ML26051A065

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Issuance of Amendment No. 294 Revisions to TS Related to Control Building Isolation, Accident Monitoring Instrumentation, Accumulator and RWST Boron Concentration Limits, and Secondary Containment SR
ML26051A065
Person / Time
Site: Millstone 
(NPF-049)
Issue date: 03/12/2026
From: Edwards T
NRC/NRR/DORL/LPL1
To: Carr E
Dominion Energy Nuclear Connecticut
Edwards T, NRR/DORL/LPL1
References
EPID L-2025-LLA-0053
Download: ML26051A065 (0)
v • d • e


Text

March 12, 2026 Mr. Eric S. Carr Chief Nuclear Officer and President Nuclear Operations & Contracted Energy Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION, UNIT NO. 3 ISSUANCE OF AMENDMENT NO. 294 RE: REVISIONS TO TECHNICAL SPECIFICATIONS RELATED TO CONTROL BUILDING ISOLATION, ACCIDENT MONITORING INSTRUMENTATION, ACCUMULATOR AND REFUELING WATER STORAGE TANK BORON CONCENTRATION LIMITS, AND SECONDARY CONTAINMENT SURVEILLANCE REQUIREMENTS (EPID L-2025-LLA-0053)

Dear Mr. Carr:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 294 to Renewed Facility Operating License No. NPF-49 for the Millstone Power Station, Unit No. 3. This amendment is in response to your application dated March 20, 2025, as supplemented by the letter dated August 14, 2025.

The amendment would provide the following changes:

revise applicable Actions in Technical Specification (TS) Table 3.3-3, Functional Unit 7.a, Manual Initiation, and Functional Unit 7.c, Automatic Actuation Logic and Actuation Relays, for Control Building Isolation (CBI), when one or both instrumentation channel(s) are inoperable, revise applicable Actions a. and b. for TS 3.3.3.6, Accident Monitoring Instrumentation, when one or both Accident Monitoring Instrumentation channel(s) are inoperable, revise the Limiting Condition for Operation (LCO) requirements for TS 3.5.1, Accumulators, and TS 3.5.4, Refueling Water Storage Tank, to relocate boron concentration limit values to the Core Operating Limit Report, and revise MPS3 TS Surveillance Requirement (SR) 4.6.6.2.2 under TS 3.6.6.2, Secondary Containment.

A copy of the related Safety Evaluation is also enclosed. The Notice of issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Theo Edwards, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423

Enclosures:

1. Amendment No. 294 to NPF-49
2. Safety Evaluation cc: Listserv DOMINION ENERGY NUCLEAR CONNECTICUT, INC., ET AL.1 DOCKET NO. 50-423 MILLSTONE POWER STATION, UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 294 Renewed License No. NPF-49
1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the applicant dated March 20, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

1 Dominion Energy Nuclear Connecticut, Inc., et al. (the licensees) consists of Dominion Energy Nuclear Connecticut, Inc.; Green Mountain Power Corporation; and Massachusetts Municipal Wholesale Electric Company. Dominion Energy Nuclear Connecticut, Inc. is authorized to act as the agency and representative of Green Mountain Power Corporation and Massachusetts Municipal Wholesale Electric Company and has exclusive responsibility and control over the physical operation and maintenance of the facility.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-49 is hereby amended to read as follows:

2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 294, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license. DENC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Undine Shoop, Acting Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 12, 2026 UNDINE SHOOP Digitally signed by UNDINE SHOOP Date: 2026.03.12 15:22:13 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 294 MILLSTONE POWER STATION, UNIT NO. 3 RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert 4

4 Replace the following pages of the Appendix A Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 3/4 3-21 3/4 3-22 3/4 3-23 3/4 3-24 3/4 3-24a 3/4 3-59 3/4 3-59a 3/4 5-1 3/4 5-2 3/4 5-9 3/4 5-10 3/4 6-21 3/4 6-22 6-19a 6-20 6-20a 6-20b 3/4 3-21 3/4 3-22 3/4 3-23 3/4 3-24 3/4 3-24a 3/4 3-59 3/4 3-59a 3/4 5-1 3/4 5-2 3/4 5-9 3/4 5-10 3/4 6-21 3/4 6-22 6-19a 6-20 6-20a 6-20b Renewed License No. NPF-49 Amendment No. 270-293, 294 (2)

Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 294 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DENC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

DENC shall not take any action that would cause Dominion Energy, Inc. or its parent companies to void, cancel, or diminish DENCs Commitment to have sufficient funds available to fund an extended plant shutdown as represented in the application for approval of the transfer of the licenses for MPS Unit No. 3 (4)

Immediately after the transfer of interests in MPS Unit No. 3 to DNC*, the amount in the decommissioning trust fund for MPS Unit No. 3 must, with respect to the interest in MPS Unit No. 3, that DNC* would then hold, be at a level no less than the formula amount under 10 CFR 50.75.

(5)

The decommissioning trust agreement for MPS Unit No. 3 at the time the transfer of the unit to DNC* is effected and thereafter is subject to the following:

(a)

The decommissioning trust agreement must be in a form acceptable to the NRC (b)

With respect to the decommissioning trust fund, investments in the securities or other obligations of Dominion Energy, Inc. or its affiliates or subsidiaries, successors, or assigns are prohibited.

Except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.

(c)

The decommissioning trust agreement for MPS Unit No. 3 must provide that no disbursements or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee until the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC.

(d)

The decommissioning trust agreement must provide that the agreement cannot be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

  • On May 12, 2017, the name Dominion Nuclear Connecticut, Inc. changed to Dominion Energy Nuclear Connecticut, Inc.

