ML25317A687

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License Amendment Request Propose Revision to Generator Water Level in Technical Specification Surveillance Requirements and Limiting Condition for Operation
ML25317A687
Person / Time
Site: Summer 
(NPF-012)
Issue date: 11/09/2025
From: James Holloway
Dominion Energy South Carolina
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
25-253
Download: ML25317A687 (1)
v • d • e


Text

Serial No.25-253 NRA/CL:

R0 Docket No.

50-395 License No.

NPF-12 LICENSE AMENDMENT REQUEST PROPOSED REVISION TO STEAM GENERATOR WATER LEVEL IN TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS AND LIMITING CONDITION FOR OPERATION Pursuant to 10 CFR 50.90, Dominion Energy South Carolina, Inc. (DESC) is hereby submitting a License Amendment Request (LAR) for Virgil C. Summer Nuclear Station (VCSNS) Unit 1.

The proposed amendment would revise the VCSNS Technical Specification (TS)

Surveillance Requirements (SR) 4.4.1.2.3, SR 4.4.1.3.2, and Limiting Condition for Operation (LCO) 3.4.1.4.1. These TS pertain to the steam generator (SG) water level required to be maintained for Modes 3, 4, and 5. The proposed change will replace the non-conservative 10% Wide Range (WR) requirement with a 25% Narrow Range (NR) requirement. provides DESCs description and assessment of the proposed change. provides the marked-up VCSNS TS pages to reflect the proposed amendment. Attachment 3 provides the revised (clean) VCSNS TS pages. Attachment 4 provides the associated TS Bases change, for information only.

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 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 stations Facility Safety Review Committee.

Dominion Energy South Carolina, Inc.

5000 Dominion Blvd.

Glen Allen, VA 23060 DominionEnergy.com U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 DOMINION ENERGY SOUTH CAROLINA, INC.

VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 November 9, 2025

~

Dominion

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Energy

Serial No.25-253 Docket No. 50-395 Page 2 of 3 James E. Holloway Vice President - Nuclear Engineering & Fleet Support Attachments:

1. Description and Assessment of Proposed Change
2. Marked-up Technical Specification Pages
3. Revised (Clean) Technical Specification Pages
4. Marked-up Technical Specification Bases Page (For Information Only)

Commitments made in this letter: None DESC requests approval of this License Amendment Request by November 30, 2026, with a 30-day implementation period.

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 Cailyn Ludwig at (804) 273-5131.

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

Executed on ___________.

Respectfully, November 9, 2025 1

1

Serial No.25-253 Docket No. 50-395 Page 3 of 3 cc:

U.S. Nuclear Regulatory Commission Region II Mr. G. Edward Miller NRC Senior Project Manager U.S. Nuclear Regulatory Commission NRC Senior Resident Inspector V. C. Summer Nuclear Station Director, Radiation Protection Program Bureau of Land and Waste Management Mr. G. J. Lindamood Santee Cooper - Nuclear Coordinator

Serial No.25-253 Docket No. 50-395 DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGE Dominion Energy South Carolina, Inc.

Virgil C. Summer Nuclear Station Unit 1

Serial No.25-253 Docket No. 50-395, Page 1 of 10 TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION.................................................................................. 2 2.0 DETAILED DESCRIPTION.................................................................................. 2 2.1 SYSTEM DESIGN AND OPERATION......................................................................... 2 2.2 CURRENT TECHNICAL SPECIFICATION REQUIREMENTS........................................... 3 2.3 REASON FOR THE PROPOSED CHANGES................................................................ 3

2.4 DESCRIPTION

OF PROPOSED CHANGES................................................................. 4

3.0 TECHNICAL EVALUATION

................................................................................. 5 3.1 TECHNICAL JUSTIFICATION................................................................................... 5

4.0 REGULATORY EVALUATION

............................................................................ 6 4.1 APPLICABLE REGULATORY REQUIREMENTS AND CRITERIA...................................... 6 4.2 PRECEDENTS...................................................................................................... 7 4.3 NO SIGNIFICANT HAZARDS CONSIDERATION.......................................................... 8

4.4 CONCLUSION

...................................................................................................... 9

5.0 ENVIRONMENTAL CONSIDERATION

............................................................. 10

6.0 REFERENCES

................................................................................................... 10

