Text

Request for Approval of Fleet Report DOM-NAF-4-P/NP, Rev 0, and Proposed License Amendment Request Moderator Temperature Coefficient Analytical .
	ML25329A238
Person / Time
Site:	Millstone, Summer, Surry, North Anna
Issue date:	11/24/2025
From:	James Holloway; Dominion Energy Nuclear Connecticut, Dominion Energy Services, Dominion Energy South Carolina, Virginia Electric & Power Co (VEPCO)
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML25329A236	List: ML25329A238;
References
	25-016 DOM-NAF-4-NP, Rev 0
	Download: ML25329A238 (0)
	v • d • e

PROPRIETARY INFORMATION-WITHHOLD UNDER 10 CFR 2.390 Dominion Energy Services, Inc.

5000 Dominion Blvd.

Glen Allen, VA 23060 DominionEnergy.com contains information that is being withheld from public disclosure under 10 CFR 2.390. Upon separation from the Enclosure 1 this letter is decontrolled.

10 CFR 2.390 10 CFR 50.90 U. S. Nuclear Regulatory Commission Serial No.:

25-016 Attention: Document Control Desk NRA/DPJ:

R0 Washington, DC 20555-0001 Docket Nos.: 50-280/281 50-338/339 50-336/423 50-395 License Nos.: NPF-4/7 DPR-32/37 DPR-65 NPF-49 NPF-12 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)

SURRY POWER STATION UNITS 1 AND 2 NORTH ANNA POWER STATION UNITS 1 AND 2 DOMINION ENERGY NUCLEAR CONNECTICUT, INC. (DENC)

MILLSTONE POWER STATION UNITS 1 AND 2 DOMINION ENERGY SOUTH CAROLINA, INC. (DESC)

VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 REQUEST FOR APPROVAL OF FLEET REPORT DOM-NAF-4-P/NP, REV 0, AND PROPOSED LICENSE AMENDMENT REQUEST MODERATOR TEMPERATURE COEFFICIENT ANALYTICAL VERIFICATION FOR TECHNICAL SPECIFICATION SURVEILANCES Virginia Electric and Power Company (Dominion Energy Virginia), Dominion Energy Nuclear Connecticut, Inc. (DENC), and Dominion Energy South Carolina, Inc. (DESC)

[collectively, Dominion Energy], is submitting two requests. Specifically:

Dominion Energy is requesting the approval of generic application of fleet report DOM-NAF-4-P/NP, Revision 0 (DOM-NAF-4) (Enclosure 1 and Enclosure 3) to Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2, Millstone Power Station Units 2 and 3, and Virgil C. Summer Nuclear Station (VCSNS) Unit 1.

DESC is requesting the approval of a plant-specific change to the VCSNS Technical Specifications (TS), pursuant to 10 CFR 50.90, to allow for analytical

Serial No.25-016 Docket No. 50-280/281, 50-338/339 50-336/423, 50-395 Page 2 of 4 verification (AV) to satisfy Moderator Temperature Coefficient (MTC) surveillance requirements.

Associated with the request for U.S. Nuclear Regulatory Commission (NRC) approval of DOM-NAF-4, the following three documents are provided as Enclosure 1, Enclosure 2, and Enclosure 3, respectively:

1. Fleet Report DOM-NAF-4-P, Revision 0 (Proprietary)

2. Application for Withholding and Affidavit

3. Fleet Report DOM-NAF-4-NP, Revision 0 (Non-proprietary) contains information proprietary to Westinghouse Electric Company LLC (Westinghouse) and is supported by an affidavit (Enclosure 2) signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commissions regulations.

Accordingly, it is respectfully requested the proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390.

The license amendment request (LAR) proposes a change to the beginning-of-cycle (BOC) and end-of-cycle (EOC) MTC measurements via a revision to VCSNS TS SR 4.1.1.3 and the associated Bases, as well as TS 6.9.1.11, allowing for the use of AV as an alternate means of satisfying the required MTC TS surveillances. The fleet report provides a detailed explanation of the methodology used to perform AV to satisfy the MTC surveillances.

provides a description and assessment of the LAR. The marked-up pages for the TS, the TS Bases, and the Updated Final Safety Analysis Report (UFSAR) are provided in Attachments 1, 2, and 3, respectively. The Bases and UFSAR changes are provided for information only. Clean TS pages are included in Attachment 4.

