Text

Proposed Amendment to Support Implementation of Framatome Gaia Fuel
	ML23304A047
Person / Time
Site:	Millstone
Issue date:	10/30/2023
From:	James Holloway; Dominion Energy Nuclear Connecticut
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
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ML23304A045	List: ML23304A047;
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	23-251
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PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 Dominion Energy Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 DominionEnergy.com October 30, 2023 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Dominion p Energy Serial No.

NRA/SS:

Docket No.

License No.23-251 RO 50-423 NPF-49 PROPOSED AMENDMENT TO SUPPORT IMPLEMENTATION OF FRAMATOME GAIA FUEL Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENG) is submitting a License Amendment Request (LAR) to the U. S. Nuclear Regulatory Commission (USNRC) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3). The proposed LAR revises the TS to support implementation of Framatome GAIA fuel, which is currently scheduled for onload during the spring 2025 refueling o"utage. The proposed TS changes include updating the Reactor Core Safety Limits (TS 2.1.1.1 ), reducing the Reactor Trip System Instrumentation Trip Setpoint for the Permissive-8 (P-8) Interlock (TS Table 2.2-1, Item 18.c), and adding to the list of approved methodologies for the Core Operating Limits Report (COLR) (TS 6.9.1.6.b).

Additionally, DENG requests USNRC approval of mixed-core Departure from Nucleate Boiling (DNB) penalties for application to retained DNB margin. The transition core penalties (TCPs) address the thermal-hydraulic effects resulting from physical differences between the resident Westinghouse fuel and prospective Framatome GAIA fuel. The LAR also seeks USNRC approval for the use of Design Basis Limits for a Fission Product Barrier (DBLFPBs) associated with MPS3 specific application of methods needed to support GAIA fuel implementation. provides the description and assessment of the proposed changes. provides the marked-up MPS3 TS pages to reflect the proposed changes. provides the Statistical Departure from Nucleate Boiling (DNB) Design Limits report for an MPS3 core with GAIA fuel assemblies. Attachment 4 provides the proprietary text excerpts which are redacted from the License Amendment Request (LAR) text in Attachment 1 and the supporting document in Attachment 3. Attachment 5 provides an Application for Withholding and Affidavit from Framatome Inc. contains information proprietary to Framatome Inc. (Framatome) and is supported by an affidavit (Attachment 5) signed by Framatome, the owner of the information.

The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations. Accordingly, it is contains information that is being withheld from public disclosure under 10 CFR 2.390. Upon separation from Attachment 4, this letter is decontrolled.

Serial No.23-251 Docket No. 50-423 Page 2 of 4 respectfully requested the proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390.

The proposed amendment does not involve a Significant Hazards Consideration under the standards set forth in 10 CFR 50.92. The basis for this determination is included in. DENC has also determined that operation with the proposed changes will not result in any significant increase in the amount of effluents that may be released offsite, or any significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 1 O CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with approval of the proposed amendment.

The proposed amendment has been reviewed and approved by the station's Facility Safety Review Committee.

DENC requests approval of this LAR by October 31, 2024, with implementation in conjunction with MPS3 fuel cycle 24 (currently planned for the spring of 2025).

In accordance with 10 CFR 50.91(b), a copy of this LAR is being provided to the State of Connecticut.

Serial No.23-251 Docket No. 50-423 Page 3 of 4 If you have any questions or require additional information, please contact Mr. Shayan Sinha at (804) 273-4687.

Sincerely, James E. Holloway Vice President - Nuclear Engineering and Fleet Support COMMONWEAL TH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by James E. Holloway who is Vice President - Nuclear Engineering and Fleet Support of Dominion Energy Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this.?P~ay of CJc...\\ob.e..r, 2023.

My Commission Expires: Ou 9wJ:- 3), '2. c:> 2 i.

I Attachments:

1. Description and Assessment of Proposed Changes

2. Marked-up Technical Specifications Pages

3. Development of Statistical Departure from Nucleate Boiling Design Limits

4. Proprietary Information Contained in License Amendment Request and Supporting Document (Proprietary)

5. Framatome Application for Withholding and Affidavit Commitments made in this letter: None

cc:

U.S. Nuclear Regulatory Commission Region I 475 Allendale Road, Suite 105 King of Prussia, PA 19406-1415 Richard V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127 Serial No.23-251 Docket No. 50-423 Page 4 of 4

Serial No.23-251 Docket No. 50-423 DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGES Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.23-251 Docket No. 50-423, Page 1 of 25 Table of Contents 1.0

SUMMARY

DESCRIPTION................................................................................... 2 2.0 DETAILED DESCRIPTION................................................................................... 3 2.1.

Current Licensing Basis Summary and Applicable Technical Specifications..... 3 2.1.1.

Departure from Nucleate Boiling (DNB) Design Basis................................. 3 2.1.2.

Reactor Trip System Interlock Permissive-8 (P-8)....................................... 5 2.1.3.

Rod Ejection................................................................................................ 5 2.2.

Reason for the Proposed Changes.................................................................... 5 2.3.

Description of Proposed Changes...................................................................... 5 2.3.1.

DOM-NAF-2-P-A and VEP-NE-2-A Application to GAIA Fuel in MPS3 Cores.......................................................................................................... 5 2.3.2.

DNBR Penalties Supporting the MPS3 Transition to GAIA Fuel.................. 7 2.3.3.

Reduction of the P-8 Power Range Neutron Flux Nominal Trip Setpoint and Allowable Value........................................................................................... 7 2.3.4.

ANP-10338-P-A Application to GAIA Fuel in MPS3 Cores.......................... 8

3.0 TECHNICAL EVALUATION

.................................................................................. 9 3.1.

DOM-NAF-2-P-A and VEP-NE-2-A Application to GAIA Fuel in MPS3 Cores... 9 3.2.

DNBR Penalties Supporting the MPS3 Transition to GAIA Fuel...................... 10 3.2.1.

General Approach..................................................................................... 10 3.2.2.

Subchannel Code Modeling Considerations.............................................. 11 3.2.3.

Mixed-Core Analysis Performed for the MPS3 GAIA Fuel Transition........ 12 3.3.

Reduction of the P-8 Power Range Neutron Flux Nominal Trip Setpoint and Allowable Value............................................................................................... 14 3.4.

ANP-10338-P-A Application to GAIA Fuel in MPS3 Cores.............................. 14

4.0 REGULATORY EVALUATION

............................................................................ 16 4.1.

Applicable Regulatory Requirements and Criteria............................................ 16 4.2.

Precedents....................................................................................................... 18 4.3.

No Significant Hazards Consideration.............................................................. 19 4.4.

Conclusions...................................................................................................... 23 5.0 Environmental Considerations............................................................................. 23 6.0 References.......................................................................................................... 24 ARCADIA, AREA, COBRA-FLX, GAIA, GALILEO, and M5 are trademarks or registered trademarks of Framatome or its affiliates in the US or other countries.

Serial No.23-251 Docket No. 50-423, Page 2 of 25 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENC) is submitting a License Amendment Request (LAR) to the U. S. Nuclear Regulatory Commission (USNRC) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3). DENC is proposing to change the following TS:

TS 2.1.1.1 Reactor Core Safety Limits TS Table 2.2-1, Item 18.c Power Range Neutron Flux, P-8 TS 6.9.1.6.b Core Operating Limits Report (COLR)

The proposed TS changes support the use of Framatome GAIA fuel with M5 fuel cladding material at MPS3, which is currently scheduled for onload during the spring 2025 refueling outage.

Pursuant to 10 CFR 50.59, DENC also requests USNRC review and approval of the following items supporting use of GAIA fuel at MPS3:

The Design Basis Limits for a Fission Product Barrier (DBLFPBs) associated with MPS3-specific application of DOM-NAF-2-P-A, Appendix F, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code [Reference 4, as supplemented by References 5 and 6], and VEP-NE-2-A, Statistical DNBR Evaluation Methodology, [Reference 2].

Mixed-core penalties for application to Departure from Nucleate Boiling Ratio (DNBR) analysis results of MPS3 cores containing both Framatome GAIA fuel and the resident Westinghouse fuel.

The DBLFPBs associated with MPS3-specific application of ANP-10338-P-A, AREA - ARCADIA Rod Ejection Accident, [Reference 3].

Serial No.23-251 Docket No. 50-423, Page 3 of 25 2.0 DETAILED DESCRIPTION 2.1. Current Licensing Basis Summary and Applicable Technical Specifications 2.1.1. Departure from Nucleate Boiling (DNB) Design Basis MPS3 Final Safety Analysis Report (FSAR) Section 4.4 describes reactor core thermal-hydraulic design criteria, evaluation methods, and DNBR limits. The DNBR design basis for the resident Westinghouse fuel product was established using USNRC-approved Fleet Report DOM-NAF-2-P-A [Reference 1]. Specifically, Appendix C of DOM-NAF-2-P-A documents the qualification of the WRB-2M Critical Heat Flux (CHF) correlation with the VIPRE-D code and includes the WRB-2M Deterministic Design Limit (DDL). Appendix D of DOM-NAF-2-P-A documents the qualification of the ABB-NV and WLOP CHF correlations with the VIPRE-D code and includes the DDLs for both CHF correlations. VIPRE-D is the version of the computer code VIPRE used by the Dominion Energy fleet to perform detailed thermal-hydraulic analyses to predict CHF and DNB of reactor cores.

Dominion Energys statistical DNBR methodology, USNRC-approved Fleet Report VEP-NE-2-A [Reference 2], is also described in MPS3 FSAR Section 4.4. DOM-NAF-2-P-A was used in conjunction with VEP-NE-2-A to calculate the Statistical Design Limits (SDLs) applicable to the VIPRE-D/WRB-2M and VIPRE-D/ABB-NV code/correlation pairs for the resident Westinghouse fuel in MPS3.

As described in MPS3 FSAR Section 4.4.1.1, the design limit DNBR values meet the 95/95 DNB design criterion. For use in safety analysis, DENC increases the design limit DNBR to a DNBR Safety Analysis Limit (SAL) that provides retained DNBR margin to offset generic DNBR penalties that may occur. This approach provides flexibility in design and operation of the plant.

FSAR Chapter 15 describes the current application of the DOM-NAF-2-P-A and VEP-NE-2-A methodologies in demonstrating acceptable results for the MPS3 accident analysis.

The following TS are related to the plant-specific application of the DOM-NAF-2-P-A and VEP-NE-2-A methodologies to the resident Westinghouse fuel product in MPS3 cores:

Serial No.23-251 Docket No. 50-423, Page 4 of 25 T.S 2.1.1.1 - Reactor Core Safety Limits (RCSLs)

The RCSLs preclude violation of fuel cladding design criteria during normal operation and Anticipated Operational Occurrences (AOOs). The current MPS3 RCSL related to DNBR is established for Westinghouse fuel and is applicable in MODES 1 and 2.