MILLSTONE - UNIT 3 TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT TOTAL NO.

OF CHANNELS CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES ACTION 6.

Auxiliary Feedwater a.

Manual Initiation 2

1 2

1, 2, 3 23 b.

Automatic Actuation Logic and Actuation Relays 2

1 2

1, 2, 3 22 c.

Stm. Gen. Water Level--

Low-Low

1) Start Motor-Driven Pumps 4/stm. gen.

2/stm. gen. in any operating stm.

gen.

3/stm. gen. in each operating stm. gen.

1, 2, 3 20A

2) Start Turbine-Driven Pump 4/stm. gen.

2/stm. gen. in any 2 operating stm.

gen.

3/stm. gen. in each operating stm. gen.

1, 2, 3 20A d.

Safety Injection Start Motor-Driven Pumps See Item 1. above for all Safety Injection initiating functions and requirements.

e.

Loss-of-Offsite Power Start Motor-Driven Pumps 2

1 2

1, 2, 3 19 Amendment No. 57, 266, 294 November 30, 2015 3/4 3-21

MILLSTONE - UNIT 3 TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT TOTAL NO.

OF CHANNELS CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES ACTION 6.

Auxiliary Feedwater (Continued) f.

Containment Depres-surization Actuation (CDA) Start Motor-Driven Pumps See Item 2. above for all CDA functions and requirements.

7.

Control Building Isolation

a. Manual Actuation 2

1 2

14 b.

Manual Safety Injection Actuation 2

1 2

1, 2, 3, 4 19 c.

Automatic Actuation Logic and Actuation Relays 2

1 2

1, 2, 3, 4 18 d.

Containment Pressure--

High-1 3

2 2

1, 2, 3 16 e.

Control Building Inlet Ventilation Radiation 2/intake 1

2/intake 18 8.

Loss of Power

a. 4 kV Bus Under-voltage-Loss of Voltage 4/bus 2/bus 3/bus 1, 2, 3, 4 27 b.

4 kV Bus Undervoltage-Grid Degraded Voltage 4/bus 2/bus 3/bus 1, 2, 3, 4 27 Amendment No. 14, 57, 203, 220, 294 3/4 3-22

MILLSTONE - UNIT 3 TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT TOTAL NO.

OF CHANNELS CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES ACTION 9.

Engineering Safety Features Actuation System Interlocks a.

Pressurizer Pressure, P-11 3

2 2

1, 2, 3 21 b.

Low-Low Tavg, P-12 4

2 3

1, 2, 3 21 c.

Reactor Trip, P-4 2

2 2

1, 2, 3 23 10.

Emergency Generator Load Sequencer 2

1 2

1, 2, 3, 4 15 11.

Cold Leg Injection Permissive, P-19 4

2 3

1, 2, 3 20 Amendment No. 70, 221, 242, 294 August 12, 2008 3/4 3-23

MILLSTONE - UNIT 3 TABLE 3.3-3 (Continued)

TABLE NOTATIONS The Steamline Isolation Logic and Safety Injection Logic for this trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.

MODES 1, 2, 3, and 4.

During movement of recently irradiated fuel assemblies.

Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked.

ACTION STATEMENTS ACTION 14 - a. With one channel inoperable, either restore the inoperable channel to operable status within 7 days or place one train of Control Room Emergency Ventilation System in the emergency mode of operation. Otherwise, immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With two channels inoperable, immediately place one train of control Room Emergency Ventilation System in the emergency mode of operation, declare one Control Room Emergency Ventilation System train inoperable, and comply with the ACTION requirements of Technical Specification 3.7.7.

Otherwise, immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

ACTION 14A -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 15 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 16 -

With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Amendment No. 57, 70, 89, 129, 203, 219, 221, 242, 243, 266, 294 3/4 3-24

MILLSTONE - UNIT 3 TABLE 3.3-3 (Continued)

ACTION STATEMENTS ACTION 17 -

With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing per Specification 4.3.2.1.

ACTION 18 - a. With one channel inoperable, either restore the inoperable channel to OPERABLE status within 7 days or place the associated train of Control Room Emergency Ventilation System in the emergency mode of operation.+

Otherwise, immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With two channels inoperable, immediately place one train of Control Room Emergency Ventilation System in the emergency mode of operation, declare one Control Room Emergency Ventilation System train inoperable, and comply with the ACTION requirements of Technical Specification 3.7.7.

Otherwise, immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

+Operation of the non-affected Control Room Emergency Ventilation System train, instead of the affected Control Room Emergency Ventilation System train, is permitted to perform required Technical Specifications 3.3.2 and 3.7.7 surveillance testing.

ACTION 19 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Amendment No. 57, 70, 89, 129, 203, 219, 221, 242, 243, 266, 272, 294 3/4 3-24a

REVERSE OF PAGE INTENTIONALLY LEFT BLANK 3/4 3-24a

MILLSTONE - UNIT 3 INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

a.

With the number of OPERABLE accident monitoring instrumentation channels except the containment area high range radiation monitor, less than the Total Number of Channels shown in Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 30 days, or prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within 14 days that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status b.

With the number of OPERABLE accident monitoring instrumentation channels except the containment area-high range radiation monitor, less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c.

With the number of OPERABLE channels for the containment area-high range radiation monitor less than required by either the total or the Minimum Channels OPERABLE requirements, initiate an alternate method of monitoring the appropriate parameter(s), within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and either restore the inoperable channel(s) to OPERABLE status within 7 days or prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within 14 days that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status.

