fire detector panels (enclosures containing tubing, filters, valves, and vacuum pumps) are active components not subject to aging management review.

Portions of the water suppression system are not credited with 10 CFR 50.48 compliance and are not within-scope, as shown on the highlighted SLR boundary drawings. The Halon systems are not credited with 10 CFR 50.48 compliance and are not within-scope.

Fire doors are addressed in the structural sections for their associated buildings. Fire dampers are addressed in the air handling system. Smoke detectors are active instruments not subject to aging management review.

System Intended Functions The fire service system performs the following safety-related functions: The system provides containment isolation. Therefore, the fire service system is within the scope of license renewal in accordance with the criteria of 10 CFR 54.4(a)(1).

The fire service system contains nonsafety-related components whose failure could prevent satisfactory accomplishment of a safety-related function. Therefore, the fire service system is within the scope of license renewal in accordance with the criterion of 1 0 CFR 54.4(a)(2) for spatial interaction and structural integrity.

Page2-97

Serial No.24-157 Supplement 2 Page 4 of 22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Electrical and Instrumentation and Control Systems 2.5.1.3 High Voltage Insulators

System Description

The electrical commodity group identified as "High-Voltage Insulators" includes those station post and suspension insulators that support overhead conductors (transmission conductors and switchyard bus) that are part of the SBO offsite power recovery path. High-voltage insulators are passive in nature. The insulating portion of high-voltage insulators is made of porcelain. The high-voltage insulator commodity group was evaluated against 10 CFR 54.21 (a)(1 )(ii) and determined to not be included in the EQ program. Therefore, high-voltage insulators are subject to aging management review.

Porcelain is a ceramics that experiences the external aging effects of reduced insulation resistance from excessive surface contamination. Porcelain relies on surface rinsing from precipitation or mechanical washing to clean contaminants from the shed surfaces. Porcelain has been in service in the utility industry for over 60 years worldwide and is considered to be a mature technology and generally standardized.

The porcelain insulators are cap and pin (suspension) insulators and experience the same aging effect of loss of material from corrosion and mechanical wear. Inspection for mechanical wear (visual) is conducted. The insulating material for porcelain has the aging effect of reduced insulation resistance from excessive surface contamination. Inspection for excessive surface contamination (visual) is conducted.

System Evaluation Boundary The high-voltage insulators commodity group includes SBO offsite power recovery path insulators that support the 115 kV switchyard bus and 230 kV transmission conductors and switchyard bus in the overhead lines that are available as the two GDC-17 offsite power paths to the ESF transformer XFT4, and the emergency auxiliary transformer XFT31. The 115 kV overhead transmission and switchyard conductors connect the PARR ESF circuit switcher (XES 4) that feeds the ESF transformer (XTF 4) high voltage side. The 7.2 kV cable bus connects the ESF transformer low voltage side to the Turbine Building 7.2 kV switchgear XSW1 DX via transfer switch XES 8 and voltage regulator XTF 6, which does not have high voltage insulators. The 230 kV overhead transmission and switchyard conductors connect the 230kV switchyard beaker (XCB 8892) that feeds the emergency auxiliary transformer (XTF-31) high voltage side. The 7.2 kV cable bus connects the emergency auxiliary transformer low voltage side to the Intermediate Building 7.2 kV switchgear XSW1 DB, which does not have high voltage insulators. The boundary for the SBO offsite power recovery path is the first active switchyard load break device (e.g.: circuit breaker or load interrupting disconnect or circuit switcher) and their associated disconnect switches downstream of 230 kV switchyard bus 3, and 115 kV Parr ESF line.

Page2-281

Serial No.24-157 Page 5 of 22 Table 3.3.2-13 Auxiliary Systems - Fire Service -Aging Management Evaluation Component Intended Material Environment Aging Effect Requiring Aging Management Programs Type Function(s)

Management Strainer element FLT Copper alloy (E) Raw water Cracking Inspection of Internal Surfaces in Miscellaneous

(>15% Zn)

Piping and Ducting Components (82.1.25)

Loss of material.:J!.Q'!t_

Fire Water System (82.1.16)

!1IQQk/.'!9!:

!.Qiiii Qf ma!i::tiii!I Selective Leaching (82.1.21)

Nickel alloy (E) Raw water Loss of material.:J!.Q'!t_

Fire Water System (82.1.16)

QIQQks!Q!il Stainless (E) Raw water Loss of material.:J!.Q'!t_

Fire Water System (82. 1.16) steel lll2~ag~

Strainer element FLT Copper alloy (E) Raw water Loss of material; flow Fire Water System (82.1.16)

(fire pump blockage suction)

Tank (diesel fire PB Steel (E) Air - outdoor Loss of material External Surfaces Monitoring of Mechanical pump fuel oil)

Components (82.1.23)

(I) Fuel oil Loss of material Fuel Oil Chemistry (82.1.18)

Tank (low PB Steel (E) Air - outdoor Loss of material External Surfaces Monitoring of Mechanical pressure carbon Components (82.1.23) dioxide)

(I) Gas None None Tank (retarding PB Gray cast (E) Air - indoor Loss of material External Surfaces Monitoring of Meehanical chamber) iron uncontrolled Components (82.1.23)

(I) Raw water Long-term loss of material One-nme Inspection (82.1.20)

Loss of material Selective Leaching (82.1.21)

Loss of material; flow Fire Water System (82.1. 16) blockage

- - - -~

Valve body LB;PB Copper alloy (E) Air - indoor None None uncontrolled (I) Raw water Loss of material; flow Fire Water System (82.1.16) blockage Virgil C. Summer Nuclear Station Page 3-378 Application for Subsequent License Renewal NUREG-2191 Table 1 Notes Item Item VII.C1.A-473b 3.3.1-160 E, 3 VII.G.AP-197 3.3.1-064 A

VII.G.A-47 3.3.1-072 A

VII.G.A-403 3.3.1-130 C

VII.G.A-55 3.3.1-066 A

VII.G.AP-197 3.3.1-064 A

VII.I.A-??

