Text

Update to Subsequent License Renewal Application (SLRA) -Response to NRC Request for Additional Information Set 1 Response to NRC Request for Confirmation of Information Set 1 and Supplement 3
	ML24155A146
Person / Time
Site:	Summer
Issue date:	05/30/2024
From:	James Holloway; Dominion Energy, Dominion Energy South Carolina
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML24155A144	List: ML24155A146;
References
	24-196A
	Download: ML24155A146 (1)
	v • d • e

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 Dominion Energy South Carolina. Inc.

5000 Dominion Boulevard, Glen Allen. VA 23060 Dominion Energy.com May 30, 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' 10 CFR 50 10 CFR 51 10 CFR 54 Serial No.: 24-196A NRA/SS: RO Docket No.: 50-395 License No.: NPF-12 UPDATE TO SUBSEQUENT LICENSE RENEWAL APPLICATION (SLRA)

RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION SET 1 RESPONSE TO NRC REQUEST FOR CONFIRMATION OF INFORMATION SET 1 AND SUPPLEMENT 3 By letter dated August 17, 2023 [Agency wide Documents Access and Manag ement 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.

The NRC staff identified areas where additional information is needed to complete their review. In an email from Marieliz Johnson (NRC) to Eric S. Carr (DESC), dated May 6, 2024 (ADAMS Package No. ML24127A110), the NRC staff transmitted requests for additional information (RAls ) to support compl etion of the Safety Review. The NRC's RAls and proprietary and non-proprietary versions of DESC's responses are provided in Enclosures 1 and 2, respectively.

Additionally, the NRC provided requests for confirmation of information (RCls) which has not been previously docketed but will likely be used by the NRC Staff in the Safety Evaluation Report for the VCS SLRA. The RC ls were transmitted in an email from Marieliz Johnson (NRC) to Eric S. Carr (DESC), dated May 8, 2024 (ADAMS Package No.

ML24129A068). The NRC's RCls and DESC's confirmation of the RCls are provided in .

Finally, the NRC staff has been conducting an Aging Management Audit since November 6, 2023, as part of their review of the VCSNS SLRA. As a result of discuss ions with the NRC staff throughout the audit, additional information which is necessary for the NRC 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 contains information that is being withheld from public disclosure under 1 0 CFR 2.390. Upon separation from Enclosure 1, this letter is decontrolled.

Serial No.: 24-196A Docket No.: 50-395

Enclosure 2

RESPONSE TO VCS SLRA REQUEST FOR CONFIRMATION OF INFORMATION SAFETY REVIEW - SET 1 ATTACHMENT TO CGE-GENW-TR-LG-000004-NP (NON-PROPRIETARY)

Dominion Energy South Carolina, Inc.

(Dominion Energy South Carolina, or DESC)

Virgil C. Summer Nuclear Station Unit 1 Westinghouse Non-Proprietary Class 3 Page 3 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 Attachment A

RAI 3.1.2.2.1-1

Regulatory Basis

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

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Background===

SLRA Table 3.1.1, Item 3.1.1-003 describes the aging management review (AMR) results for the stainless steel and nickel alloy reactor vessel internal (RVI) components that are subject to cumulative fatigue damage. This AMR item is also related to SLRA Section 3.1.2.2.1 for the further evaluation of the aging effect of cumulative fatigue damage and Section 4.3 for metal fatigue time -limited aging analysis (TLAA). SLRA Table 3.1.1 indicates that Item 3.1.1-003 is not applicable because the code of record for the RVIs is ASME Section III, Class 2, which does not specify a fatigue analysis.

Issue

In comparison, FSAR Section 3.9.3.6, Methods and Results of Blowdown Analysis (Mechanical) addresses the stress analysis of the RVI components during the postulated blowdown of the reactor coolant due to pipe breaks or loss-of-coolant accidents. FSAR Section 3.9.3.6 also indicates that the fatigue cumulative usage factor (CUF) of the RVI components during the postulated blowdown event is within the allowable limit (i.e., 1.0).

Considering that the FSAR describes the CUF analysis of the RVI components discussed above, the staff needs to clarify why SLRA Item 3.1.1-003 is not applicable to the RVIs that are subject to the existing fatigue analysis in FSAR Section 3.9.3.6.

Request

Resolve the apparent inconsistency between FSAR Section 3.9.3.6 indicating the presence of RVI CUF analysis and SLRA indicating that SLRA Item 3.1.1-003 is not applicable to the RVIs.

Dominion Response to RAI 3.1.2.2.1-1

The blowdown (due to large break LOCA) and seismic (due to safe shutdown earthquake) displacement and stress analyses contained in WCAP-7332-L-AR and WCAP-9401-P-A (referenced in FSAR Section 3.9.3.6) are for Faulted condition transients. Therefore, no calculation of long-term ASME BPV Code fatigue is included in WCAP-7332-L-AR and WCAP-9401-P-A. The acceptance criteria for the analyses in question are related to component displacement, deformation, and primary stress limits. Reference 12 of FSAR Section 3.9.5 (also referenced in FSAR Section 3.9.3.6) is a publicly available methodology paper and does not perform any generic or plant-specific analysis of record. Dominion will evaluate the FSAR Section 3.9.3.6 wording and determine if any changes are needed.

In summary, no CUF calculation or value has been discovered in the referenced analyses of FSAR Section 3.9.3.6, nor is one expected to exist. SLRA item 3.1.1-003 is not applicable to the RVI components because there is no fatigue analysis in the documents referenced by FSAR 3.9.3.6.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 4 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 RAI 4.3.1-1

Regulatory Basis

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

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Background===

SLRA Section 4.3.1 indicates that linear-rate cycle extrapolation of the total past operating period (i.e.,

cycle accumulation since the start of the operation through December 31, 2019) was used to project the number of future cycles for 80 years of operation. Th is approach is also called a full-life cycle projection.

Issue

The staff noted that there are relatively small margins for some extrapolated cycles (e.g., 9 cycles of the inadvertent auxiliary spray transient cycles for 80 years of operation compared to 10 design cycles, and 9 cycles of the reactor trip from full power with cooldown and safety injection (Case C) transient for 80 years of operation compared to 10 design cycles).

However, the applicant did not clearly discuss whether the recent cycle data (e.g., most recent 10 -year cycle data up to December 31, 2019) suggest an accelerated cycle accumulation rate for certain transients in comparison with the cycle accumulation rate based on the entire past operating period (i.e.,

cycle accumulation rate since the start of operation). Therefore, clarification of the most recent data is needed to evaluate the adequacy of the projections.

Request

Clarify whether the recent cycle data (e.g., most recent 10-year cycle data up to December 31, 2019) suggest an accelerated cycle accumulation rate for certain transients in comparison with the cycle accumulation rate based on the entire past operating per iod (i.e., cycle accumula tion rate since the start of operation). If so, provide justification for why the 80-year cycle projections do not use the most recent 10-year cycle data that involve the higher cycle accumula tion rate than the full -life cycle accumulation rate for those transients.

Dominion Response to RAI 4.3.1-1

The main cycle projection method utilized to determine the 80-year cycle counts for each transient was the full life rate method. The full life rate method calculates the projected 80 -year cycles based on an extrapolation of the cycle counts accumulated to date by the ratio of the projected number of operating years, divided by the current number of operating years (through December 31, 2019).

The table below summarizes the transient cycles recorded for Heatups, Cooldowns, and the two transients mentioned in the Issue section of this RAI from the 2010 and 2019 Virgil C. Summer Nuclear Station (VCSNS) thermal transient cycle counting yearly reviews. The rate of transient occurrence is not increasing for these four transients.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 5 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024

Heatup Time Period Cycles Rate (Cycles per Year) 1982 to 2019 68 1.84(1) 2010 to 2019 15 1.67 Cooldown Time Period Cycles Rate (Cycles per Year) 1982 to 2019 67 1.81(1) 2010 to 2019 16 1.78 Inadvertent Auxiliary Spray Time Period Cycles Rate (Cycles per Year) 1982 to 2019 4 0.11(1) 2010 to 2019 1 0.11 Reactor Trip from Full Power with Cooldown and Safety Injection (Case C)

Time Period Cycles Rate (Cycles per Year) 1982 to 2019 4 0.11(1) 2010 to 2019 0 0.00 Notes:

(1) The higher rates were used in the 80-year cycle projections.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 6 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 RAI 4.3.1-2

Regulatory Basis

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extende d operation.

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Background===

Table 3-2 of CGE-MC000-TR-LG-000010, Revision 3 lists the 80-year projected cycles of the transients for the reactor coolant system (RCS) and the auxiliary piping systems connected to the reactor coolant system.

Issue

However, the following reactor coolant system transients listed in the table are not listed as a design transient in SLRA Table 4.3.1-1: (1) feedwater cycling at hot shutdown transient; (2) reduced temperature return to power transient; (3) unit loading between 0 and 15 percent of full power transient; (4) unit unloading between 0 and 15 percent of full power transient; (5) loop out of service - normal loop shutdown transient; (6) loop out of service - normal loop startup transient; (7) boron concentration equalization transient; (8) refueling transient; (9) inadvertent reactor coolant system depressurization transient; (10) inadvertent startup of an inactive loop transient; (11) control rod drop transient; (12) inadvertent safety injection actuation transient; (13) secondary side leakage test transient; and (14) steam generator tube leakage test transient. In addition, Table 3-2 of CGE-MC000-TR-LG-000010, Revision 3 does not include the information on the inadvertent auxiliary sp ray transient listed in SLRA Table 4.3.1-1.

Request

Given that Table 3-2 of CGE-MC000-TR-LG-000010, Revision 3 describes additional RCS transients not included in SLRA Table 4.3.1-1, provide the following information.

