ML21227A000

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Issuance of Amendment No. 279 Regarding Addition of Analytical Methodology to the Core Operating Limits Report for a Large Break Loss-of-Coolant Accident
ML21227A000
Person / Time
Site: Millstone Dominion icon.png
Issue date: 10/05/2021
From: Richard Guzman
NRC/NRR/DORL/LPL1
To: Stoddard D
Dominion Energy Nuclear Connecticut
Guzman R
References
EPID L-2020-LLA-0242
Download: ML21227A000 (30)


Text

October 5, 2021 Mr. Daniel G. Stoddard Senior Vice President 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. 279 RE: ADDITION OF ANALYTICAL METHODOLOGY TO THE CORE OPERATING LIMITS REPORT FOR A LARGE BREAK LOSS-OF-COOLANT ACCIDENT (EPID L-2020-LLA-0242)

Dear Mr. Stoddard:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 279 to Renewed Facility Operating License No. NPF-49 for the Millstone Power Station, Unit No. 3 (Millstone 3), in response to your application dated November 5, 2020, as supplemented by letter dated May 20, 2021.

The amendment revises the Millstone 3 Technical Specification 6.9.1.6.b by adding topical report WCAP-16996-P-A, Revision 1, Realistic LOCA [loss-of-coolant accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (Full Spectrum LOCA Methodology), to the list of methodologies approved for reference in the core operating limits report (COLR) for Millstone 3.

The added reference identifies the analytical method used to determine the core operating limits for the large break LOCA event described in the Millstone 3 Final Safety Analysis Report, Section 15.6.5, Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary. The amendment also removes COLR Reference WCAP-12945-P-A, which is no longer being used to support the Millstone 3 core reload analysis.

A copy of the related safety evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423

Enclosures:

1. Amendment No. 279 to NPF-49
2. Safety Evaluation cc: Listserv

DOMINION ENERGY NUCLEAR CONNECTICUT, INC., ET AL DOCKET NO. 50-423 MILLSTONE POWER STATION, UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 279 Renewed License No. NPF-49

1.

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

A.

The application for amendment by Dominion Energy Nuclear Connecticut, Inc.

(DENC, the licensee), dated November 5, 2020, as supplemented by letter dated May 20, 2021, 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.

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, revised through Amendment No. 279 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 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION James G. Danna, 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: October 5, 2021 James G.

Danna Digitally signed by James G. Danna Date: 2021.10.05 13:17:29 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 279 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 a marginal line indicating the area 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 6-20 6-20 6-20a 6-20a 6-20b 6-20b 6-21 6-21 6-21a 6-21a 6-21b 6-21b (2)

Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 279 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.

Renewed License No. NPF-49 Amendment No. 270, 271, 272, 273, 275, 276, 277, 278 279

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 Rod Insertion Limit
  • 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.)

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

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

6.

WCAP-16009-P-A, REALISTIC LARGE-BREAK LOCA EVALUATION METHODOLOGY USING THE AUTOMATED STATISTICAL TREATMENT OF UNCERTAINTY METHOD (ASTRUM), (W Proprietary). (Methodology for Specification 3.2.2.1--Heat Flux Hot Channel Factor.)

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 Amendment No. 81, 170, 218, 229, 236, 253, 268, 6-20a 279

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 Rod Insertion Limit
  • 3.2.2.1 Heat Flux Hot Channel Factor
  • 3.2.3.1 Nuclear Enthalpy Rise Hot Channel Factor
  • 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, 6-20b 279

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

23.

DOM-NAF-2-P-A, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code, including Appendix C, Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code, and Appendix D, Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code. Methodology for Specifications:

  • 3.2.3.1 RCS Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor
  • 3.2.5 DNB Parameters 6.9.1.6.c The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.

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

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.

Degradation mechanisms found, c.

Nondestructive examination techniques utilized for each degradation mechanism, d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications, e.

Number of tubes plugged during the inspection outage for each degradation mechanism, f.

The number and percentage of tubes plugged to date and the effective plugging percentage in each steam generator.

Amendment No. 24, 40, 50, 69, 104, 173, 212, 215, 229, 238, 245, 249. 252, 255, 256, 6-21 279

MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT (Continued) g.

The results of condition monitoring, including the results of tube pulls and in-situ

testing, h.

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, i.

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 j.

The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

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.

Amendment No. 238, 245, 249, 252, 255, 256, 6-21a 279

MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS 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:

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 Amendment No. 245, 6-21b 279

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 279 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC., ET AL MILLSTONE POWER STATION, UNIT NO. 3 DOCKET NO. 50-423

1.0 INTRODUCTION

By letter dated November 5, 2020 (Reference 1), to the U.S. Nuclear Regulatory Commission (NRC, the Commission), as supplemented by letter dated May 20, 2021 (Reference 2),

Dominion Energy Nuclear Connecticut, Inc. (DENC, the licensee) submitted a license amendment request (LAR) for Millstone Power Station, Unit No. 3 (Millstone 3). The proposed amendment would revise Technical Specification (TS) 6.9.1.6.b by adding topical report (TR)

WCAP-16996-P-A, Revision 1, Realistic LOCA [loss-of-coolant accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (Full Spectrum LOCA Methodology)

(Reference 3), to the list of methodologies approved for reference in the Core Operating Limits Report (COLR) for Millstone 3. The reference proposed to be added identifies the analytical method used to determine the core operating limits for the large break loss-of-coolant accident (LBLOCA) event described in the Millstone 3 Final Safety Analysis Report (FSAR),

Section 15.6.5, Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary. The proposed amendment would also remove obsolete COLR Reference WCAP-12945-P-A (Reference 4), which is no longer being used to support the Millstone 3 core reload analysis.

