ML25339A001
| ML25339A001 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/16/2025 |
| From: | Edwards T NRC/NRR/DORL/LPL1 |
| To: | Carr E Dominion Energy Nuclear Connecticut |
| Edwards T, NRR/DORL/LPL1 | |
| References | |
| EPID L-2025-LLA-0075 | |
| Download: ML25339A001 (0) | |
Text
December 16, 2025 Mr. Eric S. Carr President - Nuclear Operations and Chief Nuclear Officer 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. 293 TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TECHNICAL SPECIFICATION TASK FORCE TRAVELER 577, REVISION 1, REVISED FREQUENCIES FOR STEAM GENERATOR TUBE INSPECTIONS (EPID L-2025-LLA-0075)
Dear Mr. Carr:
The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 293 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 April 22, 2025, as the amendment revises the Millstone Power Station, Unit No. 3, technical specifications to adopt Technical Specification Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections.
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. 293 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. 293 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 April 22, 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)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 293, 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, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License and Technical Specifications Date of Issuance: December 16, 2025 RICHARD GUZMAN Digitally signed by RICHARD GUZMAN Date: 2025.12.16 13:06:50 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 293 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 6-17a 6-17a 6-17b 6-17b 6-17c 6-17c 6-17d 6-17d 6-17e 6-17e 6-21a 6-21a 6-21b 6-21b 6-21c 6-21c 6-21d 6-21d Renewed License No. NPF-49 Amendment No. 270-292, 293 (2)
Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 293 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 ADMINISTRATIVE CONTROLS g.
Steam Generator (SG) Program An SG Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following:
a.
Provisions for condition monitoring assessments: Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during a SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
b.
Provisions for performance criteria for SG tube integrity: SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
- 1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or a combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
- 2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.
Amendment No. 238, 256, 293 6-17a
MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
Leakage is not to exceed 500 gpd per SG.
- 3. The operational LEAKAGE performance criterion is specified in RCS LCO 3.4.6.2, Operational LEAKAGE.
c.
Provisions for SG tube plugging criteria: Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40%
of the nominal tube wall thickness shall be plugged.
The following alternate tube plugging criteria may be applied as an alternative to the 40% depth-based criteria:
- 1. Tubes with service-induced flaws located greater than 15.2 inches below the top of the tubesheet do not require plugging.
Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 15.2 inches below the top of the tubesheet shall be plugged upon detection.
Amendment No. 238, 245, 249, 252, 255, 256, 293 6-17b
MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) d.
Provisions for SG tube inspections: Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet except for any portions of the tube that are exempt from inspection by alternate repair criteria, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
- 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
- 2. After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 54 effective full power months, which defines the inspection period. If none of the SG tubes have ever experienced cracking other than in regions that are exempt from inspection by alternate repair criteria and the SG inspection was performed with enhanced probes, the inspection period may be extended to 72 effective full power months. Enhanced probes have a capability to detect flaws of any type equivalent to or better than array probe technology. The enhanced probes shall be used from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet except any portions of the tube that are exempt from inspection by alternate repair criteria. If there are regions where enhanced probes cannot be used, the tube inspection techniques shall be capable of detecting all forms of existing and potential degradation in that region.
As approved by License Amendment No. 277, inspection of the Millstone Unit 3 SGs A and C may be deferred, on a one-time basis, from fall 2020 (Refueling Outage 20) to spring 2022 (Refueling Outage 21).
Amendment No. 69, 186, 212, 238, 243, 245, 249, 252, 255, 256, 277, 293 6-17c
MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- 3. If crack indications are found in portions of the SG tube excluding any region that is exempt from inspection by alternate repair criteria, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage, but may be deferred to the following refueling outage if the 100% inspection of all SGs was performed with enhanced probes as described in paragraph d.2. If definitive information such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not Amendment No. 256, 293 6-17d
MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) associated with a crack(s), then the indication need not be treated as a crack.
e.
Provisions for monitoring operational primary to secondary LEAKAGE.
h.
Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVs), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:
a.
The definition of the CRE and the CRE boundary.
b.
Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
c.
Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
The following are exceptions to Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0:
- 1. Appropriate application of ASTM E741 shall include the ability to take minor exceptions to the test methodology. These exceptions shall be documented in the test report, and
- 2. Vulnerability assessments for radiological, hazardous chemical and smoke, and emergency ventilation system testing were completed as documented in the UFSAR and other licensing basis documents. The exceptions to the Regulatory Guides (RG) referenced in RG 1.196 (i.e., RG 1.52, RG 1.78, and January 11, 2013 Amendment No. 243, 245, 256, 293 6-17e
MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.7 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with TS 6.8.4.g, Steam Generator (SG) Program. The report shall include:
a.