1. Provide the cumulative transient cycles (up to 12/31/2019), 80-year cycle projections and current licensing basis cycle limits (design cycles), consistent with the format of SLRA Table 4.3.1 -1, for the following transients: (1) feedwater cycling at hot shutdown transient; (2) reduced temperature return to power transient; (3) unit loading between 0 and 15 percent of full power transient; (4) unit unloading between 0 and 15 percent of full power transient; (5) loop out of service - normal loop shutdown transient; (6) loop out of service - normal loop startup transient; (7) boron concentration equalization transient; (8) refueling transient; (9) inadvertent reactor coolant system depressurization transient; (10) inadvertent startup of an inactive loop transient; (11) control rod drop transient; (12) inadvertent safety injection actuation transient; (13) secondary side leakage test transient; and (14) steam generator tube leakage test transient.

2. Clarify whether the fatigue waiver evaluation in SLRA Section 4.3.2.6 remains valid for the subsequent period of extended operation and its technical basis (e.g., the design cycles were used in the fatigue waiver evaluation and the design cycles are bounding for the 80-year projected cycles). As part of the response, clarify whether the transients discussed in Request 1 are considered in the fatigue waiver evaluation. If not, discuss why these transients do not need to be considered in the fatigue waiver evaluation. In addition, clarify whether the transients discussed in Request 1 will be also monitored by the Fatigue Monitoring AMP to ensure that the fatigue wavier evaluation remains valid. If not, discuss why such monitoring is not necessary.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 7 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024

3. Clarify whether the CUF values of the Class 1 piping and components (SLRA Sections 4.3.2.1 through 4.3.2.5, 4.3.2.7 and 4.3.2.8) continue to meet the design limit (1.0) for the subsequent period of extended operation and its technical basis (e.g., the design cycles were used in the CUF analyses and the design cycles are bounding for the 80-year projected cycle s). As part of the response, clarify whether the transients discussed in Request 1 are considered in the Class 1 fatigue analyses. If not, discuss why these transients do not need to be considered in the Class 1 fatigue analyses. In addition, clarify whether the transients discussed in Request 1 will be also monitored by the Fatigue Monitoring AMP to ensure that the CUF values of the Class 1 components (SLRA Sections 4.3.2.1 through 4.3.2.5) continue to meet the design limit. If not, discuss why such monitoring is not necessary.

4. Given that Table 3-2 of CGE-MC000-TR-LG-000010, Revision 3 does not include the information on the inadvertent auxiliary spray transient listed in SLRA Table 4.3.1-1, clarify whether the consideration of the 80-year projected cycles of the inadvertent auxiliary spray transient does not affect the applicants determination that the 80-year projected cycles of the non-Class 1 piping systems and lines evaluated in CGE-MC000-TR-LG-000010, Revision 3 do not exceed 7000 cycles.

Dominion Response to RAI 4.3.1-2

Item 1

The cumulative transient cycles (up to 12/31/2019), 80-year cycle projections and current licensing basis cycle limits (design cycles) for the transients listed in Request 1 are summarized in the table below.

Metal Fatigue Cumulative 80-Year Current Transients Transient Cycles Transient-Cycle Licensing (12/31/2019) Projections Basis (CLB)

Cycle Limit Feedwater Cycling at Hot Shutdown 1,381 1,381(1) 2,000 Reduced Temperature Return to Power 0 2 2,000 Unit Loading Between 0 and 15 percent of 106 230 500 Full Power Unit Unloading Between 0 and 15 percent 33 72 500 of Full Power Loop Out of Service - Normal Loop 3 7 80 Shutdown Loop Out of Service - Normal Loop 18 39 70 Startup Boron Concentration Equalization 1,544 1,544(3) 26,400 Refueling 26 54 80 Inadvertent RCS Depressurization 6 13 20 Inadvertent Startup of an Inactive Loop 1 3 10 Control Rod Drop 5 11 80 Inadvertent Safety Injection Actuation 8 18 60 Secondary Side Leakage Test 10 10(2) 80 Steam Generator Tube Leakage Test 180 180(2) 800 Notes:

(1) The 80-Year Transient-Cycle Projections are the same as the Cumulative Transient Cycles (12/31/2019) for this transient, since no occurrences of this transient have been recorded since monitoring with the WESTEMS software began in the early 2000s.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 8 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 (2) The 80-Year Transient-Cycle Projections are the same as the Cumulative Transient Cycles (12/31/2019) for these transients, since no future occurrences of this transient are planned.

(3) The 80-Year Transient-Cycle Projections are the same as the Cumulative Transient Cycles (12/31/2019) for this transient due to the following:

a. No separate main spray operation is initiated to equalize boron during normal steady state operation, since bypass flow is sufficient.

b. Transients for plant startup and shutdown, or any other transients that occur during normal operation that include energizing heaters that induce spray, would be captured as transient events separate from the boron concentration equalization transient.

Item 2

It is confirmed that the fatigue waivers in SLRA Section 4.3.2.6 remain valid for the subsequent period of extended operation. The fatigue waivers apply to the Steam Generator (SG) tube plug and portions of the reactor coolant pump (RCP). It is confirmed that the 40-year design cycles used in the fatigue waivers are not projected to be exceeded during 80 years of plant operation since design cycles were used in the fatigue waiver evaluation and the design cycles are bounding for the 80 -year projected cycles.

The transients mentioned in Request 1 were considered based on which transients apply to each component in the fatigue waiver evaluations of SLRA Section 4.3.2.6. The total list of transients applicable to each component was determined and the 80-year projected cycles were compared to the design cycles for each transient.

There are no plans to monitor the transients discussed in Request 1 by the Fatigue Monitoring AMP.

Explicit monitoring of these transients is not necessary since each transient falls into at least one of the following categories:

x The 80-year projections are significantly lower th an the CLB cycles (i.e., projected cycles are <

60% of the CLB cycle limit based on the VCS NS monitoring of the transients since 2001).

x There have been no occurrences of this transient since the implementation of the cycle monitoring system in 2001.

x The transient occurs on a specific interval (i.e., 18-month refueling schedule) which ensures the cycles will remain under the CLB limits.

x The transient is tracked in Dominions corrective action program.

x There are no plans to perform the test which results in this transient.

Therefore, there are no new transients that are required to be added or any required modifications to cycles in the current VCSNS Unit 1 cycle counting program as a result of SLR. The transients identified in Table 5.2-2 of the FSAR will continue to be monitored during the subsequent period of extended operation.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 9 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 Item 3

It is confirmed that the CUF values of the Class 1 piping and components (SLRA Sections 4.3.2.1 through 4.3.2.5, 4.3.2.7 and 4.3.2.8) will continue to meet the design limit (1.0) for the subsequent period of extended operation. It is confirmed that the 40-year design cycles used in the fatigue analysis are not projected to be exceeded during 80 years of plant operation, and thus, these TLAAs are dispositioned according to 10 CFR 54.21(c)(1)(i). Table 4.3.1-1, 80-Year Transient-Cycle Projections in Section 4.3.1 of the SLRA, confirms that the transients listed in Technical Specification Table 5.7 -1 and FSAR Table 5.2-2 will not be exceeded in 80 years of plant operation. The transients mentioned in Request 1 were considered based on transients that apply to each component in the Class 1 fatigue evaluations of SLRA Sections 4.3.2.1 through 4.3.2.5, 4.3.2.7, and 4.3.2.8. As discussed in Item 2 above, there are no plans to monitor the transients discussed in Request 1 by the Fatigue Monitoring AMP. See the response to Item 2 for justification related to why the transients in Item 1 are not explicitly monitored.

Item 4

Table 3-2 lists transient cycle projections for the reactor coolant system (RCS) transients. The Inadvertent Auxiliary Spray transient cycle projections are documented in Table 3-9 of CGE-MC000-TR-LG-000010 Revision 3 and are replicated below. Table 3-1 and Table 3-9 of CGE-MC000-TR-LG-000010 both show that the 80-year projected transients for the pressurizer spray and auxiliary spray lines do not exceed 7,000 cycles. The 80-year projected cycle numbers compared to 7,000 cycles in Table 3-1 and Table 3-9 include the Inadvertent Auxiliary Spray transient 80-year projected cycles. No other non-Class 1 piping systems and lines are affected by the Inadvertent Auxiliary Spray transient.

Transient VCSNS Unit 1 80-Year Projected Cycles RCS Transients 5,638

[ ] a,c,e [ ] a,c,e

[ ] a,c,e [ ] a,c,e Total 5,777 < 7,000

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 10 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 RAI 4.3.2-1

Regulatory Basis

Pursuant to 10 CFR 54.21I, the SLRA must include an evaluation of time-limited aging analyses (TLAAs).

The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or ( iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

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Background===

SLRA Section 4.3.2.8 indicates that the CUF values for the pressurizer surge line will remain less than 1.0 for the subsequent period of extended operation based on the 80-year projected cycles that are less than the design cycles as listed in SLRA Table 4.3.1-1. In addition, the following reference describes the pressurizer surge line design transients involving thermal stratification (

Reference:

WCAP -12785, Revision 0, Structural Evaluation of the Virgil C. Summer Pressurizer Surge Line, Considering the Effects of Thermal Stratification, Table 2-1). These transients are used in the existing CUF analysis for the pressurizer surge line and are not directly tied to the heatup/cooldown transients.

Issue

However, some of the transients listed in WCAP-12785, Revision 0, Table 2-1 (e.g., loop out of service shutdown active loop and loop out of service shutdown inactive loop transients) are not described in SLRA Table 4.3.1-1. In addition, the SLRA does not clearly discuss how the applicant determined that the projected cycles for 80 years of operation do not exceed the surge line design transient cycles that are not directly tied to the heatup/cooldown transients.