The supplemental letter dated May 20, 2021, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on December 29, 2020 (85 FR 85677).

2.0 REGULATORY EVALUATION

2.1 Description of Proposed TS Changes The COLR references list in TS 6.9.1.6.b lists methodologies used to determine the core operating limits for Millstone 3. The licensee proposed to remove the following methodology, which is Reference 5 in the COLR references list:

WCAP-12945-P-A, CODE QUALIFICATION DOCUMENT FOR BEST ESTIMATE LOCA ANALYSIS, (W Proprietary). (Methodology for Specification 3.2.2.1--Heat Flux Hot Channel Factor.)

The licensee proposed to add the following methodology:

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.)

The licensee stated that the current Reference 5 in the TS 6.9.1.6.b list is a legacy document that is no longer in use at Millstone 3. The licensee also stated that the LOCA methodologies of TR WCAP-16009-P-A (Reference 6), which is Reference 6 in the COLR references list, will no longer be used following the implementation of the full spectrum LOCA evaluation model (EM)

(FSLOCA EM) described in WCAP-16996-P-A (Reference 3). However, WCAP-16009-P-A (Reference 6) is proposed to be retained as a reference in the TS 6.9.1.6.b list to allow for an orderly transition to the FSLOCA EM methodology while considering the effect of thermal conductivity degradation (TCD) of the fuel pellets using the WCAP-17642-P-A, Westinghouse Performance Analysis and Design Model (PAD5) (Reference 7), methodology in a subsequent reload cycle.

2.2 Description - Plant Type As stated in FSAR, Revision 33, Chapter 1, the Millstone 3 pressurized-water reactor (PWR) incorporates a four-loop closed-cycle pressurized-water-type nuclear steam supply system.

The reactor is enclosed in a reinforced concrete containment structure maintained at a sub-atmospheric pressure during plant operation.

2.3 Applicable Regulatory Requirements and Guidance The NRC staff considered the following regulations during its review of the proposed changes.

The regulations in Title 10 of the Code of Federal Regulations (10 CFR) 50.36, Technical specifications, state, in part, that each operating license will include TSs and that the TSs will include items in specific categories, including administrative controls. 10 CFR 50.36(c)(5),

Administrative controls, defines administrative controls as 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. This applies to the list of references to approved methods to be used to determine the core operating limits contained in the COLR.

The regulations in 10 CFR 50.46(b) require that during a LOCA event, the following criteria are satisfied:

(1) Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200 degrees Fahrenheit (°F).

(2) Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

(3) Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

(4) Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

(5) Long-term cooling. After any calculated successful initial operation of the ECCS

[emergency core cooling system], the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

The regulations in 10 CFR Part 50, Appendix A, general design criterion (GDC) 35, Emergency core cooling, state, in part, that a system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

The NRC staff considered the following guidance on acceptable approaches to demonstrate that the above regulations are met.

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, Section 15.6.5, Revision 3, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, March 2007 (Reference 5).

Regulatory Guide (RG) 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, May 1989 (Reference 8).

RG 1.203, Transient and Accident Analysis Methods, December 2005 (Reference 9).

Information Notice 2011-21, Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation, December 13, 2011 (Reference 10).

Generic Letter (GL) 1988-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, October 4, 1988 (Reference 11).

3.0 TECHNICAL EVALUATION

The NRC staff evaluation of the proposed changes in the LAR (Reference 1) consisted of the following: (a) whether the proposed TS changes are acceptable, (b) whether the revised FSLOCA analysis continues to satisfy the regulations stated in Section 2.3 above and whether the FSLOCA EM is appropriately applied, (c) whether the acceptance criteria of 10 CFR 50.46(b)(1) through (4) are satisfied and whether the acceptance criterion of 10 CFR 50.46(b)(5) continues to be satisfied using the current methodology because the FSLOCA EM (Reference 3) is not approved by the NRC for LBLOCA long-term cooling analysis, (d) whether there is compliance with GDC 35, and (e) whether the limitations and conditions specified in Table 22 of the NRC staffs safety evaluation report (SER) for WCAP-16996-P-A (Reference 3) are satisfied.

3.1 Evaluation of Proposed TS Changes Millstone 3 TS 6.9.1.6.a requires that core operating limits be established and documented in the COLR before each reload cycle or any remaining part of a reload cycle. TS 6.9.1.6.b states that the analytical methods used to determine the core operating limits shall be previously reviewed and approved by the NRC, and the COLR will contain the complete identification for each of the referenced TRs used to prepare the COLR.

The proposed change would add to TS 6.9.1.6.b the WCAP-16996-P-A (Reference 3) methodology, which would become the new method for LBLOCA analysis and would replace the WCAP-16009-P-A (Reference 6) methodology after transition in a subsequent reload cycle.

The NRC staff considers the proposed TS change to be acceptable because WCAP-16996-P-A is an NRC-approved methodology and would continue to provide administrative controls consistent with 10 CFR 50.36(c)(5).