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to 1.14 for the WRB-2M DNB correlation.

TS 6.9.1.6.b - Core Operating Limits Report TS 6.9.1.6 requires core operating limits to be established for each reload cycle and contains references to the approved analytical methods used to determine the core operating limits. The TS 6.9.1.6.b COLR reference list includes documents that define the methods used to determine the core operating limits for MPS3. The existing TS 6.9.1.6.b references establishing TS 3.2.3.1 and 3.2.5 parameters related to DNB for the current Westinghouse fuel product are:

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

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 TS 6.9.1.6.b requires that the cycle-specific COLR contain the complete identification for each of the TS referenced topical reports used (i.e., report number, title, revision, date, and any supplements).

Serial No.23-251 Docket No. 50-423, Page 5 of 25 2.1.2. Reactor Trip System Interlock Permissive-8 (P-8)

The P-8 interlock blocks a reactor trip on a low reactor coolant flow in any one loop when plant thermal power is less than the P-8 setpoint listed in TS Table 2.2-1 for Reactor Trip System Instrumentation Trip Setpoints. TS Table 2.2-1, Item 18.c, contains the current nominal and allowable trip setpoints for P-8.

TS Table 2.2-1, Item 18.C - Power Range Neutron Flux, P-8 Nominal Trip Setpoint Allowable Value

18. Reactor Trip System Interlocks

c.

Power Range 50.0% of RTP**

50.6% of RTP**

Neutron Flux, P-8

- RTP = RATED THERMAL POWER 2.1.3. Rod Ejection MPS3 FSAR Section 15.4.8 describes the Spectrum of Rod Cluster Control Assembly Ejection Accidents (Rod Ejection) MPS3-specific evaluation. The rod ejection accident is classified as an American Nuclear Society (ANS) Condition IV event or a postulated accident.

2.2. Reason for the Proposed Changes The proposed changes are needed to support the use of Framatome GAIA fuel with M5 cladding at MPS3. DENC and Framatome have entered into an agreement to implement the GAIA fuel product at MPS3. The first reload batch of GAIA fuel assemblies are planned for insertion in MPS3 Cycle 24, which is currently scheduled to begin operation in the spring of 2025. The first MPS3 core full of GAIA product is expected in Cycle 26.

2.3. Description of Proposed Changes 2.3.1. DOM-NAF-2-P-A and VEP-NE-2-A Application to GAIA Fuel in MPS3 Cores DENC is requesting USNRC review and approval of the DBLFPBs associated with MPS3-specific application of DOM-NAF-2-P-A, Appendix F, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code, [Reference 4, as supplemented by References 5 and 6], and VEP-NE-2-A, Statistical DNBR Evaluation Methodology,

[Reference 2], to the Framatome GAIA fuel product. A VIPRE-D/ORFEO-GAIA

Serial No.23-251 Docket No. 50-423, Page 6 of 25 SDL of 1.26 is proposed for evaluation of the GAIA fuel product at MPS3. Similarly, a VIPRE-D/ORFEO-NMGRID SDL of 1.31 is proposed.

Changes to the following MPS3 TS associated with DOM-NAF-2-P-A and VEP-NE-2-A application are proposed. A markup of the current TS is provided in Attachment

2.

TS 2.1.1.1 - Reactor Core Safety Limits TS 2.1.1.1 currently includes the WRB-2M DDL which is the primary CHF correlation used for DNBR analysis of resident Westinghouse fuel. The proposed addition to TS 2.1.1.1 adds the VIPRE-D/ORFEO-GAIA DDL of 1.13. ORFEO-GAIA is the primary CHF correlation used in DNBR analysis of Framatome GAIA fuel. The VIPRE-D/ORFEO-GAIA DDL was submitted to the USNRC for review via Reference 4 (as supplemented by References 5 and 6). Additions to TS 2.1.1.1 are shown in bold text below.

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to 1.14 for the WRB-2M DNB correlation, or greater than or equal to 1.13 for the ORFEO-GAIA DNB correlation.

TS 2.1.1.1 will retain the WRB-2M limit to support the Westinghouse fuel product during the transition to full cores of GAIA.

TS 6.9.1.6.b - Core Operating Limits Report The change to the TS 6.9.1.6.b COLR reference list adds DOM-NAF-2-P-A, Appendix F, which describes ORFEO-GAIA and ORFEO-NMGRID correlation application with VIPRE-D. TS 6.9.1.6.b requires that the cycle-specific COLR contain the complete identification for each of the TS referenced topical reports used (i.e., report number, title, revision, date, and any supplements), therefore, only high-level reference to the applicable topical reports is provided in the TS 6.9.1.6.b list. Additions to the TS COLR reference list are shown in bold text below, and deletions are shown in strikethrough.

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

Serial No.23-251 Docket No. 50-423, Page 7 of 25 Computer Code, and Appendix F, "Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy 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 TS 6.9.1.6.b will retain DOM-NAF-2-P-A, Appendixes C and D to support the current Westinghouse fuel product during the transition to full cores of GAIA. VEP-NE-2-A is currently included as a COLR reference.

2.3.2. DNBR Penalties Supporting the MPS3 Transition to GAIA Fuel DENC is requesting USNRC approval of transition core DNBR penalties which address the thermal-hydraulic effects resulting from physical differences between the resident Westinghouse fuel and prospective Framatome GAIA fuel. A transition core penalty (TCP) of 2.4% is proposed for application to DNBR analysis results calculated using the VIPRE-D/ORFEO-GAIA code/correlation pair during the first and second transition cycles. A penalty of 2.7% is similarly proposed when using the VIPRE-D/ORFEO-NMGRID code/correlation pair above the GAIA first mixing vane spacer grid. No TCP will be applied below the GAIA first mixing vane grid or to DNBR analysis results calculated using Westinghouse CHF correlations during the transition cycles.

2.3.3. Reduction of the P-8 Power Range Neutron Flux Nominal Trip Setpoint and Allowable Value This LAR reduces the P-8 nominal trip setpoint and allowable value to meet GAIA DNBR safety limits. Additions to the TS are shown in bold text below, and deletions are shown in strikethrough. A markup of the current TS is provided in Attachment

2.

TS Table 2.2-1, Item 18.C - Power Range Neutron Flux, P-8 Nominal Trip Setpoint Allowable Value

18. Reactor Trip System Interlocks

c.

Power Range 35.050.0% of RTP**

35.650.6% of RTP**

Neutron Flux, P-8

- RTP = RATED THERMAL POWER

Serial No.23-251 Docket No. 50-423, Page 8 of 25 2.3.4. ANP-10338-P-A Application to GAIA Fuel in MPS3 Cores DENC is requesting USNRC review and approval of the DBLFPBs associated with MPS3-specific application of ANP-10338-P-A, AREA-ARCADIA Rod Ejection Accident, [Reference 3]. The MPS3-specific AREA analysis for GAIA fuel was performed against the acceptance criteria defined in USNRC Regulatory Guide (RG) 1.236 [Reference 7]. The proposed DBLFPBs are outlined in Table 1.

Framatome performed an MPS3 Rod Ejection Analysis (REA) for GAIA fuel. This included the use of Framatomes COBRA-FLX thermal-hydraulics code with the ORFEO-GAIA CHF correlation as part of the ANP-10338-P-A methodology.

Therefore, the USNRC-approved COBRA-FLX/ORFEO-GAIA DDL is provided in Table 1 to correspond with the Framatome codes and methods used in the MPS3 REA for GAIA fuel.

Table 1: MPS3-Specific AREA Acceptance Criteria RG 1.236 Fuel Cladding Failure Mechanism Parameter Limit High Temperature Peak Radial Average Enthalpy (cal/g) vs.

Cladding Pressure Differential (MPa)

RG 1.236, Figure 1 High Temperature DNBR Design Limit 1.12 for COBRA-FLX/

ORFEO-GAIA

[Reference 8]

Pellet Clad Mechanical Interaction (PCMI)

Peak Radial Average Fuel Enthalpy Rise (cal/g) vs. Excess Cladding Hydrogen (wppm)

RG 1.236, Figures 2 and 4 Molten Fuel and Core Coolability Fuel Centerline Melt (F)

Less than 4754°F, decreasing linearly by 13.7°F per 10,000 MWD/MTU of burnup; rim melt is precluded Core Coolability Peak Radial Average Enthalpy (cal/g)

Less than 230 cal/g

Serial No.23-251 Docket No. 50-423, Page 9 of 25 In accordance with RG 1.236, the peak reactor coolant pressure criterion of MPS3 FSAR Section 15.4.8 is maintained.

3.0 TECHNICAL EVALUATION

3.1. DOM-NAF-2-P-A and VEP-NE-2-A Application to GAIA Fuel in MPS3 Cores This LAR package includes the technical basis and required documentation to support the plant-specific application of the VIPRE-D thermal-hydraulics code with the ORFEO-GAIA and ORFEO-NMGRID CHF correlations to 17x17 Framatome GAIA fuel in MPS3 cores. Specifically, DENC is seeking USNRC approval for the following activities:

(1) Addition of the VIPRE-D/ORFEO-GAIA DDL to the TS 2.1.1.1 Reactor Core Safety Limits, (2) Addition of DOM-NAF-2-P-A, Appendix F to the TS 6.9.1.6.b list of COLR methodologies, and (3) The VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID SDLs documented in Attachment 3, as these values constitute DBLFPBs.

Approval of the changes will allow DENC to use the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs to perform licensing basis calculations for Framatome GAIA fuel in MPS3 cores.

Appendix F to Fleet Report DOM-NAF-2-P-A summarizes the data evaluations that qualify the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs and develops the corresponding generic DDLs. DOM-NAF-2-P-A, Appendix F, was submitted to the USNRC on December 19, 2022 [Reference 4], as supplemented by letters dated April 6 and July 26, 2023 [References 5 and 6]. The submittal is presently under USNRC review. The proposed VIPRE-D/ORFEO-GAIA DDL of 1.13 for addition to TS 2.1.1.1 is obtained from DOM-NAF-2-P-A, Appendix F. The requested addition of DOM-NAF-2-P-A, Appendix F to the TS 6.9.1.6.b COLR reference list satisfies a Limitation and Condition (L&C) of MPS3-specific application of DOM-NAF-2-P-A to GAIA fuel.

The addition of (1) the VIPRE-D/ORFEO-GAIA DDL to TS 2.1.1.1 and (2) DOM-NAF-2-P-A, Appendix F to the TS 6.9.1.6.b COLR reference list presume USNRC approval of DOM-NAF-2-P-A, Appendix F prior to implementation of the requested TS changes.