Request

Describe the 80-year projected cycles and design cycles for each of the pressurizer surge line transients that are listed in WCAP-12785, Revision 0, Table 2-1 to demonstrate that the cumulative usage factor (CUF) values of the surge line continue to meet t he design limit (1.0) for the subsequent period of extended operation. If an updated CUF analysis exists for the pressurizer surge line, provide the CUF analysis results (including the relevant transients, th eir 80 -year projected cycles and design cycles, and CUF values) to confirm that the updated CUF analysis continues to be valid for the subsequent period of extended operation.

Dominion Response to RAI 4.3.2-1

Pressurizer surge line stratification events can occur during plant heatup/cooldown transients and during non-heatup/cooldown transients. As shown in the table below, the transients used in the pressurizer surge line evaluation that are not directly tied to the heatup/cooldown transients were considered and projected to 80 years of operation, in addition to those explicitly discussed in SLRA Section 4.3.2.8 and those listed in SLRA Table 4.3.1-1. All 80-year projected transient cycles for the transients used in the pressurizer surge line evaluation are less than the corresponding design cycles. Therefore, the Class 1 CUF values for the pressurizer surge line continue to meet the design limit of 1.0 for the subsequent period of extended operation.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 11 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024

Transient 80-year Projected Design Cycles (From Cycles WCAP-12785 Table 2-1)

[ ]a,c,e [ ]a,c,e [ ] a,c,e

[ ]a,c,e [ ]a,c,e [ ]a,c,e

[ ]a,c,e [ ] a,c,e [ ] a,c,e

[ ]a,c,e [ ]a,c,e [ ] a,c,e

[ ]a,c,e [ ]a,c,e [ ]a,c,e

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 12 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 RAI 4.3.3-1

Regulatory Basis

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

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Background===

Table 3-1 of CGE-MC000-TR-LG-000010, Revision 3 (July 2023) describes the 80-year projected cycles of the following piping systems and lines to demonstrate that the total projected cycles of the non -Class 1 piping systems connected to the reactor coolant s ystem do not exceed 7000 cycles: (1) reactor coolant line (RCL); (2) residual heat removal piping; (3) safety injection accumulator pi ping; (4) cold leg safety injection piping; (5) normal/alternate charging branch line; (6) normal letdown/excess letdown b ranch line; (7) pressurizer safety and relief line; and (8) pressurizer spray and auxiliary spray line. The number of the projected cycles for the reactor coolant line is used to estimate the reactor coolant system cycles that contribute to the total projected cycles of the non-Class 1 piping systems connected to the reactor coolant system. In addition, piping-specific transient cycles are included in the total projected cycles of the non -

Class 1 piping systems and lines.

Issue

However, SLRA Table 4.3.3-1 does not clearly describe the 80-year projected cycles (i.e., RCL cycles and piping-specific cycles) for the non-Class 1 piping lines and systems and how the applicant determined these projected cycles. In addition, it is not clear to the staff whether the following non-Class 1 piping systems, which are connected to the reactor coolant line and discussed in SLRA Table 4.3.3 -1, need to consider piping-specific transient cycles in ad dition to the reactor coolant system transients: (1) extraction steam; (2) feedwater; (3) gland sealing steam; (4) main steam dump; and (5) main steam.

Request

1. Describe the 80-year projected cycles (including the RCL cycles and piping-specific cycles) and the associated transients for the following piping systems and lines: (1) reactor coolant line; (2) residual heat removal piping; (3) safety injection accumu lator pi ping; (4) cold leg safety injection piping; (5) normal/alternate charging branch line; (6) normal letdown/excess letdown branch line; (7) pressurizer safety and relief line; and (8) pressurizer spray and auxiliary spray line.

2. Describe how the applicant determined the transients and projected cycles (i.e., contributing RCL cycles and piping-specific cycles) for the piping systems and lines discussed in the background section and request 1 (e.g., based on piping system design specification and information, plant operation procedures, test requirements, FSAR information and specific system-level knowledge).

3. With respect to the total transient cycle estimation, clarify whether the following non -Class 1 piping systems, which are connected to the RCL and discussed in SLRA Table 4.3.3-1, need to consider piping-specific transient cycles in addition to the reactor coolant system transients: (1) extraction steam; (2) feedwater; (3) gland sealing steam; (4) main steam dump; and (5) main steam. If so, discuss why the consideration of the piping-specific transient cycles does not result in the 80-year transient cycles exceeding 7000 cycles. If not, provide the technical basis for why the total cycle projections do not need to consider piping-specific transient cycles in addition to the reactor coolant system cycles.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 13 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 Dominion Response to RAI 4.3.3-1

Item 1

The detailed breakdown of the 80-year projected cycles for the piping systems in Request 1 is described below.

(1) Reactor Coolant Line:

80-Year Transient Projected Cycles Normal Conditions

[ ] a,c,e [ ] a,c,e

[ ] a,c,e [ ] a,c,e Upset Conditions

[ ] a,c,e [ ] a,c,e

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 14 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 80-Year Transient Projected Cycles

[ ] a,c,e [ ] a,c,e

[ ] a,c,e [ ] a,c,e Test Conditions

[ ] a,c,e [ ] a,c,e

Total RCS Transient Cycles 5,638 Notes:

(1) [ ] a,c,e

(2) Residual Heat Removal Piping:

Transient VCSNS Unit 1 80-Year Projected Cycles RCS Transients 5,638

[ ] a,c,e [ ] a,c,e Total 5,786

(3) Safety Injection Accumulator Piping:

Transient VCSNS Unit 1 80-Year Projected Cycles RCS Transients 5,638

[ ] a,c,e (1) [ ] a,c,e

[ ] a,c,e [ ] a,c,e

[ ] a,c,e [ ] a,c,e (2)

Total 5,640 Notes:

(1) This event is a faulted event and is not included in fatigue evaluations.

(2) [

] a,c,e

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 15 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 (4) Cold Leg Safety Injection Piping:

VCSNS Unit 1 Transient 80-Year Projected Cycles RCS Transients 5,638

[ ] a,c,e [ ] a,c,e

[ [ ] a,c,e

] a,c,e

[ ] a,c,e [ ] a,c,e (1)

[ ] a,c,e [ ] a,c,e

[ ] a,c,e (2) [ ] a,c,e

[ ] a,c,e [ ] a,c,e (3)

Total 5,775 Notes:

(1) [

] a,c,e (2) This event is a faulted event and is not included in fatigue evaluations.

(3) [

] a,c,e

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 16 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 (5) Normal/Alternate Charging Branch Line:

Transient VCSNS Unit 1 80-Year Projected Cycles(1)

RCS Transients 5,638

[ [ ] a,c,e

] a,c,e

[ [ ] a,c,e

] a,c,e

[ [ ] a,c,e

] a,c,e

[ [ ] a,c,e

] a,c,e

[ [ ] a,c,e

] a,c,e

[ [ ] a,c,e

] a,c,e

[ [ ] a,c,e

] a,c,e

[ [ ] a,c,e

] a,c,e

[ [ ] a,c,e

] a,c,e Total 6,030 Notes:

(1) [

] a,c,e (2) [

] a,c,e

(6) Normal/Excess Letdown Branch Line:

Transient VCSNS Unit 1 80-Year Projected Cycles RCS Transients 5,638

[

] a,c,e [ ] a,c,e Total 5,723 Notes:

(1) [

] a,c,e

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 17 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024

(7) Pressurizer Safety and Relief Line:

Transient VCSNS Unit 1 80-Year Projected Cycles RCS Transients 5,638

[ ] a,c,e(1) [ ] a,c,e (2)

Total 5,750 Notes:

[

] a,c,e (2) [

] a,c,e

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 18 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 (8) Pressurizer Spray and Auxiliary Spray Line Transient VCSNS Unit 1 80-Year Projected Cycles RCS Transients 5,638

[ ] a,c,e [ ] a,c,e

[ ] a,c,e [ ] a,c,e Total 5,777

The total number of 80-year projected transient cycles for each non-Class 1 piping system (numbers (2) through (8) above) is less than 7,000 cycles. Therefore, the existing TLAA evaluations for these systems remain valid for the subsequent period of extended operation.

Item 2

For each non-Class 1 system, the Class 1 transients asso ciated with the piping system were also applied to the non-Class 1 piping, as indicated in the tables in the response to Request 1. This is because the Class 1 portion of each system in Request 1 will experience the RCS transients and the same transients as the non-Class 1 portion of the same system. The resulting list of transients for each non -Class 1 system is conservative because the non-Class 1 portion of each piping system will only be subject to a subset of the transients experienced by the Class 1 portion. T he projected cycles for the piping systems discussed in Request 1 were determined by summing the 80-year projected cycles for all transients applicable to that system, including both RCS transients and line-specific transients. The only RCS projected cycles that were not included in the summation were for Steady State Fluctuations because this transient is insignificant for fatigue.

Item 3 It is confirmed that the identified non-Class 1 piping systems (i.e., 1) extraction steam; (2) feedwater; (3) gland sealing steam; (4) main steam dump; and (5) main steam) are not connected to the RCL and are not subject to RCL transients. These balance of plant (BOP) systems experience different transients than the primary side RCS and auxiliary systems. However, the transient cycles experienced by the 5 piping systems above are tied to the number of fuel cycles experienced at VCSNS. Fuel c ycles are conservatively tied to Heatup and Cooldown transients for which 80-year projected cycles are provided in SLRA Table 4.3.1-1. The Heatup and Cooldown 80-year projected cycles are less than the 40-year cycles that were assumed in the original plant design. The relationship between plant fuel cycles and the BOP piping system transients has not changed since the original design. Therefore, it is concluded that the 80 -

year projected transient cycles for BOP piping system are expected to remain less than their respective design cycles. Thus, no increase in the number of full temperature range cycles experienced by these BOP piping systems is expected, and the existing TLAAs remain valid for the subsequent period of extended operation.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 19 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 RAI 4.3.4-1

Regulatory Basis

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

=

Background===

SLRA Section 4.3.4 addresses the environmentally assisted fatigue (EAF) TLAA, including the EAF screening evaluation to determine the limiting EAF locations. SLRA Section 4.3.4 indicates that the applicant calculated the most conservative Fen (environmental fatigue correction factor) values in the screening evaluation.