The proposed change would allow the LBLOCA (Region II) part of the WCAP-16996-P-A (Reference 3) methodology to be utilized for future core reload analyses. This change assures that the core operating limits have been calculated in accordance with an NRC-approved method. The currently used WCAP-16009-P-A LBLOCA analysis methodology would be retained in the TSs to allow transition to the new method between reload cycles. The NRC staff considers the proposed change to TS 6.9.1.6.b to revise the list of methodologies to reflect those necessary for future cycle-specific operating limits to be acceptable because the change would continue to provide administrative controls consistent with 10 CFR 50.36(c)(5).

The licensee stated that the current TS 6.9.1.6.b includes WCAP-12945-P-A (Reference 4),

which is no longer used to support Millstone 3 reload analyses. The licensee proposed to delete this legacy document from the TS. The NRC staff considers the proposed deletion of the unused method in the TS to be acceptable because the proposed change is administrative in nature and WCAP-12945-P-A is superseded by other NRC-approved methods for the same purpose. Therefore, the proposed change would continue to meet 10 CFR 50.36(c)(5).

Information Notice 2011-21 (Reference 10) notified addressees of concerns regarding the impact of irradiation on fuel thermal conductivity and its potential to result in significantly higher predicted peak cladding temperature (PCT) in realistic ECCS EMs. The Millstone 3 current best-estimate LBLOCA does not explicitly consider the fuel pellet TCD and peaking factor burndown. In a letter dated November 29, 2012 (Reference 12), the licensee committed to adopt an NRC-approved LBLOCA analysis method that would include the effects of TCD of fuel pellets. The proposed change, which would add the NRC-approved WCAP-16996-P-A (Reference 3) to TS 6.9.1.6.b, includes the effect of TCD. The NRC staff finds the proposed changes to the TS 6.9.1.6.b COLR references acceptable because the to-be-added WCAP-16996-P-A (Reference 3) addresses the licensees commitment of incorporating the effects of TCD of fuel pellets into the revised LBLOCA analysis.

GL 1988-16 (Reference 11) states that it is acceptable for licensees to control reactor physics parameter limits by specifying an NRC-approved calculation methodology. These parameter limits may be removed from the TSs and placed into a cycle-specific COLR that is required to be submitted to the NRC every operating cycle or each time that it is revised. The guidance in GL 1988-16 will continue to be met because the proposed changes to TS 6.9.1.6.b identify the NRC-approved methodologies that are used to determine the Millstone 3 core operating limits.

3.2 LBLOCA (Region II) Analysis The FSLOCA EM methodology divides the break spectrum into the two regions, termed Region I and Region II. The Region I analysis is for small-break LOCAs (SBLOCAs) and the Region II analysis is for LBLOCAs. According to the FSLOCA EM, the SBLOCA (Region I) and LBLOCA (Region II) analyses are independent, separable, and do not influence each other.

The licensee only provided the LBLOCA (Region II) analysis in this LAR (Reference 1).

Methodology The previous NRC-approved best-estimate LBLOCA analysis (Westinghouse) methodology described in WCAP-16009-P-A (Reference 6) is termed the Automated Statistical Treatment of Uncertainty Method (ASTRUM) EM. This methodology is applicable to the Westinghouse-designed (a) 2-loop and 4-loop PWRs with ECCS injection into the reactor coolant system (RCS) cold legs and (b) 2-loop PWRs with upper plenum injection. The ASTRUM EM uses WCOBRA/TRAC as the analysis code and is applicable only for the LBLOCA analysis with a minimum break size of 1.0 square-foot (ft2). The Millstone 3 current LBLOCA licensing analysis is based on ASTRUM EM.

As stated in Section 1.2.2 of the NRC staffs SER for WCAP-16996-P-A (Reference 3), the FSLOCA EM is built on the ASTRUM EM by extending the applicability of the WCOBRA/TRAC code to include the full spectrum of break sizes postulated in the RCS cold leg. The break sizes considered include any size in which the break flow is beyond the capacity of the normal charging pumps, up to and including a double ended guillotine rupture in the RCS cold leg with a break flow area equal to two times the pipe flow area.

The FSLOCA EM methodology uses the WCOBRA/TRAC-TF2 code to analyze the RCS thermal-hydraulic response for the full spectrum of break sizes. This methodology is applicable to Westinghouse 2-, 3-, and 4-loop plants with RCS cold leg injections. Since Millstone 3 is a Westinghouse-designed 4-loop plant with ECCS injection into the cold legs of the RCS, the FSLOCA EM is applicable.

The licensee proposed to use the FSLOCA EM methodology to meet its commitment in letter dated November 29, 2012 (Reference 12), by updating its licensing basis analysis to account for the TCD for the LBLOCA analyses only. The FSLOCA EM explicitly accounts for the effects of fuel pellet TCD by using the fuel rod performance input data generated by the PAD5 code (Reference 7). The fuel pellet thermal conductivity input to the WCOBRA/TRAC-TF2 code used in the FSLOCA EM accounts for the TCD of fuel pellets, thereby addressing Information Notice 2011-21 (Reference 10) and satisfying the licensees commitment in letter dated November 29, 2012 (Reference 12).