The scope of inspections performed on each SG; b.
The nondestructive examination techniques utilized for tubes with increased degradation susceptibility; c.
For each degradation mechanism found:
1.
The nondestructive examination techniques utilized; 2.
The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported; 3.
A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and 4.
The number of tubes plugged during the inspection outage.
d.
An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results; e.
The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG.
f.
The results of any SG secondary side inspections; g.
The primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report; Amendment No. 238, 245, 249, 252, 255, 256, 279, 289, 290, 293 6-21a
MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT (Continued) h.
The calculated accident induced leakage rate from the portion of the tubes below 15.2 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.49 times the maximum operational primary to secondary leakage rate; the report should describe how it was determined; and i.
The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.
Amendment No. 238, 245, 249, 252, 255, 256, 279, 289, 290, 293 6-21b
MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator Region I, and one copy to the NRC Resident Inspector, within the time period specified for each report.
6.10 Deleted.
6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.
6.12 HIGH RADIATION AREA As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20:
6.12.1 High Radiation Areas with Dose Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation a.
Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
b.
Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent; that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
c.
Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
d.
Each individual or group entering such an area shall possess:
June 4, 2024 Amendment No. 245, 279, 289, 290, 293 6-21c
MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA (cont.)
1.
A radiation monitoring device that continuously displays radiation dose rates in the area, or 2.
A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the devices dose alarm setpoint is reached, with an appropriate alarm setpoint, or 3.
A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or June 4, 2024 Amendment No. 245, 279, 289, 290, 293 6-21d SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 293 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 April 22, 2025, Dominion Energy Nuclear Connecticut (DENC, the licensee) requested changes to the technical specifications (TSs) for Millstone Power Station, Unit 3, (MPS3), by license amendment request (application). In its April 22, 2025 (Agencywide Documents Access and Management System (Accession No. ML25113A202), letter, the licensee requested that the U.S. Nuclear Regulatory Commission (NRC, the Commission) process the proposed amendment under the Consolidated Line-Item Improvement Process (CLIIP). The proposed changes would revise the Steam Generator (SG) Program and the Steam Generator Tube Inspection Report TSs based on Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections (TSTF-577) (ML21060B434), and the associated NRC staff safety evaluation (SE) of TSTF-577 (ML21098A188).
The tubes within a SG function as an integral part of the reactor coolant pressure boundary and, in addition, isolate fission products in the primary coolant from the secondary coolant and the environment. SG tube integrity means that the tubes are capable of performing this safety function in accordance with the plant design and licensing basis.
MPS3 SGs have Alloy 600 thermally treated (Alloy 600TT) tubes.
1.1 Proposed Technical Specification (TS) Changes to Adopt TSTF-577 In accordance with NRC staff-approved TSTF-577, the licensee proposed changes that would revise MPS3 TS 6.8.4.g, Steam Generator (SG) Program, and MPS3 TS 6.9.1.7, Steam Generator Tube Inspection Report. Specifically, the licensee proposed the following changes to adopt TSTF-577:
TS 6.8.4.g, Steam Generator (SG) Program:
TS 6.8.4.g introductory paragraph and paragraph b.1 would be revised by replacing Steam Generator or steam generator with SG in a few instances.
TS 6.8.4.g.d would be revised by adding a phrase regarding portions of the tube that are exempt from inspection by alternate repair criteria.
TS 6.8.4.g.d.2 would be revised by deleting the requirement to base inspection frequency on the more restrictive metric between either the effective full power months (EFPM) or refueling outage and to use just the EFPM metric.
TS 6.8.4.g.d.2 would be revised by changing the requirement to inspect 100 percent of the tubes at periods of 120, 96, and 72 EFPM to 54 EFPM. A 72 EFPM inspection period would be permitted if SG tubing has never experienced cracking (not including regions exempt from inspection by alternate repair criteria) and the SG inspection was performed with enhanced probes. A description of the enhanced probe inspection would be added.
TS 6.8.4.g.d.2 would be revised by deleting the allowance to extend the inspection period by 3 months and by deleting the discussion of prorating inspections.
TS 6.8.4.g.d.3 would be revised by replacing shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections) with shall be at the next refueling outage.
TS 6.8.4.g.d.3 would be revised by adding a phrase regarding portions of the tube that are exempt from inspection by alternate repair criteria. An additional phrase would be added that permits deferring SG inspections after cracking indications are found if the 100 percent inspection was performed with enhanced probes.