Issue

SLRA Section 4.3.4 does not clearly describe how the applicant calculated the most conservative Fen values in the EAF screening evaluation (e.g., how the applicant determined the temperature, strain rate, and sulfur content, as applicable, in the Fen calculations). In addition, the SLRA does not clearly discuss how the applicant refined the screening CUFen (environmentally adjusted cumulative usage factor) values to remove the associated conservatism after the screening CUFen calculations.

Request

1. Clarify how the applicant calculated the most conservative Fen values in the EAF screening evaluation in terms of temperature, strain rate, and sulfur content, as applicable, in the Fen.

2. Describe how the applicant removed the conservatisms from the screening CUFen values to determine the refined CUFen.

Dominion Response to RAI 4.3.4-1

Item 1

The maximum (most conservative) Fen values for the materials in the EAF screening evaluation were determined by choosing the transformed temperature (T*), transformed sulfur content (S*), as applicable, and the transformed strain rate (*) that would maximize the Fen value for each material type, based on the equations in NUREG/CR-6909 Revision 1.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 20 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 Item 2

The assessment of environmental effects on fatigue was performed according to NUREG/CR-6909, Revision 1 per SLRA Section B3.1, page B-218, and is consistent with the calculation of Fen factors described in SLRA Section 4.3.4.

Each location is reviewed to see if minor conservatisms can be removed. The following are examples of minor conservatisms that were removed:

o Ungrouping of grouped transients o Application of more refined allowable stresses in elastic/plastic penalty factors based on detailed Code rules.

o Application of new fatigue curves for carbon and low alloy steel provided in NUREG -6909, Revision 1, which is allowed per Regulatory Guide 1.207.

If the minor conservative reductions were insufficient to lower the CUFen to less than 1.0, then a finite element analysis was performed to refine the derived transient stresses and the corresponding CUF value.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 21 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 RAI 4.3.4-2

Regulatory Basis

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

=

Background===

SLRA Section 4.3.4 explains that, among the limiting EAF locations listed in SLRA Table 4.3.4 -1, flaw tolerance evaluations were performed on the normal and alternate charging cold leg nozzles and pressurizer surge line hot leg nozzle in accordance with ASME Code Section XI, Appendix L. SLRA Section B2.1.1 for the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD aging management program (AMP) also discusses the program enhancement regarding the inspections of these nozzles.

The staff noted that the corrective actions program element of GALL-SLR AMP X.M1 states that a flaw tolerance analysis with appropriate (e.g., inclusion of environmental effects) crack growth rate curves and associated inspections performed in accordance with Appendix L of ASME Code Section XI is an acceptable corrective action. With respect to the inspections, ASME Code Section XI, Appendix L, L -

3420, Successive Examinations specifies that, if the allowable operating period per the flaw tolerance analysis is equal to or greater than 10 years, the locations subject to the flaw tolerance analysis be inspected at the end of each inservice inspection interval (i.e., 10-year inspection interval).

Issue

However, the SLRA does not clearly address the following items: (1) whether the limiting locations subject to flaw tolerance evaluations are inspected by using ultrasonic testing examination; (2) whether these nozzles are included in the examination item f or thermal fatigue in accordance with the risk -informed inservice inspection program based on ASME Code Case N-716-1 (SLRA Section B2.1.1); (3) when the baseline inspection for the flaw tolerance evaluations will be performed to confirm the absence of an unacceptable flaw per Appendix L, L3410 prior to the subsequent period of extended operation; and (4) whether the inspections will be performed at the end of each 10-year inservice inspection interval in accordance with the ASME Code Section XI, Appendix L provisions (L-3420, Successive Examinations) and, if not, whether these nozzles will be inspected at least once during the subsequent period of extended operation.

Request

1. Provide the following information: (1) whether the limiting locations subject to flaw tolerance evaluations are inspected by using ultrasonic testing examination; (2) whether these nozzles are included in the examination item for thermal fatigue in accordance with the risk-informed inservice inspection program based on ASME Code Case N-716-1 (SLRA Section B2.1.1); (3) when the baseline inspection for the flaw tolerance evaluations will be performed to confirm the absence of an unacceptable flaw per Appendix L, L-3410 prior to the subsequent period of extended operation; and (4) whether the inspections will be performed at the end of each 10-year inservice inspection interval in accordance with the ASME Code Section XI, Appendix L provisions (L-3420, Successive Examinations) and, if not, whether these nozzles will be inspected at least once during the subsequent period of extended operation.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 22 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024

2. If the baseline and successive inspections for the charging and pressurizer surge line nozzles are not consistent with ASME Code Section XI, Appendix L, provide justification for why such inspections are acceptable as a corrective action to manage the aging effects of environmentally assisted fatigue.

3. Provide justification for why the applicants inspection approach is not identified as an exception to the corrective actions program element of GALL-SLR AMP X.M1 even though it may be different from the inspection approach per ASME Code Section XI, Appendix L specified in GALL-SLR AMP X.M1 (e.g.,

inspections at least at the end of each 10-year inservice inspection interval per Appendix L, L -3420). If it cannot be justified, identify an exception of the Fatigue Monitoring AMP to GALL -SLR AMP X.M1.

4. Revise the SLRA (e.g., SLRA Sections B2.1.1, B3.1, 4.3.4 and the related FSAR supplements), as needed.

Dominion Response to RAI 4.3.4-2

Item 1

(1)

The VCSNS Inservice Inspection (ISI) Program is written and implemented every 10 years. The current version of the ISI Program Plan is for the 5th 10-Year Interval.

The limiting locations subject to flaw tolerance evaluations are the Alternate Charging Nozzle -to-Pipe weld (CGE-1-4106A-10), the Pressurizer Surge Line Hot Leg weld (CGE-1-4500A-13) and the Normal Charging Nozzle-to-Pipe weld (CGE-1-4205A-12). The 5th 10-Year ISI Interval plan includes an ultrasonic inspection (UT) of the Alternate Charging Nozzle-to-Pipe weld (CGE-1-4106A-10) during refueling outage RF-31 (Spring 2029), but has no UT planned for the Pressurizer Surge Line Hot Leg weld (CGE-1-4500A-

13) or the Normal Charging Nozzle-to-Pipe weld (CGE-1-4205A-12).

(2)

Going forward, there are plans for performing UT on two of these three locations, with the exception being the Normal Charging Nozzle. However, the planned inspections will not be conducted on a 10-year ISI Interval frequency. The reason for performing inspections for all three of these locations would be related to satisfying the stability criteria in the flaw tolerance evaluation and implementation of the normal ISI program. For the welds that attach the piping to the Normal and Alternate Charging nozzles, the flaw tolerance evaluation confirmed the piping to be flaw tolerant for more than 80 years of plant operation and no future inspections would be necessary for manage ment of EAF at these two locations. However, the 5th interval ISI plan already has UT of the Appendix L flaw tolerance location of the Alternate Charging Nozzle-to-Pipe weld scheduled for refueling outage RF-31 (Spring 2029). The Alternate Charging and the Normal Charging lines experience a similar level of loading during operation, thus, the inspection of the Alternate Charging line is representative of the Normal Charging line. The 5th ISI plan also includes volumetric inspection of the Normal Charging line at other field welds per Code Case N-716-2, which manages thermal fatigue based on a risk-informed inspection plan. These other field welds on the Normal Charging line are in the same region as the analyzed flaw tolerance location (i.e. , the Normal Charging Nozzle to branch line weld).

The Pressurizer Surge Line Hot Leg Nozzle is included in Code Case N -716-2 inspection program but is not scheduled to be inspected during the 5th 10-Year ISI Interval.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 23 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 (3)

Dominion confirms that the limiting location for the Pressurizer Surge Line Hot Leg Nozzle to Surge Line field weld (CGE-1-4500A-13) has been previously ultrasonically inspected on March 31, 1993. Dominion also confirms that the Loop B Cold Leg Normal Charging Nozzle -to-Pipe Stainless Steel Weld and Loop A Cold Leg Alternate Charging Nozzle-to-Pipe Stainless Steel Weld have not been ultrasonically inspected.

Dominion has credited the preservice inspections for all three locations to confirm the absence of unacceptable flaws. For the Normal and Alternate Charging Nozzle-to-Pipe welds, the liquid penetrant (PT) examinations (performed in 2015) and the radiography testing (RT) examination (pre -service performed in 1978) showed no recordable indications. The Pressurizer Surge Line Hot Leg Surge Nozzle to Pipe Stainless Steel weld was UT during the preservice inspections, with no recordable indications.

The most recent UT and PT for the Pressurizer Surge Line Hot Leg Surge Nozzle to Pipe Stainless Steel weld was performed in 1993, with no recordable indications.

(4)

For the Pressurizer Surge Line Hot Leg Nozzle weld, the flaw tolerance evaluation confirmed that re -

inspection is needed after 48 years of plant operation. The Pressurizer Surge Line Hot Leg Nozzle weld was last ultrasonically inspected on March 31, 1993. Supplement 1 of the SLRA confirms that Dominion will re-inspect the Pressurizer Surge Line Hot Leg weld in 1st quarter of 2041. This date is before the plant enters the subsequent period of extended operation.

In addition, Dominion has proactively scheduled UT of the Alternate Charging Nozzle-to-Pipe weld during refueling outage RF-31 (Spring 2029).

While not required to manage EAF because the stability criteria demonstrate that the piping is flaw tolerant for greater than 80 years, Dominion confirms that the Alternate Charging Nozzle and Normal Charging Nozzle will be included in the scope of the Code Case N-716-2 program as potential inspection locations.