For the calculation of the transient containment minimum back pressure at the break as an input to the LBLOCA analysis, the licensee used the NRC-approved Westinghouse Containment Pressure Analysis Code, COCO, methodology (Reference 13) by integrating it into the WCOBRA/TRAC-TF2 code. The mass and energy (M&E) release calculated by the WCOBRA/TRAC-TF2 code at the end of each timestep in the LBLOCA transient is transferred as an input to the COCO for calculating the containment back pressure as a boundary condition at the break. The COCO described in WCAP-8327 (Reference 13) is an NRC-approved code for calculating the containment pressure response.

Based on the above, the NRC staff finds that the licensee: (a) appropriately applied the proposed FSLOCA EM for the LBLOCA analysis in order to satisfy 10 CFR 50.46(b)(1) through (b)(4) and (b) used the NRC-approved COCO for calculating the transient minimum containment back pressure at the break.

Analysis The licensee performed the LBLOCA analyses assuming both LOOP and offsite power available (OPA). The licensee listed the following in the LAR (Reference 1), Attachment 3: (a) the major plant parameters and inputs in Tables 1 through 4, (b) the sequence of events and the sampled values of the decay heat uncertainty multiplier used in the analysis in Tables 6 and 7, respectively, and (c) the input data for calculating the containment minimum back pressure as a boundary condition at the break in Tables 2 and 3. Consistent with the FSLOCA EM, while considering the heat sinks and all trains of the pressure reducing (containment spray) system and using the COCO, the licensee calculated a conservative minimum transient containment back pressure at the break.

In its supplemental letter dated May 20, 2021 (Reference 2), for the LBLOCA uncertainty analysis, the licensee referred to Section 30 of WCAP-16996-P-A (Reference 3) for the method used. The licensee used a Monte Carlo sampling of all uncertainty contributors that led to the results from which the upper tolerance limits are derived for the PCT, maximum local oxidation (MLO), and core-wide oxidation (CWO). In this letter, the licensee provided some of the sampled uncertainty contributors used in the statistical analysis to determine the 95/95 upper tolerance limits on these parameters.

Westinghouse letters to the NRC dated July 18, 2018 (Reference 14), and February 7, 2019 (Reference 15), reported errors and some changes in the FSLOCA EM. The NRC staff noted that these letters did not report the error that impacts the gamma energy redistribution multiplier identified in a Virginia Electric and Power Company letter to the NRC dated August 31, 2020 (Reference 16). In its supplemental letter dated May 20, 2021 (Reference 2), the licensee confirmed that the FSLOCA EM for Millstone 3 used the WCOBRA/TRAC-TF2 code, which incorporates the error corrections and changes identified in the above Westinghouse letters.

Regarding the error that impacts the gamma energy redistribution multiplier, the licensee stated that this was an error in the North Anna, Unit Nos. 1 and 2, plant-specific implementation of the methodology and not in the FSLOCA EM or the WCOBRA/TRAC-TF2 code. Therefore, this error was not reported in the above Westinghouse letters as an FSLOCA EM methodology error. The licensee stated that this error did not occur in the Millstone 3 analysis; therefore, the NRC-approved FSLOCA EM methodology related to gamma energy redistribution multiplier was correctly applied.

The NRC staff finds that the licensees LBLOCA analysis is acceptable because of the following:

The assumptions and key inputs, which include core parameters, RCS parameters, and containment parameters, are consistent with the plant configuration and current licensing basis.

The uncertainty analysis is consistent with the FSLOCA EM.

The licensee used the corrected FSLOCA EM after removing the errors reported in Westinghouse letters to the NRC dated July 18, 2018 (Reference 14), and February 7, 2019 (Reference 15).

The error that impacts the gamma energy redistribution multiplier identified in a Virginia Electric and Power Company letter to the NRC dated August 31, 2020 (Reference 16),

does not apply to Millstone 3.

Results Table 1 below (same as Table 5 in Attachment 3 to the LAR (Reference 1)) summarizes the Millstone 3 LBLOCA analysis results for PCT, MLO, and CWO under the conditions of LOOP and OPA. The maximum values of PCT, MLO, and CWO are 1712 °F, 7.82 percent, and 0.12 percent, respectively, and are in the OPA case. The licensee stated that these results were determined using the calculated total MLO (pre-transient plus transient oxidation), the PCT as shown in the supplemental letter dated May 20, 2021 (Reference 2), Figures Request for Additional Information (RAI)-2-1 and RAl-2-3, and the calculated CWO. These maximum values are less than their respective acceptance criteria in 10 CFR 50.46.

The containment back pressure for the transient that produced the analysis PCT result is provided in the LAR (Reference 1), Attachment 3, Figure 9, for the case assuming OPA.

Table 1: Results Results LBLOCA Analysis Value Under LOOP LBLOCA Analysis Value Under OPA Acceptance Criteria in 10 CFR 50.46 95/95 PCT 1661°F 1712°F (b)(1) 2200°F 95/95 MLO 7.81%

7.82%

(b)(2) 17%

95/95 CWO 0.08%

0.12%

(b)(3) 1%

3.3 Compliance with 10 CFR 50.46 The results in Table 1 for the PCT, MLO, and CWO show significant margins from the acceptable values given in 10 CFR 50.46(b)(1), (b)(2), and (b)(3), respectively. Regarding compliance with 10 CFR 50.46(b)(4), the NRC staff determined that coolable core geometry is maintained in the LBLOCA because: (i) the acceptance criteria 10 CFR 50.46(b)(1) through (b)(3) are satisfied and (ii) based on the information in the supplemental letter dated May 20, 2021 (Reference 2), which states that, based on Section 32.1 of the FSLOCA EM, the effects of LOCA and seismic loads on core geometry do not need to be considered unless the fuel assembly grid deformation extends beyond the core periphery (i.e., deformation in a fuel assembly with no sides adjacent to the core baffle plates). The licensee stated that the current core grid analysis of record related to the combined LOCA and seismic loads is not affected, and this conclusion is retained in the Millstone 3 FSAR, Revision 33, Section 4.2.3.4.1, which states as follows:

The maximum grid impact forces for both the seismic accident and asymmetric LOCA occur at the peripheral fuel assembly locations adjacent to the baffle wall.