3.3.1-078 A

VII.G.AP-234a 3.3.1-070 A

VII.I.A-??

3.3.1-078 A

VII.J,AP-6 3.3.1-121 A

VII.I.A-77 3.3.1-078 A

VII.G.A-532 3.3.1-193 A

VII.G.A-51 3.3.1-072 A

VII.G.A-33 3.3.1-064 A

VII.J.AP-144 3.3.1-114 A

VII.G.AP-197 3.3.1-064 A, 1 Supplement 2

Serial No.24-157 Supplement 2 Page 6 of 22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Aging Management Review potential radiation damage, the PSW demand is almost wholly dictated by the RV support load in this region.

The PSW contains embedded RV support anchorages. The maximum fluence at 72 EFPY on the RV support anchorage embedded steel (within the PSW), adjusted to add 20% for analytical uncertainty, is 8.82 x 1017 n/cm2 (E > 1.0 MeV). This is less than the damage threshold for the steel of 1.00 x 1019 n/cm2 (E > 1.0 MeV), per EPRI Report 3002013084.

The conservatisms in the evaluation were as follows:

• Exposures were based on 72 EFPY

• Future projections included a 10% positive bias on the peripheral and re-entrant corner assemblies on the projection fuel cycle.

• The loss of strength in the PSW concrete as a result of gamma dose incident on the PSW was assumed to apply to the full thickness to the point where the gamma dose falls below the NUREG-2192 damage threshold, when in reality the gamma dose effect would reduce in an approximately linear fashion from the outside surface.

• The latest research data presented in EPRI Report 3002011710 indicated that the threshold for damage to concrete from gamma dose may be higher than 1 x 108 Gy.

• The IR does not include any reduction in PSW demand to account for LBB implementation.

• Reduction in LOCA loads from LBB evaluation not taken into account.

As evidenced by the evaluation described above, and considering the integrated effects of neutron fluence, gamma dose, and gamma heating, the PSW is capable of carrying the loads of the RV at the end of 80 years of plant operation. Therefore, the PSW will continue to satisfy its design criteria considering the long-term radiation effects and a plant specific AMP or enhancements to an existing AMP is not required.

Gamma Heating The impact of gamma heating on the PSW has been evaluated and it was concluded that the maximum PSW concrete temperature would be less than ~

145°F. This temperature is bounded by the long-term PSW concrete temperature limit of 150°F reported in Section 3.8.1.5.1.2 of the FSAR. Therefore, gamma heating is not an issue for the PSW concrete.

Secondary Shield Wan Evaluation The neutron fluence and gamma dose threshold limits of NUREG-2192 are not exceeded beyond the first 10 inches of the PSW. The SSW is physically external to the PSW. The entire SSW is further from the core at all points compared to the PSW, meaning that the neutron fluence and gamma dose is higher in the PSW than the SSW. Therefore, since the NUREG-2192 threshold limits are not exceeded external to the PSW and the SSW is physically external to the PSW, the Page 3-614

Serial No.24-157 Supplement 2 Page 7 of22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Aging Management Review I

3.6 AGING MANAGEMENT OF ELECTRICAL AND INSTRUMENTATION AND CONTROLS 3.

6.1 INTRODUCTION

This section provides the results of the aging management review for components and commodities identified in Section 2.5.1, Electrical Component Groups as being subject to aging management review. Components and commodities addressed in this section are described in the indicated sections.

• Cable Bus (Section 2.5.1.1)

• Cables And Connections (Section 2.5.1.2)

• High Voltage Insulators (Section 2.5.1.3) 3.6.2 RESULTS The following tables summarize the results of the aging management review for Electrical and Instrumentation and Controls.

• Table 3.6.2-1, Electrical and Instrumentation and Controls - Cable Bus -Aging Management Evaluation

• Table 3.6.2-2, Electrical and Instrumentation and Controls - Cables And Connections -Aging Management Evaluation

• Table 3.6.2-3, Electrical and Instrumentation and Controls - High Voltage Insulators - Aging Management Evaluation

• Table 3.6.2-4. Electrical and Instrumentation and Controls - Electrical Equipment Subject to 1 O CFR 50.49 Environmental Qualification - Aging Management Evaluation Page3-711

Serial No.24-157 Page 8 of 22 Table 3.6.2-4 Electrical and Instrumentation and Controls - Electrical Equipment Subject to 10 CFR 50.49 Environmental Qualification - Aging Management Evaluation Subcom122nent Intended Ma~rii!I Envjronment Aging Effect Requiring Aging Management Programs NUREG-2191 Ti!ble 1 I~

EunctJ2n(!i:)

Mi!ni!gemen!

J1m Jwn

!;;l!igris;al QE

~ (El Ar!ii!S Qf lh!l VariQl.!:i aging !lff!i!.l!i (fQr

!;nvirQnmenl!i!I Ql.!!i!lifiQ!i!!iQD Qf Elegris; i;;!JUill!De l Vl B L-05 J,§, 1-001 8

l;gui12m1::n1

~ Ql!i!n! !h!i!! !.QUl!l bl::

EQl

~

Subieg !Q m!i!terials

iUbieg !Q h2rah 10 QFR fiO,~l,il eovirQnm1;1nl!i!I Eovi(.Qntnen!!i!I effeg!i Q.t !i! IQ!i!i Qf Qu2IifiQii!liQn s;QQlii!Ol as;s;igen!

(LQQ.8), high

!lO!l(gll lio!l br!liil~

Q[ i:lQS! LQQ8 1;1nvirQnm1;1nj.