If USNRC approval is not received for DOM-NAF-2-P-A, Appendix F, this LAR will be supplemented accordingly. The current LAR and its supporting documentation meet the L&Cs of DOM-NAF-2-P-A, Appendix F as presently submitted. Should the

Serial No.23-251 Docket No. 50-423, Page 10 of 25 USNRC impose additional restrictions in the final Safety Evaluation for DOM-NAF P-A, Appendix F, DENC will assess the added L&Cs for impact on this LAR and supplement as necessary.

The technical basis supporting implementation of DOM-NAF-2-P-A and VEP-NE-2-A for the Framatome GAIA fuel product at MPS3 is summarized in Attachment 3. A discussion of the work performed to qualify the ORFEO-GAIA and ORFEO-NMGRID CHF correlations for statistical DNB evaluation is included therein. The SDLs obtained by methodology application and requested for USNRC-approval are 1.26 for VIPRE-D/ORFEO-GAIA and 1.31 for VIPRE-D/ORFEO-NMGRID. The statistical limits are consistent with the DDLs submitted in DOM-NAF-2-P-A, Appendix F.

Further, this plant and fuel specific implementation of DOM-NAF-2-P-A requires a change to TS Table 2.2-1, Item 18.C, Permissive-8 Power Range Neutron Flux setpoints to meet applicable DNBR limits for the GAIA fuel product. Section 3.3 provides the technical evaluation supporting the P-8 change.

3.2. DNBR Penalties Supporting the MPS3 Transition to GAIA Fuel This LAR package requests USNRC approval of mixed-core DNBR penalties supporting the transition from MPS3s resident Westinghouse fuel product to Framatome GAIA fuel. Specifically, a TCP of 2.4% is proposed for application to DNBR analysis results calculated using the VIPRE-D/ORFEO-GAIA code/correlation pair. A penalty of 2.7% is similarly proposed when using the VIPRE-D/ORFEO-NMGRID code/correlation pair above the GAIA first mixing vane spacer grid. Approval of the penalties allows allocation of retained DNBR margin to address the thermal-hydraulic effects of adjacent dissimilar fuels during the transition cycles.

DENC investigated the thermal-hydraulic effects from the physical differences between the resident Westinghouse and Framatome GAIA fuel products in mixed-core configurations at MPS3. DNBR penalties were developed for both the Westinghouse and Framatome fuel products using the VIPRE-D code and applicable CHF correlations in accordance with the USNRC-approved DOM-NAF-2-P-A methodology in support of mixed-core analysis. TCPs will be accommodated within the retained DNBR margin described in Attachment 3. DENCs mixed-core DNBR evaluation supporting the GAIA fuel transition is described below.

3.2.1. General Approach Dominion Energys mixed-core DNBR analysis approach is to model each fuel product type residing in the mixed-core as if it were a full-core of that fuel product type, and then apply a penalty to retained DNBR margin to account for the effects

Serial No.23-251 Docket No. 50-423, Page 11 of 25 of the mixed-core configuration. This process is performed for each fuel design that will be placed into the core.

The DNBR penalty is determined from a subchannel code analysis of the mixed-core configuration. An exact transition or limiting mixed-core pattern (e.g., one fuel assembly type in the center with the remaining core represented by the other fuel assembly type) can be modeled. DNBR results from the selected mixed-core configuration are compared to calculations performed at identical conditions from the applicable full-core subchannel model of resident or prospective fuel. Each CHF correlation applicable to the transition cycle is assessed. Penalties are set by the largest DNBR differential for the evaluated range of initial conditions. This penalty is referred to as the TCP and is defined in Equation 1 below. The penalty is then included in the retained DNBR margin evaluation for the respective cycle.

100 1

TCPs developed from exact transition core patterns produce a pattern-specific penalty applicable to a single cycle. When the limiting core loading pattern is modeled, the calculated TCPs bound the entire core transition period to provide loading flexibility for core designers. If mixed-core DNBR calculations show improved CHF performance for an applicable code/correlation pair compared to full core DNB calculations, no TCP is applied to retained DNBR margin.

3.2.2. Subchannel Code Modeling Considerations Dominion Energys core thermal-hydraulics code, VIPRE-D, is used to perform mixed-core DNBR analysis. A key requirement of the Dominion Energy mixed-core DNB evaluation is accurate modeling of the flow redistribution between adjacent hydraulically dissimilar fuel assemblies. Flow redistribution among dissimilar fuel assemblies induces a lateral flow component to the more dominant axial fluid velocity. The lateral component is commonly called fuel assembly crossflow. The following modeling considerations, in addition to the approved modeling considerations outlined in the USNRC-approved DOM-NAF-2-P-A topical report, ensure predictions of crossflow from higher pressure drop regions to lower pressure drop regions are simulated correctly.

Radial Nodalization - The level-of-detail in the mixed-core radial nodalization accounts for the hydraulic differences between the dissimilar fuel assemblies.

This requires the 1/8th symmetric Power Limiting Bundle (PLB) be modeled on

Serial No.23-251 Docket No. 50-423, Page 12 of 25 a pin-by-pin subchannel basis to capture the flow redistribution within the PLB and into/out of the PLB from the adjacent fuel assemblies. Each fuel designs radial thermal-hydraulic characteristics are preserved (e.g., computation of flow areas, wetted and heated perimeters, crossflow gaps, centroid distance, etc.).

Axial Nodalization - Although the subchannel code limitations only allow one axial nodal scheme, the mixed-core axial nodalization accounts for the physical location of fuel assembly components for both fuel designs (e.g. spacer grid locations, etc.). The fuel assembly modeled in the PLB dictates the mixed-core axial nodalization when this modeling satisfies the nodalization criteria given in DOM-NAF-2-P-A for both fuel products.

Axial Hydraulic Form Losses - Each fuel products hydraulic losses are preserved in the mixed-core models. The fuel vendor provides hydraulic form losses for spacer grids and end fittings in addition to fuel rod friction loss. If necessary, this information is adjusted to correct for differences in the experimental data reduction techniques applied by the different fuel vendors when calculating hydraulic losses (e.g., vendor dependent hydraulic data reduction may use a different rod friction correlation in their differential pressure testing). The correction puts these parameters on an even basis using the same bare rod friction correlation in the approved subchannel code.

Critical Heat Flux Correlations - Calculations are performed using the USNRC-approved CHF correlations for each fuel product. Multiple cases are analyzed to assess each fuel product with their qualified CHF correlations placed in the PLB.

These modeling considerations were included in a VIPRE-D model that was benchmarked against a fuel vendor test between two hydraulically dissimilar fuel designs. The benchmark concluded that the void drift modeled in VIPRE-D responds consistently with different void propagations across the boundaries. It also concluded that the gap to centroid lateral scaling consistently reflected the flow redistribution (i.e., crossflow velocities) when the PLB had a different form loss pressure drop than the surrounding assemblies. The benchmark did not change based on the model size (i.e., number of assemblies or pins within an assembly),

however, it did demonstrate that the PLB pin-by-pin model most accurately predicted the crossflow velocities of the test. Therefore, a pin-by-pin PLB model is used to generate GAIA TCPs.

3.2.3. Mixed-Core Analysis Performed for the MPS3 GAIA Fuel Transition DENC investigated the thermal-hydraulic effects from the physical differences between the resident Westinghouse and prospective Framatome GAIA fuel products in mixed-core configurations at MPS3. Calculations were performed with

Serial No.23-251 Docket No. 50-423, Page 13 of 25 the VIPRE-D code and applicable CHF correlations for each fuel product in accordance with the USNRC-approved DOM-NAF-2-P-A methodology in support of mixed-core analysis. Calculations also follow the mixed-core modeling considerations outlined above.

To develop the TCPs, the DNBR results from mixed-core configurations are compared to calculations performed at identical conditions with either a full-core of resident Westinghouse fuel model or a full-core GAIA fuel model. The penalty is set by the largest DNBR differential for the evaluated range of initial conditions. The examined initial conditions cover the operating space where the CHF correlations have been applied historically at MPS3. The bounding initial condition may vary for each CHF correlation.

A limiting mixed-core loading pattern is used to develop the TCPs. A representative GAIA mixed-core pattern was examined and confirmed the limiting (i.e., largest)

TCPs are calculated for the case of a single assembly of one type in a full-core of the alternate type. The PLB is modeled on a pin-by-pin subchannel basis to accurately capture the flow redistribution. Since the axial location of spacer grids and intermediate flow mixing grids between the two fuel products are similar, the PLB dictated the mixed-core axial nodalization. This modeling satisfies the nodalization criteria of DOM-NAF-2-P-A for both fuel products. Applicable adjustments to vendor supplied hydraulic form losses were performed to accommodate differences in experimental data reduction techniques. Multiple cases were analyzed to assess each fuel product with their qualified CHF correlations placed in the PLB.

Based on the above, a TCP of 2.4% was derived for application to DNBR analysis results calculated using the VIPRE-D/ORFEO-GAIA code/correlation pair. A penalty of 2.7% was similarly derived for application when using the VIPRE-D/ORFEO-NMGRID code/correlation pair above the GAIA first mixing vane spacer grid. The mixed-core configuration used to calculate these TCPs consisted of a single GAIA assembly placed in a core full of the resident Westinghouse fuel.

Therefore, the penalties are applicable to both the first and second fuel transition cycles and may be removed once a full core of GAIA is attained.

For the resident Westinghouse fuel, mixed-core CHF calculations showed improved DNBR performance for all applicable code/correlation pairs when compared to full-core CHF calculations of the resident fuel. Improved CHF performance was also observed below the GAIA first mixing vane spacer grid for applicable correlations.

Serial No.23-251 Docket No. 50-423, Page 14 of 25 No TCP will be applied to DNBR analysis results for the resident Westinghouse fuel or below the GAIA first mixing vane grid in GAIA transition cycles.

TCPs will be accommodated within the DNBR retained margin described in.

3.3. Reduction of the P-8 Power Range Neutron Flux Nominal Trip Setpoint and Allowable Value DENC is requesting USNRC approval of a reduction to the P-8 nominal trip setpoint and allowable value listed in MPS3 TS Table 2.2-1, Item 18.c for Reactor Trip System Instrumentation Trip Setpoints. Specifically, DENC proposes a P-8 nominal setpoint of 35% RTP and an allowable value of 35.6% RTP. The proposed earlier (i.e., lower power) actuation of the permissive results in more restrictive plant operations and is necessary to meet the GAIA DNBR safety limits.