The maximum grid impact forces result from postulated LOCA and seismic loadings, and are required to be less than the allowable grid crush strength. A calculation of the maximum LOCA and seismic grid impact forces, combined with using the square root sum of the squares method in accordance with SRP 4.2, Appendix A, demonstrated that the maximum value is below the allowable grid strength.

For compliance with 10 CFR 50.46(b)(5), the licensee did not utilize the FSLOCA EM because it is not NRC-approved for LBLOCA long term analyses. Therefore, for continued compliance with 10 CFR 50.46(b)(5), the current LBLOCA analysis for long term cooling performed using the NRC-approved ASTRUM WCAP-16009-P-A (Reference 6) methodology is maintained as the licensing basis as stated in Millstone 3 FSAR, Revision 33, Section 15.6.5.2.6, under compliance with 10 CFR 50.46(b)(5).

Based on the above, the NRC staff finds that the proposed LBLOCA (Region II) analysis results for Millstone 3 are in compliance with 10 CFR 50.46(b)(1) through (b)(4) and, therefore, acceptable. For the LBLOCA long term cooling, the NRC staff finds it acceptable that the licensee proposes to maintain the current analysis and its results as the licensing basis because they are in compliance with 10 CFR 50.46(b)(5).

3.4 Compliance with GDC 35 Section 6.3 of the Millstone 3 FSAR, Revision 33, describes the ECCS. The NRC staff finds that the conformance with GDC 35 described in Section 3.1.2.35 of the FSAR, Revision 33, is unaffected by the proposed change because during a LBLOCA, as determined by the FSLOCA EM and the currently used ASTRUM (Reference 6) methodology for long term core cooling, effective core cooling will continue while the cladding oxidation due to cladding metal-water reaction will be within the acceptable limits as shown in Table 1 above.

3.5 Limitations and Conditions The NRC staffs SER for WCAP-16996-P-A (Reference 3), Table 22 provides limitations and conditions required to be satisfied in order for the licensee to implement the NRC-approved FSLOCA EM. The licensee summarized each limitation and condition in the LAR (Reference 1),

which are provided below. The NRC staff reviewed the licensees summaries against the actual limitations and conditions documented in Table 22 of the above NRC staff SER and determined that the summaries are acceptable because the licensee appropriately described the specific requirement of each limitation and condition. The NRC staffs evaluation of whether each limitation and condition is satisfied for the Millstone 3 LBLOCA analysis is provided below.

Limitation and Condition 1: FSLOCA EM Applicability with Regard to LOCA Transient Phases Summary:

The FSLOCA EM is not approved to demonstrate compliance with 10 CFR 50.46 acceptance criterion (b)(5) related to the long-term cooling.

NRC Staff Evaluation

The licensees analysis using the FSLOCA EM is only used to demonstrate compliance with 10 CFR 50.46(b)(1) through (b)(4). The FSLOCA EM does not treat boric acid precipitation and therefore lacks the capability to adequately address post-LOCA long-term core cooling. The NRC staff determined that the licensee met this limitation and condition because the FSLOCA EM is not used to demonstrate compliance with 10 CFR 50.46(b)(5) for long-term core cooling.

Limitation and Condition 2: FSLOCA EM Applicability with Regard to Type of PWR Plants Summary:

The FSLOCA EM is approved for the analysis of Westinghouse-designed 3-loop and 4-loop PWRs with cold-side injection. Analyses should be executed consistent with the approved method, or any deviations from the approved method should be described and justified.

NRC Staff Evaluation

The NRC staff determined that the licensee met this limitation and condition because the FSLOCA EM is an NRC-approved methodology used for the LBLOCA analysis without any deviations for Millstone 3, which is a Westinghouse-designed 4-loop PWR with cold-side injection.

Limitation and Condition 3: FSLOCA EM Applicability for Containment Pressure Modeling Summary:

For Region II, the containment pressure calculation will be executed in a manner consistent with the approved methodology (i.e., the COCO or LOTIC2 model will be based on appropriate plant-specific design parameters and conditions, and engineered safety features which can reduce pressure are modeled). This includes utilizing a plant-specific initial containment temperature, and only taking credit for containment coatings which are qualified and outside of the break zone-of-influence.

NRC Staff Evaluation

The NRC staff reviewed the information provided in Attachment 3 of the LAR (Reference 1) and determined that the LOCA containment pressure response analysis is performed using NRC-approved Westinghouse COCO methodology (Reference 13). For this analysis, the licensee integrated the COCO into the WCOBRA/TRAC-TF2 thermal-hydraulic code. The licensee stated that appropriate input parameters were used for a conservatively low containment back pressure as a boundary condition at the break. A plant-specific minimum initial temperature associated with normal full-power operating conditions was modeled and, conservatively, no coatings were credited on any of the containment structures.