AQver!ie IQQi;!li7.!:lQ

!;lnvirQment (e g,

lillllll!lril!Ur!l rsiQiii!liQO, Qr mQiSl!.!r!l)

IN

~ (El Areas Qf tb!:l VariQUS aging effes.ts (fQr EovirQnmentii!I Qu!i!lifis;a!iQ Qf Elegri!. Eaui12rnent Vl,B,L-05 M,l-QQ1 A

QQIJlm!:lris; 12lii!nt tbat !.Ql.!ls:l bi:1 fQl

~

m!i!terials siibie!.l IQ IJ!i!t!ib envirQ mentii!I effes.l!i Qf !! IQS!i Qf QQQli;l t !i!QQiQ!lnt (I.QQ8l high 1::!lr.<Jll lioe brea~

QrllQ!il L,QQ8 eovirQtrl!l I 8QV!:l[!i!:l IQs;ali1::!:lQ envirQnment (e,g l!:lmll!;lrl!!l.!re r!i!!1ia!iQn, Qr mQiS!ur!:l)

Table 3.6.2-4 Plant-Specific Notes: None Virgil C. Summer Nuclear Station Page 3-741 Supplement 2 Application for Subsequent License Renewal

Serial No.24-157 Supplement 2 Page 9 of22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Time-Limited Aging Analyses 4.3.5 HIGH-ENERGY LINE BREAKANALYSIS TLAA

Description:

As stated in Table 4.1-2 of NUREG-2192 (Reference 1.7-5), a high-energy line-break (HELB}

analysis is considered one of the potential TLAAs for oonsideration when assessing the subsequent period of extended operation. As indicated in FSAR Section 3.6.2 (Reference 1.7-18), high-energy systems that require analysis for the consequences of pipe break were identified based on the fluid in the pipe, the pressure, and the temperature during normal station operation. The lines that were both high-temperature and high-pressure were postulated to experience a longitudinal or circumferential break, and were analyzed for pipe whip, jet impingement, and environmental effects. A HELB is not required to be postulated at a given piping location if the design CUF calculated in accordance with ASME Code, Section Ill for that location is less than or equal to 0.1.

Therefore, these HELB evaluations excluded Class 1 locations within each high-energy piping system that have a CUF value of 0.1 or less, consistent with the HELB screening criterion ts specified in FSAR Section 3.6.2.1.2, "High Energy System Piping Inside Containment."

Now that plant operation will be extended through 80 years, it is possible that one or more of these locations could see an increase in fatigue usage, potentially above 0.1, that would require the location to be evaluated as a potential break location. Since the ASME Code, Section Ill, Class 1 piping fatigue analyses that provided the CUF values less than 0.1 are based upon 40-year design transients, these HELB analyses of Class 1 piping systems have been identified as TLAAs that require evaluation for subsequent license renewal. The l=lELB e*,aluations for the ASME Gode, Section Ill, Class 2 an1 a piping do not involve a time limited assumption and are not identified as TL/V\\s.

The HELB analyses also postulate breaks to occur in ASME Code, Class 2 and 3 piping and branch runs at the following locations:

1.

At terminal ends.

2.

At intermediate locations selected by either one of the following criteria:

(a) At each pipe fitting.

(b) At each location where the pipe stresses exceed 0.8{1.2Sh + S:i), where Sa is the allowable stress.

The allowable stress value, Sa, is equal to f (1.25 (SJ.+ 0.25 Su), where f is the allowable stress reduction factor that is based upon the total number of full-temperature-range transient cycles for a component through its service life. Sc is the basic material allowable stress (from ASME Section II) at the minimum (cold} temperature, and Sh is the basic material allowable stress {from ASME Section II) at the maximum (hot} temperature. The value off is 1.0 for 7,000 cycles or less. If more than 7,000 cycles are predicted, the factor is reduced. depending upon the numbers of cycles.

Since the number of cycles is a time-limited assumption, the value of S~ is based upon a time-limited assumption, and the HELB analyses for Class 2 and 3 piping systems evaluated using Page4-84

Serial No.24-157 Supplement 2 Page 10 of 22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Time-Limited Aging Analyses the method described in 2b, above, have been identified as TLAAs that reguire evaluation for the SPEO.

In addition, FSAR Section 3.6.2.1.2 explains that specific non-nuclear safety class piping has been classified as Quality Related piping for the purpose of minimizing postulated pipe break locations.

This Quality Related piping is designed in accordance with Code requirements. A rigorous analysis of this Quality Related piping was performed and breaks were then postulated based on the criteria noted above for Code Class piping. Since these analyses are also based upon the number of cycles, which is a time-limited assumption. these HELB analyses of non-nuclear safety class piping have also been identified as TLAAs that reguire evaluation for the SPEO.

TLAA Evaluation:

The CUF analyses for the high-energy piping systems were developed in accordance with the requirements of the ASME Code, Section Ill rules (i.e., NB-3222.4(e)), based upon the 40-year design transients listed in Table 4.3.1-1. Existing TLAAs for the Class 1 piping locations were reviewed to compare the applicable transients and cycles with the 80-year transient cycle projections. The 80 year transient cyole pr9:ieetions for eaeh of the transients applieable to the Glass 1 piping components under in\\*estigation is oonsistent 'Nith the transients identif:ied for subsequent lioense renmval.

The 40-year design transient cycles analyzed in the ASME Code, Section Ill fatigue analyses of record for the ASME Class 1 piping bound the projected 80-year transient cycles, as shown in Table 4.3.1-1. The CUF values will remain less than 1.0 for all locations and no CUF values increase. Therefore, the original Class 1 HELB locations identified through the HELB screening process remain unchanged and no new HELB locations must be postulated for the subsequent period of extended operation.