Amendment No. 47, 57, 76, 224, 294 3/4 3-59

MILLSTONE - UNIT 3 LIMITING CONDITION FOR OPERATION (Continued)

SURVEILLANCE REQUIREMENTS 4.3.3.6.1 Each required accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 4.3-7.

4.3.3.6.2 Deleted Amendment No. 47, 57, 76, 142, 224, 258, 285, 3/4 3-59a 294

MILLSTONE - UNIT 3 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:

a.

The isolation valve open and power removed, b.

A contained borated water volume of between 6618 and 7030 gallons, c.

A boron concentration within limits specified in the COLR, and d.

A nitrogen cover-pressure of between 636 and 694 psia.

APPLICABILITY:

MODES 1, 2, and 3*.

ACTION:

a.

With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE:

a.

At the frequency specified in the Surveillance Frequency Control Program by:

1)

Verifying that the contained borated water volume and nitrogen cover-pressure in the tanks are within their limits, and 2)

Verifying that each accumulator isolation valve is open.

b.

At the frequency specified in the Surveillance Frequency Control Program and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of greater than or equal to 1 of tank volume by verifying the boron concentration of the accumulator solution.

This surveillance is not required when the volume increase makeup source is the RWST.

Pressurizer pressure above 1000 psig.

Amendment No. 12, 51, 60, 100, 258, 294 3/4 5-1

MILLSTONE - UNIT 3 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c.

At the frequency specified in the Surveillance Frequency Control Program when the RCS pressure is above 1000 psig by verifying that the associated circuit breakers are locked in a deenergized position or removed.

February 25, 2014 Amendment No. 100, 121, 258, 294 3/4 5-2

MILLSTONE - UNIT 3 EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A contained borated water volume between 1,166,000 and 1,207,000 gallons, b.

A boron concentration within limits specified in the COLR, c.

A minimum solution temperature of 42F, and d.

A maximum solution temperature of 73F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:

a.

At the frequency specified in the Surveillance Frequency Control Program by:

1)

Verifying the contained borated water volume in the tank, and 2)

Verifying the boron concentration of the water.

b.

At the frequency specified in the Surveillance Frequency Control Program by verifying the RWST temperature.

Amendment No. 12, 60, 258, 262, 294 3/4 5-9

MILLSTONE - UNIT 3 EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS LIMITING CONDITION FOR OPERATION 3.5.5 The trisodium phosphate (TSP) dodecahydrate Storage Baskets shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3 and 4 ACTION:

With the TSP Storage Baskets inoperable, restore the system TSP Storage Baskets to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.5 The TSP Storage Baskets shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying that a minimum total of 974 cubic feet of TSP is contained in the TSP Storage Baskets.

February 25, 2014 Amendment No. 115, 206, 258, 294 3/4 5-10

MILLSTONE - UNIT 3 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) e.

After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 7600 cfm to 9800 cfm; and f.

After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 7600 cfm to 9800 cfm.

January 3, 1995 Amendment No. 2, 53, 87, 100, 294 3/4 6-21

MILLSTONE - UNIT 3 CONTAINMENT SYSTEMS SECONDARY CONTAINMENT LIMITING CONDITION FOR OPERATION 3.6.6.2 Secondary Containment shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With Secondary Containment inoperable, restore Secondary Containment to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENT 4.6.6.2.1 OPERABILITY of Secondary Containment shall be demonstrated at the frequency specified in the Surveillance Frequency Control Program by verifying that each door in each access opening is closed except when the access opening is being used for normal transit entry and exit.

4.6.6.2.2 At the frequency specified in the Surveillance Frequency Control Program, verify one Supplementary Leak Collection and Release System produces a negative pressure of greater than or equal to 0.4 inch water gauge in the Auxiliary Building at 24-6 elevation within 120 seconds after a start signal.

Amendment No. 87, 100, 123, 126, 206, 258, 3/4 6-22 294

MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.5 Deleted CORE OPERATING LIMITS REPORT 6.9.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1.

Reactor Core Safety Limit for Specification 2.1.1.

2.

Overtemperature T and Overpower T setpoint parameters for Specification 2.2.1.

3.

SHUTDOWN MARGIN for Specifications 3/4.1.1.1.1, 3/4.1.1.1.2, and 3/4.1.1.2.

4.

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

5.

Shutdown Bank Insertion Limits for Specification 3/4.1.3.5.

6.

Control Bank Insertion Limits for Specification 3/4.1.3.6.

7.

AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1.1.

8.

Heat Flux Hot Channel Factor Limits for Specification 3/4.2.2.1.

9.

RCS Total Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for Specification 3/4.2.3.1.

10. DNB Parameters for Specification 3/4.2.5.
11. Shutdown Margin Monitor minimum count rate for Specification 3/4.3.5.
12. Accumulator boron concentration limits for Specification 3/4.5.1.
13. Refueling Water Storage Tank boron concentration limits for Specification 3/4.5.4.
14. Boron Concentration for Specification 3/4.9.1.1.

Amendment No. 24, 37, 69, 86, 188, 218, 223, 229, 236, 268, 282, 6-19a 294

MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)

6.9.1.6.b The analytical methods used to determine the core operating limits in Specification 6.9.1.6.a shall be those previously reviewed and approved by the NRC and identified below. The CORE OPERATING LIMITS REPORT will contain the complete identification for each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT (i.e.,

report number, title, revision, date, and any supplements).

1.