The proposed amendment does not involve a Significant Hazards Consideration under the standards set forth in 10 CFR 50.92. The basis for this determination is included in. DESC has also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released offsite, or any significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR

Serial No.25-016 Docket No. 50-280/281, 50-338/339 50-336/423, 50-395 Page 3 of 4 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change.

The proposed amendment has been reviewed and approved by the station's Facility Safety Review Committee.

DESC requests approval of this license amendment request by October 31, 2026, with a 60-day implementation period to support EOC MTC surveillances at VCSNS.

In accordance with 10 CFR 50.91 (b), a copy of this license amendment request is being provided to the State of South Carolina.

If you have any questions or require additional information, please contact Mr. Daniel Johnson at (804) 273-2381.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on _____________.

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

Enclosures:

1. Fleet Report DOM-NAF-4-P, Revision 0 (Proprietary)

2. Application for Withholding and Affidavit

3. Fleet Report DOM-NAF-4-NP, Revision 0 (Non-proprietary)

4. Description and Assessment for Proposed License Amendment Attachments to Enclosure 4:

1. Marked-up Technical Specifications Pages

2. Marked-up Technical Specifications Bases Pages (For Information Only)

3. Marked-up UFSAR Pages (For Information Only)

4. Clean Technical Specification Pages Commitments made in this letter: None 5

Serial No.25-016 Docket No. 50-280/281, 50-338/339 50-336/423, 50-395 Page 4 of 4 cc:

U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, GA 30303-1257 Mr. G. Edward Miller NRC Senior Project Manager - Dominion Energy Nuclear Fleet U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738 Director, Radiation Protection Program Bureau of Land and Waste Management 2600 Bull Street Columbia, SC 29201 Mr. G. J. Lindamood Santee Cooper - Nuclear Coordinator 1 Riverwood Drive Moncks Corner, SC 29461 NRC Senior Resident Inspector V. C. Summer Nuclear Station

PROPRIETARY INFORMATION-WITHHOLD UNDER 10 CFR 2.390 Serial No.: 25-016 Docket Nos.: 50-280/281, 50-338/339 50-336/423, 50-395 Enclosure contains information that is being withheld from public disclosure under 10 CFR 2.390. Upon separation from the Enclosure this page is decontrolled.

Enclosure FLEET REPORT DOM-NAF-4-P, REVISION 0 (Proprietary)

Dominion Energy Services, Inc.

Virgil C. Summer Nuclear Station Unit 1 Millstone Power Station Units 2 and 3 Surry Power Station Units 1 and 2 North Anna Power Station Units 1 and 2

Serial No.: 25-016 Docket Nos.: 50-280/281, 50-338/339 50-336/423, 50-395 Enclosure APPLICATION FOR WITHHOLDING AND AFFIDAVIT Dominion Energy Services, Inc.

Virgil C. Summer Nuclear Station Unit 1 Millstone Power Station Units 2 and 3 Surry Power Station Units 1 and 2 North Anna Power Station Units 1 and 2

Serial No.: 25-016 Docket Nos.: 50-280/281, 50-338/339 50-336/423, 50-395

Enclosure FLEET REPORT DOM-NAF-4-NP, REVISION 0 (Non-proprietary)

Dominion Energy South Carolina, Inc.

Virgil C. Summer Nuclear Station Unit 1 Serial No.25-016 Docket No. 50-280/281, 50-338/339 50-336/423, 50-395

Serial No.25-016 Docket No. 50-280/281, 50-338/339 50-336/423, 50-395

Serial No.: 25-016 Docket No.: 50-395 DESCRIPTION AND ASSESSMENT Dominion Energy South Carolina, Inc.

Virgil C. Summer Nuclear Station Unit 1

Serial No.: 25-016 Docket No.: 50-395 Page 1 of 8 DESCRIPTION AND ASSESSMENT 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Dominion Energy South Carolina, Inc. (DESC) is submitting this license amendment request to revise the Technical Specifications (TS) for V.C.

Summer Nuclear Station (VCSNS) Unit 1. DESC is proposing changes to TS 4.1.1.3.a/b, Reactivity Control Systems - Surveillance Requirements, and TS 6.9.1.11, Core Operating Limits Report.