A deterministic DNBR evaluation for GAIA fuel, performed in accordance with the DOM-NAF-2-P-A methodology, establishes the proposed P-8 allowable value. This DNBR analysis assumes a low flow condition consistent with the P-8 interlock logic (i.e., loss of a Reactor Coolant Pump (RCP)). The analyses performed for GAIA fuel determined that an upper limit of 45 percent power is required to ensure the P-8 setpoint maintains DNB within acceptable limits. A 10 percent allowance was conservatively applied in the analysis to establish a new nominal trip setpoint of 35 percent RTP. Evaluation results meet the ORFEO-GAIA Safety Analysis Limit (SAL) identified in Attachment 3 and confirm positive DNBR margins for the current Westinghouse fuel.

3.4. ANP-10338-P-A Application to GAIA Fuel in MPS3 Cores This LAR package requests USNRC review and approval for the DBLFBPs associated with MPS3-specific application of Framatomes ARCADIA Rod Ejection Accident (AREA) Methodology described in ANP-10338-P-A. Specifically, DENC seeks USNRC-approval of the rod ejection-specific DBLFPBs provided in Table 1 above.

Approval of the Table 1 limits will allow DENC to implement the MPS3 REA analysis for GAIA fuel in accordance with 10 CFR 50.59.

Application of the AREA methodology satisfies a L&C of ANP-10342-P-A, GAIA Fuel Assembly Mechanical Design, [Reference 9], which was proposed for addition to the MPS3 TS 6.9.1.6.b COLR reference list by letter dated May 23, 2023 [Reference 10].

The ANP-10342-P-A condition states the licensee should consider the most up-to-date guidance and analytical limits when demonstrating acceptable performance of GAIA under reactivity-initiated accident conditions. The recently issued REA guidance

Serial No.23-251 Docket No. 50-423, Page 15 of 25 found in RG 1.236 (formerly Draft Regulatory Guide 1327) is highlighted in the L&C.

MPS3 application of the AREA methodology is compliant with RG 1.236.

Framatome performed a MPS3-specific AREA analysis for GAIA fuel using the Table 1 acceptance criteria. This included use of Framatomes COBRA-FLX thermal-hydraulics code with the ORFEO-GAIA CHF correlation as part of the ANP-10338-P-A methodology. Therefore, the USNRC-approved COBRA-FLX/ORFEO-GAIA DDL

[Reference 8] is proposed as the AREA-specific DNBR acceptance criterion to correspond with the Framatome codes and methods used in the MPS3 REA for GAIA fuel.

Additionally, the Table 1 fuel centerline melt limit is specific to AREA application. The MPS3 AREA analysis used only the GALILEO code to derive thermal-mechanical properties. The AREA centerline melt limit includes adjustments to account for fuel melt temperature measurement uncertainty and code temperature prediction uncertainty. Adjustments for burnable poison content have been accounted for. This is different from the proposed Framatome fuel centerline melt addition to MPS3 TS 2.1.1.2 by letter dated May 23, 2023 [Reference 10], which reflects a bounding treatment of uncertainty based on the COPERNIC fuel rod performance code and its corresponding application methodology.

The COBRA-FLX/ORFEO-GAIA DDL and the AREA fuel centerline melt limit are not proposed for addition to the MPS3 TS 2.1.1 RCSLs. TS 2.1.1 is applicable to normal operation and AOOs. Rod Ejection is an ANS Condition IV event also referred to as a postulated accident.

The completed MPS3 Rod Ejection analysis complies with the L&Cs of the AREA topical report and those of the approved ARCADIA codes, including the GALILEO code for thermal-mechanical properties. One deviation was taken from the USNRC-approved AREA methodology. [(See Attachment 4, Proprietary INSERT 1)].

MPS3 AREA analysis results demonstrate margin to the limits for fuel temperature, fuel rim temperature, enthalpy, and enthalpy rise. DNB fuel rod failures are predicted, but the failure total remains within the assumptions of the current REA radiological analysis described in MPS3 FSAR Chapter 15.4.8.

Reactor coolant system overpressure was not evaluated using ANP-10338-P-A. No aspect of the Framatome fuel affects the severity of the rod ejection overpressure analysis and does not require reanalysis. If the AREA overpressure analysis were performed, it would produce a reduced pressure response compared to the

Serial No.23-251 Docket No. 50-423, Page 16 of 25 existing analysis described in the FSAR. The current MPS3 FSAR REA overpressure analysis remains bounding of operation with the GAIA fuel product at MPS3.

Mixed-core application of the MPS3 AREA analysis is addressed in the development of parameter biasing to account for the impact of cycle-to-cycle changes including cycle designs with co-resident fuel assemblies. Additionally, the MPS3 REA analysis incorporates a conservative thermal-hydraulic penalty that accommodates mixed-core changes in flow distribution.

The report summarizing the results of the MPS3 AREA analysis for GAIA fuel are available for USNRC audit.

4.0 REGULATORY EVALUATION

4.1. Applicable Regulatory Requirements and Criteria The regulatory requirements and/or guidance documents associated with this LAR include the following:

10 CFR 50.36 - Technical Specifications 10 CFR 50.36(c) requires the TS include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation, (3) surveillance requirements; (4) design features; and (5) administrative controls. The TS changes described in this amendment request are related to categories (1) and (5). The supporting assessments described herein ensure 10 CFR 50.36 requirements continue to be met through the specification of appropriate safety limits and limiting safety systems settings, and the establishment of administrative controls. The evaluations described in this attachment and Attachment 3 demonstrate the adequacy of the proposed TS changes.

USNRC Generic Letter (GL) 88-16 USNRC GL 88-16 states that it is acceptable for licensees to control reactor physics parameter limits by specifying an USNRC-approved calculation methodology. These parameter limits may be removed from the TS and placed in a cycle-specific COLR, which is required to be submitted to the USNRC every operating cycle or each time it is revised.

TS 6.9.1.6.b identifies the USNRC-approved analytical methodologies that are used to determine the core operating limits for MPS3. The guidance in USNRC

Serial No.23-251 Docket No. 50-423, Page 17 of 25 GL 88-16 continues to be met since the proposed changes will continue to specify USNRC-approved methodologies used to determine the core operating limits.

10 CFR 50 Appendix A - General Design Criterion (GDC) and NUREG-0800 -

Standard Review Plan (SRP)

Review of the 10 CFR 50 Appendix A, GDC identified the following specific criterion as relevant to this submittal. GDC 10, Reactor Design, requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs). SAFDLs are established to ensure fuel is not damaged (i.e., fuel rods do not fail, fuel system dimensions remain within operational tolerances, and functional capabilities are not reduced below those assumed in the safety analysis). Compliance with GDC 10 provides assurance that the integrity of the fuel and cladding will be maintained, thus preventing the potential for release of fission products during normal operation or AOOs.

GDC 28, Reactivity Limits, requires reactivity control systems be designed with appropriate limits to assure that the effects of postulated reactivity accidents, including a rod ejection, can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. Compliance with GDC 28 provides assurance that the integrity of the reactor coolant pressure boundary and core coolability will be maintained.

Review of the NUREG-0800 SRP identified Chapters 4 and 15 as relevant to this submittal. Chapter 4 includes Section 4.2 (Fuel System Design), Section 4.3 (Nuclear Design), and Section 4.4 (Thermal and Hydraulic Design). The relevant portions of Chapter 15 include Section 15.4.8 (PWR Rod Ejection Accidents).

SRP Section 4.2 describes fuel damage criteria. SRP Section 4.3 describes reactivity criteria related to the analysis of reactivity events. SRP Section 4.4 provides specific thermal-hydraulic criteria for the core and reactor coolant system. SRP Section 15.4.8 discusses the postulated control rod ejection accident and associated criteria for evaluation of reactor coolant pressure boundary damage and cooling flow impairment.

Serial No.23-251 Docket No. 50-423, Page 18 of 25 This attachment and Attachment 3 provide the thermal-hydraulic basis for the proposed GAIA fuel SAFDLs, setpoint changes, and generic penalties. These evaluations address the DNBR Acceptance Criteria defined in SRP Section 4.4, which includes GDC 10. In addition, SRP Sections 4.2, 4.3, and 15.4.8, and GDC 28 were reviewed as they relate to the GAIA fuel rod ejection evaluation. DENC elected to apply the RG 1.236 guidance to demonstrate the relevant requirements of GDC 28 and SRP Sections 4.2, 4.3 and 15.4.8 are met. This attachment provides a summary of the rod ejection evaluation for the GAIA design and the resultant SAFDLs. DENC has determined that the proposed changes meet the current regulatory requirements, and applicable design criteria are met under normal, upset, and faulted operating conditions.

There are no changes being proposed in this LAR that would challenge the conformance or commitments to regulatory and/or guidance documents described above. The evaluations documented herein confirm that MPS3 will continue to comply with its applicable regulatory requirements.

4.2. Precedents The following precedent submittals and USNRC approvals are identified to support the proposed GAIA fuel transition licensing activities.

DENC submitted an LAR to adopt Dominion Energy core design and safety analysis methods at MPS3 (ML15134A244), which received USNRC approval in ML16131A728. The LAR requested application of the DOM-NAF-2-P-A VIPRE-D and VEP-NE-2-A statistical DNBR methodologies to the resident Westinghouse fuel loaded in MPS3 cores. Approval allowed DENC to use the VIPRE-D/WRB-2M, VIPRE-D/ABB-NV, and VIPRE-D/WLOP code/correlation pairs to perform licensing basis DNBR calculations for Westinghouse fuel in MPS3 cores. The requested licensing actions supporting plant-specific application DOM-NAF-2-P-A and VEP-NE-2-A to GAIA fuel in MPS3 cores, specifically, use of the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlations, are fundamentally the same.

Virginia Electric and Power Company (VEPCO) submitted an LAR (ML020930212) to permit use of Framatome ANP Advanced Mark-BW (AMBW) fuel with M5 cladding at North Anna Power Station, which was approved by the USNRC in 2004 (ML040960040 and ML042330659). The North Anna submittal included a discussion of the mixed-cored analysis methodology and DNBR evaluation supporting the AMBW fuel transition. Similarly, this LAR describes Dominion Energys approach to mixed-cored DNBR analysis and requests USNRC approval of the GAIA fuel TCPs.

Serial No.23-251 Docket No. 50-423, Page 19 of 25 DENC requested an increase to the P-8 setpoint at MPS3 as part of the Stretch Power Uprate (SPU) Licensing Report (ML072000400), which was USNRC-approved in MPS3 Amendment 242 (ADAMS Package No. ML082180137). The proposed P-8 change is similar to the SPU setpoint request with exception to the direction of change (setpoint decrease versus increase).