The TLAAs for the ASME Section 111, Class 2 and 3 piping and non-nuclear safety class piping discussed in FSAR Section 3.6.2.1.2 were evaluated by comparing the total number of applicable transient cycles in the 80-year cycle projections to the 7,000-cycle limit associated with the stress range reductjon factor of 1.0 used in these TLAAs, as shown on Table 4.3.3-1. Since the number of proiected cycles remains below 7,000 cycles for each of the non-Class 1 piping locations. the stress range reduction factor remains equal to 1.0, and the non-Class 1 HELB analyses remain valid.

Therefore. the original non-Class 1 locations identified through the HELB screening process remain unchanged and no new HELB locations must be postulated for the subsequent period of extended operation.

TLAA Disposition: 10 CFR 54.21(c)(1)(i)

The current HELB analyses remain valid for the subsequent period of extended operation and are dispositioned in accordance with 10 CFR 54.21(c)(1)(i).

Page4-85

Serial No.24-157 Page 11 of 22 Supplement 2 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix A - FSAR Supplement A1.15 Fire Protection The Fire Protection program is an existing condition monitoring and performance monitoring program that requires periodic visual inspections of fire barrier components, and functional testing of fire-rated doors and the carbon dioxide (CO2~} fire suppression system. The program manages:

Loss of material for fire-rated doors <including roll-up fire window shutter), fire damper housingsassemblies, and steel seismic gap covers and the CO2 fire suppression system Loss of material or cracking for concrete structures, including fire barrier walls, ceilings, and floors Loss of material or cracking for the CO2 fire suppression system Loss of material, cracking/delamination, change in material properties, and separation for non-elastomer fire barrier penetration seals, fire stops, radiant energy shields, and fire wraps and coatings Shrinkage... aREJ-loss of strength. and hardening. or any other sign of degradation. for elastomeric fire barrier penetration seals and for seismic gap elastomersfillers The Fire Protection program is a risk-informed, performance-based program built upon National Fire Protection Association (NFPA) 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. 2001 Edition." Adoption of NFPA 805 in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.48(c) serves as the method of satisfying 10 CFR 50.48(a) and General Design Criterion 3.

A 1.16 Fire Water System The Fire Water System program is an existing condition monitoring program that manages aging effects associated with water-based fire protection system components. This program manages loss of material and flow blockage due to fouling by conducting periodic visual inspections, tests, and flushes performed in accordance with the NFPA 25 (2011 Edition). Testing or replacement of sprinklers that have been in place for 50 years is performed in accordance with NFPA 25.

In addition to NFPA codes and standards, portions of the water-based fire protection system that are: (a) normally dry but periodically subjected to flow and (b} cannot be drained or allow water to collect are subjected to augmented testing beyond that specified in NFPA 25, including: (a) periodic system full flow tests at the design pressure and flow rate or internal visual inspections and (b) piping volumetric wall-thickness examinations.

The water-based fire protection system is normally maintained at required operating pressure and is monitored such that loss of system pressure is immediately 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. Inspections and tests follow site procedures that include PageA-11

Serial No.24-157 Supplement 2 A3.3.5 Page 12 of22 High-Energy Line-Break Analyses Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix A - FSAR Supplement A high-energy line break is not required to be postulated at a given high-energy piping location if the design CUF for that location, calculated in accordance with ASME Code, Section Ill, is less than or equal to 0.1. The Class 1 piping fatigue analyses that provided the CUF values less than 0.1 are based on 40-year design transient cycles. For Class 2, 3 and specific non-nuclear safety class piping. HELB analyses also postulate breaks at the following locations:

1.

At terminal ends.

2.

At intermediate locations selected by either one of the following criteria:

(a) At each pipe fitting.

(b) At each location where the pipe stresses exceed 0.8(1.2Sh + S8}, where Sa is the allowable stress.

The allowable stress value. S8

• is equal to f (1.25 S~ + 0.25 Sh), where f is the allowable stress reduction factor that is based upon the total number of full-temperature-range transient cycles for a component through its service life. The value off is 1.0 for 7,000 cycles or less. The 80-year projected cycles are bounded by the 40-year design cycles and are bound by the 7,000 cycles.

Since the CUF values for this piping remain unchanged and the f factor is 1.0, no new HELB locations are required for the subsequent period of extended operation. The 40 year design cyeles bound the 80 year projeeted eyeles, so no GUF values inerease. Sinee the CUF values for this piping remain unehanged, no new HELB locations are required for the subsequent period of extended operation. Therefore, the original HELB analyses remain valid for the subsequent period of extended operation and are dispositioned in accordance with 10 CFR 54.21 (c)(1)(i).

A3.4 Environmental Qualification of Electric Equipment Thermal, radiation, and cyclical aging analyses of electrical and l&C components, developed to meet 10 CFR 50.49 requirements, have been identified as time-limited aging analyses (TLAAs).

The NRC nuclear station environmental qualification (EQ) requirements in 10 CFR 50.49 require that an EQ program be established to demonstrate that certain electrical equipment located in harsh environments is qualified to perform applicable safety functions in those harsh environments after the effects of in-service aging. Harsh environments are defined as those areas of the plant that could be subject to the harsh environmental effects of a loss-of-coolant accident (LOCA), high energy line break (H~LB) or post-LOCA radiation. 1 O CFR 50.49 requires that the effects of significant aging mechanisms be addressed as part of environmental qualification.