The TS changes support use of Analytical Verification (AV) as an alternate means of satisfying the required beginning-of-cycle (BOC) and end-of-cycle (EOC) Moderator Temperature Coefficient (MTC) surveillances.

Pursuant to 10 CFR 50.59, DESC is requesting approval for the generic use of fleet report DOM-NAF-4-P, Revision 0 (DOM-NAF-4) (Enclosure 1) as the methodology of the AV to satisfy the MTC surveillances.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation VCSNS Unit 1 follows the reload design methodology described in the NRC approved WCAP-9272-P-A (Ref. 5), which utilizes the Westinghouse Advanced Nodal Code (ANC) system of codes (WCAP-10965-P-A, Ref. 9) to perform neutronic calculations (including thermal hydraulic feedback) for the design, optimization, and confirmation that the reload core is bounded by the assumptions of the safety analyses.

Reactivity coefficients and parameters for each reload design are calculated on a core wide basis for a representative range of core conditions at the beginning, middle, and end of each cycle. These parameters include the MTC and the Isothermal Temperature Coefficient (ITC). Specifically, for a designed loading pattern, the MTC at operating conditions must meet the Core Operating Limits Report (COLR) limits. The limiting values for safety analyses are determined using the appropriate conditions on an event-by-event basis (for example, the most negative MTC is the limiting condition for a steamline break accident) and are calculated for the most conservative conditions. Predicted reload design parameters required for startup are also calculated. An example of this is that the critical boron concentration is measured during startup, which is combined with control rod position and temperature, and then compared to the predicted critical conditions that were determined using the neutronics code.

Serial No.: 25-016 Docket No.: 50-395 Page 2 of 8 2.2 Current Technical Specifications Requirements The MTC is defined as the change in reactivity per degree of change in the moderator temperature. There are limits for the MTC stated in the TS at different power levels and cycle burnup throughout core life. The most positive MTC limit applies to BOC, and the most negative MTC limit applies to EOC. These requirements are based on the most limiting accident scenarios in the safety analyses, and the limits are required to be confirmed via measurement per current TS Surveillance Requirement (SR) 4.1.1.3. This language is aligned with the generic language given in NUREG-0452 (Ref. 3). Note that for the EOC surveillance, TS SR 4.1.1.3.b has an asterisk that allows measurement of the MTC to be suspended, provided that the benchmark criteria in WCAP-13749-P-A (Ref. 10) and the revised prediction specified in the COLR are satisfied. The TS limits are currently confirmed by measurement of temperature coefficients and using predicted coefficient values from a neutronics code to calculate the measured MTC. This method of verification would remain valid and would in fact be necessary under certain conditions described in DOM-NAF-4.

2.3 Reason for the Proposed Change Currently, the TS requires measurement at BOC to verify the most positive MTC limit and near-EOC to verify the most negative MTC limit. The EOC test requires that plant systems be operated in a mode or condition not typical of steady state operation at Hot Full Power (HFP) conditions, as the plant is cooled down several degrees from its normal operating temperature by adjusting boron and turbine load while power is held at a constant level.

This non-routine plant manipulation increases the likelihood of creating an undesirable plant transient due to an equipment malfunction or human performance error. While the plant is more stable during the BOC measurement, this manipulation also poses similar risks as the EOC measurement. The proposed change to allow for AV to satisfy MTC TS surveillances will reduce this operational risk.

2.4 Description of Proposed Changes The proposed change would allow for the use of AV as an alternate means of satisfying the required MTC TS SRs. The MTC AV will be performed in accordance with DOM-NAF-

4. Implementing the AV methodology from DOM-NAF-4 will require changes to TS SR 4.1.1.3 and the associated Bases, as well as TS 6.9.1.11.

TS SR 4.1.1.3.a will be revised to read, The MTC shall be measured verified by measurement or analytical confirmation and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

Serial No.: 25-016 Docket No.: 50-395 Page 3 of 8 TS SR 4.1.1.3.b will be revised to read, The MTC shall be measured verified by measurement or analytical confirmation at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm*. In the event this comparison indicates the MTC is more negative than the 300 ppm surveillance limit specified in the COLR, the MTC shall be remeasured verified by measurement, and compared to the EOL MTC limit specified in the COLR, at least once per 14 EFPD during the remainder of the fuel cycle.