Ameren Missouri submitted an LAR to allow the use of a limited number of GAIA fuel assemblies at Callaway Plant Unit 1 (ADAMS Package No. ML22285A115, as supplemented by ML22335A498 and ML23215A195), which was accepted for review by the USNRC in 2022 (ADAMS Accession No. ML22348A116). The Ameren Missouri LAR described use of the AREA methodology in support of the use of GAIA fuel assemblies at Callaway. This LAR proposes use of the AREA methodology and requests USNRC approval of its associated DBLFPBs to support the loading of GAIA fuel in MPS3 cores.

4.3. No Significant Hazards Consideration Dominion Energy Nuclear Connecticut, Inc. (DENC) proposes changes to the Millstone Power Station Unit 3 (MPS3) Technical Specifications (TS) to support the use of Framatome GAIA fuel with M5 cladding. The proposed TS changes include updating the Reactor Core Safety Limits (RCSLs) (TS 2.1.1.1), the Permissive 8 (P-

8) setpoint (TS Table 2.2-1, Item 18.c), and the list of approved methodologies for the Core Operating Limits Report (COLR) (TS 6.9.1.6.b).

DENC additionally requests USNRC approval of mixed-core Departure from Nucleate Boiling (DNBR) penalties for application to retained DNBR margin. The transition core penalties (TCPs) address the thermal-hydraulic effects resulting from physical differences between the resident Westinghouse fuel and prospective Framatome GAIA fuel. Further, the License Amendment Request (LAR) seeks USNRC approval for the use of Design Basis Limits for a Fission Product Barrier (DBLFPBs) associated with MPS3-specific application of the following methodologies: DOM-NAF-2-P-A (specifically Appendix F), VEP-NE-2-A, and ANP-10338-P-A. The request related to DOM-NAF-2-P-A and VEP-NE-2-A will allow DENC to perform licensing basis DNBR calculations for GAIA fuel in MPS3 cores. The request related to ANP-10338-P-A will permit DENC implementation of the MPS3 Rod Ejection Accident (REA) analysis for GAIA fuel.

It should be noted that the submittal for DOM-NAF-2-P-A, Appendix F is presently under USNRC review. The proposed updates to the TS 2.1.1.1 RCSLs and TS 6.9.1.6.b COLR reference list presume that USNRC approval for DOM-NAF-2-P-A,

Serial No.23-251 Docket No. 50-423, Page 20 of 25 Appendix F will be received prior to implementation of the TS changes. If USNRC approval is not received as anticipated, this LAR will be supplemented accordingly.

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

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed license amendment modifies the TS to (1) include a critical heat flux (CHF) limit applicable to GAIA fuel, (2) reduce the plant permissive setpoint for P-8 Power Range Neutron Flux, and (3) allow the use of the DOM-NAF-2-P-A, Appendix F method. Additionally, the proposed license amendment requests approval for (4) GAIA mixed-core DNBR penalties and (5) the DBLFPBs associated with approved (5a) Dominion Energy DNBR and (5b) Framatome REA methodologies.

The proposed activities are similar as these activities establish or ensure conformance with Specified Acceptable Fuel Design Limits (SAFDLs).

SAFDLs ensure the fuel is not damaged or its functional capabilities are not reduced below those assumed in the safety analysis.

Proposed activities (1), (3), (4) and (5) do not require physical changes to plant systems, structures, or components (SSCs) for implementation.

Proposed change (2) reduces the Power Range Neutron Flux, P-8 setpoint to a value that results in earlier (i.e. lower power) actuation of the permissive.

The P-8 interlock blocks a reactor trip on a low reactor coolant flow in any one loop when plant thermal power is less than the P-8 setpoint. The P-8 setpoint reduction ensures protection of the GAIA fuel in operating conditions allowed by the MPS3 TS.

The proposed activities do not impact the performance of any equipment used to mitigate the consequences of an analyzed accident. The changes do not impact any accident initiators. All proposed changes supporting GAIA implementation ensure fuel integrity will be maintained during normal

Serial No.23-251 Docket No. 50-423, Page 21 of 25 operations and anticipated operational transients, and that acceptable consequences are obtained for postulated accidents.

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

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

Response: No.

The proposed license amendment modifies the TS to (1) include a CHF limit applicable to GAIA fuel, (2) reduce the plant permissive setpoint for P-8 Power Range Neutron Flux, and (3) allow the use of the DOM-NAF-2, Appendix F method. Additionally, the proposed license amendment requests approval of (4) GAIA mixed-core DNBR penalties and (5) the DBLFPBs associated with approved (5a) Dominion Energy DNBR and (5b) Framatome REA methodologies.

The proposed activities are similar as these activities establish or ensure conformance with SAFDLs. SAFDLs ensure the fuel is not damaged or its functional capabilities are not reduced below those assumed in the safety analysis.

Proposed activities (1), (3), (4) and (5) do not require physical changes to plant SSCs for implementation. No new or different type of equipment will be installed which would create a new or different kind of accident. The proposed changes do not impose any new or different operating requirements.

No new or different accidents result from the item (2) proposed P-8 change.

The change does not require the installation of new reactor protection equipment that would create new accident initiators or failure modes. While the P-8 change alters a plant setpoint, the proposed earlier (i.e., lower power) actuation of the permissive results in more restrictive plant operations following the loss of one reactor coolant pump.

Therefore, it is concluded that the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Serial No.23-251 Docket No. 50-423, Page 22 of 25

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed license amendment modifies the TS to (1) include a CHF limit applicable to GAIA fuel, (2) reduce the plant permissive setpoint for P-8 Power Range Neutron Flux, and (3) allow the use of the DOM-NAF-2-P-A, Appendix F method. Additionally, the proposed license amendment requests approval for (4) GAIA mixed-core DNBR penalties and (5) the DBLFPBs associated with USNRC-approved (5a) Dominion Energy DNBR and (5b)

Framatome REA methodologies. The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated.

Proposed change (1) incorporates a new MPS3 TS safety limit for CHF for Framatome fuel. The CHF safety limit provides assurance that Framatome fuel cladding will perform within applicable acceptance criteria for normal operation and anticipated operational occurrences. No reduction in the margin of safety occurs with this proposed change.

Proposed change (2) reduces the MPS3 TS P-8 setpoint. The P-8 setpoint sets a maximum Power Range Neutron Flux at which MPS3 can operate at with the loss of one reactor coolant pump without an automatic reactor trip.

The setpoint reduction assures the M5 fuel cladding will perform within applicable acceptance criteria at reduced flow conditions. The lower setpoint maintains the margin of safety for GAIA fuel.

Proposed change (3) adds a new TS method for establishing core operating limits. The added DOM-NAF-2-P-A, Appendix F methodology will be used to ensure the MPS3 plant continues to meet applicable design criteria and safety analysis acceptance criteria with the loading of GAIA fuel. The reactor will continue to operate within its analyzed operating and design envelope.

Thus, no reduction in the margin of safety occurs with this proposed change.

Proposed change (4) requests approval of mixed-core DNBR penalties for application to retained DNBR margin for GAIA transition cycles. The assignment of DNBR margin to address the thermal-hydraulic impacts of adjacently placed dissimilar fuels ensures the fuel cladding does not violate

Serial No.23-251 Docket No. 50-423, Page 23 of 25 DNBR acceptance criteria. No reduction in the margin of safety occurs with this proposed change.

For the requested approval of DBLFPBs associated with (1) DOM-NAF-2-P-A, Appendix F, (2) VEP-NE-2-A, and (3) ANP-10338-P-A, the DBLFPBs constitute the safety analysis acceptance criteria. The associated USNRC-approved methodologies will be used to ensure the DPLFPBs are not exceeded during normal operation and anticipated operational occurrences, and that acceptable consequences are obtained for postulated accidents.

The reactor will continue to operate within its analyzed operating and design envelope. Thus, no reduction in the margin of safety occurs with this proposed change.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above information, DENC concludes that the proposed changes do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

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

5.0 Environmental Considerations The proposed license amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Serial No.23-251 Docket No. 50-423, Page 24 of 25 6.0 References

1. Dominion Energy Fleet Report, DOM-NAF-2-P-A, Revision 0, Minor Revision 4, Reactor Core Thermal-Hydraulics using the VIPRE-D Computer Code, with Appendixes A, B, C, D, and E, March 2023. ML23103A225 (Proprietary, Non-Public), ML23103A228 (Public).

2. Dominion Energy Fleet Report, VEP-NE-2-A, Revision 1, Statistical DNBR Evaluation Methodology, June 1987. ML101330527 (Non-Public).

3. Framatome Topical Report, ANP-10338-P-A, Revision 0, AREA-ARCADIA Rod Ejection Accident, December 2017.

4. Letter from J. Holloway (Dominion Energy) to USNRC, Virginia Electric and Power Company (Dominion Energy Virginia), North Anna and Surry Power Stations Units 1 and 2, Dominion Energy Nuclear Connecticut, Inc. (DENC), Millstone Power Station Units 2 and 3, Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code, December 19, 2022.

ML22353A619 (Proprietary, Non-Public), ML22353A620 (Public).

5. Letter from J. Holloway (Dominion Energy) to USNRC, Virginia Electric and Power Company (Dominion Energy Virginia), North Anna and Surry Power Stations Units 1 and 2, Dominion Energy Nuclear Connecticut, Inc. (DENC), Millstone Power Station Units 2 and 3, Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code Response to Request for Additional Information, April 6, 2023. ML23096A297 (Proprietary, Non-Public),

ML23096A298 (Public).

6. Letter from J. Holloway (Dominion Energy) to USNRC, Virginia Electric and Power Company (Dominion Energy Virginia), North Anna and Surry Power Stations Units 1 and 2, Dominion Energy Nuclear Connecticut, Inc. (DENC), Millstone Power Station Units 2 and 3, Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code Response to Second Request for Additional Information, July 26, 2023. ML23208A091 (Proprietary, Non-Public), ML23208A092 (Public).

7. USNRC Regulatory Guide, RG 1.236, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents, June 2020; ML20055F490.

8. Framatome Topical Report, ANP-10341-P-A, Revision 0, The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations, September 2018; including Errata 1P-000, October 2019.

Serial No.23-251 Docket No. 50-423, Page 25 of 25

9. Framatome Topical Report, ANP-10342-P-A, Revision 0, GAIA Fuel Assembly Mechanical Design, September 2019.

10. Letter from J. Holloway (Dominion Energy) to USNRC, Dominion Energy Nuclear Connecticut, Inc., Millstone Power Station Unit 3, Proposed Amendment to Revise Technical Specifications for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report related to Framatome GAIA Fuel, May 23, 2023.

ML23145A194 (Proprietary, Non-Public), ML23145A195 (Public).