The Environmental Qualification of Electric Equipment program (A2.3) will manage the effects of aging for EQ equipment through the subsequent period of extended operation in accordance with 10 CFR 50.49(c)(1)(iii). The program meets the requirements of 10 CFR 50.49 for the applicable electrical equipment important to safety. Reanalysis of an aging evaluation to extend the qualifications of equipment is performed on a routine basis as part of the EQ program. Important PageA-43

Serial No.24-157 Page 13 of 22 Table A4.0-1 Subsequent License Renewal Commitments Program Commitment The ln§12eQtiQn Qf Qverhesi!Q HealQl LOsi!Q ang Light LQad (Relate!:l 1Q Refyeling) Hang ling ~J!§tem§ 12rogram i§ an exi§ting !:,QnditiQn mQnitQring 12rQgram 1h21 i§ ~regited. +l=!e I-R5f:)ee#eA ef. g,,,eFReaEJ l-lea*,1y* l:.eaf. aAef.

Inspection of l:.igM l:.eaef. (Re/Jtef. te Refl:JeliRf/1 l=laAef.liRg S;'6teff/5 !;)FO§FaFA is aR elEistiR§ 69RElitioR Ffl9RitoFiR§ !;lF9§FaFA Overhead Heavy tl=!at will be eRl=laRsed as follows:

Load and Light 13 Load (Related to 1.. PFoseEluFe(s) will be Fe,*iscd te Fequirn:

• l,lis1:1al iRS!;)estieRS of Fails, bFid§CS, stFustural FAeFAbCFS, a Rel stF1;1st1:1ral 60FA!;l9ReAts foF less ef FAaterial Refueling) d1:1e to §OAOFal GOFFOsieR; ElefoFFRatioR; srasl~iR§, aREl wear.

Handling

• l,,£isual iRS!;lOStioRS ef Fails, salted 60RROstioAS fer loss of FAaterial Elt1e to §OROFal 60FFOSieR; srasl~iR§;

Systems program aRd loose or FAissiA§ salts oF At1ts, aRd otl=leF seRElitioRs iAdisatiYe of loss of !;)Feload.(Completed -

~y12121ement 2l The Compressed Air Monitoring program is an existing preventive and condition monitoring program that will be enhanced as follows:

Compressed Air

1. Procedure(s) will be revised to require Turbine Building instrument air dryer outlet dew point readings 14 Monitoring greater than zero be documented in the Corrective Action Program and evaluations performed for program results that do not satisfy established criteria as identified in the applicable procedures.

2. Procedure(s) will be revised to specify inspections and tests be performed by personnel qualified in accordance with site procedures and programs to perform the specified task.

Virgil C. Summer Nuclear Station PageA-54 Application for Subsequent License Renewal Appendix A-UFSAR Supplement AMP Implementation n

,...... L,..,..

.:11

"~ - -

B2.1.13 iFA!;llCFAeRteEl e FAORtl=ls !;)Fior to tl=!e susseeiueRt !;)OFieEl ef OlEtCAded o!;)eFatieA.Ongoing Program enhancements for SLR will be implemented 6 months prior to the B2.1.14 subsequent period of extended operation.

Supplement 2

Serial No.24-157 Page 14 of 22 Table A4.0-1 Subsequent License Renewal Commitments Program Commitment The Fire Protection program is an existing condition monitoring and performance monitoring program that will be enhanced as follows:

1. Procedure(s) will be revised to provide guidance for detection of loss of material, cracking, holes, and gaps during the visual inspections of fire dampers to ensure that any deficiencies are noted on a condition report. and to determine the acceptability of the findings. (Revised - Supplement 2) 2... PrQQeciyre(s) will tie reviseci tQ regyire th12t in§Qe!:;tions Qf flee tiarrier eli;lstQmeriQ QenetratiQn seals and seismic gap filler identify shrinkage, IQ§S of strength, and hardening, or any other signs of ciegradation.

(Added - SUQQlement 2)

~ PrQ!:,edure(§) will tie revi§eg to sQeQiO£ that insQe!:;tiQn result§ of materials susceQti!:2Ie tQ gelaminatiQn,

!:;hi;lnge in material QrQQ!;!!:tie§ seg5JratiQn, inQrea§eg h!i!rgne§s §hrio~i:lge Qr IQ§§ Qf §trength will be trend!;!d. Where pragtical, identified d!:!gradation will be grojected until the neig; scheduled inSQ!;!!:,tion.

Re:;;ults will be evaluated against acceQtance criteria tQ confirm that the timing of subseguent insg!;!ctions will maintain the comgQnents' intended functiQns throughout the subsegu!;!nt geriQd of

!';lXt!;!nd!:!d QQecation b12lil!ilQ on th!:! QrQi!il!::ted r12t!il Qf d!';lgraci!i!tion. (Renumb!:!reci and Revised -

SupQlement 2}

15 Fire Protection PFeseel1:1Fe(s) will ee Fe11iseel ta 13Fe,,.iele §l:lielaRee te eRs1:1Fe tl'la~ iRs13eetieR Fes1:1lts feF fiFe 13FeteetieR program eemJ:leAeAts aFe tFeReleel te f:IF8 1.,iae feF timely EletestieA ef a§iR§ effeets se tl'lat tl'le a131:1Fe13Fiate eeFFeeti1~e aetieAs saR ee tal,eR. 1JVl:ieFe 13Fastieal, ieleAtifieel ele§!FaelatieR will ee (:IFejeeteel 1:1Atil tl'le Reie!

ssl'leEl1:1leel iAs(:lesl;ieA. Res1:1lts aFe ei.*al1:1ateEl agaiRst aeee1=1taAee sFiteFia te seAfiFFA U:iat tl'le timiR§ ef s1:1eseei1:1eRt iRSJ:ISStieRS will A'laiRtaiR tl=ie 69A'l(:19ReRtS' iAteReleEl fl:IAStieAs tRF91:1JR91:1t tl=ie s1:1eseeic:1eAt

(:leFieel ef eie!eAeleel e(:leFatieA easeel eR tl=ie (:IFejesl;eel r:ate ef elegr:adatieA.