WCAP-9272-P-A, WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY, (W Proprietary). Methodology for Specifications:

  • 2.1.1 Reactor Core Safety Limits
  • 3.1.1.3 Moderator Temperature Coefficient
  • 3.1.3.5 Shutdown Bank Insertion Limits
  • 3.1.3.6 Control Bank Insertion Limits
  • 3.2.1.1 AXIAL FLUX DIFFERENCE
  • 3.2.2.1 Heat Flux Hot Channel Factor
  • 3.2.3.1 RCS Total Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor
  • 3.9.1.1 REFUELING Boron Concentration
  • 3.2.5 DNB Parameters

Deleted 3.

Deleted 4.

WCAP-10216-P-A-R1A, RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION, (W Proprietary). (Methodology for Specifications 3.2.1.1--AXIAL FLUX DIFFERENCE and 3.2.2.1--Heat Flux Hot Channel Factor) 5.

WCAP-16996-P-A, REALISTIC LOCA EVALUATION METHODOLOGY APPLIED TO THE FULL SPECTRUM OF BREAK SIZES (FULL SPECTRUM LOCA METHODOLOGY), (W Proprietary) (Methodology for Specification 3.2.2.1--Heat Flux Hot Channel Factor.)

February 16, 2022 Amendment No. 24, 37, 60, 69, 81, 120, 170, 218, 229, 236, 242, 253, 268, 279, 282, 294 6-20

MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)

6.

Deleted 7.

WCAP-11946, Safety Evaluation Supporting a More Negative EOL Moderator Temperature Coefficient Technical Specification for the Millstone Nuclear Power Station Unit 3, (W Proprietary). Methodology for Specification:

  • 3.1.1.3 - Moderator Temperature Coefficient 8.

WCAP-10054-P-A, WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE, (W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)

9.

WCAP-10079-P-A, NOTRUMP - A NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE, (W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)

10.

WCAP-12610, VANTAGE+ Fuel Assembly Report, (W Proprietary).

(Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)

11.

Deleted 12.

Deleted 13.

Deleted 14.

Deleted 15.

Deleted 16.

WCAP-8301, LOCTA-IV Program: Loss-of-Coolant Transient Analysis.

Methodology for Specification:

  • 3.2.2.1 - Heat Flux Hot Channel Factor 17.

WCAP-10054-P-A, Addendum 2, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model. Methodology for Specification:

  • 3.2.2.1 - Heat Flux Hot Channel Factor May 21, 2024 Amendment No. 81, 170, 218, 229, 236, 253, 268, 279, 289, 294 6-20a

MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)

18.

WCAP-8745-P-A, Design Bases for the Thermal Overpower T and Thermal Overtemperature DT Trip Functions, (Westinghouse Proprietary Class 2).

(Methodology for Specifications 2.2.1 -- Overtemperature T and Overpower T Setpoints.)

19.

WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO',

(W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)

20.

VEP-FRD-42-A, Reload Nuclear Design Methodology. Methodology for Specifications:

  • 2.1.1 Reactor Core Safety Limits
  • 3.1.1.3 Moderator Temperature Coefficient
  • 3.1.3.5 Shutdown Bank Insertion Limits
  • 3.1.3.6 Control Bank Insertion Limits
  • 3.2.2.1 Heat Flux Hot Channel Factor
  • 3.2.3.1 Nuclear Enthalpy Rise Hot Channel Factor
  • 3.5.4 Refueling Water Storage Tank
  • 3.9.1.1 REFUELING Boron Concentration 21.

VEP-NE-1-A, Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications. Methodology for Specifications:

  • 3.2.1.1 AXIAL FLUX DIFFERENCE
  • 3.2.2.1 Heat Flux Hot Channel Factor 22.

VEP-NE-2-A, Statistical DNBR Evaluation Methodology. Methodology for Specifications:

  • 3.2.3.1 RCS Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor
  • 3.2.5 DNB Parameters Amendment No. 268, 279, 282, 294 6-20b

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 294 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION, UNIT NO. 3 DOCKET NO. 50-423

1.0 INTRODUCTION

By letter dated March 20, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25079A225), and supplemented by letter dated August 14, 2025 (ML25226A079), Dominion Energy Nuclear Connecticut, Inc. (Dominion, the licensee) submitted a license amendment request (LAR), to the U.S. Nuclear Regulatory Commission (NRC, the Commission), that contained proposed amendments to the Technical Specifications (TS) for the Millstone Power Station, Unit No. 3 (MPS3).

1.1 System Description and Operation In section 2.1 of the LARs attachment, the licensee provided descriptions of the following systems: Control Building Isolation, Accident Monitoring Instrumentation, Accumulator and Refueling Water Storage Tank (RWST) Boron Concentration, and Secondary Containment Surveillance Requirements.

1.2 Description of Proposed Changes 1.2.1 Changes to Table 3.3-3, Functional Unit 7.a, Manual Initiation, and Functional Unit 7.c Automatic Actuation Logic and Actuation Relays, The proposed change revises the applicable Action for Functional Unit 7.a from ACTION 19 to Action 14, which is also being updated. It also revises the applicable Action for Functional Unit 7.c from Action 14 to Action 18.

The proposed change to Action 14 for TS Table 3.3-3 deletes the existing wording and replaces it with the following:

a. With one channel inoperable, either restore the inoperable channel to operable status within 7 days or place one train of Control Room Emergency Ventilation System in the emergency mode of operation. Otherwise, immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With two channels inoperable, immediately place one train of Control Room Emergency Ventilation System in the emergency mode of operation, declare one Control Room Emergency Ventilation System train inoperable, and comply with the ACTION requirements of Technical Specification 3. 7. 7. Otherwise, immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The proposed wording for Action 14 is consistent with Action 18, with the exception that one train of Control Room Emergency Ventilation System (CREVS) be placed in emergency mode of operation (instead of the associated train), and the footnote related to Surveillance Testing is omitted.

