ML25279A166

Annual Submittal of Technical Specification Bases Changes Pursuant to Technical Specification 5.5.13.d

ML25279A166
Person / Time
Site:	North Anna
Issue date:	10/06/2025
From:	Bugas T Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
25-239
Download: ML25279A166 (1)

Text

Dominion Energy Virginia North Anna Power Station 1022 Haley Drive, Mineral, VA 23117 Dominion Energy.com October 6, 2025 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Serial No.25-239 NAPS/RAP RO Docket Nos.

50-338 50-339 License Nos.

NPF-4 NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)

NORTH ANNA POWER STATION UNITS 1 AND 2 ANNUAL SUBMITTAL OF TECHNICAL SPECIFICATION BASES CHANGES PURSUANT TO TECHNICAL SPECIFICATION 5.5.13.d Pursuant to Technical Specification 5.5.13.d, "Technical Specifications (TS) Bases Control Program," Dominion Energy Virginia hereby submits changes to the Bases of the Technical Specifications implemented during the period of October 1, 2024, through September 30, 2025.

A summary of these TS Bases changes is provided in the Attachment to this letter.

Enclosed is a copy of the affected TS Bases pages for Revisions 75 and 76, for your information.

Bases changes to the Technical Specifications that were submitted to the NRC for information with their associated License Amendment Request (LAR) transmittals, submitted pursuant to 10 CFR 50.90, were reviewed, and approved with the LAR by the Facility Safety Review Committee. These TS Bases changes have been implemented with the respective approved License Amendments. A summary of these TS Bases changes associated with License Amendments is also provided in the Attachment to this letter.

If you have any questions regarding this submittal, please contact Mr. Matt Hayes at (540) 894-2100.

T. R. Bugas Director, Plant Support

Attachment:

Summary of TS Bases Changes Serial No.25-239 Docket Nos.: 50-338/339 Annual Summary of TS Bases Changes Page 2 of 2

Enclosure:

Affected TS Bases Pages for Revisions 75 and 76 Commitments made in this letter: None cc:

J. Lara - NRC Region II G. E. Miller - NRC Project Manager L. John Klos - NRC Project Manager, Surry NRC Senior Resident Inspector Mr. Chris Cosby (without Enclosure)

Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.

Suite 300 Glen Allen, Virginia 23060 State Health Commissioner (without Enclosure)

Virginia Department of Health James Madison Building - 7th Floor 109 Governor Street Room 730 Richmond, Virginia 23219

ATTACHMENT

SUMMARY

OF TS BASES CHANGES Serial No.24-239 Docket Nos.: 50-338/339 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)

NORTH ANNA POWER STATION UNITS 1 AND 2

Serial No.25-239 Docket Nos.: 50-338/339 TS BASES CHANGES ASSOCIATED WITH LICENSE AMENDMENTS None.

TS BASES CHANGES NOT ASSOCIATED WITH LICENSE AMENDMENTS TS Bases 3.1.7, Rod Position Indication The purpose of this TS Bases change was to an additional means for monitoring control rod position. The change explicitly states that the Rod Position Indication (RPI) system indication can be made via the control board rod position indicators or a continuously displayed Plant Computer System (PCS) indication. The axial position of control and shutdown rods is indicated by two separate and independent systems: the Bank Demand Position Indication System (commonly called group step counters) and the RPI system.

The Bank Demand Position Indication System counts pulses from the Rod Control System. There is one step counter for each group of 4 rods. Individual rods in a group all receive the same signal to move and should be at the same position. The Bank Demand Position Indication System is considered highly precise(+/- 1 step).

The RPI System provides an accurate indication of actual rod position, but at a lower precision than the step counters. The system is based on inductive analog signals from a series of coils spaced along a hollow tube. The analog signal is produced for each rod by a linear variable differential transformer. Direct continuous readout of each rod is provided by control board rod position indicators. The analog signal is also monitored by PCS. The RPI System is capable of monitoring rod position within at least +/- 12 steps.

Additionally, both the control board rod position indicators and PCS points are verified to be correct during the Rod Position Indication System Channel Calibration, 1-PT-25 and 2-PT-25.

The RPI Vendor Technical Manual (VTM) from Westinghouse also specifies that "The detector A.C. position signal enters the signal conditioning module through its TEST/OPERA TE switch. The position signal is converted into a D.C. analog position signal, buffered, and sent to the control board rod position indicator. A second isolated D.C. analog position signal is sent to the plant computer."

The PCS is an acceptable means of indicating actual rod position for TS 3.1.7. It's a reliable system that serves as the station's Emergency Response Facility Computer System, fulfilling the requirements of NUREG-0737, Supplement 1, and the guidance of NUREG-0696. It meets the overall 99% availability requirement in NUREG-0696, and it's already used to fulfill other TS requirements, such as calculating containment sump in-leakage. PCS receives an isolated analog signal that is produced by the RPI system for each rod and can provide continuous readout of control rod position.