The NRC staff determined that the licensee met limitation and condition 3 because: (a) the licensee used an acceptable plant-specific initial temperature for the containment pressure response, (b) the licensee used NRC-approved methodology for the LBLOCA (Region II) containment pressure calculation using inputs that minimized the containment back pressure as a boundary condition to the break for a conservative PCT calculation, and (c) the licensee did not credit any coatings on any of the containment structures.

Limitation and Condition 4: Decay Heat Modeling in FSLOCA EM Applications Summary:

The decay heat uncertainty multiplier will be sampled consistent with the NRC-approved methodology for the FSLOCA EM. The analysis simulations for the FSLOCA EM will not be executed for longer than 10,000 seconds following reactor trip unless the decay heat model is appropriately justified. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results will be provided in the analysis submittal in units of sigma and absolute units.

NRC Staff Evaluation

The NRC staff reviewed Table 7 in Attachment 3 of the LAR (Reference 1) and confirmed that the licensee provided the sampled value of the decay heat uncertainty multiplier (DECY_HT),

which are absolute values in units of sigma (). The analysis simulations were all executed for less than 10,000 seconds following reactor trip.

The NRC staff determined that that the licensee appropriately modeled decay heat per limitation and condition 4 and correctly reported the resulting sampled values in units of and absolute units for the limiting cases and, therefore, the licensee met this limitation and condition.

Limitation and Condition 5: Fuel Burnup Limits in FSLOCA EM Applications Summary:

The maximum assembly and rod length-average burnup must remain below the limits contained in the NRC-approved methodology for the FSLOCA EM.

NRC Staff Evaluation

In Attachment 3 of the LAR (Reference 1), the licensee stated that the maximum analyzed assembly and rod length-average burnups for the Millstone 3 analysis are less than or equal to the limits contained in the NRC-approved methodology for the FSLOCA EM.

Based on the above, the NRC staff determined that the licensee met limitation and condition 5.

Limitation and Condition 6: WCOBRA/TRAC-TF2 Interface with PAD 5.0 in the FSLOCA EM Summary:

The fuel performance data for analyses with the FSLOCA EM should be based on the PAD5 code (at present), which includes the effect of thermal conductivity degradation. The nominal fuel pellet average temperatures and rod internal pressures should be the maximum values, and the generation of all the PAD5 fuel performance data should adhere to the NRC-approved PAD5 methodology.

NRC Staff Evaluation

The NRC staff reviewed Attachment 3 of LAR (Reference 1) and confirmed that the licensee used PAD5 (Reference 7) in the Millstone 3 analysis with the FSLOCA EM. PAD5 is the latest version of the fuel performance code that explicitly models TCD and is benchmarked to high burnup data in Reference 7. The licensee stated that the FSLOCA EM considers the effects of fuel pellet TCD and other burnup-related effects by initializing to fuel rod performance data input generated by the PAD5 code. In the analysis, the fuel pellet average temperatures conservatively bound the maximum values calculated in accordance with Section 7.5.1, Maximum Fuel Temperatures, of Reference 7. The analyzed rod internal pressures were calculated in accordance with Section 7.5.2, Rod Internal Pressure, of Reference 7.

The NRC staff determined that the Millstone 3 analysis for LBLOCA met limitation and condition 6 because the licensee used PAD5, which is the latest version of NRC-approved fuel performance code, and explicitly includes the effect of TCD using conservative inputs.

Limitation and Condition 7: Interfacial Drag Uncertainty in FSLOCA EM Region I Analyses Summary:

The YDRAG uncertainty parameter for Region I analyses should be treated as specified in the NRC-approved methodology for the FSLOCA EM.

NRC Staff Evaluation

The NRC staff determined that limitation and condition 7 is not applicable because it applies to the Region I analysis only, which is not within the scope of the review for this LAR (Reference 1).

Limitation and Condition 8: Biased Uncertainty Contributors in FSLOCA EM Region I Analyses Summary:

Certain uncertainty contributors will be treated for Region I analyses as specified in the NRC-approved methodology for the FSLOCA EM.

NRC Staff Evaluation

The NRC staff determined that limitation and condition 8 is not applicable because it applies to the Region I analysis only, which is not within the scope of the review for this LAR (Reference 1).

Limitation and Condition 9: Effect of Bias in FSLOCA EM Applications for Region I Summary:

For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to confirm modeling selections for Region I analyses.

NRC Staff Evaluation

The NRC staff determined that limitation and condition 9 is not applicable because it applies to the Region I analysis only, which is not within the scope of the review for this LAR (Reference 1).

Limitation and Condition 10: Boundary Between FSLOCA EM Region I and Region II Breaks Summary:

For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to demonstrate that the applied break size boundary for Region I analyses serves the intended goal.

Additionally, the minimum sampled break area for the analysis of Region II should be 1 ft2.

NRC Staff Evaluation

The NRC staff notes that the items to be confirmed in this limitation and condition are: (a) for non-Westinghouse 3-loop PWRs, there is no unexplained behavior in the predicted safety criteria, including PCT, that occurs across the boundary between Region I and Region II and (b) for Region II analysis, the sampled break size should be 1 ft2.