Serial No.23-251 Docket No. 50-423 MARKED-UP TECHNICAL SPECIFICATIONS PAGES Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

MILLSTONE - UNIT 3 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, Reactor Coolant System highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT; and the following Safety Limits shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to 1.14 for the WRB-2M DNB correlation.

2.1.1.2 The peak fuel centerline temperature shall be maintained less than 5080qF, decreasing by 9qF per 10,000 MWD/MTU of burnup.

APPLICABILITY:

MODES 1 and 2.

ACTION:

Whenever the Reactor Core Safety Limit is violated, restore compliance and be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY:

MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

November 9, 2021 Amendment No. 173, 217, 236, 242, 281, 280 2-1 Serial No.23-251 Docket No. 50-423, Page 1 of 3 By submittal dated May 23, 2023, DENC requested NRC approval to add the Framatome-specific peak fuel centerline temperature limit to TS 2.1.1.2. DENC shall affirm the correct presentation of the changes proposed below to this page in conjunction with the changes requested in the May 23, 2023 letter following receipt of a NRC Safety Evaluation for both changes.

, or greater than or equal to 1.13 for the ORFEO-GAIA DNB correlation

MILLSTONE - UNIT 3 TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT NOMINAL TRIP SETPOINT ALLOWABLE VALUE 16.

Turbine Trip a.

Low Fluid Oil Pressure 500 psig t 450 psig b.

Turbine Stop Valve Closure 1 open t 1 open 17.

Safety Injection Input from ESF N.A.

N.A.

18.

Reactor Trip System Interlocks a.

Intermediate Range Neutron Flux, P-6 1 x 10-10 amp t 9.0 x 10-11 amp b.

Low Power Reactor Trips Block, P-7

1) Power Range Neutron Flux, P-10 input (Note 5) 11 of RTP**

d 11.6 of RTP**

2) Turbine Impulse Chamber Pressure, P-13 input 10 RTP** Turbine Impulse Pressure Equivalent d 10.6 RTP**

Turbine Impulse Pressure Equivalent c.

Power Range Neutron Flux, P-8 50.0% of RTP**

d 50.6% of RTP**

- RTP = RATED THERMAL POWER August 12, 2008 2-7 Amendment No. 60, 85, 159, 217, 220, 242 Serial No.23-251 Docket No. 50-423, Page 2 of 3 35.0 35.6

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.

October 5, 2021 Amendment No. 24, 40, 50, 69, 104, 173, 212, 215, 229, 238, 245, 249. 252, 255, 256, 279 6-21 Serial No.23-251 Docket No. 50-423, Page 3 of 3

," and Appendix F, "Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code By submittals dated May 2, 2023 & May 23, 2023, DENC requested NRC approval of the addition of new TS 6.9.1.6.b COLR references 24, 25, 26, and 27. DENC shall affirm the correct presentation of the changes proposed below to this page in conjunction with changes requested in the May 2, 2023 and May 23, 2023 letters following receipt of NRC Safety Evaluations for all changes.

Serial No.23-251 Docket No. 50-423 DEVELOPMENT OF STATISTICAL DEPARTURE FROM NUCLEATE BOILING DESIGN LIMITS Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.23-251 Docket No. 50-423, Page 1 of 29 Table of Contents 1

INTRODUCTION................................................................................................................................... 2 2

BACKGROUND..................................................................................................................................... 3 2.1 FLEET REPORT DOM-NAF-2-P-A.................................................................................................................... 3 2.2 TOPICAL REPORT VEP-NE-2-A....................................................................................................................... 4 3

IMPLEMENTATION OF THE STATISTICAL DNBR EVALUATION METHODOLOGY....................... 5 3.1 METHODOLOGY REVIEW................................................................................................................................ 5 3.2 UNCERTAINTY ANALYSIS............................................................................................................................... 5 3.3 CHF CORRELATIONS....................................................................................................................................... 7 3.4 MODEL UNCERTAINTY TERM......................................................................................................................... 8 3.5 CODE UNCERTAINTY....................................................................................................................................... 8 3.6 MONTE CARLO CALCULATIONS..................................................................................................................... 8 3.7 FULL CORE DNB PROBABILITY SUMMATION............................................................................................. 11 3.8 VERIFICATION OF NOMINAL STATEPOINTS............................................................................................... 16 3.9 SCOPE OF APPLICABILITY............................................................................................................................ 19 3.10

SUMMARY

OF ANALYSIS............................................................................................................................... 21 4

APPLICATION OF VIPRE-D/ORFEO-GAIA AND VIPRE-D/ORFEO-NMGRID TO MPS3................ 23 4.1 VIPRE-D/ORFEO-GAIA AND VIPRE-D/ORFEO-NMGRID SDLS FOR MPS3............................................... 23 4.2 SAFETY ANALYSIS LIMITS (SAL).................................................................................................................. 23 4.3 RETAINED DNBR MARGIN............................................................................................................................. 25 4.4 VERIFICATION OF EXISTING REACTOR CORE SAFETY LIMITS, PROTECTION SETPOINTS AND MPS3 UFSAR CHAPTER 15 EVENTS...................................................................................................................... 26 5

CONCLUSIONS.................................................................................................................................. 27 6

REFERENCES.................................................................................................................................... 28

Serial No.23-251 Docket No. 50-423, Page 2 of 29 1 Introduction This report provides the plant specific application of the Statistical DNBR Methodology for Millstone Power Station Unit 3 (MPS3) core containing Framatome 17x17 GAIA fuel assemblies. The GAIA fuel assembly for MPS3 will be equipped with structural grids (GAIA) and mid-span-mixing grids, called intermediate GAIA mixers (IGMs). The GAIA and IGM grids have mixing devices (vanes). The lower end grid and the upper end grid are high mechanical performance (HMP) grids, which do not have mixing devices. Specifically, this report supports the application of U.S. Nuclear Regulatory Commission (USNRC) approved Dominion Energy Topical Report VEP-NE-2-A, Statistical DNBR Evaluation Methodology (Reference 1) to MPS3, where DNBR stands for Departure from Nucleate Boiling Ratio. It provides the technical basis and documentation required by the USNRC to evaluate the plant specific application of VEP-NE-2-A methods to MPS3. This application employs the VIPRE-D thermal-hydraulic computer code (DOM-NAF-2-P-A) with the Framatome ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux (CHF) correlations (VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs) for the thermal-hydraulic analysis of Framatome 17x17 GAIA fuel assemblies at MPS3. In particular, Dominion Energy requests the review and approval of the Statistical Design Limits (SDLs) documented herein per 10 CFR 50.59(c)(2)(vii) as they constitute Design Basis Limits for a Fission Product Barrier (DBLFPB).

Dominion Energy is seeking approval for the inclusion of Appendix F of Fleet Report DOM-NAF-2-P-A (Reference 2) to the Technical Specification (T.S.) 6.9.1.6.b list of USNRC-approved methodologies used to determine core operating limits (i.e., the reference list of the MPS3 Core Operating Limits Report (COLR)).

This would allow Dominion Energy the use of the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs to perform licensing calculations for Framatome 17x17 GAIA fuel in MPS3 core, using the deterministic design limits (DDLs) qualified in Appendix F (Reference 3, as supplemented by References 4 and 5) of the DOM-NAF-2-P-A Fleet Report, and the Statistical Design Limits (SDLs) identified herein.

With these approvals, Dominion Energy will be licensed to perform in-house Departure from Nucleate Boiling (DNB) analyses for the intended uses described in Fleet Report DOM-NAF-2-P-A to support Millstone Power Station Unit 3 with the Framatome 17x17 GAIA fuel.

Serial No.23-251 Docket No. 50-423, Page 3 of 29

2 Background

Dominion Energy is planning to purchase fuel assemblies from Framatome for use at MPS3. These assemblies are scheduled to be inserted in Unit 3, commencing with Cycle 24. The fuel assemblies are designated as the Framatome 17x17 GAIA fuel. These assemblies are a replacement for the resident fuel product, which is the Westinghouse RFA-2 fuel product.

Dominion Energy submitted to the USNRC in Reference 6, and received approval from the USNRC in Reference 7, a request to adopt the Dominion Energys Core Design and Safety Analysis Methods at MPS3.

The Dominion Energys USNRC-approved methods applied to MPS3 include Fleet Report DOM-NAF-2-P-A (Reference 2) and Topical Report VEP-NE-2-A (Reference 1).

2.1 Fleet Report DOM-NAF-2-P-A The computer code VIPRE (Versatile Internals and Components Program for Reactors - EPRI) was developed for the Electric Power Research Institute (EPRI) by Battelle Pacific Northwest Laboratories to perform detailed thermal-hydraulic analyses to predict CHF and DNBR of reactor cores. VIPRE-01 was approved by the USNRC in References 9 and 10 for referencing in licensing applications. VIPRE-D is the Dominion Energy version of the VIPRE computer code based upon VIPRE-01. VIPRE-D was developed to fit the specific needs of Dominion Energys nuclear plants and fuel products by adding vendor specific CHF correlations and customizing its input and output. Dominion Energy has not made any modifications to the USNRC-approved constitutive models and algorithms contained in VIPRE-01.

Dominion Energys approved Fleet Report DOM-NAF-2-P-A (Reference 2) has been reviewed and approved by the USNRC. DOM-NAF-2-P-A provided the necessary documentation to describe Dominion Energys use of the VIPRE-D code, including modeling and qualification for Pressurized Water Reactors (PWR) thermal-hydraulic design and demonstrated that the VIPRE-D methodology is appropriate for PWR licensing applications. Appendix F qualified the ORFEO-GAIA and the ORFEO-NMGRID CHF correlations with the VIPRE-D code and listed the deterministic code/correlation DNBR limits. Appendix F to DOM-NAF-2-P-A has been submitted to the USNRC and is pending approval (Reference 3, as supplemented by References 4 and 5). The ORFEO-GAIA and ORFEO-NMGRID CHF correlations are applicable for the DNBR evaluation of the Framatome 17x17 GAIA fuel product.

In addition, Section 2.1 of Fleet Report DOM-NAF-2-P-A listed the information to be provided to the USNRC by Dominion Energy for the review and approval of any plant specific application of the VIPRE-D code:

1) Technical Specifications Change Request to add DOM-NAF-2-P-A and relevant Appendices to the plants COLR list.

2) Statistical Design Limit(s) for the relevant code/correlation(s) (Section 4.1).

3) Any technical specification changes related to thermal over-temperature T (OTT), over-power T (OPT), axial power distribution (FI), enthalpy rise factor (FH) or other reactor protection function, as well as revised Reactor Core Safety Limits (Section 4.4).

Serial No.23-251 Docket No. 50-423, Page 4 of 29

4) List of FSAR transients for which the code/correlations will be applied (Section 3.9).