Serial No.25-253 Docket No. 50-395, Page 2 of 10 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Dominion Energy South Carolina, Inc. (DESC) requests an amendment to Renewed Facility Operating License NPF-12 for Virgil C. Summer Nuclear Station (VCSNS) Unit 1. The proposed change would revise the VCSNS Operating License to change the Steam Generator (SG) level requirement for Technical Specification (TS) Surveillance Requirements (SR) 4.4.1.2.3 and SR 4.4.1.3.2 and Limiting Condition for Operation (LCO) 3.4.1.4.1. The change ensures adequate SG secondary side inventory is available to maintain a viable heat sink during Modes 3, 4, and 5. A SG level of 25% Narrow Range (NR), which immerses the SG tubes completely, is proposed for use in the applicable TS.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The function of the SGs is to transfer heat produced by the fission process in the reactor core to the secondary side of the plant. To assure that the SGs are able to perform this function, the water level on the secondary side of the SG is to be maintained above the tops of the SG tubes whenever the SG is required to be operable. In Modes 3, 4, and 5, the operability of the SGs is tied to SG level. SG water level preserves the SG heat sink for removal of long-term residual heat. This is of particular importance during Modes 3, 4, and 5 to mitigate the consequences of any unanticipated power producing transients that could occur during shutdown.

VCSNS has three Westinghouse D-75 SGs to remove heat. Water level in the SGs is indicated by two scales: Wide Range and Narrow Range. The Wide Range scale spans most of the height of a SG, while the Narrow Range scale spans a tighter window starting near the vicinity of the top of the SG tubes and ending near the 100% Wide Range elevation.

Wide Range level provides a broad indication of SG water inventory, ensuring there is enough water for heat removal during abnormal or shutdown conditions. It is primarily used to confirm the SG can serve as an effective heat sink for decay heat during extended shutdown or loss of Residual Heat Removal (RHR) system scenarios.

Narrow Range level is used for more precise control of SG water level during normal operations and transients, focusing on the operational band where the SG is actively transferring heat. The Narrow Range ensures the tubes are covered for effective heat transfer and helps prevent issues like tube uncovering or overfill.

Serial No.25-253 Docket No. 50-395, Page 3 of 10 2.2 Current Technical Specification Requirements VCSNS Unit 1 TS LCO 3.4.1.2 is applicable to Mode 3 and requires at least two of the Reactor Coolant loops be operable and at least one in operation. TS SR 4.4.1.2.3 considers a SG operable when the secondary side water level is greater than or equal to 10% of wide range indication. TS Bases 3/4.1.1 indicates a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require two loops be operable.

TS LCO 3.4.1.3 is applicable to Mode 4 and requires at least two of the Reactor Coolant and/or residual heat removal (RHR) loops be operable and at least one in operation. TS SR 4.4.1.3.2 considers a SG operable when the secondary side water level is greater than or equal to 10% of wide range indication. TS Bases 3/4.1.1 indicates a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require at least two loops be operable.

TS LCO 3.4.1.4.1 is applicable to Mode 5 with reactor coolant loops filled and requires that at least RHR loop shall be operable and in operation, and either: (a) one additional RHR loop shall be operable, or (b) the secondary side water level of at least two steam generators shall be greater than 10 percent of wide range indication. TS Bases 3/4.1.1 indicates a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat in Mode with reactor coolant loops filled; however, single failure considerations require at least two loops be operable.

2.3 Reason for the Proposed Changes A non-conservatism was identified in the VCSNS TS surveillances related to the required SG level, which would allow a SG to be operable and viable as a heat sink in Modes 3, 4, and 5. The current VCSNS TS Modes 3 and 4 SRs and Mode 5 LCO 10% Wide Range (WR) minimum covers approximately five feet of the SG tubes.

The proposed 25% Narrow Range (NR) minimum immerses the tubes completely (approximately 30 feet above the tube sheet). The effectiveness of the SGs as heat sinks is dependent on how much liquid immerses the SG tubes (more wetted, submerged, heat transfer area). Tube coverage promotes more efficient heat transfer and increases margin to dryout. These are ensured by maintaining level in the NR scale.

The Technical Specification Bases for the above SRs and LCO do not provide the rationale for the existing 10% WR value. A review of NUREG-0452 [Reference 1],

the Technical Specification standard to which VCSNS was originally licensed, shows a SG level of 17% in SRs 4.4.1.2.2 (Mode 3) and 4.4.1.3.2 (Mode 4) and in LCO 3.4.1.4.1 (Mode 5). However, neither the Wide Range nor Narrow Range scale was

Serial No.25-253 Docket No. 50-395, Page 4 of 10 specified. As such, this issue appears to be an oversight from the original licensing of VCSNS and a non-conservative legacy error.