~ PrQ!:,edur!il(l:l} will be revis!:!d tQ §pegify that insgeQtiQn§ will be i;ierfQrmed tQ id!:!nti:t! gracking and lorl!s of material for COz. fire grot!;!Ction system gomQQnents, and that those results are trended sind aggroi;irisite cgrr!ilctive agtions icientified, if ne!:,e§sa!:)!. Where Qractical identifi§d Q!:!gradation will be grojected until the next sghedul!:!d insgegtiQn. Results will be evaluated against ac!:,egtangg criteria tQ

~Qnfirm tbat th~ timing of ~ubs!:!gu~nt inSQ!:!CtiQn§ will maintain th~ !:,omgon!:!nts' i t!ind!:!Q fu s.tions thrQughout the subsegu~ ! Q!:!riQd Qf eig;ended QQeri:ltion t2i:l§eci on the grQjected rat~ of gegrad12!iQn (Added - Sugplement 2)

&-§..Procedure(s) will be revised to si;iegi:t! that fQr insi;iection r§rl!ults 19Fe*o1iae §t1ielaAse feF e1o1alc:1atiR§

(:IFejesteel iRsf:lestieR Fes1:1lts. FeF Fes1:1lts that will fail to meet acceptance criteria prior to the next scheduled inspection, inspection frequencies will be adjusted as determined by the Corrective Action Program. (Renumbered sind Revis~ci - Sui;1glement 2)

Virgil C. Summer Nuclear Station PageA-55 Application for Subsequent License Renewal Appendix A-UFSAR Supplement AMP Implementation Program enhancements for SLR will be B2.1.15 implemented 6 months prior to the subsequent period of extended operation.

Supplement 2

Serial No.24-157 Page 15 of 22 Table A4.0-1 Subsequent License Renewal Commitments Program Commitment The Fatigue Monitoring program is an existing preventive program that will be enhanced as follows:

1. Procedure(s) will be revised to require:

a.Transient cycles associated with the ASME Code,Section XI, Appendix A and L fatigue-sensitive locations be identified and tracked each ten-year interval.

b.A surveillance limit be established for transient cycles associated with the ASME Code, Fatigue Section XI, Appendix A and L fatigue-sensitive locations and corrective actions be initiated prior to 46 Monitoring exceeding the ASME Code,Section XI, Appendix A or L analyses transient cycle assumptions.

program

2. Procedure(s) will be revised to include component repair, component replacement, performance of a more rigorous analysis, performance of an ASME Code,Section XI, Appendix A or L flaw-tolerance analysis, or scope expansion that considers other locations with the highest expected CUFen values, as corrective action considerations when a cycle counting surveillance limit is exceeded.

3. Procedure(s) will be revised to require that when a cycle-counting surveillance limit is reached, action will be taken to ensure that the analytical bases of the high-energy line break (HELB) locations are maintained.

Neutron F/uence 47 Monitoring The Neutron Fluence Monitoring program is an existing condition monitoring program that is credited.

program Environmental Qualification of The Environmental Qualification of Electric Equipment program is an existing condition monitoring program 48 Electric Equipment that is credited.

program The Concrete Containment Unbonded Tendon Prestress program is an existing condition monitoring Concrete program that will be enhanced as follows:

Containment

1. Procedure(s) will be revised to specify that the trend analyses of tendon prestress loss will include 49 Unbonded trends projected through the end of the subsequent period of extended operation.

Tendon Prestress

2. Procedure(s) will be revised to specify that for each surveillance interval, the predicted lower limit, program minimum required value, and trending lines will be developed for the subsequent period of extended operation as part of the regression analysis for each tendon group.

The diesel fire gumg engine jacket water heat exchanger core will be reglaced at least once eve!Y 20 50 N/A years. (Added - Sugglement 2}

Virgil C. Summer Nuclear Station PageA-76 Application for Subsequent License Renewal Appendix A* UFSAR Supplement AMP Implementation Program enhancements for SLR will be 83.1 implemented 6 months prior to the subsequent period of extended operation.

83.2 Ongoing 83.3 Ongoing Program enhancements for SLR will be implemented 6 months prior to the 83.4 subsequent period of extended operation.

Proceg!,!re§ will Qe imglementeg fl month§ grior to the §!.!bseguent geriog N/A of extenged ogerl;!tion to reglace the diesel fire gumg engine j2Qket water heat exchanger core 2t least onQe eve0£ 2Q years.

Supplement 2

Serial No.24-157 Supplement 2 Page 16 of22 Table 82-1 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix B - Aging Management Programs VCSNS Program Consistency with NUREG-2191 Program Appendix Existing Pro~ram has Program has or New NUREG-2191 Program B

VCSNS NU EG-2191 Exce~tions to Reference Program Enhancements NUR G-2191 Inspection of Overhead Heavy Load and Light Load (Related 82.1.13 Existing X

to Refueling) Handling Systems Compressed Air Monitoring 82.1.14 Existing X

Fire Protection 82.1.15 Existing X

Fire Water System 82.1.16 Existing X

Outdoor and Large Atmospheric Metallic Storage B2.1.17 New Tanks Fuel Oil Chemistry B2.1.18 Existing X

Reactor Vessel Material 82.1.19 Existing Surveillance One-Time Inspection B2.1.20 New Selective Leaching B2.1.21 New ASME Code Class 1 B2.1.22 Existing X