WCAP-13749-P-A (Ref. 10) is currently referenced in TS 6.9.1.11, as well as in an asterisk referenced in a footnote to TS SR 4.1.1.3.b. This proposed change will delete both references.

The methodology of DOM-NAF-4 will be used when performing AV for satisfying BOC or EOC surveillances. A change to Updated Final Safety Analysis Report (UFSAR) Section 4.3.2.3.2.1 will be implemented upon approval of the LAR. The UFSAR will be revised to state that DOM-NAF-4 is the approved method of evaluation for MTC analytical confirmation. Therefore, the need for prior NRC approval of any revisions to DOM-NAF-4 will be evaluated in accordance with the provisions of 10 CFR 50.59.

3.0 TECHNICAL EVALUATION

DOM-NAF-4 provides the technical justification for the proposed changes. The report describes a methodology for satisfying TS surveillance requirements for MTC limits through analytical verification. The methodology is referred to as AV and involves application of an appropriate uncertainty term to the most limiting value determined by a neutronics code (such as ANC) to demonstrate required limits are met throughout core life. Allowing for AV will prevent the need to perform low value plant manipulations (which increase the risk of equipment malfunction or human performance error) at BOC and EOC. The conditions for use of AV are described in DOM-NAF-4.

WCAP-9272-P-A (Ref. 5) includes recurrent validation of nuclear design analytical predictions through comparison with reload core measurement data. Using AV, this condition is still met through the verification of the initial critical condition through the Total Core Reactivity Check performed as part of startup physics testing, as well as the flux maps taken during initial power ascension and periodically throughout the cycle.

Additionally, boron letdown follow and reactivity balance checks are periodically performed throughout the cycle. These comparisons confirm that the core was manufactured and loaded per the design and implicitly incorporate thermal hydraulic feedback affects (including the effects of MTC) within the overall acceptance criteria.

Additionally, WCAP-9272-P-A (Ref. 5) establishes the core operating limits, which are documented in each cycle-specific COLR. Since DOM-NAF-4 is a surveillance method

Serial No.: 25-016 Docket No.: 50-395 Page 4 of 8 for confirming that the core is operating within the bounds of the safety analysis and does not establish a safety limit, DOM-NAF-4 is added as a reference within the stations UFSAR as opposed to the stations COLR, as established by Generic Letter (GL) 88-16.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria MTC is one of the controlling parameters for power and reactivity changes. The requirements of the TS ensure that the MTC remains within the bounds used in UFSAR Chapter 4. In turn, this ensures inherently stable power operations during normal operation and accident conditions. This TS requirement is derived from the following regulatory basis.

Regulation 10 CFR 50 has no specific requirements for MTC limits, but MTC limitations are included in the TS per 10 CFR 50.36. Regulation 10 CFR 50.36(c), Paragraph (3),

states that surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. The proposed change allows for a new surveillance method to be performed if specific conditions (described in DOM-NAF-4) are met and is in compliance with 10 CFR 50.36(c)(3).

10 CFR 50 Appendix A has two relevant General Design Criteria (GDC). Note that the measurement of MTC was not mentioned in 10 CFR 50.

Criterion 11Reactor inherent protection. The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

Criterion 28Reactivity limits. The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.

NUREG-0800 Standard Review Plan, Chapter 4.3 (Ref. 4), Acceptance Criteria 2 further discusses the requirements for the temperature coefficients.

Serial No.: 25-016 Docket No.: 50-395 Page 5 of 8 The only directly applicable GDC in the area of reactivity coefficients is GDC 11, which states...the net effect of the prompt inherent nuclear feedback characteristics tend to compensate for a rapid increase in reactivity, and is considered to be satisfied in light water reactors (LWRs) by the existence of the Doppler and negative power coefficients. There are no criteria that explicitly establish acceptable ranges of coefficient values or preclude the acceptability of a positive moderator temperature coefficient (MTC) such as may exist in PWRs at beginning of core life.

The acceptability of the coefficients in a particular case is determined in the reviews of the analyses in which they are used, e.g., control requirement analyses, stability analyses, and transient and accident analyses. The use of spatial effects such as weighting approximations as appropriate for individual transients are included in the analysis reviews. The judgement to be made under this SRP section is whether the reactivity coefficients have been assigned suitably conservative values by the applicant. The basis for that judgment includes the use to be made of a coefficient, i.e., the analyses in which it is important; the state of the art for calculation of the coefficient; the uncertainty associated with such calculations, experimental checks of the coefficient in operating reactors; and any required checks of the coefficient in the startup program of the reactor under review.