This report provides the technical basis (Items 1 through 4 above) for the USNRC review and approval of the implementation of Appendix F of Fleet Report DOM-NAF-2-P-A to analyze the Framatome 17x17 GAIA fuel at MPS3 with the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs, as well as the SDLs obtained by this implementation (DOM-NAF-2-P-A Item 2). This report also documents that the existing Reactor Core Safety Limits and protection functions (OTT, OPT, FI, etc.) will be evaluated as a consequence of this implementation (DOM-NAF-2-P-A Item 3). The list of FSAR transients for which the code/correlation pair will be applied is also included herein (DOM-NAF-2-P-A Item 4).

2.2 Topical Report VEP-NE-2-A In 1985, Virginia Power (Dominion Energy) submitted to the USNRC Topical Report VEP-NE-2-A (Reference

1) describing a proposed methodology for the statistical treatment of key uncertainties in core thermal-hydraulic DNBR analysis. The methodology provided DNBR margin through the use of statistical rather than deterministic uncertainty treatment. The methodology was reviewed and approved by the USNRC in May 1987, and the Safety Evaluation Report (SER) provided by the USNRC listed the following conditions for its use (Reference 11):

1) The selection and justification of the Nominal Statepoints used to perform the plant specific implementation must be included in the submittal (Sections 3.6 and 3.8).

2) Justification of the distribution, mean, and standard deviation for all the statistically treated parameters must be included in the submittal (Section 3.2).

3) Justification of the value of model uncertainty must be included in the plant specific submittal (Section 3.4).

4) For the relevant CHF correlations, justification of the 95/95 DNBR limit and the normality of the M/P distribution, its mean and standard deviation must be included in the submittal, unless there is an approved Topical Report documenting these (such as Reference 2 and Reference 3, as supplemented by References 4 and 5).

This report provides the technical basis (Conditions 1 through 4 above) for the USNRC review and approval of the implementation of the Dominion Energy Statistical DNBR Evaluation Methodology for Framatome 17x17 GAIA fuel at MPS3 with the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs (VEP-NE-2-A Condition 1). This report documents the selection and justification of the Nominal Statepoints for MPS3 (VEP-NE-2-A Condition 1), the justification of the distribution, mean and standard deviation for all the statistically treated parameters (VEP-NE-2-A Condition 2), and the justification of the value of model uncertainty for MPS3 (VEP-NE-2-A Condition 3).

Serial No.23-251 Docket No. 50-423, Page 5 of 29 3 Implementation of the Statistical DNBR Evaluation Methodology 3.1 Methodology Review In Appendix F to Fleet Report DOM-NAF-2-P-A (Reference 3, as supplemented by References 4 and 5),

Dominion Energy calculated DDLs for the VIPRE-D/ORFEO-GAIA and the VIPRE-D/ORFEO-NMGRID code/correlation pairs. The Statistical DNBR Evaluation Methodology (Reference 1) is employed herein to determine SDLs for VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs for Framatome GAIA fuel at MPS3. These new limits combine the correlation uncertainty with the DNBR sensitivities to uncertainties in key DNBR analysis input parameters. Even though the new DNBR limits (the SDLs) are larger than the deterministic code/correlation design limits, its use is advantageous as the Statistical DNBR Evaluation Methodology permits the use of nominal values for operating initial conditions instead of requiring the application of evaluated uncertainties to the initial conditions for statepoint and transient analysis.

The SDLs are developed by means of a Monte Carlo analysis. The variation of actual operating conditions about nominal statepoints due to parameter measurement and other key DNB uncertainties is modeled through the use of a random number generator. Two thousand random statepoints are generated for each nominal statepoint. The random statepoints are then supplied to the thermal-hydraulics code VIPRE-D, which calculates the minimum DNBR (MDNBR) for each statepoint. Each MDNBR is randomized by a code/correlation uncertainty factor as described in Reference 1 using the upper 95% confidence limit on the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pair measured-to-predicted (M/P)

CHF ratio standard deviation (Reference 3). The standard deviation of the resultant randomized DNBR distribution is increased by a small sample correction factor to obtain a 95% upper confidence limit and is then combined Root-Sum-Square with code and model uncertainties to obtain a total DNBR standard deviation (stotal). In accordance with Reference 1, the SDL is then calculated as:

SDL = 1 + 1.645

stotal

[Equation 3.1]

in which the 1.645 multiplier is the z-value for the one-sided 95% probability of a normal distribution. This SDL provides peak fuel rod DNB protection at greater than 95/95.

As an additional criterion, the SDL is tested to determine the full core DNB probability when the peak pin reaches the SDL. This process is performed by summing the DNB probability of each rod in the core, using a bounding fuel rod census curve and the DNB sensitivity to rod power. If necessary, the SDL is increased to reduce the full core DNB probability to 0.1% or less.

3.2 Uncertainty Analysis This section is included herein to satisfy Condition 2 in the SER (Reference 11) of VEP-NE-2-A (Reference 1).

Serial No.23-251 Docket No. 50-423, Page 6 of 29 Consistent with VEP-NE-2-A, inlet temperature, pressurizer pressure, core thermal power, reactor vessel flow rate, core bypass flow, the nuclear enthalpy rise factor and the engineering enthalpy rise factor were selected as the statistically treated parameters in the implementation analysis. The magnitudes and functional forms of the uncertainties for the statistically treated parameters were derived in a rigorous analysis of plant hardware and measurement/calibration procedures and have been summarized in Table 3.2-1.

The uncertainties for core thermal power, vessel flow rate, pressurizer pressure and inlet temperature were quantified using all sensor, rack, and other components of a total uncertainty and combined in a manner consistent with their relative dependence or independence. Total uncertainties were quantified at the 2 level, corresponding to two-sided 95% probability. Margin was included in these uncertainties to allow for future changes in plant hardware or calibration procedures without invalidating the analysis. The standard deviations,

, were obtained by dividing the total uncertainty by 1.96, which is the z-value for the two-sided 95% probability of a normal distribution.

The magnitude and distribution of uncertainty for pressurizer pressure (system pressure) per the pressurizer pressure control system were quantified. The pressurizer pressure uncertainty is a normal, two-sided, 95%

probability distribution. The calculations of the MPS3 SDLs use a standard deviation () of 30.0 psia.

The magnitude and distribution of uncertainty for the average temperature (TAVG) per the Tavg rod control system were quantified. The TAVG uncertainty is a normal, two-sided, 95% probability distribution. The calculations of the MPS3 SDLs use a standard deviation () of 3.0 F.

The core thermal power uncertainty is defined as a normal, two-sided, 95% probability distribution. The calculations of the MPS3 SDLs use a standard deviation () of 1.0%.

The reactor coolant system (RCS) flow uncertainty is defined as a normal, two-sided, 95% probability distribution. The calculations of the MPS3 SDLs use a standard deviation () of 1.5%.

The measured FHN uncertainty is defined as a normal, two-sided, 95% probability distribution. The calculations of the MPS3 SDLs use a standard deviation () of 2.04%.

The magnitude and distribution of uncertainty on the engineering hot channel factor, FHE, is defined as a uniform probability distribution with a magnitude of +/-4.0%. The Statistical DNBR Evaluation Methodology (Reference 1) treats the FHE uncertainty as a uniform probability distribution.

The total core bypass flow consists of separate flow paths through the thimble tubes, direct leakage to the outlet nozzle, baffle joint leakage flow, upper head spray flow and core-baffle gap flow. These five components were each quantified based on the current MPS3 core configuration, their uncertainties conservatively modeled, and the flows and uncertainties totaled. The Monte Carlo analysis ultimately used a best estimate bypass flow of 7.6% with an uncertainty of 1.0%. The analysis assumed that the probability was uniformly distributed.

Serial No.23-251 Docket No. 50-423, Page 7 of 29 Table 3.2-1: Millstone Unit 3 Parameter Uncertainties PARAMETER NOMINAL VALUE STANDARD DEVIATION UNCERTAINTY DISTRIBUTION Pressure

[psia]

2250 30.0 psia

+/-58.8 psi at 2 Normal Temperature

[°F]

557.06 3.0°F

+/-5.88°F at 2 Normal Power [MWt]

3,712 1.0%

+/-1.96% at 2 Normal Flow [gpm]

379,200 1.5%

+/-2.94% at 2 Normal FHN 1.635 2.04%

+/-4.0% at 2 Normal FHE 1.0 N/A

+/-4.0%

Uniform Bypass [%]

7.6 N/A

+/-1.0%

Uniform 3.3 CHF Correlations The VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID CHF code/correlation pairs are used for the calculation of DNBRs in the Framatome 17x17 GAIA fuel product within the restrictions of Reference 3, as supplemented by References 4 and 5, and are applicable to the operating conditions for which the Statistical DNBR Evaluation Methodology applies.

Table 3.3-1 presents the correlation data for the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs.

The USNRC raised a concern on the use of the calculated CHF correlation statistics for the determination of statistical DNB limits in USNRC Information Notice IN-2014-01 (Reference 12). Per the methodology of VEP-NE-2-A (Reference 1), the use of the calculated standard deviation is acceptable (References 6, 7, and 8).

The acceptability of the use of the calculated standard deviation is based on the use of a 95% upper confidence factor that is essentially equivalent to the Owens tables for ensuring a 95% probability at a 95% confidence limit. However, the VIPRE-D/ORFEO-GAIA DDL includes a 0.01 bias due to fuel design variations (Reference 3, as supplemented by References 4 and 5). The VIPRE-D/ORFEO-NMGRID DDL includes no bias. For consistency, the development of the ORFEO-GAIA SDL and the ORFEO-NMGRID SDL herein used the back calculated standard deviations (S(M/P) in Table 3.3-1) based on the CHF statistics and the DDLs submitted for approval in DOM-NAF-2-P-A, Appendix F (Reference 3, as supplemented by References 4 and 5). This provides additional conservatism if no bias is included into the DDL.

Serial No.23-251 Docket No. 50-423, Page 8 of 29 Table 3.3-1: CHF Code/Correlation Data (Reference 3, as supplemented by References 4 and 5)

[(See Attachment 4, Proprietary INSERT 2)]

3.4 Model Uncertainty Term This section is included herein to satisfy Condition 3 in the SER (Reference 11) of the Statistical DNBR Evaluation Methodology Topical Report (Reference 1).

The VIPRE-D 21-channel production model for MPS3 with the Framatome 17x17 GAIA fuel product was used in the development of the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pair SDLs for MPS3. Since the production model that Dominion Energy intends to use for all MPS3 evaluations was used to develop the SDLs, there is no additional uncertainty associated with the use of this model. In summary, it is concluded that no correction for model uncertainty is necessary, and the model uncertainty term is set to zero for the calculation of the total DNBR standard deviation.