TS 6.9.1.7, Steam Generator Tube Inspection Report:
Existing reporting requirement b. would be renumbered as c. and be revised by editorial and punctuation changes.
New reporting requirement b. would be added to require the nondestructive examination (NDE) techniques utilized for tubes with increased degradation susceptibility be reported.
Existing reporting requirement c. would be renumbered as c.1. and be revised by editorial and punctuation changes.
Existing reporting requirement d. would be renumbered as c.2. and be revised to note that the location, orientation (if linear), measured size (if available), and voltage response do not need to be reported for tube wear indications at support structures that are less than 20 percent through-wall. However, the total number of tube wear indications at support structures that are less than 20 percent through-wall would be reported.
New reporting requirement d. would be added to require an analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection relative to the applicable performance criteria, including the analysis methodology, inputs, and results.
Existing reporting requirement e. would be renumbered as c.4. and be revised by editorial and punctuation changes.
Existing reporting requirement f. would be renumbered as e. and be revised by editorial and punctuation changes.
New reporting requirement f. would be added to require the results of any SG secondary side inspections be reported.
Existing reporting requirement g. would be renumbered as c.3. and be revised to add the requirements to report a description of the condition monitoring assessment, the margin to the tube integrity performance criteria, and a comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment. In addition, the requirement to report the results of tube pulls and in-situ testing would be deleted.
Existing reporting requirements h, i, and j would be renumbered to g, h, and i, and be revised by editorial and punctuation changes.
1.2 Additional Proposed TS Changes In addition to the changes proposed consistent with the traveler discussed in section 1.1, the licensee proposed the following variations.
1.2.1 Editorial Variations
- 1) MPS3 TSs have different numbering than standard technical specifications (STSs) on which TSTF577 was based. Specifically, the Steam Generator (SG) Program is numbered 6.8.4.g in MPS3 TSs rather than 5.5.9 as stated in the TSTF. In addition, the Steam Generator Tube Inspection Report is numbered 6.9.1.7 in MPS3 TSs rather than 5.6.7 as stated in the TSTF. Furthermore, the TS for RCS [reactor coolant system]
Operational LEAKAGE is numbered 3.4.6.2 in MPS3 TSs rather than 3.4.13 as stated in the TSTF.
- 2) MPS3 TS 6.9.1.7 is revised to place the title of TS 6.8.4.g in quotes, which is consistent with TSTF577.
1.2.2 Other Variations
- 1) MPS3 TSs contain a requirement the differs from the STSs on which TSTF-577 was based. Specifically, MPS3 TS 6.8.4.g Steam Generator (SG) Program, paragraph b.2, states, Leakage is not to exceed 500 gpd [gallons per day] per SG rather than, Leakage is not to exceed [1 gpm (gallon per minute)] per SG.
- 2) MPS3 TSs currently contain a provision for alternate tube plugging criteria. The licensee noted that the description of the alternate tube plugging criteria in the proposed change is equivalent to the description in the current TSs.
- 3) MPS3 TS 6.8.4.g.c currently states, in part, The following alternate tube plugging criteria shall be applied as an alternative to the 40% depth based criteria (emphasis added). However, TSTF-577 states: The following alternate tube plugging [or repair]
criteria may be applied as an alternative to the 40% depth based criteria (emphasis added). The licensee proposed to change the word shall to may in MPS3 TS 6.8.4.g.c which aligns with the wording in TSTF577.
2.0 REGULATORY EVALUATION
The regulations in Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.36(c)(5),
Administrative controls, state 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. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in § 50.4.
Technical Specification Section 5.0, Administrative Controls, requires that an SG Program be established and implemented to ensure that SG tube integrity is maintained. Programs established by the licensee, including the SG Program, are listed in the administrative controls section of the TS to operate the facility in a safe manner.
The NRC staffs guidance for the review of TSs is in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]
Edition (SRP), Chapter 16.0, Technical Specifications, Revision 3, dated March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STSs for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with NUREG-14311 as modified by NRC-approved travelers.
TSTF-577 revised the STSs related to SG tube inspections and SG tube inspection reporting requirements. The NRC approved TSTF-577, under the CLIIP on April 14, 2021 (ML21099A086).