Small-Bore Piping External Surfaces Monitoring of B2.1.23 Existing X

Mechanical Components Flux Thimble Tube Inspection B2.1.24 Existing X

Inspection of Internal Surfaces in Miscellaneous Piping and B2.1.25 Existing X

Ducting Components Lubricating Oil Analysis B2.1.26 Existing X

Monitoring of Neutron-Absorbing Materials B2.1.27 Existing Other Than Boraflex Buried and Underground Piping B2.1.28 Existing X

and Tanks PageB-12

Serial No.24-157 Supplement 2 Page 17 of 22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix B -Aging Management Programs 82.1.13 Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program Description The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems program is an existing condition monitoring program that manages loss of material due to general corrosion and wear, deformed or cracked rails, bridges, structural members, and structural components; and loss of material due to general corrosion, cracking and loss of preload on bolted connections for cranes and hoists within the scope of subsequent license renewal. The inspection and maintenance activities specified in this program are consistent with the following requirements identified in FSAR Section 3.12.4.1:

ANSI 830.2 1976, "Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder}"

NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" The cranes and hoists within the scope of subsequent license renewal include those previously evaluated as part of compliance with NUREG-0612, as well as other equipment handling systems operating over safety-related equipment. Also, within the scope of subsequent license renewal are fuel and equipment handling systems that handle loads over fuel and safety-related equipment.

The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems program uses periodic visual inspections and non-destructive examination (NOE) examinations, as needed if identified during any follow-up under the Corrective Action Program, to manage cracking and loss of material. Structural bolting is also monitored for loss of preload by inspecting for loose or missing bolts, or nuts. Inspection frequencies are consistent with the recommendations within the ASME/ANSI 830 series of standards. For handling systems that are infrequently in service, such as those only used during refueling outages, periodic inspections are performed prior to use. Cranes and hoist inspections do not include inspection of the structures that support the cranes. The individual structures and structural components are examined by the Structures Monitoring program (82.1.35).

NUREG-2191 Consistency The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems program is an existing program that, following enhaneement, 1.vill be i.§_consistent with NUREG-2191,Section XI.M23, "Inspection of Overhead Heavy Load and Light Load {Related to Refueling) Handling Systems."

Exception Summary None PageB-79

Serial No.24-157 Supplement 2 Enhancements Page 18 of 22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix B -Aging Management Programs NonePrior to the subsequent period of extended operation, the follo*..ving enhancements will be implemented in the follo*Ning program element(s):

Scope of Program (Element 1); Parameters Monitored I Inspected (Element 3); Detection of Aging Effects (Element 4); and /\\cceptance Criteria (Element 6) 4-:-

Procedure(s) will be re*tised to require:

Visual inspections of rails, bridges, structural members, and structural components for loss of material due to general corrosion; deformation; oracking, and 1Near.

Visual inspections of rails, bolted connections for loss of material due to general corrosion; oracking; and loose or missing bolts or nuts, and other conditions indicative of loss of preload.(Completed - Supplement 2)

Operating Experience Summary The following examples of operating experience provide objective evidence that the Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems program has been, and will be effective in managing the aging effects for SSCs within the scope of the program so that the intended functions will be maintained consistent with the current licensing basis during the subsequent period of extended operation.

1.

In April 2016, it was identified that the spent fuel pit bridge crane was generating foreign material along the rail as it moved through a section of its normal travel path adjacent to the spent fuel pool. Operation of the crane revealed that the foreign material was due to loss of material due to wear. The cause of the wear was attributed to misalignment of the guide rollers used to assist in maintaining the crane along the proper orientation. This misalignment resulted in contact between the rails and the seismic restraint brackets. New brackets were fabricated and installed for the guide rollers and the guide rollers were properly adjusted. The clearance between the seismic restraint brackets and the rails was increased to prevent grinding against the rail during operation. The results of subsequent inspections of the spent fuel pit bridge crane in 2019 and 2021 were satisfactory.

2.

In May 2017, during the Reactor Building closeout inspection following a refueling outage, metal flakes were identified along the reactor building polar crane trolley rails. This condition was reviewed by the reactor building polar crane subject matter expert who determined that the small amount of metal flakes was a normal byproduct of metal-to-metal wear between the wheels and the rails and that no immediate corrective actions were required. The results of subsequent inspections of the reactor building polar crane in 2018 and 2020 were satisfactory.

PageB-80

Serial No.24-157 Supplement 2 Page 19 of22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix B - Aging Management Programs The above examples of operating experience provide objective evidence that the Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems program includes activities to perform visual inspections to identify loss of material due to general corrosion or wear of the rails, bridges, structural members, and structural components; and identify deformation or cracking of the rails, bridges, structural members, and structural components. In addition, bolted connections are monitored for loss of material, cracking, and loose bolts, missing or loose nuts, and other conditions indicative of loss of preload and to initiate corrective actions. Occurrences identified under the Inspection of Overhead Heavy Load and Light Load (Related to Refueling)

Handling Systems 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 Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems program, following enhancement, will effectively manage aging prior to a loss of intended function.

Conclusion The continued implementation of the Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems program, following enhancement, provides reasonable assurance that aging effects will be managed such that the components within the scope of this program will continue to perform their intended functions consistent with the current licensing basis during the subsequent period of extended operation.

PageB-81

Serial No.24-157 Page 20 of22 Supplement 2 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix B -Aging Management Programs B2.1.15 Fire Protection Program Description The Fire Protection program is an existing condition monitoring and performance monitoring program that requires periodic visual inspections of fire barrier components, and functional testing of fire-rated doors and the carbon dioxide (CO2) fire suppression system. The program manages:

Loss of material for fire-rated doors (including roll-up fire window shutter). fire damper housings.assemblies, and steel seismic gap covers, and the GO~ fire suppression system 0

Loss of material or cracking for concrete structures, including fire barrier walls, ceilings, and floors Loss of material or cracking for the COi fire suppression system Shrinkage and loss of strength for elastomerie fire barrier penetration seals and for seismio gap elastomers Change in material properties, cracking/delamination, loss of material, and separation for non-elastomer fire barrier penetration seals, fire stops, radiant energy shields, and fire wraps and coatings Shrinkage, loss of strength, and hardening, or any other sign of degradation, for elastomeric fire barrier penetration seals and for seismic gap fillers The Fire Protection program requires periodic visual inspections of the penetration seals, and the fire barrier walls, ceilings and ~oors in structures within the scope of subsequent license renewal.