For item (a), in Reference 17 the licensee reported the results of a sensitivity study by applying the FSLOCA EM to a 4-loop Westinghouse PWR and confirmed that there is no unexplained behavior in the predicted safety criteria, including PCT, that occurs across the boundary between Region I and Region II. For item (b), by reviewing the LAR (Reference 1), the NRC staff confirmed that the licensee performed the Region II analysis for a minimum break area of 1 ft2. Therefore, the licensee met limitation and condition 10.

Limitation and Condition 11: Limitation and Condition related to FSLOCA EM Uncertainty Analyses for Region II and Documentation of Reanalysis Results for Region I and Region II Summary:

There are various aspects of this Limitation and Condition, which are summarized below:

1. Certain information regarding the Region I and Region II analyses must be declared and documented prior to performing the uncertainty analysis, and will not be changed throughout the remainder of the analysis once they have been declared and documented.
2. If the analysis inputs are changed after they have been declared and documented, for the intended purpose of demonstrating compliance with the applicable acceptance criteria, then the changes and associated rationale for the changes will be provided in the analysis submittal.

Additionally, the preliminary values for PCT, MLO, and CWO which caused the input changes will be provided. These preliminary values are not subject to Appendix B verification, and archival of the supporting information for these preliminary values is not required.

3. Plant operating ranges which are sampled within the uncertainty analysis will be provided in the analysis submittal for both regions.

NRC Staff Evaluation

The licensee confirmed that the information specified for the uncertainty analysis in the NRC-approved FSLOCA EM, including the analysis inputs, was declared and documented prior to analysis and was not changed after being declared and documented. The licensee provided the Millstone 3 plant operating ranges that were sampled within the uncertainty analysis in Table 1 in Attachment 3 of the LAR (Reference 1). Based on this information, the NRC staff determined that the licensee met limitation and condition 11.

Limitation and Condition 12: Steam Generator Heat Removal During SBLOCAs Summary:

The plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves must be adequately accounted for in analysis with the FSLOCA EM.

NRC Staff Evaluation

The licensee stated that this limitation and condition applies to Region I (SBLOCA) transients only and does not apply to Region II transients. The NRC staff agrees with this statement because the Region II (LBLOCA) transient would not result in secondary-side pressurization such that the main steam safety valve setpoint pressures would be reached. Therefore, the NRC staff determined that limitation and condition 12 is not applicable because it applies to the Region I analysis only, which is not within the scope of the review for this LAR (Reference 1).

Limitation and Condition 13: Upper Head Spray Nozzle Loss Coefficient Summary:

In plant-specific models for analysis with the FSLOCA EM, specific modeling considerations for the upper head spray nozzles should be followed as required by the NRC-approved methodology.

NRC Staff Evaluation

The licensee stated that the specific requirements for the upper head spray nozzles were adhered to in the Millstone 3 Region II analysis. The NRC staff therefore determined that the licensee met limitation and condition 13.

Limitation and Condition 14: Correlation for Oxidation Summary:

For analyses with the FSLOCA EM to demonstrate compliance against the current 10 CFR 50.46 oxidation criterion, the transient time-at-temperature will be converted to an equivalent cladding reacted (ECR) using either the Baker-Just or the Cathcart-Pawel correlation. In either case, the pre-transient corrosion will be summed with the LOCA transient oxidation. If the Cathcart-Pawel correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 13 percent limit. If the Baker-Just correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 17 percent limit.

NRC Staff Evaluation

The licensee used the Baker-Just correlation to convert the LOCA transient time-at-temperature into an ECR. The licensee met the 10 CFR 50.46(b)(2) MLO acceptance criterion of 17 percent after summing up the pre-existing corrosion with the resulting LOCA transient ECR. Therefore, the NRC staff determined that the licensee met limitation and condition 14.

Limitation and Condition 15: LOOP versus OPA Treatment in FSLOCA EM Uncertainty Analyses for Region II Summary:

The Region II analysis will be executed twice; once assuming loss-of-offsite power (LOOP) and once assuming offsite power available (OPA). The results from both analysis executions should be shown to be in compliance with the 10 CFR 50.46 acceptance criteria.

The statistical analysis must adhere to the Limitation and Condition as specified in the NRC-approved methodology for the FSLOCA EM.

NRC Staff Evaluation

The NRC staff review of LAR (Reference 1), Attachment 3, confirmed that the Region II analysis for Millstone 3 was performed for LOOP and OPA cases. The licensee performed the statistical analysis for both cases in accordance with the FSLOCA EM. The results from both analyses are in compliance with the 10 CFR 50.46 acceptance criteria as evaluated in Section 3.3 above.

Therefore, the NRC staff determined that the licensee met limitation and condition 15.

3.5 NRC Staff Conclusion

The licensee proposed to modify TS 6.9.1.6.b by replacing the existing reference for TR WCAP-12945-P-A (Reference 4) with the FSLOCA EM. Based on the above evaluation of the proposed change, the NRC staff concludes:

The licensee has appropriately applied the FSLOCA EM and the analysis results satisfy the 10 CFR 50.46(b)(1) through (b)(4) requirements. All applicable limitations and conditions for the LBLOCA analysis are met.

The licensee did not implement the FSLOCA EM to confirm compliance with 10 CFR 50.46(b)(5) because it is not approved for long-term core cooling. The current compliance with 10 CFR 50.46(b)(5) based on the currently used ASTRUM (Reference 6) methodology remains as the licensing basis.