GDC 11 and NUREG-0800 emphasize the importance of MTC to nuclear safety and discuss the judgement process for determining the MTC limit values. NUREG-0800 states that the basis of this judgement should, in part, use the uncertainty associated with such calculations, experimental checks of the coefficient in operating reactors Implementing this proposed change does not alter how the MTC limits are determined or how the safety analyses are performed. The limits continue to be determined in accordance with WCAP-9272-P-A (Ref. 5), and safety analyses assumptions continue to account for appropriate uncertainties. DOM-NAF-4 (Enclosure 1) does not establish the MTC limits assumed in the safety analyses. The methodology is only a surveillance method that confirms the core is operating within the assumptions of the safety analyses through AV, which accounts for appropriate code and surveillance uncertainties.

NUREG-0800 also says that the basis of this judgement should, in part, use any required checks of the coefficient in the startup program of the reactor. This element continues to be met because any cycle that has plant modifications that may significantly alter nuclear, thermal, or hydraulic performance will verify the MTC by measurement. Therefore, the AV process complies with GDC 11 and NUREG-0800.

GDC 28 stated that the limits must be met for the different plausible accident scenarios, which is why there is both a most positive BOC and most negative EOC limit. As previously stated, this proposed change does not alter how the MTC limits are determined

Serial No.: 25-016 Docket No.: 50-395 Page 6 of 8 in accordance with WCAP-9272-P-A (Ref. 5), and the methodology described in DOM-NAF-4 (Enclosure 1) is only a surveillance method for verifying the MTC remains within the required limits. Therefore, the AV process will comply GDC 28.

4.2 No Significant Hazards Consideration Pursuant to 10 CFR 50.90, Dominion Energy South Carolina, Inc. (DESC) is submitting this license amendment request to revise the Technical Specifications (TS) for V.C.

Summer Nuclear Station (VCSNS) Unit 1. DESC is proposing changes to TS 4.1.1.3.a/b, Reactivity Control Systems - Surveillance Requirements, and TS 6.9.1.11, Core Operating Limits Report.

The TS changes support use of Analytical Verification (AV) as an alternate means of satisfying the required beginning-of-cycle (BOC) and end-of-cycle (EOC) Moderator Temperature Coefficient (MTC) surveillances.

DESC has performed the significant hazards consideration for the proposed changes by addressing the three standards set forth in 10 CFR 50.92, Issuance of Amendment, as discussed below:

1.

Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change allows for the use of AV based on the methodology of DOM-NAF-4 as an alternate means of satisfying the required MTC TS surveillance requirements. AV is a conservative surveillance process that demonstrates the core is operating with the required TS MTC limits and does not change how the limits are determined. The MTC AV involves application of an appropriate uncertainty term to the most limiting value determined by a neutronics code, such as ANC, to demonstrate required limits are met throughout core life. The use of MTC AV does not involve a significant increase in the probability or consequence of an accident previously evaluated. The AV process does not change or impact any accident event initiating conditions.

2.

Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change involves satisfying TS requirements using analytical verification instead of making a plant manipulation to facilitate measurement.

Serial No.: 25-016 Docket No.: 50-395 Page 7 of 8 There is no change to any plant equipment, including how equipment is operated and maintained. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Specific criteria must be met for each cycle in order for AV to be used, otherwise the TS surveillances will be performed via measurement. The current reload process, which is not being changed, includes recurrent validation of nuclear design analytical predictions through comparison with reload core measurement data. These comparisons confirm that the core was manufactured and loaded per the design and implicitly incorporate thermal hydraulic affects (including the effects of the MTC) within the overall acceptance criteria.

For the EOC surveillance using AV, operational correction factors are applied to make the revised value more limiting based on how the plant has operated over the course of the cycle. These corrections are based on core leakage, core burnup, and hot full power axial flux difference. Predicted MTC values for each cycle are based on the original model, and do not include adjustments through a core reactivity adjustment process that could gain margin to the EOC MTC limit.

Therefore, all the margins of safety are maintained, and the proposed change does not involve a significant reduction in a margin of safety.