The proposed change to ACTION 18 for MPS3 TS Table 3.3-3 makes the following revisions:

a. With one Control Building Inlet Ventilation Radiation Monitor channel inoperable, either restore the inoperable channel to operable status within 7 days or place the associated train of Control Room Emergency Ventilation System in the emergency mode of operation.+ Otherwise, immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With two Control Building Inlet Ventilation Radiation Monitor channels inoperable, immediately place one train of Control Room Emergency Ventilation System in the emergency mode of operation, declare one Control Room Emergency Ventilation System train inoperable, and comply with the ACTION requirements of Technical Specification 3. 7. 7. Otherwise, immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The words specifying Control Building Inlet Radiation Monitor are proposed for deletion since the ACTION would also apply to Automatic Actuation Logic and Actuation Relays.

1.2.2 Changes to Actions a. and b. for TS 3.3.3.6 The proposed change to Action a. for TS Limiting Condition for Operation (LCO) 3.3.3.6 makes the following revisions:

With the number of OPERABLE accident monitoring instrumentation channels except the containment area high range radiation monitor, and reactor vessel water level, less than the Total Number of Channels shown in Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 30 days, or prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within 14 days that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status.

The proposed change to Action b. for TS LCO 3.3.3.6 will state:

With the number of OPERABLE accident monitoring instrumentation channels except the containment area-high range radiation monitor, and reactor vessel water level less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Actions d., e., and f. are also being proposed for deletion in their entirety.

1.2.3 Changes to LCO Requirements for TS 3.5.1 and TS 3.5.4 The proposed change to LCO 3.5.1 makes the following revision to item c.:

A boron concentration of between 2600 and 2900 ppm within limits specified in the COLR, and The proposed change to LCO 3.5.4 makes the following revision to item b.:

A boron concentration between 2700 and 2900 ppm of boron within limits specified in the COLR, The proposed change to TS 6.9.1.6.a makes the following revisions:

Current item 12, Boron Concentration for Specification 3/4.9.1.1., is renumbered to item 14.

Items 12 and 13 are proposed to state:

12.

Accumulator boron concentration limits for 3/4.5.1.

13.

Refueling Water Storage Tank boron concentration limits for 3/4.5.4.

The proposed change to TS 6.9.1.6.b makes the following revisions:

Item 20, VEP-FRD-42-A, Reload Nuclear Design Methodology. Methodology for Specifications, is updated to include 3.5.1 Accumulators and 3.5.4 Refueling Water Storage Tank.

1.2.4 Changes to SR 4.6.6.2.2 The proposed change to Surveillance Requirement (SR) 4.6.6.2.2 makes the following revisions:

At the frequency specified in the Surveillance Frequency Control Program, verify each one Supplementary Leak Collection and Release System produces a negative pressure of greater than or equal to 0.4 inch water gauge in the Auxiliary Building at 24-6 elevation within 120 seconds after a start signal.

2.0 REGULATORY EVALUATION

The NRC staff considered the following NRC regulations and guidance in its review of the proposed LAR:

2.1 Regulatory Requirements The NRCs regulatory requirements related to the content of the TS are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical specifications. This regulation requires that the TSs include items in, among other things, the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2)

LCOs; (3) SRs; (4) design features; and (5) administrative controls.

Regulations in 10 CFR 50.36(c)(2) state that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility, and when an LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

Regulations in 10 CFR 50.36(c)(3) state that SRs 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.

Regulations in 10 CFR 50.36(c)(5) state in part that administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

2.2 Design Basis Appendix A to 10 CFR Part 50 provides General Design Criteria (GDC) for nuclear power plants. Plant-specific design criteria are described in the plants Final Safety Analysis Report (FSAR). Specifically, Section 3.1 of Millstones FSAR (ML25177D082) discusses conformance with the following GDC:

GDC 10, Reactor Design, states:

The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

GDC 13, Instrumentation and Control, states:

Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

GDC 16, Containment Design, states:

Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

GDC 19, Control Room, states in part:

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss of coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of five rem whole body, or its equivalent to any part of the body, for the duration of the accident.

GDC 20, Protection System Functions, states:

The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

2.3 Regulatory Guidance Regulatory Guide (RG) 1.97, Revision 3, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident, May 1983 (ML003740282)

NUREG-0737, Clarification of TMI Action Plan Requirements, November 1980 (ML051400209)

Generic Letter (GL) 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications (ML031200485)

Standard Technical Specifications (STS) 3.37, Control Room Emergency Filtration System (CREFS) Actuation Instrumentation in NUREG 1431, Revision 5

3.0 TECHNICAL EVALUATION

3.1 Review of Proposed Changes 3.1.1 Changes to Table 3.3-3, Functional Unit 7.a, Control Building Isolation Manual actuation and Functional Unit 7.c Control Building Isolation Automatic Actuation Logic and Actuation Relays The NRC staff reviewed the proposed changes to ACTION 14 and ACTION 18; proposed changes to the action associated with Function Unit 7.a from ACTION 19 to ACTION 14 that is applicable during MODES 1, 2, 3, 4, and movement of recently irradiated fuel assemblies; and the proposed changes to the action associated with Function Unit 7.c from ACTION 14 to ACTION 18 that is applicable during MODES 1, 2, 3, 4.