Taken together, these actions provide reasonable assurance that environmental fatigue is being managed for the Normal and Alternate Charging Nozzles and Pressurizer Surge Line Hot Leg nozzle.

Item 2

Baseline inspections have been completed for all three locations via:

1) Completion of preservice inspections,

2) Completion of inservice inspections (UT) for the Pressurizer Surge Line Hot Leg Nozzle, and

3) Completion of inservice inspections (PT) for the Normal and Alternate Charging Nozzle.

The need for successive examinations is tied to exceeding an acceptance criterion. For Appendix L, the acceptance criteria include a CUF value greater than 1.0 and a demonstration of flaw stability. For all three locations, the acceptance criteria are satisfied. Nonetheless, Dominion has scheduled successive examinations for the Pressurizer Surge Line Hot Leg Nozzle for management of EAF based upon the criteria for time between inspections from Appendix L. Also, the Alternate Charging Nozzle is scheduled for examination during the Spring 2029 refueling outage. The Alternate Charging Nozzle is representative of the Normal Charging Nozzle.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 24 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 Dominion has confirmed that ASME Section XI, Appendix L-3410(a) permits both surface examination and volumetric inspection techniques for confirming the absence of an unacceptable flaw. Appendix L -

3410(b) permits examination in accordance with IWA-2220, IWA-2230, or IWA-2240, as applicable.

ASME Section XI IWA-2220 discusses the various surface examinations, such as liquid penetrant (IWA-2222), which was used for the Pressurizer Surge and Charging Line Nozzle examinations.

The flaw tolerance evaluation performed per Appendix L is based on fracture mechanics guidance that provides a deterministic procedure to manage thermal fatigue issues and should be considered as a supplemental inspection management plan similar to Code Case N-716-2. The conclusions of the Appendix L evaluations, coupled with the inspection plans in Code Case N-716-2, work together to adequately manage the various thermal fatigue-sensitive locations in the plant. The successive examination requirements in Appendix L-3420 state that the inspection period shall not exceed what is specified in Table L-3420-1 or IWB-2410. ASME Appendix L-3420 is invoked when the CUF value is projected to exceed 1.0. For the Pressurizer Surge Line and Charging Lines, the CUF value continues to be less than 1.0 throughout the 80-year life of the plant. In addition, the transients for these piping systems are not projected to exceed the 40-year design transient limits within 80 years of plant operation.

Since the CUF values for these systems remain less than 1.0, Dominion has elected to use the formulas in Appendix L to verify the flaw tolerance and to determine the time between inspections . However, the successive inspection frequency identified in ASME Appendix L, Table L-3420-1 for aging management of the locations would not be invoked. Thus, Dominion maintains that the successive inspection criteria in Appendix L-3420 are neither applicable nor mandatory.

The requirements in IWB-2410 (Inspection program), and IWB-2420 (Successive Inspections) form the basis of the plant inspection plan. However, it should be noted that Code Case N-716-2 provides alternate requirements for successive examinations of IWB-2420 of risk-informed piping such as the Pressurizer Surge line and Charging Nozzle lines (Inspection category R-A per Code Case N-716-2).

Therefore, the Appendix L-3420 provisions for inspection are deferred to the inspection requirements of Code Case N-716-2. Thus, if a particular piping system is in the scope of the Code Case N-716-2 inspection program (as they are for VCSNS), then specific locations within that system (such as Pressurizer Surge Line Hot Leg Nozzles and Charging Lines Nozzles) would be potential inspection locations as determined through future Code Case N-716-2 program revisions for the 6 th, 7th, and 8th 10-Year ISI Intervals. As described in the response to Item 1, Dominion has elected to perform volumetric examinations for the Pressurizer Surge Line Hot Leg Nozzle weld and the Alternate Charging Nozzle-to-Pipe weld.

In summation, ASME Section XI, Appendix L, Paragraph L-3420 states that the successive inspection period shall not exceed that specified in Table L-3420-1 or IWB-2410. Since the acceptance criteria has not been exceeded, Dominion has elected to use the provisions in IWB -2410 and Code Case N-716-2.

Therefore, Dominion has scheduled successive inspections for the Pressurizer Surge Line Hot Leg Nozzle and Alternate Charging Nozzle per the requirements of ASME Section XI, IWB -2410, IWB-2420, and Code Case N-716-2.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 25 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 Item 3

Dominion has reviewed the corrective actions program element of GALL -SLR AMP X.M1 and confirmed that it permits use of a flaw tolerance evaluation:

In addition, a flaw tolerance analysis with appropriate (e.g., inclusion of environmental effects) crack growth rate curves and associated inspections performed in accordance with Appendix L of ASME Code Section XI is an acceptable correction [sic] action.

Furthermore, while Dominion has confirmed the absence of an unacceptable flaw at these locations, both the NRC GALL-SLR report (NUREG-2191, Vol. 2) and ASME Section XI allow for flaw tolerance evaluations to be performed in accordance with ASME Section XI Nonmandatory Appendix C Evaluation of Flaws in Piping. Since there are no flaws at these locations, there is no requirement to perform successive inspections to monitor flaw growth. Consistent with GALL-SLR AMP X.M1, Dominion will monitor the transients used in the flaw tolerance evaluation to confirm that the evaluation remains bounding.

The successive inspection for the Pressurizer Surge Line Nozzle per ASME Section XI, Appendix L is every 48 years. The time between inspections for the normal and alternate charging nozzles is greater than 80 years. The successive inspections for the normal and alternate charging nozzles are performed according to Code Case N-716-2. The 5th Interval ISI Plan has the alternate charging nozzle scheduled for inspection in 2029.

An exception to GALL-SLR AMP X.M1, Element 7 Corrective Acti ons is not required because Dominion is basing successive inspections as determined using ASME Section XI Appendix L stability limitations.

Furthermore, successive inspections are scheduled for both the Pressurizer Surge Line Nozzle and the Alternate Charging Nozzle, which represents the normal charging nozzle.

Item 4

Supplement 1 of the SLRA [ML24095A207] replaces the branch weld numbers with the field weld numbers and provides the inspection dates for the most recent inspection of the field welds for the Pressurizer Surge Line Hot Leg Nozzle, Normal Charging Nozzle, and Alternate Charging Nozzle.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 26 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 RAI 4.3.4-3

Regulatory Basis

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

=

Background===

SLRA Section 4.3.4 indicates that, among the limiting EAF locations listed in SLRA Table 4.3.4-1, a flaw tolerance evaluation was performed on normal and alternate charging cold leg nozzles and pressurizer surge line hot leg nozzle in accordance with Non-mandatory Appendix L of ASME Code Section XI. The staff also noted that the flaw tolerance evaluation calculated the fatigue crack growth rate in accordance with ASME Code Case N-809 (

Reference:

WCAP18790-P/NP, Revision 2, V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Charging Line Nozzle and Hot Leg Surge Line Nozzle Flaw Tolerance Evaluation Using ASME Section XI Appendix L Methodology).

Issue

In Code Case N-809, the parameter defining the effect of metal temperature on fatigue crack growth rate (ST) has a minimum value at 300 °F for austenitic stainless steels. Therefore, how the metal temperature is defined in the calculations of ST values may result in non -conservative fatigue crack growth rates. The staff needs clarification of whether the applicant determined the ST values for the fatigue transients in such a manner to estimate conservatively bounding ST values.

Request

Describe how the applicant determined the metal temperature in the calculation of the ST value for each fatigue transient. As part of the response, clarify whether the applicant determined the metal temperature for each fatigue transient in such a manner to estimate conservatively bounding ST value for the transient.

Dominion Response RAI 4.3.4-3

The Fatigue Crack Growth (FCG) evaluation in WCAP-18790 Rev. 2 is consistent with ASME Section XI Code Case N-809. WCAP-18790 has been revised from Revision 2 to Revision 3. In WCAP-18790, Revision 3 Section 3.2.2, a note was added that describes how conservative temperatures used in calculating FCG per Code Case N-809 were also considered in the evaluation. Specifically, [

] a,c,e

The FCG analysis in WCAP-18790 Revision 3 appropriately recognizes that the Code Case N-809 ST correlation reaches a minimum value at 300°F with respect to temperature . Accordingly, the appropriate temperatures were considered in WCAP-18790 Revision 3 to provide limiting FCG results when determining the ST parameter from Code Case N-809.

The conclusions and results in WCAP-18790, Revision 3 remain the same as Revision 2.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 27 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 RAI 4.3.4-4

Regulatory Basis

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or ( iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

=

Background===

The following reference indicates that the pressurizer lower head at the penetration (low alloy steel lower head location) has an 80-year projected CUFen is 0.994 (

Reference:

CGE-MC000-TR-LG-000011, Revision 3, V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Environmentally Assisted Fatigue Evaluations Summary Report, Table 4-1).

Issue

However, SLRA Table 4.3.4-1 does not identify the pressurizer lower head at the heater penetration (low alloy steel location) as a limiting EAF (environmentally assisted fatigue) location. Even though the 80 -year CUFen for the pressurizer lower head location is almost identical to 1.0, the SLRA does not clearly discuss the following information: (1) whether the CUFen value for the pressurizer lower head location is based on the design cycles or 80-year projected cycles; (2) why the pressurizer lower head location is not identified as a limiting EAF location in SLRA Section 4.3.4; (3) whether the location will be monitored by the Fatigue Monitoring AMP and, if not, how the applicant will ensure the CUFen of the location will not exceed the design limit of 1.0; and (4) if the CUFen value approaches the design limit, whether a corrective action (e.g., refinement of CUFen or repair/replacement activities) will be performed to ensure that the CUFen does not exceed the design limit.

In addition, note (12) of CGE-MC000-TR-LG-000011, Table 4-1 indicates that the pressurizer lower head location was evaluated by using a flaw tolerance evaluation. However, the SLRA does not discuss the flaw tolerance evaluation.