The 10% WR minimum has been recognized as non-conservative and administrative controls have been in place to ensure sufficient heat transfer capability during Modes 3, 4, and 5. The proposed change would allow the administrative controls to be eliminated and ensure that the affected TS are fully qualified.

2.4 Description of Proposed Changes The proposed changes to VCSNS Unit 1 SR 4.4.1.2.3, SR 4.4.1.3.2, and LCO 3.4.1.4.1 will replace 10% wide range with 25% narrow range for the SG level indication. The proposed changes are bolded and underlined below. The proposed deletions are strikethrough:

VCSNS Unit 1 TS SR 4.4.1.2.3 for Mode 3:

The required steam generator(s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 10% of wide range 25 percent of Narrow Range indication in accordance with the Surveillance Frequency Control Program.

VCSNS Unit 1 TS SR 4.4.1.3.2 for Mode 4:

The required steam generator(s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 10% of wide range 25 percent of Narrow Range indication in accordance with the Surveillance Frequency Control Program.

VCSNS Unit 1 TS LCO 3.4.1.4.1 for Mode 5:

At least one residual heat removal (RHR) loop shall be OPERABLE and in operation*, and either:

a. One additional RHR loop shall be OPERABLE#, or
b. The secondary side water level of at least two steam generators shall be greater than 10 percent of wide range 25 percent of Narrow Range indication.

Markups of the proposed TS changes are provided in Attachment 2 and a revised version of the TS is provided in Attachment 3. The corresponding TS Bases change is provided in Attachment 4, for information only. The update to the TS Bases clarifies the steam generator serves as a heat sink in Modes 3, 4, and 5.

Serial No.25-253 Docket No. 50-395, Page 5 of 10

3.0 TECHNICAL EVALUATION

3.1 Technical Justification The immersion of the SG tubes provides the most effective heat transfer from the primary to the secondary. To establish adequate heat sink, at least half of the SGs must be filled. The SG level should also be maintained above the top of the tubes to gain the maximum effectiveness of the SG, especially in reflux cooling mode.

The proposed changes to TS SR 4.4.1.2.3, SR 4.4.1.3.2, and LCO 3.4.1.4.1 are based on heat transfer evaluations. The proposed 25% NR minimum SG level ensures adequate SG secondary side inventory is available to maintain a viable heat sink to transport core heat during Modes 3, 4, and 5. This was proven through calculations of decay heat, heat transfer coefficients based on SG performance, and the feedwater requirements for decay heat removal at these Modes. With adequate liquid, the heat sink function of the steam generators would be preserved during lower Modes when the RCPs may be secured or if a loss of Residual Heat Removal system operation occurs.

The calculational inputs for SG heat removal capability include decay heat loads, RCS mass flow rates, SG and RCS temperatures, and feedwater flow rates. The methodology applies standard heat transfer and fluid flow equations, with all results validated against as-built plant limits. The acceptance criteria for the calculations are based on providing adequate cooling for Mode 3, 4, and 5 operation. Specifically, the SG level that covers the tubes in each operating configuration must allow for adequate cooling, which is defined as removal of all decay heat as well as any heat input from operating RCPs. The conclusions for each Mode are as follows:

For Mode 3 (TS 3.4.1.2), one reactor coolant loop in operation where the SG level covers the tube at 25% narrow range provided sufficient cooling.

The SG was able to take on the decay heat load and the heat input from the RCP with no operating parameters exceeding their as-built limits.

For Mode 4 (TS 3.4.1.3), the limiting condition is the operation of one RHR loop or one reactor coolant loop with associated SG and reactor coolant pump. With the use of one SG where the level covers the tubes at 25%

narrow range, there was sufficient cooling. The SG was able to take on the decay heat load and the heat input from the RCP with no operating parameters exceeding their as-built limits.

For Mode 5 (TS 3.4.1.4.1), the limiting condition is one RHR loop in operation with two SGs having a level covering the tubes in natural

Serial No.25-253 Docket No. 50-395, Page 6 of 10 circulation. However, to ensure consistency with the other Technical Specifications, as well as to improve defense in depth in the unlikely event of a RHR, a level of greater than 25% narrow range and the low RCS temperature (200°F) in Mode 5 would provide adequate time to restore the RHR.