Periodic visual inspections of fire barriers include determining the condition of fire wraps, fire-rated doors (including any shutter for a fire window). and fire damper housings3ssemblies. These periodic inspections are performed every 18 months.

For the penetration seals, a 10% sample of each seal type is visually inspected each refueling outage. If abnormal.my sign of degradation is observed, an additional 10% sample population is inspected. The process for 10% sample population expansions continues until no abnormalsigns of degradation isare observed.

The Fire Protection program monitors the external surfaces of the CO2 fire suppression system components for aging effects during plant walkdowns. Visual inspections during the walkdowns identify cracking or corrosion that may lead to [loss of material. Periodic functional testing is performed to confirm the operability of the CO2 fire suppression system. The halon fire suppression system is used only for the Secondary Alarm Station and does not require aging management for SLR.

Personnel performing inspections are qualified and trained to perform the inspection activities.

Unacceptable conditions are entered into the Corrective Action Program for disposition.

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Serial No.24-157 Supplement 2 Page 21 of22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix B -Aging Management Programs The Fire Protection program is a risk-informed, performance-based program built upon National Fire Protection Association (NFPA) 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition." Adoption of NFPA 805 in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.48{c) serves as the method of satisfying 10 CFR 50.48{a) and General Design Criterion 3.

NUREG-2191 Consistency The Fire Protection program is an existing program that, following enhancement, will be consistent with NUREG-2191,Section XI.M26, "Fire Protection,"

as modified by SLR-ISG-2021-02-MECHANICAL, "Updated Aging Management Criteria for Mechanical Portions of the Subsequent License Renewal Guidance."

Exception Summary None Enhancements Prior to the subsequent period of extended operation, the following enhancements will be implemented in the following program elements:

Scope of Program (Element 1), Parameters Monitored/ Inspected (Element 3); Detection of Aging Effects (Element 4); and Acceptance Criteria (Element 6)

1.

Procedure(s) will be revised to provide guidance for detection of loss of material, cracking, holes, and gaps during the visual inspections of fire dampers to ensure that any deficiencies are noted on a condition report, and to determine the acceptability of the findings. (Revised -

Supplement 2)

2.

Procedure(s) will be revised to require that inspections of fire barrier elastomeric penetration seals and seismic gap filler identify shrinkage, loss of strength, and hardening, or any other signs of degradation. (Added - Supplement 2)

Monitoring and Trending (Element 5) 6-3. Procedure(s) will be revised to specify that inspection results of materials susceptible to delamination, change in material properties, separation, increased hardness, shrinkage, or loss of strength will be trended. Where practical, identified degradation will be projected until the next scheduled inspection. Results will be evaluated against acceptance criteria to confirm that the timing of subsequent inspections will maintain the components' intended functions throughout the subsequent period of extended operation based on the projected rate of degradation. (Renumbered and Revised Supplement 2)Proceduro(s) 111ill be revised to provide guidance to ensure that inspection results for fire protection components are trended to provide for timely detection of aging effects so that the appropriate corrective actions can be taken. VI/here practical, identified degradation 1Nill be projected until the next scheduled PageB-86

Serial No.24-157 Supplement 2 Page 22 of 22 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix B -Aging Management Programs inspection. Results arc evaluated against acceptance criteria to confirm that the timing of subsequent inspections 'Nill maintain the components' intended functions throughout the subsequent period of extended operation based on the prOjected rate of degradation.

4.

Procedure(s) will be revised to specify that inspections will be performed to identify cracking and loss of material for CO2 fire protection system components. and that those results are trended and appropriate corrective actions identified, if necessary. Where practical. identified degradation will be projected until the next scheduled inspection. Results will be evaluated against acceptance criteria to confirm that the timing of subsequent inspections will maintain the components' intended functions throughout the subsequent period of extended operation based on the projected rate of degradation. {Added - Supplement 2}

Corrective Actions (Element 7)

&.-5. Procedure(s) will be revised to provide guidance for evaluating prOjected inspection results.

For rcsultsspecify that for inspection results that will fail to meet acceptance criteria prior to the next scheduled inspection, inspection frequencies will be adjusted as determined by the Corrective Action Program. (Renumbered and Revised - Supplement 2)

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

1.

In July 2014, a 1-inch diameter hole was identified on a gypsum fire barrier wall in the switchgear room. The hole penetrated the gypsum (fire barrier) but not the corrugated steel on the opposite side (pressure boundary). The pressure boundary was intact, but the fire barrier was degraded. The hole in the gypsum board was patched with joint compound to correct the condition and restore the function of the gypsum wall as a fire barrier.

2.

In February 2015, during a plant walkdown, a fire barrier penetration seal was found to contain a void. The penetration was supposed to contain 1 O inches of foam seal material. The degradation was 3-2/3 inches deep within the 10-inch foam, and was approximately 6 inches high. It appeared to be due to shrinkage rather than any external damage. The void in the penetration seal was filled with the required amount of additional foam sealant to correct the condition, and to restore the penetration seal as a fire barrier.

3.

In March 2016, during scheduled inspections of fire barriers in the Control Building, it was identified that the Kaowool fire barrier for one of the penetrations was degraded due to a gap between the Kaowool triple wrap and the adjacent drywall. The gap was closed using RTV fire-resistant foam to reconnect the Kaowool triple wrap to the drywall, thus correcting the condition and restoring the fire barrier function.

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