Request

1. Provide the following information: (1) whether the CUFen value for the pressurizer lower head location is based on the 80-year projected cycles or design cycles; (2) why the pressurizer lower head location is not identified as a limiting EAF location in SLRA Section 4.3.4; (3) whether the location will be monitored by the Fatigue Monitoring AMP and, if not, how the applicant will ensure that the CUFen of the location will not exceed the design limit of 1.0; and (4) if the CUFen value approaches the desig n limit, whether a corrective action (e.g., refinement of CUFen or repair/replacement activities) will be performed to ensure that the CUFen does not exceed the design limit.

2. Clarify whether a flaw tolerance evaluation was performed on the pressurizer lower head at the heater penetration (low alloy steel location) to manage the aging effects of EAF. If so, discuss the methodology and results of the flaw tolerance evaluation, including the inspections for the location.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 28 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 Dominion Response to RAI 4.3.4-4

Item 1

(1) The CUFen value for the pressurizer lower head is based on design cycles.

(2) The CUFen at the pressurizer lower head at the heater penetration is 0.994 , which is less than the limit of 1.0 (unity). Therefore, Dominion did not include the pressurizer lower head at the heater penetration as a limiting EAF location in SLRA Section 4.3.4.

(3) The CUFen at the pressurizer lower head at the heater penetration is 0.994 , which is less than the limit of 1.0 (unity). Although not necessary for aging management of th e TLAA for this location, the Fatigue Monitoring AMP will continue to monitor the transients listed in Technical Specification Table 5.7-1 and FSAR Table 5.2-2.

(4) A corrective action will be taken (consistent with the guidance of NUREG-2191, Vol. 2,Section X.M1 Fatigue Monitoring) if the transients listed in Technical Specification Table 5.7-1 and FSAR Table 5.2-2 approach their respective cycle limits.

Item 2

Dominion created Table 4-1: Final Equipment Sentinel Location EAF Results in the SLRA. Table 4-1 in the SLRA does not credit a flaw tolerance evaluation for the pressurizer lower head; therefore, note 12 from CGE-MC000-TR-LG-000011, Revision 3, Table 4-1 does not apply.

A flaw tolerance evaluation was not performed or credited for the VCSNS pressurizer lower head at the heater penetration to manage the aging effects of EAF .

Note that a sample flaw tolerance evaluation, using the available fracture mechanics methodology available in 1998, was performed in WCAP-14950 for this location to demonstrate feasibility for plant specific applications.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 29 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 RAI 4.3.4-5

Regulatory Basis

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

=

Background===

SLRA Table 4.3.4-1 and WCAP-18772-P, Table 4-26 address the CUFen values for the following EAF locations: (1) safety injection 6-inch reactor coolan t line (RCL) cold-leg nozzle and (2) residual heat removal 6-inch hot-leg nozzle.

Issue

However, the applicant did not credit the Fatigue Monitoring AMP to manage the effects of EAF for these safety injection and residual heat removal nozzle locations. In addi tion, SLRA Section 4.3.4 does not clearly discuss the following items related to these locations: (1) whether the design cycles or 80-year projected cycles are used in the CUFen calculations; (2) why the Fatigue Monitoring AMP is not credited to manage the aging effects of EAF for these components; (3) how the applicant will ensure the CU Fen of these locations does not exceed the design limit of 1.0; and (4) whether a corrective action will be taken (e.g., refinement of CUFen or repair/replacement activities) if the CUFen of locations approaches the design limit of 1.0.

Request

Describe the following items regarding the safety injection 6 -inch RCL cold-leg nozzle and residual heat removal 6-inch hot-leg nozzle: (1) whether the design cycles or 80-year projected cycles are used in the CUFen calculations; (2) why the Fatigue Monitoring AMP is not credited to manage the aging effects of EAF for these components; (3) how the applicant will ensure the CUFen of these locations does not exceed the design limit of 1.0; and (4) whether a corrective action will be taken (e.g., refinement of CUFen or repair/replacement activities) if the CUFen of these locations approaches the design limit of 1.0.

Dominion Response to RAI 4.3.4-5

(1) It is confirmed that the design cycles were used to produce the CUFen values for the safety injection 6-inch RCL cold-leg nozzle and the residual heat removal 6-inch hot-leg nozzle.

(2) The fatigue monitoring program is not credited for managing the aging effects of EAF for these components because the CUFen values are acceptable, since they are less than unity. Although not necessary for aging management of this TLAA, the Fatigue Monitoring AMP will continue to monitor the transients listed in Technical Specification Table 5.7-1 and FSAR Table 5.2-2. For transients not monitored by the Fatigue Monitoring AMP, the response to RAI 4.3.1-2, Item 2 provides the basis for not monitoring.

(3) It has been confirmed that the CUFen values for these locations will not exceed unity since the transients used to compute the 40-year design cycles are not projected to be exceeded during 80 years of plant operation. The transients in FSAR Table 5.2-2 are being monitored by the Fatigue Monitoring AMP to ensure that operational cycles remain below the respective cycle limits.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 30 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 (4) A corrective action will be taken (consistent with the guidance of NUREG-2191, Vol. 2,Section X.M1 Fatigue Monitoring) if a design transient listed in Technical Specification Table 5.7 -1 and FSAR Table 5.2-2 approaches its cycle limit.

RAI 4.3.5-1

Regulatory Basis

Pursuant to 10 CFR 54.21(c), the SLRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

=

Background===

SLRA Section 4.3.5 explains that the high-energy line-break (HELB) analysis for the ASME Code Class 1 piping is based on a cumulative usage factor (CUF) threshold of 0.1 for postulating the intermediate -

piping break locations in the HELB analysis. In compa rison, FSAR Section 3.6.2.1.2 also indicates that the postulation of HELB locations for ASME Code Class 2 and 3 piping is, in part, based on the allowable stress range for expansion stress (SA), consistent with Branch Technical Position MEB 3 -1 (ADAMS Accession No. ML19137A335).

Issue

SA may need to be adjusted by a stress range reduction factor that is determined by the number of thermal cycles, as addressed in the implicit fatigue analysis in SLRA Section 4.3.3. However, SLRA Sections 4.3.5 and A3.3.5 (FSAR supplement) do not clearly discuss whether the HELB location postulation for the Class 2 and 3 piping, which involves SA, is a basis for identifying the HELB analysis as a TLAA.

The staff also needs to clarify whether Enhancement 3 of the Fatigue Monitoring AMP (SLRA Section B3.1) for the corrective actions regarding HELB analysis is applied to the Class 2 and 3 piping HELB analysis as well as to the Class 1 piping HELB analysis.

Request

1. Provide justification for why SLRA Sections 4.3.5 and A3.3.5 do not identify the HELB analysis for the Class 2 and 3 piping systems as a TLAA even though the screening criteria of the HELB location postulation involve the dependency on the transient cyc les. If justification cannot be provided, identify the HELB analysis for the non-Class 1 piping systems as a TLAA in a consistent manner with the identification of non-Class 1 fatigue TLAA in SLRA Section 4.3.3 and discuss the disposition of the TLAA.

In addition, clarify whether the FSAR supplement (e.g., SLRA Section A3.3.5) needs to be revised accordingly.

2. Clarify whether Enhancement 3 of the Fatigue Monitoring AMP (SLRA Section B3.1) for the corrective actions regarding HELB analysis is applied to the Class 2 and 3 piping HELB analysis as well as to the Class 1 piping HELB analysis. If not, describe the technical basis of the scope and implementation of the program enhancement (i.e. corrective actions applicable only to Class 1 piping HELB analysis).

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 31 of 33 Our ref: CGE-GENW-TR-LG-000004-R2-NP-1, Rev. 2 May 24, 2024 Dominion Response to RAI 4.3.5-1

Item 1

Dominion concurs that SLRA Sections 4.3.5 and A3.3.5 did not identify the HELB analysis for the Class 2 and 3 piping systems as a TLAA. SLRA Section 4.3.3 confirms that the projected cycles for the HELB Class 2 and 3 piping would remain less than 7,000 cyc les. Nonetheless, Dominion has transmitted Supplement 2 (ADAMS Accession Number ML24129A200) of the SLRA to the NRC to identify the HELB analysis for the Class 2 and 3 piping systems as a TLAA. Analysis of HELB for the Class 2 and 3 piping confirms that the transients will remain less than 7,000 cycle s throughout the 80-year life of the plant.

Item 2

As described in Supplement 2 (ADAMS Accession Number ML24129A200) of the SLRA, Dominion confirms that Enhancement 3 of the of the Fatigue Monitoring AMP (SLRA Section B3.1) for the corrective actions regarding HELB analysis is applied to the Class 2 and 3 piping HELB analysis , as well as to the Class 1 piping HELB analysis.

- - This record was final approved on 05/24/2024 22:54:22. (This statement was added by the PRIME system upon its validation)

Serial No.: 24-196A Docket No.: 50-395

Enclosure 3

WESTINGHOUSE APPLICATION FOR WITHHOLDING AND AFFIDAVIT

Dominion Energy South Carolina, Inc.

(Dominion Energy South Carolina, or DESC)

Virgil C. Summer Nuclear Station Unit 1 Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-24-029 Page 1 of 3

Commonwealth of Pennsylvania:

County of Butler:

(1) I, Zachary Harper, Senior Manager, Licensing, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am requesting the proprietary portions of CGE-GENW-TR-LG-000004 Rev. 2 be withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.

(iii) Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

- - This record was final approved on 05/28/2024 13:44:42. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-24-029 Page 2 of 3

(5) Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(6) The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit.