3.0 TECHNICAL EVALUATION
3.1 Proposed TS Changes to Adopt TSTF-577 The NRC staff compared the licensees proposed TS changes in section 1.1 of this SE against the changes approved in TSTF-577. In accordance with SRP Chapter 16.0, the NRC staff determined that the STS changes approved in TSTF-577 are applicable because MPS3 is a pressurized water reactor (PWR) design plant, and the NRC staff approved the TSTF-577 changes for PWR designs. The NRC staff finds that the licensees proposed changes to the MPS3 TSs in section 1.1 of this SE are consistent with those found acceptable in TSTF-577.
In the SE of TSTF-577, the NRC staff concluded that the TSTF-577 changes to STS 5.5.9, Steam Generator (SG) Program, and STS 5.6.7, Steam Generator Tube Inspection Report, 1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 5, September 2021 (ML21259A155 and ML21259A159, respectively).
were acceptable because, as discussed in section 3.0 of that SE, they continued to ensure SG tube integrity and, therefore, protected the public health and safety. In particular, the structural integrity performance criterion and accident-induced leakage performance criterion (explained in STS 5.5.9.b, items 1 and 2, respectively) will continue to be met with the proposed revised SG inspection intervals (maximum allowable time between SG inspections) and inspection periods (maximum allowable time between 100 percent of SG tubes inspections). Additionally, the proposed changes to the reporting requirements will provide more detailed and consistent information to the NRC. Therefore, the NRC staff found that the proposed changes to the SG Program and inspection reporting requirements were acceptable because they continued to meet the requirements of 10 CFR 50.36(c)(5) by providing administrative controls necessary to assure operation of the facility in a safe manner. For these same reasons, the NRC staff concludes that the corresponding proposed changes to the MPS3 TSs in section 1.1 of this SE continue to meet the requirements of 10 CFR 50.36(c)(5).
3.2 Additional Proposed TS Changes 3.2.1 Editorial Variations The editorial variations are described in section 1.2.1 of this SE. The NRC staff finds that these variations (i.e., TS numbering differences and placing a title in quotes) are acceptable because these variations do not substantively alter TS requirements.
3.2.2 Other Variations There are three variations identified in section 1.2.2 of this SE.
For the first variation, the licensee explained that the MPS3 SG Program TSs currently contain a requirement that differs from the STS on which TSTF577 was based. Specifically, the last sentence of MPS3 TS 6.8.4.g.b.2 states, Leakage is not to exceed 500 gpd per SG rather than Leakage is not to exceed [1 gpm] per SG. The NRC staff notes that the MPS3 TS 6.8.4.g.b.2 leakage rate requirements are more restrictive than the STS. As part of the request to adopt TSTF-577, the licensee did not propose any changes to these leakage rate requirements.
Therefore, the NRC staff considers this licensee identified variation acceptable.
For the second variation, the licensee explained that the MPS3 SG Program TSs currently contain a provision for alternate tube plugging criteria. As part of the request to adopt TSTF-577, the licensee did not propose any changes to these criteria. Therefore, the NRC staff considers this licensee identified variation acceptable.
For the third variation, the licensee noted that the MPS3 SG Program TSs currently contain a requirement that differs from the STS on which TSTF577 was based. Specifically, MPS3 TS 6.8.4.g.c currently states, in part: The following alternate tube plugging criteria shall be applied as an alternative to the 40% depth based criteria (emphasis added). However, TSTF-577 states: The following alternate tube plugging [or repair] criteria may be applied as an alternative to the 40% depth based criteria (emphasis added). In the application, the licensee explained that alternate tube plugging criteria are an alternative to the 40% tube plugging criterion, therefore, the word may is the proper word to use. Changing, the word shall to may in MPS3 TS 6.8.4.g.c also aligns with the STS wording contained in TSTF-577.
The NRC staff finds the change from shall to may acceptable based on the explanation provided by the licensee and because it is consistent with STS SG Program guidance.
3.3 TS Change Consistency The NRC staff reviewed the proposed TS changes for technical clarity and consistency with the existing requirements for customary terminology and formatting. The NRC staff finds that the proposed changes are consistent with Chapter 16.0 of the SRP and are therefore acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Connecticut State official was notified of the proposed issuance of the amendment on November 21, 2025. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 [and changes surveillance requirements.] The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding (90 FR 24420). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Clint Ashley, NRR Date: December 16, 2025
ML25339A001 NRR-058 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DSS/STSB/BC NAME TEdwards KEntz SMehta DATE 12/5/2025 12/5/2025 12/9/2025 OFFICE NRR/DORL/LPL1/BC (A)
NRR/DORL/LPL1/PM NAME UShoop (RGuzman for)
TEdwards DATE 12/16/2025 12/16/2025