TS Bases 3.3.1, Reactor Trip System (RTS) Instrumentation Serial No.25-239 Docket Nos.: 50-338/339 An error was discovered in the North Anna Power Station Technical Specification Bases B 3.3.1, Function 7 incorrectly states that the Reactor Trip System Overpower delta T function protects against the 1 % cladding strain limit. The overpower and overtemperature delta T trip functions in Westinghouse reactors are designed to provide primary protection against departure from nucleate boiling (DNB) (overtemperature delta T), and fuel centerline melt ( overpower delta T) through excessive linear heat generation rates (LHGR) during postulated transients.

Transient cladding strain analyses are performed to demonstrate that the design basis limit of the fission product barrier (i.e., the fuel cladding) is met during Condition II overpower transients, such as feedwater malfunction and loss of normal feedwater.

Restricting the allowable cladding strain during a Condition 11 transient event precludes fuel failure due to excessive strain. Strain is a measure of the deformation of the cladding relative to its original size, calculated as the ratio of the change in diameter to its original diameter.

The OP delta T reactor trip function and cladding strain criterion each protect different aspect of the reactor. The OP delta T reactor trip function ensures that operations remain within the fuel melt temperature design bases and is typically checked in house (using Dominion methods); while the cladding strain limit ensures fuel cladding integrity stays within the allowable expansion limit and is typically verified by Westinghouse.

TS Bases 3.7.6, Emergency Condensate Storage Tank (ECST)

The NAPS TS 3.7.6, Emergency Condensate Storage Tank (ECST), Bases are being revised to provide clarifying language with regards to the specific sizing requirements for the ECST, specific assumptions relevant to the sizing requirements for the ECST, and relevant UFSAR Chapter 15 analyses. Specifically, the Applicable Safety Analyses and Limiting Conditions for Operation (LCO) are being revised for clarity.

The current version of the NAPS TS 3.7.6, "ECST," Bases originates from the generic Westinghouse improved Standard Technical Specifications (STS). The Westinghouse improved STS reflects many plant designs (e.g., three/four loop plants, one/two containment sumps, dedicated versus common auxiliary feedwater (AFW) pump configurations, etc.). Initial revisions were made to the TS 3.7.6 Bases during the TS conversion project to reflect the North Anna plant design with respect to the ECST and, to a lesser extent, the AFW system.

While the TS 3.7.6 Bases were adequately revised to reflect the physical AFW and ECST configuration at North Anna, little to no consideration was given to the discussion relevant to the sizing basis for the ECST and the supporting analyses. As a result, statements are made in the current TS 3.7.6 Bases which are not reflective of correspondence with, and commitments made to the NRC.

ENCLOSURE Serial No.25-239 Docket Nos.: 50-338/339 AFFECTED TS BASES PAGES FOR REVISIONS 75 AND 76 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)

NORTH ANNA POWER STATION UNITS 1 AND 2

BASES BACKGROUND

( continued)

APPLICABLE SAFETY ANALYSES Rod Position Indication B3.1.7 The axial position of shutdown rods and control rods are determined by two separate and independent systems: the Bank Demand Position Indication System ( commonly called group step counters) and the Rod Position Indication (RPI) System.

The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or+/- 5/8 inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.

The RPI System provides a highly accurate indication of actual rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube. This signal is sent to the control board rod position indicators and the Plant Computer System. The RPI System is capable of monitoring rod position within at least+/- 12 steps.

Control and shutdown rod position accuracy is essential during power operation. Power peaking, ejected rod worth, or SDM limits may be violated in the event of a Design Basis Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected rod worth, and with minimum SDM (LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits"). The rod positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.4, "Rod Group Alignment Limits"). Control rod positions are continuously monitored to provide operators with information that ensures the unit is operating within the bounds of the accident analysis assumptions.

The control rod position indicator channels satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

North Anna Units 1 and 2 B 3.1.7-2 Revision 75

BASES LCO Rod Position Indication B 3.1.7 LCO 3.1. 7 specifies that the RPI System and the Bank Demand Position Indication System be OPERABLE for each rod. For the rod position indicators to be OPERABLE requires meeting the SR of the LCO and the following:

a. The RPI System indicates within 12 or 24 steps of the group step counter demand position as required by LCO 3.1.4, "Rod Group Alignment Limits". RPI System indication may be monitored on the control board rod position indicators or a continuously displayed Plant Computer System indication. In addition, rod position indication for a single rod can be determined by measuring control rod drive mechanism (CRDM) stationary gripper coil voltage using a temporary monitoring device that provides alarm capability; and

b. The Bank Demand Indication System has been calibrated either in the fully inserted position or to the RPI System.