Additionally, the Reactor Protection System and Engineering Safeguards Action System, which are part of the Stations Limiting System Safety Settings, remain unchanged by the proposed 25% NR SG water level.

While procedural measures are already in place to assure a viable heat sink, TS SR 4.4.1.2.3, SR 4.4.1.3.2, and LCO 3.4.1.4 require correction with respect to the minimum SG level required. The proposed change supports the SGs ability to mitigate the consequences of any unanticipated power producing transients that could occur during shutdown conditions. With the correction, the non-conservative TS value is addressed.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. Conformance with the General Design Criteria (GDC) is described in Section 3.1.2 of the VCSNS FSAR [Reference 2].

GDC Criterion 13 - Instrumentation and Control Plant instrumentation and control systems are provided to monitor significant variables. Steam generator water inventory on secondary side is maintained.

GDC Criterion 14 - Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

GDC Criterion 15 - Reactor Coolant System Design The Reactor Coolant System and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

Serial No.25-253 Docket No. 50-395, Page 7 of 10 GDC Criterion 34 - Residual Heat Removal A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core.

GDC Criterion 44 - Cooling Water A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided.

10 CFR 50.36(c)(2)

The TSs must include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. That provision also requires that when an LCO of a nuclear reactor is not met, the licensee must shut down the reactor or follow any remedial action permitted by the TSs until the condition is met.

10 CFR 50.36(c)(3)

The TSs include SRs, which 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 LCOs will be met.

These criteria were reviewed and were found to not be adversely affected by the proposed change.

4.2 Precedents

This amendment request is similar in nature to the Watts Bar Nuclear Plant Technical Specification change for SG secondary side narrow range level.

  • Watts Bar Nuclear Plant, Unit 1 - Issuance of Amendment Regarding the Change in Steam Generator Narrow Range Level Requirements to Accommodate the Replacement Steam Generators at Watts Bar Nuclear Plant (ADAMS Accession No. ML060960075)
  • Watts Bar Nuclear Plant, Unit 2 - Issuance of Amendment No. 57 to Revise Technical Specifications to Change the Steam Generator Secondary Side Water Level (ADAMS Accession No. ML21260A210)

Serial No.25-253 Docket No. 50-395, Page 8 of 10 4.3 No Significant Hazards Consideration Dominion Energy South Carolina (DESC) proposes to amend Virgil C. Summer Nuclear Station (VCSNS) Unit 1 Technical Specification Surveillance Requirements (SR) 4.4.1.2.3, SR 4.4.1.3.2, and Limiting Condition for Operation (LCO) 3.4.1.4.1.

The Technical Specification SRs and LCO provide verification of steam generator level which demonstrates the steam generator is operable and able to act as a heat sink. The amendment will increase the minimum steam generator (SG) level in Modes 3, 4, and 5.

DESC has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment, as discussed below:

1.

Does the proposed amendment involve a signi"cant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change involves a revision to Technical Specification Surveillance Requirement (SR) 4.4.1.2.3, SR 4.4.1.3.2, and Limiting Condition for Operation (LCO) 3.4.1.4.1. The proposed change to the SRs and LCO involves an increase in minimum steam generator level during Modes 3, 4, and 5, increasing the wetted surface of the steam generator tubes. There are no accidents previously evaluated that initiate from these Modes where steam generator level is a key initial condition such that radiological releases would result. The minimum water level ensures the steam generators remain a viable heat sink in these Modes. As such, the activity does not alter the probability of accident initiation nor the mitigation of its consequences.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

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

Response: No.

The proposed change involves a revision to Technical Specification SR 4.4.1.2.3, SR 4.4.1.3.2, and LCO 3.4.1.4.1. The proposed change only corrects the minimum steam generator level Technical Specification. The change does not involve any credible new failure mechanisms, malfunctions, or accident

Serial No.25-253 Docket No. 50-395, Page 9 of 10 initiators not previously considered. There are not any physical modifications to the plant, and there are not any changes to the way the plant is being operated.

Therefore, the proposed change does not create the possibility of a new or dierent kind of accident from any previously evaluated.

3.

Does the proposed change involve a signi"cant reduction in a margin of safety?

Response: No.