Based on the above information, DESC concludes that the proposed change involves no significant hazards consideration under the criteria set forth in 10 CFR 50.92(c) and, accordingly, a finding of no significant hazards consideration is justified.

4.3 Conclusion Based on the considerations discussed above, there is reasonable assurance that (1) the health and safety of the public will not be endangered by the proposed change, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of a requested license amendment to implement the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Serial No.: 25-016 Docket No.: 50-395 Page 8 of 8

5.0 ENVIRONMENTAL CONSIDERATION

S DESC has reviewed the proposed change for environmental considerations. The proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

6.0 REFERENCES

1. Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, last updated August 2007.

2. Fleet report DOM-NAF-4-P and NP, Rev. 0, Minor Rev. 0, Moderator Temperature Coefficient Analytical Verification, July 2025.

3. NUREG-0452, Rev. 4, Standard Technical Specifications for Westinghouse Pressurized Water Reactors, 1981.

4. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 4.3, Rev. 3, Nuclear Design, March 2007.

5. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985.

6. Code of Federal Regulations, Title 10, Part 50, Section 36, Technical specifications, last updated January 2025.

7. Code of Federal Regulations, Title 10, Part 50, Section 92, Issuance of Amendment, last updated August 2007.

8. Code of Federal Regulations, Title 10, Part 51, Section 22, Criterion for Categorical Exclusion; Identification of Licensing and Regulatory Actions Eligible for Categorical Exclusion or Otherwise Not Requiring Environmental Review, last updated October 2020.

9. WCAP-10965-P-A, ANC: A Westinghouse Advanced Nodal Code, September 1986.

10. WCAP-13749-P-A, Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement, March 1997.

Attachment MARKED-UP TECHNICAL SPECIFICATIONS PAGES Dominion Energy South Carolina, Inc.

Virgil C. Summer Nuclear Station Unit 1 Serial No.: 25-016 Docket No.: 50-395 Serial No.: 25-016 Docket No.: 50-395 Page 1 of 3

SUMMER - UNIT 1 6-16 Amendment No. 79, 88, 104, ADMINISTRATIVE CONTROLS 6.9.1.9 Not used.

6.9.1.10 Not used.

CORE OPERATING LIMITS REPORT 6.9.1.11 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

a.

Moderator Temperature Coefficient BOL and EOL Limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, b.

Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, c.

Control Rod Insertion Limits for Specification 3/4.1.3.6, d.

Axial Flux Difference Limits, target band, and APLND for Specification 3/4.2.1, e.

Heat Flux Hot Channel Factor,

, K(z), W(z), APLND, W(z)BL, and FQ(z) manufacturing/measurement uncertainties for Specification 3/4.2.2, H

RTP H

H N

f.

Nuclear Enthalpy Rise Hot Channel Factor, F

, Power Factor Multiplier, PF and F measurement uncertainties limits for Specification 3/4.2.3.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a.

WCAP-9272-P-A, WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY, July 1985 (W Proprietary).

(Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Rod Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -

RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor.)

b.

WCAP-10216-P-A, Rev. 1A, RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION, February 1994 (W Proprietary).

(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (FQ Methodology for W(z) surveillance requirements).)

Serial No.: 25-016 Docket No.: 50-395 Page of 3 Serial No.: 25-016 Docket No.: 50-395 Page of 3

Attachment MARKED-UP TECHNICAL SPECIFICATIONS BASES PAGES Dominion Energy South Carolina, Inc.

Virgil C. Summer Nuclear Station Unit 1 Serial No.: 25-016 Docket No.: 50-395 Serial No.: 25-016 Docket No.: 50-395 Page 1 of 2

SUMMER - UNIT 1 B 3/4 1-2 Amendment No. 52, 75, 88, BRN 05-001 REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting End of Cycle Life (EOL) MTC value. The 300 ppm surveillance limit MTC value represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting EOL MTC value.

The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551 F. This limitation is required to ensure

1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 4) the reactor pressure vessel is above its minimum RTNDT temperature.

3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, and 5) an emergency power supply from OPERABLE diesel generators.

Those valves that can stop or throttle flow coming from or going to its intended location are flowpath valves. These include the valves that provide for the minimum required pump flow capability. Diversion valves that can pass significant flow and render the function inoperable are also to be considered flowpath valves.