STS 3.3.7 specifies actions when the required number of channels for the CREFS actuation instrumentation are not operable. With one channel inoperable, either restore the inoperable channel to operable status within 7 days OR place one train of CREVS in the emergency mode of operation, OR be in MODE 3 (Hot Standby) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND be in MODE 4 (Hot Shutdown) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, OR if applicable, immediately suspend movement of recently irradiated fuel assemblies. With two channels inoperable, immediately place one train of CREFS in the emergency mode of operation AND immediately enter applicable Conditions and Required Actions for one CREFS train made inoperable by inoperable CREFS actuation instrumentation, OR immediately place both trains of CREFS in emergency mode, OR be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND be in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, OR if applicable, immediately suspend movement of recently irradiated fuel assemblies.

The NRC staff compared the STS actions with the proposed MPS3 TS ACTION 14, which provides the required actions when the minimum number of operable channels specified is not met for Functional Unit 7.a, Control Building Isolation Manual Actuation, in MPS3 TS table 3.3-

3. The staff found that proposed ACTION 14 specifies equivalent actions to those specified for Conditions A, B, C and D of STS 3.3.7 as follows; With one channel inoperable, both TSs allow a period of 7 days to restore the inoperable channel to operable status before requiring one train of the CREVS to be placed in its emergency mode of operation, OR if required action associated with completion time is not met, immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, which is same as the MODE 5 (Cold Shutdown) completion time specified in the STS Section 1.3.

With two channels inoperable, both TSs require immediately placing one train of the CREVS in the emergency mode of operation, declaring one CREVS train inoperable and complying with the ACTION requirements of Technical Specification 3.7.7, otherwise immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The NRC staff compared the STS specified actions with the proposed MPS3 TS ACTION 18, which provides the required actions when the minimum number of operable channels specified is not met for Functional Unit 7.c, Control Building Isolation Automatic Actuation Logic and Actuation Relays, in MPS3 TS Table 3.3-3. The staff found that proposed ACTION 18 specifies equivalent actions to those specified for Conditions A, B, C, and D of STS 3.3.7 as follows:

With one channel inoperable, both TSs allow a period of 7 days to restore the inoperable channel to operable status before requiring the associated train of CREVS to be placed in the emergency mode of operation, OR if required action associated with completion time is not met, immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With two channels inoperable, both TSs require immediately placing one train of the CREVS in the emergency mode of operation, declaring one CREVS train inoperable and complying with the ACTION requirements of Technical Specification 3.7.7, otherwise immediately suspend movement of recently irradiated fuel assemblies, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The proposed MPS3 TS ACTION 14 and ACTION 18 are consistent with the corresponding STS actions and provide for reasonable assurance of safe operations. With two CBI channels inoperable, both proposed ACTION 14 and ACTION 18 require immediate placement of one CREVS train in the emergency mode of operation, which completes the safety function of supplying filtered air to the control room. If the plant is operating in MODEs 1, 2, 3, or 4, and the licensee fails to either immediately place one CREVS train in the emergency mode of operation or restore the inoperable CREVS train to operable status within 7 days, the plant must be placed in a mode where the LCO no longer applies. These actions provide reasonable assurance that one train of CREVS would reliably be operating in the emergency mode of operation to protect control room operators from a potential release of radioactive material or that the plant would be promptly placed in an operating mode where an energetic release of radioactive material would not be credible (Cold Shutdown). Therefore, proposed changes to ACTION 14 for Functional Unit 7.a, Control Building Isolation Manual Actuation, and ACTION 18 for Functional Unit 7.c, Control Building Isolation Automatic Actuation Logic and Actuation Relays are acceptable.

3.1.2 Changes to MPS3 TS ACTIONs a., b., e., and f. for TS LCO 3.3.3.6, Accident Monitoring Instrumentation Channels The NRC staff reviewed the proposed changes to ACTIONs a., b., e., and f. of MPS3 TS LCO 3.3.3.6 - related to operability of the accident monitoring instrumentation channels. The proposed changes involve extending the allowed outages times (AOTs) for inoperable accident monitoring instrumentation channels and eliminating specific actions that are considered redundant.

MPS3 TS 3.3.3.6, ACTION a., for One Inoperable Channel Current MPS3 TS 3.3.3.6 ACTION a. requires restoring an inoperable accident monitoring instrument to operable status within 7 days when the number of Operable channels is less than the total number of channels, and the requirement for minimum channels Operable is met.

Following expiration of the 7-day AOT, the current TS 3.3.3.6 also requires a plant shutdown.

The proposed change to ACTION a. revises the AOT to allow 30 days for restoration of the inoperable channel, and eliminates the shutdown requirement and adds a requirement for submitting a special report to the NRC, pursuant to Specification 6.9.2, within 14 days. This report will contain the actions being taken, the cause of the inoperability, and the plans and schedule for restoring the instrument to Operable status.

The NRC staff compared the STS 3.3.3 specified actions with the above proposed changes to MPS3 TS 3.3.3.6 ACTION a. The staff found that proposed ACTION a. of the MPS3 TS 3.3.3.6 specifies equivalent actions to those specified for Conditions A and B of the STS 3.3.3. Both allow a period of 30 days to restore the inoperable channel to operable status before requiring the submission of a Special Report to the Commission and thereby provide reasonable assurance of safe operations. If the inoperable channel is restored to operable status within 30 days, the instrument would perform its intended function of monitoring and assessing plant parameters during and following an accident. Similarly, if a special report is required to be submitted, it provides actions taken, the cause of inoperability, and plans and schedule for restoring the channel to operable status. Thus, the specified actions result in identical end states at similar times, with the completion times proposed ACTION a. of the MPS3 TS3.3.3.6 being consistent with the applicable STS 3.3.3 specified actions.