The proposed change involves a revision to Technical Specification SR 4.4.1.2.3, SR 4.4.1.3.2, and LCO 3.4.1.4.1. The acceptance criterion is to provide an indicated level that will ensure sufficient heat removal capability for removing decay heat. The minimum steam generator level will ensure the steam generator tubes are covered so the steam generator can serve as an effective heat sink for decay heat and continues to meet the acceptance criterion. The proposed change supports the SGs ability to mitigate the consequences of any unanticipated power producing transients that could occur during shutdown.

Therefore, the proposed changes do not involve a signi"cant reduction in a margin of safety.

Based on the above, DESC concludes that the proposed amendment does not involve a signi"cant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a "nding of no signi"cant hazards consideration is justi"ed.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Serial No.25-253 Docket No. 50-395, Page 10 of 10

5.0 ENVIRONMENTAL CONSIDERATION

Dominion Energy South Carolina has reviewed the proposed license amendment for environmental considerations in accordance with 10 CFR 51.22. The proposed license amendment does not involve: (i) a signi"cant hazards consideration, (ii) a signi"cant change in the types or signi"cant increase in the amounts of any euent that may be released osite, or (iii) a signi"cant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion 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 amendment.

6.0 REFERENCES

1.

NUREG-0452, Revision 4,

Standard Technical Specifications for Westinghouse Pressurized Water Reactors (Revision Issued Fall 1981).

2.

V. C. Summer Final Safety Analysis Report, Revision 25.

Serial No.25-253 Docket No. 50-395 MARKED-UP TECHNICAL SPECIFICATION PAGES Dominion Energy South Carolina, Inc.

Virgil C. Summer Nuclear Station Unit 1

Serial No.25-253 Docket No. 50-395, Page 1 of 4 REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the Reactor Coolant loops !listed below shall be OPERABLE and at least one of these Reactor Coolant loops shall be in operation.'

a.

Reactor Coolant Loop A and its associated steam generator and Reactor Coolant

Pump,
b.

Reactor Coolant Loop B and its associated steam generator and Reactor Coolant

Pump,
c.

Reactor Coolant Loop C and its associated steam generator and Reactor Coolant

Pump, APPLICABILITY:

MODE 3 ACTIONS:

a. Wllh less than the above required Reactor Coolant loops OPERABLE, restore tlhe required loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b. Wllh no Reactor Coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the reqUJired coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required Reactor Coolant pumps, if not in operation, shall be determined OPERABLE in accordance with the surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one cooling loops shall be verified to operation and circulating reac.tor coolant in acoordance with the Surveillance Frequency Control Program.

4.4.1.2.3 The required steam generator(s) shall be determined OPERABLE by verifying secondary side water level to be greater than or eqUJal to 1 ~% ol 110~0,aAge indication in accordance with the Surveillance Frequency Control Program.

~ ~

-1REPLACE w/:

25% of Narrow Range

-:.A=u -::;R~e~ac~t~or:--:c"'oo

= 1a~n~t-::p~u~m~p~s-::m:--:iJ:,J= be= d:-:e~-e:--:n~e".:'.rg~iz~ed

= 1~or=up=10~1:-::-ho~u'.'.:r"'.:p".:'.ro~v:;-:id:-:

no opera,ons are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 1 O'F below saturation temperature.

SUMMER - UNIT 1 3/4 4-2 Amendment No. 222

Serial No.25-253 Docket No. 50-395, Page 2 of 4 REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the Reactor Coolant ancVor residual heat removal (RHR) loops listed below shall be OPERABLE and at least one of these React()( Coolant and/or RHR loops shall be in operation.-

a.

Reactor Coolant Loop A and its associated steam generator and Reactor Coolant Pump;

b.

Reactor Coolant Loop Band its associated steam generator and Reactor Coolant Pump;

c.

Reactor Coolant Loop C and its associated steam generator and Reactor Coolant Pump;

d.

Residual Heat Removal Loop A,

e.

Residual Heat Removal Loop B, APPLICABILITY:

MODE 4 ACTIONS:

NO CHANGES.

Included for completeness only.

a.

With less than the above required React()( Coolant and/or RHR loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is an RIHR loop, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

With no Reactor Coolant loop or RHR loop in ope<ation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

'A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 300°F unless 1) the pressurizer water volume is less than 1288 cubic feet ancVor 2) the secondary water temperature of each steam generator is less than 50°F above each of the RCS cold leg temperatures.