- - This record was final approved on 05/28/2024 13:44:42. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-24-029 Page 3 of 3

I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 5/28/2024 _____________________________ _____________________________________________________________________________________________________________________________________________________

Signed Signedededededededeededededelectronically electronically byby Zachary Harper

- - This record was final approved on 05/28/2024 13:44:42. (This statement was added by the PRIME system upon its validation)

Serial No.: 24-196A Docket No.: 50-395

Enclosure 4

RESPONSE TO VCS SLRA REQUEST FOR CONFIRMATION OF INFORMATION SAFETY REVIEW - SET 1

Dominion Energy South Carolina, Inc.

(Dominion Energy South Carolina, or DESC)

Virgil C. Summer Nuclear Station Unit Serial No.: 24-196A Docket No.: 50-395 Enclosure 4, Page 2 of 5 The NRC provided requests for confirmation of information (RCIs) which has not been previously docketed but will likely be used by the NRC Staff in the Safety Evaluation Report for the VCS SLRA. The RCIs were transmitted in an email from Marieliz Johnson (NRC) to Eric S. Carr (DESC), dated May 8, 2024 (ADAMS Package No. ML24129A068).

The NRCs RCIs and DESCs confirmation of the RCIs are provided in this enclosure.

RCI-B2.1.28-1 During the audit, the staff reviewed several documents that contain information which will likely be used in conclusions documented in the Safety Evaluation (SE). To the best of the staff's knowledge, this information is not on the docket. Any information used to reach a conclusion in the SER must be included on the docket by the applicant. We request that you submit confirmation that the information gathered from the documents and listed below is correct or provide the associated corrected information. During its audit, the staff reviewed operating experience associated with a buried piping break (CR-17-01949) and buried piping leak (CR 21 01475). During a breakout session associated with SLRA Section B2.1.28 (specifically Question No. 1), the applicant informed the staff that the subject break and leak were due to mechanical damage of polyvinyl chloride (PVC) piping. Confirm that mechanical damage of PVC piping was the cause of the subject break and leak.

Dominion Response to RCI-B2.1.28-1 This information has been confirmed to be correct as stated.

RCI-3.3.1-1 During the audit, the staff reviewed several documents that contain information which will likely be used in conclusions documented in the Safety Evaluation (SE). To the best of the staff's knowledge, this information is not on the docket. Any information used to reach a conclusion in the SER must be included on the docket by the applicant. We request that you submit confirmation that the information gathered from the documents and listed below is correct or provide the associated corrected information. VII Table G, AMR Item VII.G.A-90 (SRP Item 3.3-1, 060) in Volume 1 of NUREG-2191 (GALL-SLR Report) manages cracking due to chemical reaction, weathering, settlement, or corrosion of reinforcement; and loss of material due to delamination, exfoliation, spalling, popout, or scaling of reinforced concrete structural fire barriers exposed to air by both the Fire Protection and Structures Monitoring programs. VII Table G, AMR Item VII.G.A-626 (SRP Item 3.3-1, 179) in Volume 1 of NUREG-2191 manages cracking due to restraint shrinkage, creep, aggressive environment; and loss of material (spalling, scaling) and cracking due to freeze-thaw of masonry walls that are considered fire barriers by both the Fire Protection and Masonry Walls programs. SLRA Tables 3.5.2-2, 3.5.2-4, 3.5.2-6, 3.5.2-7, 3.5.2-12, 3.5.2-13, 3.5.2-18, and 3.5.2-20 cite SRP Item 3.3-1, 060 for managing cracking and loss of material of reinforced concrete concrete elements exposed to air by both the Fire Protection and Structures Monitoring programs. SLRA Table 3.5.2-1 cites SRP Item 3.3-1, 060, with standard note E (Consistent with NUREG-2191 item for material, environment, and aging effect, but a different AMP is credited or NUREG-2191 identifies a plant-specific Serial No.: 24-196A Docket No.: 50-395 Enclosure 4, Page 3 of 5 AMP), for managing cracking and loss of material of reinforced concrete concrete elements exposed to air by both the Fire Protection Program and ASME Section XI, Subsection IWL (substituted for Structures Monitoring program) programs. In addition, SLRA Tables 3.5.2-4 and 3.5.2-20 cite SRP Item 3.3-1, 179 for managing cracking and loss of material of masonry walls masonry block walls exposed to air by both the Fire Protection and Masonry Walls programs. However, the SLRA tables noted above also included concrete concrete elements and concrete block masonry block walls where only the Structures Monitoring or Masonry Walls programs were credited to manage the effects of aging. During the audit of the Fire Protection AMP, it was discussed whether the concrete concrete elements and concrete block masonry block walls managed only by the Structures Monitoring or Masonry Walls programs have a fire barrier intended function and should also be managed by the Fire Protection program. The applicant stated during the audit that some of the concrete concrete elements and concrete block masonry block walls have a fire barrier intended function and are also managed by the Fire Protection program. The applicant also stated that the material terms of reinforced concrete and masonry walls are consistent with AMR items in the GALL-SLR report and that SRP Items 3.3-1, 060 and 3.3-1, 179 apply to concrete elements and masonry block walls, respectively, with a fire barrier intended function. Please confirm, consistent with the GALL-SLR Report, that the effects of aging for concrete concrete elements with a fire barrier intended function (except concrete concrete elements associated with the Containment Structure) are managed by both the Fire Protection and Structures Monitoring programs. For the concrete concrete elements with a fire barrier intended function associated with the Containment Structure, please confirm that the effects of aging are managed by both the Fire Protection and ASME Section XI, Subsection IWL programs. In addition, please confirm, consistent with the GALL-SLR Report, that the effects of aging for concrete block masonry block walls with a fire barrier intended function are managed by both the Fire Protection and Masonry Walls programs.

Dominion Response to RCI-3.3.1-1 This information has been confirmed to be correct as stated.

RCI-3.3.1-2 During the audit, the staff reviewed several documents that contain information which will likely be used in conclusions documented in the Safety Evaluation (SE). To the best of the staff's knowledge, this information is not on the docket. Any information used to reach a conclusion in the SER must be included on the docket by the applicant. We request that you submit confirmation that the information gathered from the documents and listed below is correct or provide the associated corrected information. AMR Item VII.G.A-19 (SRP Item 3.3-1, 057) in Volume 1 of NUREG-2191 manages hardening, loss of strength, and shrinkage due to elastomer degradation of elastomer fire barrier penetration seals exposed to air and condensation by the Fire Protection program. SLRA Table 3.5.2-14 cites AMR Item VII.G.A-19 (SRP Item 3.3-1, 057) for managing hardening, loss of strength, shrinkage of elastomer penetration seals and elastomer seismic gap filler material exposed to air by only the Fire Protection program. SLRA Table 3.5.2-14 includes elastomer, rubber and other similar materials penetration seals and elastomer, rubber and other similar Serial No.: 24-196A Docket No.: 50-395 Enclosure 4, Page 4 of 5 materials seismic gap filler material where only the Structures Monitoring program is credited to manage the effects of aging (loss of sealing). SLRA Table 3.5.2-14 appears to indicate that the elastomer and elastomer, rubber and other similar materials penetration seals and seismic gap filler material have the following intended functions: enclosure protection, fire barrier, flood barrier, and pressure boundary. During the audit of the Fire Protection AMP, it was discussed whether the elastomer and elastomer, rubber and other similar materials penetration seals and seismic gap filler material have all of the intended functions noted above. The applicant stated during the audit that the "elastomer and elastomer, rubber and other similar materials penetration seals and seismic gap filler material may have one or more of the cited intended functions, however, the Fire Protection program will manage the effects of aging for all elastomer and elastomer, rubber and other similar materials penetration seals and seismic gap filler material with a fire barrier intended function. In addition, if an elastomer and elastomer, rubber and other similar materials penetration seals and seismic gap filler material has any of the other cited intended functions, in addition to the fire barrier intended function, it would be managed by both the Fire Protection and Structures Monitoring programs. Please confirm that the Fire Protection program will manage the effects of aging for all elastomer and elastomer, rubber and other similar materials penetration seals and seismic gap filler material with a fire barrier intended function. In addition, please confirm that both the Fire Protection and Structures Monitoring program will manage the effects of aging for elastomer and elastomer, rubber and other similar materials penetration seals and seismic gap filler material with other intended functions (i.e., enclosure protection, flood barrier, and/or pressure boundary), in addition to the fire barrier intended function.

Dominion Response to RCI-3.3.1-2

This information has been confirmed to be correct as stated.

RCI-4.2.1-1 During the audit, the staff reviewed several documents that contain information which will likely be used in conclusions documented in the Safety Evaluation (SE). To the best of the staff's knowledge, this information is not on the docket. Any information used to reach a conclusion in the SER must be included on the docket by the applicant. We request that you submit confirmation that the information gathered from the documents and listed below is correct or provide the associated corrected information. In Section 4.2.1 of the SLRA the licensee states used of RAPTOR-M3G and the BUGLE 96 cross section library in accordance with the methodology described in WCAP-18124-NP-A, Fluence Determination with RAPTOR-M3G and FERRET and WCAP-18124-NP A Supplement 1-NP-A, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials. NRC approved the use of RAPTOR-M3G and FERRET for determination of reactor pressure vessel (RPV) fluence provided that the two limitations and conditions in the staff safety evaluation for the topical report WCAP 18124-NP-A are met. a) As part of the regulatory audit, the licensee provided information that WCAP 18124-NP-A Revision 0, Supplement 1-NP-A, Revision 0 provides the conditions necessary for meeting the Limitation 1 and allows for application of RAPTOR-M3G method to the RPV extended beltline region on a generic basis. The licensee stated that except for lower shell Serial No.: 24-196A Docket No.: 50-395 Enclosure 4, Page 5 of 5 to bottom head circumferential weld, these conditions are met for the plant specific neutron exposure measurements. To satisfy Limitation 1, please confirm that the lower shell to bottom head circumferential weld fast neutron (E > 1.0 MeV) fluence exposures for the SLRA for 72 EFPY, along with any increase in non calculated analytical uncertainty associated with these exposures, would not result in exposures greater than 1.0E+17 n/cm2. b) To meet the Limitation 2, please confirm that the least-squares analyses were not used to adjust any calculated RPV or surveillance capsule neutron exposure, and were only used as a supplemental check on the results of the dosimetry analyses.