Valves that pass minimum diversion flow, such as vents, drains, instrument root valves, and sample valves are not flowpath valves. Flow diversion through such valves would be generally self evident and readily detectable allowing for prompt corrective action.

Serial No.: 25-016 Docket No.: 50-395 Page of 2

Attachment MARKED-UP UFSAR PAGES Dominion Energy South Carolina, Inc.

Virgil C. Summer Nuclear Station Unit 1 Serial No.: 25-016 Docket No.: 50-395

VC SUMMER FSAR 4.3-22 variation of fuel temperature is taken into account by calculating the effective fuel temperature as a function of power density as discussed in Section 4.3.3.1.

Typical Doppler temperature coefficients are shown in Figure 4.3-31 as a function of the effective fuel temperature (at BOL and EOL conditions). The effective fuel temperature is lower than the volume averaged fuel temperature since the neutron flux distribution is non-uniform through the pellet and gives preferential weight to the surface temperature. An example of Doppler-only contribution to the power coefficient, defined later, is shown in Figure 4.3-32 as a function of relative core power. The integral of the differential curve on Figure 4.3-32 is an example of the Doppler contribution to the power defect and is shown in Figure 4.3-33 as a function of relative power. The Doppler coefficient becomes more negative as a function of life as the Pu-240 content increases, thus increasing the Pu-240 resonance absorption but less negative as the fuel temperature changes with burnup as described inSection 4.3.3.1. The upper and lower limits of Doppler coefficient used in accident analyses are given in Chapter 15.0.

4.3.2.3.2 Moderator Coefficients The moderator coefficient is a measure of the change in reactivity due to a change in specific coolant parameters such as density, temperature, pressure or void. The coefficients obtained are moderator density, temperature, pressure, and void coefficients.

4.3.2.3.2.1 Moderator Density and Temperature Coefficients The moderator temperature (density) coefficient is defined as the change in reactivity per degree change in the moderator temperature. Generally, the effect of the changes in moderator density as well as the temperature are considered together. A decrease in moderator density means less moderation which results in a negative moderator coefficient. An increase in coolant temperature, keeping the density constant, leads to a hardened neutron spectrum and results in an increase in resonance absorption in U-238, Pu-240 and other isotopes. The hardened spectrum also causes a decrease in the fission-to-capture ratio in U-235 and Pu-239. Both of these effects make the moderator coefficient more negative. Since water density changes more rapidly with temperature as temperature increases, the moderator temperature (density) coefficient becomes more negative with increasing temperature.

The soluble boron used in the reactor as a means of reactivity control also has an effect on moderator density coefficient since the soluble boron absorber density, as well as, the water density is decreased when the coolant temperature rises. A decrease in the soluble absorber concentration introduces a positive component in the moderator coefficient.

Thus, if the concentration of soluble absorber is large enough, the net value of the coefficient may be positive. With the burnable absorber rods present, however, the initial hot boron concentration is sufficiently low that the moderator temperature coefficient is negative at full power operating conditions. The effect of control rods is to make the moderator coefficient more negative by reducing the required soluble boron concentration and by increasing the ³leakage' of the core.

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Revision 24--06/17/24 VC SUMMER FSAR 4.3-23 With burnup, the moderator coefficient becomes more negative primarily as a result of boric acid dilution but also to a significant extent from the effects of the buildup of plutonium and fission products.

The moderator coefficient is calculated for the various plant conditions discussed above by performing 2-group two or three dimensional calculations, varying the moderator temperature (and density) by about +/- 5 °F about each of the mean temperatures. The moderator coefficient is shown as a function of core temperature and boron concentration for the unrodded and rodded core in Figures 4.3-34 through 4.3-36. The temperature range covered is from cold (68 °F) to about 600 °F. The contribution due to Doppler coefficient (because of change in moderator temperature) has been subtracted from these results. Figure 4.3-37 shows the hot, full power moderator temperature coefficient plotted as a function of cycle lifetime at the critical boron concentration for a transition cycle and a typical equilibrium reload cycle.

The moderator coefficients presented here are calculated on a corewise basis, since they are used to describe the core behavior in normal and accident situations when the moderator temperature changes can be considered to affect the whole core.