-All Reactor Coolant pumps and decay heat removal pumps may be de-energized f()( up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided 1) no operation are pennitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) C()(e outlet temperature is maintained at least 10°F below saturation temperature.

SUMMER - UNIT 1 3/4 4-3

Serial No.25-253 Docket No. 50-395, Page 3 of 4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required Reactor Coolant pump(s), if not in operation, shall be determined OPERABLE in accordance willl the Suiveillance f requency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.3.2 The required steam generator(s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 19% el \YiSe FaR~e indication in accordance with the Suiveillance Frequency Control Program.

~ - --'REPLACE wl:

4.4.1.3.3 At least one Reactor Coolant or RHR loop shall be verified to be in 25% of Narrow Range circulatimg reactor coolant in accordance with the Suiveillance Frequency Control Program.

I 4.4.1.3.4 Verify RHR loop locations susceptible to gas accumulation are sufficiently filled with water in accoraance with the Surveillance Frrequency Control Program.*

  • Not r,equired to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> afler entering MODE 4.

SUMMER - UNIT 1 3/4 4-4 Amendment No. i!S-4; 222

Serial No.25-253 Docket No. 50-395, Page 4 of 4 REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4. 1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation*, and either:

a.

One additional RHR loop shall be OPERABLEi, or

b.

The secondary side water le~~eas than 1Q persant 9f "'iEla fiiR~ ndication.

__ _,REPLACE wJ:

25% of Narrow Range APPLICABILITY:

MODE 5 with Reactor Coolant loops filted##.

ACTION:

a.

Wrth less than the above required loops OPERABLE and/or with less than the required steam generator level, immediately initiate corrective action to return the required loops to OPERABLE status or to restore the required level as soon as possible.

b.

Wrth no residual heat removal loop in operation, suspend all operations invotving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required residual heat removal loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits in accordance with the Surveillance Frequency Control Program.

4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

4.4.1.4.1.3 Verify RHR loop locations susceptible to gas accumulation are sufficiently fined with water in accordance with the Surveillance Frequency Control Program.

One residual heat removal loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RHR loop is OPERABLE and in operation.

A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 300°F unless 1) the pressurizer water volume is less than 1288 cubic feet and/or 2) the secondary water temperature of each steam generator is less than S<rF above each of the Reactor Coolant System cold leg temperatures.

The RHR pump may be de--e-nergized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 1ocF below saturation temperature.

SUMMER - UNIT 1 3/4 4-5 Amendment ~

!Q4, 222

Serial No.25-253 Docket No. 50-395 REVISED (CLEAN) TECHNICAL SPECIFICATION PAGES Dominion Energy South Carolina, Inc.

Virgil C. Summer Nuclear Station Unit 1

Serial No.25-253 Docket No. 50-395, Page 1 of 3 Revised Technical Specification Page REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the Reactor Coolant loops listed below shall be OPERABLE and at least one of these Reactor Coolant loops shall be in operation.*

a.

Reactor Coolant Loop A and its associated steam generator and Reactor Coolant

pump,
b.

Reactor Coolant Loop Band its associated steam generator and Reactor Coolant

pump,
c.

Reactor Coolant Loop C and rts associated steam generator and Reactor Coolant

pump, APPLICABILITY:

MODE 3 ACTIONS:

a.

Wrth less than the above required Reactor Coolant loops OPERABLE, restore the required loops to OPERABLE status wrthin 72 hoors or be in HOT SHUTDOWN wrthin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

Wrth no Reactor Coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required Reactor Coolant pumps, if not in operation, shall be determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one cooling loop shall be verified to operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

4.4.1.2.3 The required steam generator(s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 25% of Narrow Range indication in accordance with the Surveillance Frequency Control Program.

'All Reactor Coolant pumps may be de-energized for up to *1 hour provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 10°F below saturation temperature.

SUMMER - UNIT 1 3/4 4-2 Amendment No. :lJ:!.

Serial No.25-253 Docket No. 50-395, Page 2 of 3 Revised Technical Specification Page REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The reqlired Reactor Coolant pump(s), d llOI i'1 operaoon, shall be deteimined OPERABLE in a,ccordance witll the Sll'Veil!ance Frequency Control Program by verifying correct breaker alignments and indicated pcM'er availabiliy.