The 30-day AOT considers the remaining operable channel(s) or other non-RG 1.97 instrumentation, and the low probability of an event requiring accident monitoring instrumentation during the 30-day period. It also considers the passive nature of Post Accident Monitoring (PAM) instrumentation, which does not provide inputs to automatic protective actions.

MPS3 TS 3.3.3.6, ACTION b., for Two Inoperable Channels Current MPS3 TS 3.3.3.6 ACTION b. requires restoring inoperable accident monitoring instruments to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> when the number of Operable channels is less than the minimum channels Operable requirements. The proposed change to ACTION b.

revises the AOT to allow 7 days for restoration of one of the inoperable channels thus making the action consistent with condition C of STS 3.3.3.

Similarly, the staff compared the STS 3.3.3 specified actions with the above proposed changes to ACTION b. of the specifies equivalent action. Both STS 3.3.3, Conditions C and D, and MPS TS ACTION b. require restoring the inoperable channel(s) to operable status within 7 days or be in at least Hot Standby (MODE 3) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least Hot Shutdown (MODE 4) within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The 7-day AOT considers availability of other non-RG 1.97 instrumentation, and the low probability of an event requiring accident monitoring instrumentation during the 7-day period. It also considers the passive nature of PAM instrumentation, which does not provide inputs to automatic protective actions.

The proposed MPS3 TS 3.3.3.6 ACTIONs a. and b. are consistent with the corresponding STS 3.3.3 actions and provide reasonable assurance of safe operations. The extended allowed outage times consider the low probability of an event requiring accident monitoring instrumentation during the extended period and the availability of alternate means to obtain required information. Proposed changes to the MPS3 TS 3.3.3.6 provide reasonable assurance that the accident monitoring instrumentation will be restored to operable status in a timely manner or that the plant will be placed in a safe condition if the instrumentation cannot be restored.

The staff determined that the deletion of MPS3 TS 3.3.3.6 ACTIONs e. and f. (specific to accident monitoring instrumentation channels for Reactor Vessel Level) is acceptable, because proposed changes to MPS3 TS 3.3.3.6, ACTIONs a. and b. make ACTIONs e. and f. redundant and therefore no longer necessary.

3.1.3 Changes to LCO Requirements for TS 3.5.1, Accumulators, and TS 3.5.4, Refueling Water Storage Tank The NRC staff reviewed the proposed changes to the LCO requirements for TS 3.5.1 and TS 3.5.4. The proposed changes would relocate fixed boron concentration limits for the accumulators and RWST to the COLR, Current Limiting LCO 3.5.1.c requires each reactor coolant system (RCS) accumulator to be OPERABLE with a boron concentration between 2600 and 2900 ppm.

Current LCO 3.5.4.b requires the RWST to be OPERABLE with a boron concentration between 2700 and 2900 ppm of boron.

The proposed LCOs relocate these boron concentration limits to the COLR and the revised LCO be read as follows:

Proposed LCO 3.5.1.c requires that each RCS accumulator be OPERABLE with a boron concentration within limits specified in the COLR.

Proposed LCO 3.5.4.b requires that the RWST be OPERABLE with a boron concentration within limits specified in the COLR.

These boron concentration limits are cycle-specific parameters. Relocating them to COLR allows for updates on a cycle-specific basis without requiring a license amendment. The revised TSs will reference the COLR for the applicable limits.

The licensee confirmed that the boron concentration limits are determined using NRC-approved methodology documented in Topical Report (TR) VEP-FRD-42-A Revision 2, Minor Revision 2, (ML18012A098). This methodology was previously approved for Framatome ANP Advanced Mark-BW and Westinghouse fuel types (ML16131A728), with a condition requiring verification of applicability when used with other fuel types.

During the review, the NRC staff requested additional information to confirm compliance with this condition.

In response to the NRC staffs request for additional information (RAI), the licensee confirmed (ML25226A079) that MPS3 has been operating with Framatome GAIA fuel following the completion of the spring 2025 refueling outage. The transition from Westinghouse RFA-2 fuel to GAIA fuel was implemented under 10 CFR 50.59 and supported by NRC-approved Amendment Nos. 287 (ML23226A005), 289 (ML24109A002), 290 (ML24128A276), and 291 (ML24346A253).

The NRC-approved methodology documented in VEP-FRD-42-A includes a condition requiring verification of applicability when used with fuel types other than Westinghouse or Framatome ANP Advanced Mark-BW. The licensee evaluated the applicability of this methodology to GAIA fuel and concluded that no changes were necessary. This evaluation was performed under 10 CFR 50.59.

The licensees assessment addressed the following three major analytical models used in the methodology of VEP-FRD-42-A:

Neutronics (CMS5) Model: The CMS5 model is described in NRC-approved TR SSP P01/028-TR-P-A (ML17279A986), which contains conditions restricting the applicability of the model to fuel features. The licensee confirmed that the design features of the NRC-approved GAIA fuel at MPS3, relating to the fuel and poison characteristics, M5 cladding, structural materials, and burnup limits, meet the modeling limitations described in NRC-approved Topical Report SSP-14-P01/028-TR-P-A. No changes to VEP-FRD-42-A requiring NRC review of SSP-14-P01/028-TR-P-A were identified.

Non-Loss of Coolant Accident (LOCA) System Transients (RETRAN) Model: The RETRAN model described in the NRC-approved TR VEP-FRD-41-P-A (ML19141A148) uses reactivity inputs that are bounding of the reload cores derived from the NRC-approved CMS5 code. Fuel-specific material properties were incorporated using data determined from the fuel vendors approved fuel performance code, consistent with VEP-FRD-41-P-A.