4.3.2.3.2.2 Moderator Pressure Coefficient The moderator pressure coefficient relates the change in moderator density, resulting from a reactor coolant pressure change, to the corresponding effect on neutron production. This coefficient is much less significant in comparison with the moderator temperature coefficient. A change of 50 psi in pressure has approximately the same effect on reactivity as a half-degree change in moderator temperature. This coefficient can be determined from the moderator temperature coefficient by relating change in pressure to the corresponding change in density. The moderator pressure coefficient is negative over a portion of the moderator temperature range at BOL (-0.004 pcm/psi, BOL) but is always positive at operating conditions and becomes more positive during life (+0.5 pcm/psi, EOL).

4.3.2.3.2.3 Moderator Void Coefficient The moderator void coefficient relates the change in neutron multiplication to the presence of voids in the moderator. In a PWR this coefficient is not very significant because of the low void content in the coolant. The core void content is less than 1/2 of 1% and is due to local or statistical boiling. The void coefficient varies from 50 pcm/percent void at BOL, and at low temperatures to

-250 pcm/percent void at EOL and at operating temperatures. The negative void coefficient at operating temperature becomes more negative with fuel burnup.

4.3.2.3.3 Power Coefficient The combined effect of moderator temperature and fuel temperature change as the core power level changes is called the total power coefficient and is expressed in terms of reactivity change per percent power change. The power coefficient at BOL and EOL conditions for a typical reload cycle is given in Figure 4.3-38.

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Attachment Dominion Energy South Carolina, Inc.

Virgil C. Summer Nuclear Station Unit 1 Serial No.25-016 Docket No. 50-395

SUMMER - UNIT 1 3/4 1-5 Amendment No. 75, 88, 169 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a.

The MTC shall be verified by measurement or analytical confirmation and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

b.

The MTC shall be verified by measurement or analytical confirmation at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than the 300 ppm surveillance limit specified in the COLR, the MTC shall be verified by measurement, and compared to the EOL MTC limit specified the COLR, at least once per 14 EFPD during the remainder of the fuel cycle.

Serial No.25-016 Docket No. 50-395 Page 1 of 3

SUMMER - UNIT 1 6-16 Amendment No. 79, 88, 104, 117, 142, 175, 182 ADMINISTRATIVE CONTROLS 6.9.1.9 Not used.

6.9.1.10 Not used.

CORE OPERATING LIMITS REPORT 6.9.1.11 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

a.

Moderator Temperature Coefficient BOL and EOL Limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, b.

Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, c.

Control Rod Insertion Limits for Specification 3/4.1.3.6, d.

Axial Flux Difference Limits, target band, and APLND for Specification 3/4.2.1, e.

Heat Flux Hot Channel Factor, RTP Q

F

, K(z), W(z), APLND, W(z)BL, and FQ(z) manufacturing/measurement uncertainties for Specification 3/4.2.2, f.

Nuclear Enthalpy Rise Hot Channel Factor, F H RTP, Power Factor Multiplier, PF H,

and F H N measurement uncertainties limits for Specification 3/4.2.3.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a.

WCAP-9272-P-A, WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY, July 1985 (W Proprietary).

(Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Rod Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -

RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor.)

b.

WCAP-10216-P-A, Rev. 1A, RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION, February 1994 (W Proprietary).

(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (FQ Methodology for W(z) surveillance requirements).)

Serial No.25-016 Docket No. 50-395 Page of 3

SUMMER - UNIT 1 6-16a Amendment No. 219 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) c.

WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applies to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology),

November 2016, (Westinghouse Proprietary) d.

WCAP-12472-P-A, BEACON CORE MONITORING AND OPERATIONS SUPPORT SYSTEM, August 1994, (W Proprietary).

(Methodology for Specifications 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 -

RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.4 -

Quadrant Power Tilt Ratio.)

e.

WCAP-12610-P-A, VANTAGE + Fuel Assembly Reference Core Report, April 1995 (W Proprietary). WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLOTM, July 2006 (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements there to shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

Serial No.25-016 Docket No. 50-395 Page of 3

v t e Letter Sequence Request
EPID:L-2025-LLA-0178, () EPID:L-2025-LLT-0000, ()
Initiation Acceptance Administration Questions Answers Results Approval... Other:
MONTHYEAR2025-12-19 [Table View]

ML25329A238

Text

Enclosures:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

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