Conformance with GDC 35 is unaffected.

The licensee has implemented the Information Notice 2011-21 (Reference 10) on the TCD for the LBLOCA analysis while confirming compliance with 10 CFR 50.46(b)(1) on PCT.

The proposed deletion of the obsolete COLR reference WCAP-12945-P-A (Reference 4) is acceptable because it is no longer used.

The guidance in GL 1988-16 (Reference 11) continues to be implemented because the proposed change specifies the NRC-approved methodology for the determination of core operating limits.

The 10 CFR 50.36(c)(5) requirement is satisfied because the licensee added the NRC-approved FSLOCA EM in the TS COLR reference list as a provision for administrative controls.

Based on the above conclusions, the NRC staff finds the proposed change to TS 6.9.1.6.b acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Connecticut State official was notified on August 10, 2021, of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. 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 NRC has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on this finding (December 29, 2020; 85 FR 85677). Accordingly, the amendment meets 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 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.

7.0 REFERENCES

1.

Dominion Energy Nuclear Connecticut, Inc. letter to NRC, Millstone Power Station Unit 3, Proposed License Amendment Request, Addition of Analytical Methodology to the Core Operating Limits Report for a Large Break Loss of Coolant Accident (LBLOCA),

November 5, 2020 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML20310A324).

2.

Dominion Energy Nuclear Connecticut, Inc. letter to NRC, Millstone Power Station Unit 3, Response to Request for Additional Information for Proposed License Amendment Request to Add an Analytical Methodology to the Core Operating Limits Report for a Large Break Loss of Coolant Accident, May 20, 2021 (ADAMS Accession Nos. ML21140A299 and ML21140A300 (non-publicly available)).

3.

Westinghouse Electric Company (Westinghouse) letter to NRC, Submittal of WCAP-16996-P-A/WCAP-16996-NP-A, Volumes I, II, III and Appendices, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), October 2, 2017 (ADAMS Accession Nos.

ML17277A131, ML17277A132/ML17277A143 (non-publicly available),

ML17277A133/ML17277A144 (non-publicly available), ML17277A134/ML17277A149 (non-publicly available), and ML17277A135/ML17277A150 (non-publicly available)).

4.

Westinghouse letter to NRC, Transmittal of WCAP-12945-P-A and WCAP-14449-P-A, Addendum 1-A, Revision 0, Method for Satisfying 10 CFR 50.46 Reanalysis Requirements for Best-Estimate LOCA Evaluation Models, December 13, 2004 (ADAMS Accession No. ML043550249).

5.

NRC, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, Section 15.6.5, Revision 3, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, March 2007 (ADAMS Accession No. ML070550016).

6.

Westinghouse letter to NRC, Transmittal of WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM), March 11, 2005 (ADAMS Accession No. ML050910157).

7.

Westinghouse letter to NRC, Submittal of WCAP-17642-P-A/WCAP-17642-NP-A, Revision 1 Westinghouse Performance Analysis and Design Model (PAD5),

(Proprietary/Non-Proprietary), November 27, 2017 (ADAMS Package Accession No. ML17335A334).

8.

NRC, Regulatory Guide 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, May 1989 (ADAMS Accession No. ML003739584).

9.

NRC, Regulatory Guide 1.203, Transient and Accident Analysis Methods, December 2005 (ADAMS Accession No. ML053500170).

10. NRC, Information Notice 2011-21, Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation, December 13, 2011 (ADAMS Accession No. ML113430785).
11. NRC, GL 1988-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, October 4, 1988 (ADAMS Accession No. ML031130447).
12. Dominion Nuclear Connecticut, Inc. letter to NRC, Millstone Power Station Unit 3, 30-Day Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the Requirements of 10 CFR 50.46, November 29, 2012 (ADAMS Accession No. ML12340A010).
13. Westinghouse, WCAP-8327, Containment Pressure Analysis Code (COCO), July 1974 (ADAMS Accession No. ML092460709 (non-publicly available)).
14. Westinghouse letter to NRC, U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2017, July 18, 2018 (ADAMS Accession No. ML19288A174).
15. Westinghouse letter to NRC, U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2018, February 7, 2019 (ADAMS Package Accession No. ML19042A378).
16. Virginia Electric and Power Company letter to NRC, North Anna Power Station Units 1 and 2, Proposed License Amendment Request, Addition of Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident (FSLOCA)

Gamma Energy Redistribution Information, August 31, 2020 (ADAMS Accession No. ML20244A336).

17. Westinghouse letter to NRC, Information to Satisfy the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology Plant Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs) (Proprietary/Non-Proprietary), July 13, 2018 (ADAMS Package Accession No. ML18198A038).

Principal Contributor:

A. Sallman Date of Issuance: October 5, 2021

ML21227A000 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LAiT NRR/DORL/LPL3/LA NAME RGuzman KEntz SRohrer DATE 8/18/2021 8/30/2021 8/31/2021 OFFICE NRR/DSS/SNSB/BC NRR/DSS/STSB/BC (A)

OGC - NLO NAME SKrepel NJordan JWachutka DATE 7/9/2021 8/25/2021 09/15/2021 OFFICE NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME JDanna RGuzman DATE 10/5/2021 10/5/2021