Thermal-Hydraulics (VIPRE-D) Model: The licensee uses the VIPRE-D model described in the NRC-approved TR DOM-NAF-2-P-A, including GAIA-specific critical heat flux correlations, as described in DOM-NAF-2-P-A, Appendix F (ML24170B053). This application was previously approved by the NRC in Amendment No. 291 for MPS3.

The licensee also addressed event-specific considerations, including rod ejection and LOCA analyses. These events were analyzed using NRC-approved Framatome methodologies and previously approved by NRC (ML24296B234 and ML24109A002). The reload verification parameters (e.g., reactivity coefficients, peaking factors, axial power shapes) remain valid under the existing methodology in VEP-FRD-42-A.

Based on this assessment above-mentioned, the licensee concluded that the application of VEP-FRD-42-A to reload cores containing GAIA fuel at MPS3 is within the scope of the NRC-approved methodology and its associated conditions. Therefore, no prior NRC review or approval was required under 10 CFR 50.59.

The NRC staff reviewed the licensees RAI response and determined that it adequately addressed the condition in the VEP-FRD-42-A safety evaluation report. Accordingly, the NRC staff determined the application of VEP-FRD-42-A acceptable for determining boron concentration limits in the accumulators and RWST at MPS3.

This satisfies Condition 1 of GL 88-16.

To satisfy Condition 2 of GL 88-16, the licensee proposed the following additions:

TS 6.9.1.6.a: Add items 12 and 13 for accumulator and RWST boron concentration limits documented in the COLR.

TS 6.9.1.6.b: Add TS 3.5.1 and TS 3.5.4 to the list of the parameters determined using the VEP-FRD-42-A methodology.

The NRC staff found that the proposed changes are consistent with the format of existing TSs and appropriately reference the NRC-approved analytical methodology and thus determined that the changes meet Condition 2 of GL 88-16.

The licensee will maintain the boron concentration limits in the COLR and submit the COLR to the NRC for information, consistent with Condition 3 of GL 88-16.

An editorial change is proposed to renumber item 12 to item 14 in TS 6.9.1.6.a. This change is administrative and does not affect the technical content of the TSs.

3.1.4 Changes to SR 4.6.6.2.2 The NRC staff reviewed the language change that was made to SR 4.6.6.2.2, as described in this evaluation. Only one Supplementary Leak Collection and Release System is needed to provide the minimum postulated iodine removal assumed in the safety analysis and the licensee states that the successful completion of the surveillance for the opposite train would validate that there is no concurrent single failure. Therefore, from an operations standpoint, the staff finds the change acceptable since the requirements of 10 CFR 50.36 would continue to be satisfied.

Additionally, the licensee states that the proposed change would align the TS with the latest version of the STS (NUREG-1431, Revision 5). The staff confirmed that the proposed change would align the SR text with the text in SR 3.6.8.4 of NUREG-1431.

The staff finds the proposed change to SR 4.6.6.2.2 acceptable because only one system is needed to perform the specified safety function as required by the safety analysis which ensures the requirements of 10 CFR 50.36 continue to be met and the SR language would further align with that of the STS.

3.2 Technical Summary The staff reviewed the following proposed changes:

revise applicable Actions in TS Table 3.3-3, Functional Unit 7.a, Manual Initiation, and Functional Unit 7.c, Automatic Actuation Logic and Actuation Relays, for CBI, when one or both instrumentation channel(s) are inoperable, revise applicable Actions a. and b. for TS 3.3.3.6, Accident Monitoring Instrumentation, when one or both accident monitoring instrumentation channel(s) are inoperable, revise the LCO requirements for TS 3.5.1, Accumulators, and TS 3.5.4, Refueling Water Storage Tank, to relocate boron concentration limit values to the Core Operating Limit Report, and revise MPS3 TS SR 4.6.6.2.2 under TS 3.6.6.2, Secondary Containment.

For reasons described in this technical evaluation, the staff finds the changes acceptable because the requirements of 10 CFR 50.36 and GDCs 10, 13, 16, 19, and 20 continue to be met.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Connecticut State official was notified of the proposed issuance of the amendments on January 12, 2026, and the State of Connecticut had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on May 13, 2025(90 FR 20516).

Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations; and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: J. Monachan, NRR D. Scully, NRR S. Sun, NRR T. Sweat, NRR D. Taneja, NRR Date of Issuance: March 12, 2026

SUBJECT:

MILLSTONE POWER STATION, UNIT NO. 3 ISSUANCE OF AMENDMENT NO. 294 RE: REVISIONS TO TECHNICAL SPECIFICATIONS RELATED TO CONTROL BUILDING ISOLATION, ACCIDENT MONITORING INSTRUMENTATION, ACCUMULATOR AND REFUELING WATER STORAGE TANK BORON CONCENTRATION LIMITS, AND SECONDARY CONTAINMENT SURVEILLANCE REQUIREMENTS (EPID L-2025-LLA-0053)

DATED MARCH 12, 2026 DISTRIBUTION:

Public RidsRgn1MailCenter Resource RidsACRS_MailCTR Resource RidsNrrDexEicb Resource RidsNrrDssStsb Resource RidsNrrDssScpb Resource RidsNrrDssSnsb Resource RidsNrrDorlLpl1 Resource RidsNrrPMMillstone Resource RidsNrrLASLent Resource ADAMS Accession Nos.:

Package: ML26070A282 Amendment: ML26051A065 eConcurrence Case: 20260220-50008

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