The 12 step agreement limit between the Bank Demand Position Indication System and the RPI System indicates that the Bank Demand Position Indication System is adequately calibrated, and can be used for indication of the measurement of rod bank position.

A deviation of less than the allowable limit, given in LCO 3.1.4, in position indication for a single rod, ensures high confidence that the position uncertainty of the corresponding rod group is within the assumed values used in the analysis (that specified rod group insertion limits).

These requirements ensure that rod position indication during power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged.

OPERABILITY of the position indicator channels ensures that inoperable, misaligned, or mispositioned rods can be detected. Therefore, power peaking, ejected rod worth, and SDM can be controlled within acceptable limits.

APPLICABILITY The requirements on the RPI and step counters are only applicable in MODES 1 and 2 (consistent with LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6),

because these are the only MODES in which power is generated, and the OPERABILITY and alignment of rods have the potential to affect the safety of the unit. In the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System.

North Anna Units 1 and 2 B 3.1.7-3 Revision 75

BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

6.

Overtemperature ~T (continued)

RTS Instrumentation B 3.3.1 channels shared with other RTS Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions.

In MODE 1 or 2, the Overtemperature ~T trip must be OPERABLE to prevent DNB. In MODE 3, 4, 5, or 6, this trip Function does not have to be OPERABLE because the reactor is not operating and there is insufficient heat production to be concerned about DNB.

7.

Overpower ~T The Overpower ~T trip Function ensures that protection is provided to ensure the integrity of the fuel (i.e., no fuel pellet melting) under all possible overpower conditions. This trip Function also limits the required range of the Overtemperature ~T trip Function and provides a backup to the Power Range Neutron Flux-High Setpoint trip. The Overpower ~T trip Function ensures that the allowable heat generation rate (kW/ft) of the fuel is not exceeded. It uses the ~T of each loop as a measure of reactor power with a setpoint that is automatically varied with the following parameters:

• reactor coolant average temperature-the trip setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature; and rate of change of reactor coolant average temperature-including dynamic compensation for the delays between the core and the temperature measurement system. The function generated by the rate lag controller for Tavg dynamic compensation is represented by the expression: 13s/1+13s. The time

~onstant utilized in the rate lag controller for Tavg lS 13.

The Overpower ~T trip Function is calculated for each loop as per Note 2 of Table 3.3.1-1. Trip occurs if Overpower ~Tis indicated in two loops. Note that this Function also provides a signal to generate a turbine runback prior to reaching the Allowable Value. A turbine runback will reduce turbine power and reactor power.

(continued)

North Anna Units 1 and 2 B 3.3.1-18 Revision 76

ECST B 3.7.6 B 3.7 PLANT SYSTEMS B 3.7.6 Emergency Condensate Storage Tank (ECST)

BASES BACKGROUND The ECST provides a safety grade source of water to the steam generators for removing decay and sensible heat from the Reactor Coolant System (RCS). The ECST provides a passive flow of water, by gravity, to the Auxiliary Feedwater (AFW)

System (LCO 3.7.5). The steam produced is released to the atmosphere by the main steam safety valves (MSSVs) or the steam generator power operated relief valves (SG PORVs). The AFW pumps operate with a continuous recirculation to the ECST.

When the main steam trip valves are open, the preferred means of heat removal is to discharge steam to the condenser by the nonsafety grade path of the steam dump valves. The condensed steam is returned to the hotwell and is pumped to the 300,000 gallon condensate storage tank which can be aligned to gravity feed the ECST. This has the advantage of conserving condensate while minimizing releases to the environment.

Because the ECST is a principal component in removing residual heat from the RCS, it is designed to withstand earthquakes and other natural phenomena, including missiles that might be generated by natural phenomena. The ECST is designed to Seismic Category I to ensure availability of the feedwater supply. Feedwater is also available from alternate sources.

A description of the ECST is found in the UFSAR, Section 9.2.4 (Ref. 1).

APPLICABLE The ECST provides cooling water to remove decay heat and to SAFETY ANALYSES cool down the unit following all events in the accident analysis as discussed in the UFSAR, Chapters 6 and 15 (Refs. 2 and 3, respectively). Analyses have been performed for the limiting transients which define the Auxiliary Feedwater (AFW) system performance requirements. As the AFW system and the ECST are interrelated, the limiting transients for the AFW system apply to the ECST basis for OPERABILITY.

(continued)

North Anna Units 1 and 2 B 3.7.6-1 Revision 76

BASES ECST B 3.7.6 APPLICABLE Specifically, these include:

SAFETY ANALYSES (continued)

a. Rupture of a Main Feedwater Pipe

b. Rupture of a Main Steam Pipe Inside Containment

c. Plant Cooldown Each of the above transients is used in a specific manner for setting a portion of the AFW system performance requirements.