4.4.1.3.2 The reqlired steam generator(s) shall be determined OPERABLE by verifying secondary side water level to be g eater than or equal to 25% of Narrow Range indication in accordance with the Surveillance Frequency Control Program.

4.4.1.3.3 At least one Reactor Coolant or RHR loop shall be verified to be in operation and circulating reactor coolant in accordance with tile surveillance Frequency Control Program.

4.4.1.3.4 Verily RHR loop locations susceptible to gas accll'llulation are sufficienl!y filled v.'ith water in acoordanoe with the SUiveillance Frequency Control Prograrn.*

  • Not required fo be perfOITlled until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 4.

SUMMER - LtJIT 1 314 4-4 Amendment No. 404, ~

Serial No.25-253 Docket No. 50-395, Page 3 of 3 Revised Technical Specification Page REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation*, and either:

a.

One additional RHR loop shall be OPERABLE", or

b.

The secondary side water level of at least two steam generators shall be greater than 25% of Narrow Range indication.

APPLICABILITY:

MODE 5 wrth Reactor Coolant loops filled"".

ACTION:

a.

With less than the above requira<l loops OPERABLE and/or with less than the required steam generator level, immediately initiate corrective action to return the required loops to OPERABLE status or to restore the required level as soon as possible.

b.

With no residual heat removal loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required residual heat removal loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limrts in accordance with the Surveillance Frequency Control Program.

4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant in accordance with the SurvEillance Frequency Control Program.

4.4.1.4.1.3 Verify RHR loop locations susceptible to gas accumulation are sufficiently filled with water in accordance with the Surveillance Frequency Control Program.

One residual heat removal loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RHR loop is OPERABLE and in operation.

A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 300°F unless 1) the pressurizer water volume is less than 1288 cubic feet an(IJor 2) the secondary water temperature of each steam generator is less than 50°F above each of the Reactor Coolant System cold leg temperatures.

The RHR pump may be de-energized tor up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10°F below saturation temperature.

SUMMER - UNIT 1 3/4 4-5 Amendment 1Q, ~04, J~

Serial No.25-253 Docket No. 50-395 MARKED-UP TECHNICAL SPECIFICATION BASES PAGE (FOR INFORMATION ONLY)

Dominion Energy South Carolina, Inc.

Virgil C. Summer Nuclear Station

Serial No.25-253 Docket No. 50-395, Page 1 of 1 FOR INFORMATION ONLY 3.4.4 REAC I OR COOLANT SVSTE BASES 3/4.4_1 REACTOR COOLANT LOOPS A D COOLANT CIRCULATION The plant is designed o opera e wi h all reactor coolan loops in operation, and maint -n BR in the core at or abo e the design lim-during all normal operations and an -cipa ed transients. In ODES 1 and 2 -

one reactor coo nt loop not in opera ion this specification requires that the plant be in at least HOT STANDBY

  • in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In ODE 3, a single reactor coolant loop pro - es sufficien hea removal capab-lity for remo ing decay heat; ho ever, s*ng failure con -dera *ons require that two loops be OPERAB E.

In ODE 4, and in ODE 5

anagement of gas oids is importan to RHR Sys em OPERABILITY.

In ODE 5 with reactor coolant loops not lled, a s-ng RHR loop pro - es suffic-nt heat remo al capability for remo *ng decay hea bu s*ngle failure con -dera *ons, and the una ailability of he steam generators as a heat remo

  • g componen, require tha at ast o RHR loops be OPERABLE.

anagement o gas oids is important o RHR S stem OPERABILITY_

- ---1 11~SERT:

The ope adequate flow t 5, a steam generat changes during se<:andary side I reactivity chang boron reduction e

capability of operator recognition and control.

The restrictions on starting a Reactor Coolan Pump with one or more RCS cold legs ss than or equ to 300°F are pro ided o prevent RCS pressure transients, caused by energy additions from the secondary sys em, ich could exceed the limits of Appendix G to 1 0 CFR Part SO. The RCS

-11 be protected again overpressure transients and -u no exceed the limits of Appendix G by either (1) restric - g the water volume in the pressurizer and reby pro "ding a olume or the primary coolant to e pand into, or (2) by restric *ng starting of the RCPs o hen the secondary wa er temperature of each s earn genera or is less than S0°F above each of the RCS cold leg temperatures.

SU ER-U IT 1 B 3/4 4-1 Amendment o. 83, 204

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