Dominion Response RCI-4.2.1-1

This information has been confirmed to be correct as stated.

Serial No.: 24-196A Docket No.: 50-395

Enclosure 5

TOPICS THAT REQUIRE A SLRA SUPPLEMENT SUPPLEMENT 3

Dominion Energy South Carolina, Inc.

(Dominion Energy South Carolina, or DESC)

Virgil C. Summer Nuclear Station Unit Supplement 3 Serial No.: 24-196A VCS SLRA Enclosure 5 Page 2 of 3

The following two topics require the SLRA to be supplemented:

1. Reactor Vessel Internals: SLRA Table 3.1.2-2 Update

2. SLRA Table A4.0 Commitment #41 Clarification Supplement 3 Serial No.: 24-196A VCS SLRA Enclosure 5 Page 3 of 3

The following two topics require the SLRA to be supplemented:

1. Reactor Vessel Internals: SLRA Table 3.1.2-2 Update

As a result of industry operating experience, the Core barrel assembly (upper girth weld) component in the Reactor Vessel Internals system will be categorized in a different inspection category in the PWR Vessel Internals (B2.1.7) program. This component is being elevated from the Expansion component inspection category (3.1.1-053b) to the Primary component inspection category (3.1.1-053a). SLRA Table 3.1.2-2 is being updated to reflect this change.

Based on the above, the SLRA is supplemented as shown in Enclosure 2, to update the NUREG-2191 Item, Table 1 Item and Notes references, as shown in the following:

SLRA Table 3.1.2-2

2. SLRA Table A4.0 Commitment #41 Clarification

SLRA Table A4.0-1, Commitment #41, is being modified to include the precise wording from SLR-ISG-2021-04-ELECTRICAL to describe applicability to electrical insulation that is potentially exposed to significant moisture. SLRA Sections A1.41, A1.42, and B2.1.41 will be clarified similarly.

Based on the above, the SLRA is supplemented as shown in Enclosure 2, to ensure consistency with SLR-ISG-2021-04-ELECTRICAL regarding significant moisture, as shown in the following:

SLRA Section SLRA Table A1.41 A4.0-1 #41 A1.42 B2.1.41 Serial No.: 24-196A Docket No.: 50-395

Enclosure 6

SLRA MARK-UPS SUPPLEMENT 3

Dominion Energy South Carolina, Inc.

(Dominion Energy South Carolina, or DESC)

Virgil C. Summer Nuclear Station Unit

Enclosure 6 Serial No.: 24-196A Page 3 of 6 Virgil C. Summer Nuclear Station Application for Subseq uent License Renewal Supplement 3 Appendix A - FSAR Supplement

A1.40 Electrical Insulation For Inaccessible Medium-Voltage Power Cables Not Subject To 10 CFR 50.49 Environmental Qualification Requirements The Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements program is an existing condition monitoring program that manages the aging effect of reduce d electrical insulation resistance or degraded dielectric strength of non-EQ inaccessible medium-voltage power cables exposed to significant moisture.

The program applies to inaccessible or underground (e.g., installed in buried conduits, embedded raceway, duct banks, underground vaults, manholes, cable trenches, or direct buried installations) non-EQ medium-voltage power (operating voltage of 2kV to 35kV) cables within the scope of subsequent license renewal expose d to significant moisture. Sign ificant moisture is defined as exposure to moisture that lasts more than three days (i.e., long term wetting or submergence over a continuous period), that if left unmanaged, could potentially lead to a loss of intended function.

Periodic actions are taken to prevent non-EQ inaccessible medium-voltage power cables from being exposed to significant moisture. Acce ssible cable conduit endpoints and manholes associated with in-scope non-EQ inaccessible me dium-voltage power cables included in this program (i.e., installed in duct banks and manholes) are inspected for water accumulation, and the water is drained, as necessary.

A1.41 Electrical Insulation For Inaccessible Instrument And Control Cables Not Subject To 10 CFR 50.49 Environmental Qualification Requirements The Electrical Insulation for Inacce ssible Instrument and Control Ca bles Not Subject to 10 CFR 50.49 Environmental Q ualification Requirements program is a new condition monitoring program that will manage the aging effect of reduced electrical insulation resistance or degraded dielectric strength of non-EQ inaccessible and underground (e.g ., installed in buried conduit, embedded raceway, cable trenches, cable troughs, duct banks, vaults, manholes, or direct buried installations) instrument and control cables that are within the scope of subsequent license renewal and potentially exposed to significant moisture. Significant moisture is defined as exposure to moisture that lasts more than three days (i.e., long term wetting or submergence over a continuous period),

that if left unmanaged, could potentially lead to a loss of intended function.

Periodic actions are taken to prevent non-EQ inaccessible and underground instrument and control cables from being exposed to significant moisture. Accessible manholes/vaults associated with the cables included in this progra m are inspected for water accumulat ion and the water is drained, as necessary. The inspections are performed based on actual plant-specific operating experience over time with the inspection frequency being at least once per year and after event-driven occurrences such as heavy rain, rapid thawing of ice and snow, or flooding.

Page A-27

Enclosure 6 Serial No.: 24-196A Page 4 of 6 Virgil C. Summer Nuclear Station Application for Subseq uent License Renewal Supplement 3 Appendix A - FSAR Supplement

Industry and plant-specific operating experience will be evaluated in the development and implementation of this program.

A1.42 Electrical Insulation For Inaccessible Low-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements The Electrical Insulation for Inaccessible Low-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements program is a new condition monitoring program that will manage the aging effect of reduced electrical insu lation resistance or degr aded dielectric strength of non-EQ inaccessible and underground (e.g., installed in buried conduit, embedded raceway, cable trenches, cable troughs, duct banks, vault s, manholes, or direct buried) low-voltage power (operating voltage less than 2kV) cables that are within the scope of subsequent license renewal and potentially exposed to significant moisture. Significant moisture is defined as exposure to moisture that lasts more than three days (i.e., long term wetting or submergence over a continuous period), that if left unmanaged, could potentially lead to a loss of intended function.

Periodic actions are taken to prevent non-EQ inaccessible and underground low-voltage power cables from being exposed to significant moisture. Accessible manholes/vaults associated with the cables included in this progra m are inspected for water accumulat ion and the water is drained, as necessary. The inspections are performed based on actual plant-specific operating experience over time with the inspection frequency being at least once per year and after event-driven occurrences such as heavy rain, rapid thawing of ice and snow, or flooding.

Industry and plant-specific operating experience will be evaluated in the development and implementation of this program.

A1.43 Fuse Holders The Fuse Holders program is a new condition monitoring program that will manage increased electrical resistance of connection of the me tallic clamps and reduced electrical insulation resistance of the fuse holder electrical insulation material.

This program applies to fuse holders within t he scope of subsequent license renewal located outside of active equipment that require aging management.

Visual inspection and testing will be utilized to identify age related degradation for both fuse holder metallic clamps and fuse holder electrical insulation material.

Industry and plant-specific operating experience will be evaluated in the development and implementation of this program.

Page A-28

Enclosure 6 Serial No.: 24-196A Page 6 of 6 Virgil C. Summer Nuclear Station Application for Subseq uent License Renewal Supplement 3 Appendix B - Aging Management Programs

B2.1.41 Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program Description The Electrical Insulation for Inacce ssible Instrument and Control Ca bles Not Subject to 10 CFR 50.49 Environmental Q ualification Requirements program is a new condition monitoring program that will manage reduced electrical insulation resistance or degraded dielectric strength leading to electrical failure of in-scope non-EQ inaccessible and underground (e.g., installed in buried conduit, embedded raceway, cable trenche s, cable troughs, duct banks, va ults, manholes, or direct buried installations) instrument and control (I&C) cables that are within the scope of subsequent license renewal and potentially exposed to significant moisture, including cables designed for continuous wetting or submergence. Significant moisture is defined as exposure to moisture that lasts more than three days (i.e., long term wetting or submergence over a continuous period), that if left unmanaged, could potentially lead to a loss of intended function. Cable wetting or submergence resulting from event driven occurrences and mitigat ed by either automatic or passive drains is not considered significant moisture.

Periodic actions will be taken to prevent inaccessible and underground I&C cables from being exposed to significant moisture. Manholes associated with in-scope non-EQ inaccessible and underground I&C cables will be inspected for water collection , and the water will be drained, as necessary. Inspections will confi rm that cables are not wetted or submerged. The inspection and water removal will be performed based on actual plant-specific experience over time with inspection frequency being at least once pe r year and after event-driven o ccurrences (such as heavy rain, rapid thawing of ice and snow, or flooding). The first inspection will occur before the subsequent period of extended operation. Additional inspections will be conducted if one of the inspections does not meet the acceptance criterion due to current or projected degradation (i.e., trending). The number of increased inspections will be determined in accordance with the Corrective Action Program. Additional samples will be inspected for any recurring degradation to ensure corrective actions appropriately address the associated causes.

In-scope, non-EQ, inaccessible and underground I&C cables exposed to significant moisture will be evaluated to determine if testing is required. If required, initial testing will be performed on a sample population to determine the condition of the electrical insulation. One or more tests may be required due to the cable type, application, and electrical insulation to determine degradation of the electrical insulation. A one-time test prior to the subsequent period of extended operation will be performed for cable exposed to significant moisture if the cable insulation type is known to degrade with continuous exposure to moisture or if operating experience indicates insulation degradation resulting from continuous exposur e to moisture. Tests may include combinations of in-situ or laboratory, electrical, physical, or chemical tests. The need for additional peri odic tests and

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