Rupture of a Main Feedwater Pipe This transient establishes requirements for alignment to preclude indefinite loss of AFW to the postulated break and establishes train association requirements for equipment such that the AFW system can deliver the minimum flow required within one minute following actuation assuming the worst single failure. This transient is NOT considered for minimum tank size requirements.

Rupture of a Main Steam Pipe Inside Containment This transient establishes the maximum allowable AFW flow rate to a single faulted steam generator (assuming all pumps are operating), establishes the basis for pump runout protection, and establishes alignment requirements so that flow requirements may be met considering the worst single failure. This transient is NOT considered for minimum tank size requirements.

Plant Cooldown This operation is analyzed to establish minimum tank size requirements. For anticipated operational occurrences and accidents that do not affect the OPERABILITY of the steam generators, the analysis assumption is holding the unit in MODE 3 for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, OR maintaining the unit in MODE 3 for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> followed by a 4-hour cooldown to RHR entry conditions within the limit of 100 °F/hour, assuming a coincident loss of offsite power. In addition to meeting both of these conditions, the ECST must retain sufficient water to ensure adequate net positive suction head for the AFW pumps, as well as account for any losses due to other constraints (continued)

North Anna Units 1 and 2 B 3.7.6-2 Revision 76

BASES APPLICABLE SAFETY ANALYSES (continued)

LCO ECST B 3.7.6 (e.g., ECST level measurement/transmitter uncertainty, adequate water level above suction centerline to prevent adverse air entrainment into the pump suction line due to formation of vortices, etc.).

The large feedwater line break event, while limiting for the condensate volume, is not considered in the determination of the minimum required volume for ECST OPERABILITY (i.e., ECST sizing), as it affects the OPERABILITY of a steam generator due to a severed feedwater line. The ECST sizing analysis assumptions require the steam generators to be OPERABLE.

As described in the Updated Final Safety Analysis Report (UFSAR), the ECST was not designed to cool down the plant by itself to a cold shutdown condition. Should the ECST volume be exhausted, plant cool down may be achieved by transferring to various backup sources. The backup sources include the 300,000 gallon Condensate Storage Tank (CST), the Fire Protection (FP) system, and the Service Water (SW) system.

The ECST satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

The ECST level required is equivalent to a contained volume of~ 110,000 gallons, which is based on (i) holding the unit in MODE 3 for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following a reactor trip from 100.37%

RTP, OR (ii) maintaining the unit in MODE 3 for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following a reactor trip from 100.37% RTP followed by a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> cooldown to RHR entry conditions within the limit of 100

°F/hour, assuming a coincident loss of offsite power in both scenarios. In addition to meeting both of these conditions, the ECST must retain sufficient water to ensure adequate net positive suction head for the AFW pumps, as well as account for any losses due to other constraints (e.g., ECST level measurement/transmitter uncertainty, adequate water level above suction centerline to prevent adverse air entrainment into the pump suction line due to formation of vortices, etc.).

The OPERABILITY of the ECST is determined by maintaining the tank level at or above the minimum required level to ensure the minimum volume of water.

North Anna Units 1 and 2 B 3.7.6-3 Revision 76

BASES APPLICABILITY ACTIONS ECST B 3.7.6 In MODES 1, 2, and 3, and in MODE 4, when steam generator is being relied upon for heat removal, the ECST is required to be OPERABLE.

In MODE 5 or 6, the ECST is not required because the AFW System is not required.

A.1 and A.2 If the ECST is not OPERABLE, the OPERABILITY of the backup supply, the Condensate Storage Tank, should be verified by administrative means within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter. OPERABILITY of the backup feedwater supply must include verification that the flow paths from the backup water supply to the AFW pumps are OPERABLE, and that the backup supply has the required volume of water available.

The ECST must be restored to OPERABLE status within 7 days, because the backup supply may be performing this function in addition to its normal functions. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time is reasonable, based on operating experience, to verify the OPERABILITY of the backup water supply. Additionally, verifying the backup water supply every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is adequate to ensure the backup water supply continues to be available.

The 7 day Completion Time is reasonable, based on an OPERABLE backup water supply being available, and the low probability of an event occurring during this time period requiring the ECST.

B.1 and B.2 If the ECST cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4, without reliance on the steam generator for heat removal, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

North Anna Units 1 and 2 B 3.7.6-4 Revision 76

BASES SURVEILLANCE REQUIREMENTS REFERENCES SR 3.7.6.1 ECST B 3.7.6 This SR verifies that the ECST contains the required volume of cooling water. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

1. UFSAR, Section 9.2.4.

2. UFSAR, Chapter 6.

3. UFSAR, Chapter 15.

North Anna Units 1 and 2 B 3.7.6-5 Revision 76