3.5 Code Uncertainty The code uncertainty accounts for any differences between Dominion Energys VIPRE-D and other thermal-hydraulic codes, with which the ORFEO-GAIA and ORFEO-NMGRID CHF data were correlated, and any effect due to the modeling of a full core with a correlation based upon bundle test data. These uncertainties are clearly independent of the correlation, the model, and parameter induced uncertainties. The code uncertainty was quantified at 5%, consistent with the factors specified for other thermal/hydraulic codes in Reference 1. The basis for this uncertainty is described in detail by USNRC Staff in Reference 11. In Reference 11, the USNRC Staff refers to the 5% uncertainty as being a 2 value. The 5% code uncertainty serves as a conservative factor that may be shown to be wholly or partially unnecessary at a later time. A one-sided 95%

confidence level on the code uncertainty is then 3.04% ( = ( 5.0%) 1.645 ). The use of the 1.645 divisor (the one-sided 95% tolerance interval multiplier) is conservative since the USNRC Staff considers the 5.0%

uncertainty to be a 2 value.

3.6 Monte Carlo Calculations In order to perform the Monte Carlo analysis, nine Nominal Statepoints covering the full range of normal operation and anticipated transient conditions were selected for both the ORFEO-GAIA and ORFEO-NMGRID CHF correlations. These statepoints must span the range of conditions over which the statistical methodology

Serial No.23-251 Docket No. 50-423, Page 9 of 29 will be applied. The Nominal Statepoints were selected to cover the DNB limiting range of the Reactor Core Safety Limits (RCSL) and within the validation range of applicability of the associated correlations. In order to apply the methodology to low flow events, a low flow statepoint is also included. The selected Nominal Statepoints are listed in Tables 3.6-1 and 3.6-2.

Table 3.6-1: Nominal Statepoints for Framatome 17x17 GAIA Fuel at MPS3 with VIPRE-D/ORFEO-GAIA STATEPOINT PRESSURIZER PRESSURE

[psia]

INLET TEMPERATURE

[F]

POWER

[%]

FLOW

[%]

FHN MDNBR A

2425 574.47 121 100 1.635 1.230 B

2425 635.51 70 100 1.7822 1.230 C

2250 564.48 121 100 1.635 1.230 D

2250 619.08 76 100 1.7527 1.230 E

2000 548.82 121 100 1.635 1.230 F

2000 596.02 85 100 1.7086 1.230 G

1840 539.01 121 100 1.635 1.230 H

1840 578.12 95 100 1.6595 1.230 I

2250 557.06 100 73.90 1.635 1.230 Table 3.6-2: Nominal Statepoints for Framatome 17x17 GAIA Fuel at MPS3 with VIPRE-D/ORFEO-NMGRID STATEPOINT PRESSURIZER PRESSURE

[psia]

INLET TEMPERATURE

[F]

POWER

[%]

FLOW

[%]

FHN MDNBR A

2425 455.78 121 100 1.635 1.300 B

2425 555.34 80 100 1.7331 1.300 C

2250 438.95 121 100 1.635 1.300 D

2250 544.38 80 100 1.7331 1.300 E

2000 417.65 121 100 1.635 1.300 F

2000 529.49 80 100 1.7331 1.300 G

1840 405.96 121 100 1.635 1.300 H

1840 521.07 80 100 1.7331 1.300 I

2250 480.00 90 75.64 1.6841 1.300

Serial No.23-251 Docket No. 50-423, Page 10 of 29 The Monte Carlo analysis itself consisted of 2000 calculations performed around each of the nine Nominal Statepoints for each CHF correlation. As described in Section 3.1, the DNBR standard deviation at each Nominal Statepoint was augmented by the code/correlation uncertainty, the small sample correction factor, and the code uncertainty to obtain a total DNBR standard deviation.

The Total sTotal, is obtained using the Root-Sum-Square method according to Equation 3.2:

1.0

[Equation 3.2]

where:

sDNBR is the standard deviation for the Randomized DNBR distribution.

The factor

1.0 is the uncertainty in the standard deviation of the 2,000 Monte Carlo simulations and provides a 95% upper confidence limit on the standard deviation.

1 is the uncertainty in the mean of the correlation. N is the number of degrees of freedom in the original correlation database (given in Reference 3 for ORFEO-GAIA and ORFEO-NMGRID).

FC is the code uncertainty, that has been defined as 5% (2 value), i.e., 5.0%

1.645

3.04% (1 value). See Section 2.5 in Reference 1.

FM is the model uncertainty, which is 0.0 since the Monte Carlo simulation is run with the production model.

Note that this equation differs slightly from the equation listed in Reference 1. It has an additional factor applied to the Randomized DNBR sDNBR, the 1 factor to correct for the uncertainty in the mean of the correlation.

This factor has been used in previous implementations of the Statistical DNBR Evaluation Methodology, such as References 6, 7, and 8.

The limiting peak fuel rod SDL was calculated to be 1.243 for the VIPRE-D/ORFEO-GAIA code/correlation pair and 1.261 for the VIPRE-D/ORFEO-NMGRID code/correlation pair. The Monte Carlo Statepoint analysis is summarized in Tables 3.6-3 and 3.6-4.

Serial No.23-251 Docket No. 50-423, Page 11 of 29 Table 3.6-3: Peak Pin SDL Results for Framatome 17x17 GAIA Fuel at MPS3 with VIPRE-D/ORFEO-GAIA STATEPOINT Randomized DNB sDNBR Total DNB sTOTAL Pin Peak SDL95/95 A

0.1390 0.1479 1.243 B

0.1211 0.1297 1.213 C

0.1363 0.1451 1.239 D

0.1218 0.1304 1.215 E

0.1338 0.1427 1.235 F

0.1149 0.1235 1.203 G

0.1288 0.1375 1.226 H

0.1148 0.1234 1.203 I

0.1359 0.1448 1.238 Table 3.6-4: Peak Pin SDL Results for Framatome 17x17 GAIA Fuel at MPS3 with VIPRE-D/ORFEO-NMGRID STATEPOINT Randomized DNB sDNBR Total DNB sTOTAL Pin Peak SDL95/95 A

0.1457 0.1542 1.254 B

0.1498 0.1584 1.261 C

0.1464 0.1549 1.255 D

0.1448 0.1533 1.252 E

0.1437 0.1522 1.250 F

0.1464 0.1549 1.255 G

0.1388 0.1472 1.242 H

0.1428 0.1512 1.249 I

0.1435 0.1520 1.250 3.7 Full Core DNB Probability Summation After the development of the peak pin 95/95 DNBR limit, the data statistics are used to determine the number of rods expected in DNB. The DNB sensitivity to rod power is estimated as (DNBR)/ (1/FH). The specific values of (DNBR)/ (1/FH), denoted, are listed in Tables 3.7-1 and 3.7-2.

Serial No.23-251 Docket No. 50-423, Page 12 of 29 To ensure that the calculations are conservative, a one-sided tolerance limit of is used:

in which:

is the one-sided tolerance limit on t() is the T-statistic with significance level and degrees of freedom. For 2,000 observations at a 0.05 level of significance t(0.05,2000) = 1.645.

se() is the standard error of.

The resulting linear regression model, which includes the independent variable 1/FH, yields R2 values larger than 99%.

Table 3.7-1: (DNBR)/ (1/FH) Estimation for ORFEO-GAIA STATEPOINT se()

R2 A

3.4299 0.010 3.413450 99.8%

B 2.7091 0.008 2.695940 99.8%

C 3.4175 0.011 3.399405 99.7%

D 2.7105 0.008 2.697340 99.8%

E 3.3608 0.016 3.334480 99.4%

F 2.5954 0.010 2.578950 99.6%

G 3.1330 0.017 3.105035 99.2%

H 2.4723 0.010 2.455850 99.5%

I 3.4752 0.016 3.448880 99.5%

Serial No.23-251 Docket No. 50-423, Page 13 of 29 Table 3.7-2: (DNBR)/ (1/FH) Estimation for ORFEO-NMGRID STATEPOINT se()

R2 A

2.4771 0.002 2.473810 99.9%

B 2.6920 0.002 2.688710 99.9%

C 2.4603 0.002 2.457010 100.0%

D 2.6916 0.002 2.688310 99.9%

E 2.4435 0.002 2.440210 100.0%

F 2.6613 0.002 2.658010 100.0%

G 2.4294 0.001 2.427755 100.0%

H 2.6490 0.002 2.645710 100.0%

I 2.6004 0.002 2.597110 99.9%

A representative fuel rod census curve used for the determination of the SDL is listed in Table 3.7-3. The full core DNB probability summation is evaluated on a reload basis to verify the applicability of the fuel rod census (FHN versus % of core with FHN greater than or equal to a given FH limit) used in the implementation analysis.

The limiting full-core DNB probability summation resulted in an SDL of 1.251 for ORFEO-GAIA and 1.298 for ORFEO-NMGRID.

The DNB probability summations for VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs are summarized in Tables 3.7-4 and 3.7-5, respectively.

Serial No.23-251 Docket No. 50-423, Page 14 of 29 Table 3.7-3: Representative Fuel Rod Census for a Maximum Peaking Factor FH = 1.635 MAXIMUM % OF FUEL RODS IN CORE WITH FH to:

FH LIMIT 0.0 1.635 0.1 1.633 0.2 1.630 0.3 1.627 0.4 1.625 0.5 1.622 0.6 1.617 0.7 1.615 0.8 1.612 0.9 1.608 1.0 1.604 1.5 1.593 2.0 1.581 2.5 1.571 3.0 1.562 4.0 1.549 5.0 1.538 6.0 1.527 7.0 1.515 8.0 1.505 9.0 1.496 10.0 1.490 20.0 1.458 30.0 1.411 40.0 1.349 PEAK 1.635

Serial No.23-251 Docket No. 50-423, Page 15 of 29 Table 3.7-4: Full Core DNB Probability Summation for Framatome 17x17 GAIA Fuel at MPS3 with VIPRE-D/ORFEO-GAIA STATEPOINT sTOTAL

% of Rods in DNB Full Core SDL99.9 A

0.1479 0.0989 1.251 B

0.1297 0.0989 1.228 C

0.1451 0.0988 1.245 D

0.1304 0.0988 1.230 E

0.1427 0.0990 1.241 F

0.1235 0.0988 1.217 G

0.1375 0.0990 1.235 H

0.1234 0.0990 1.220 I

0.1448 0.0989 1.243 Table 3.7-5: Full Core DNB Probability Summation for Framatome 17x17 GAIA Fuel at MPS3 with VIPRE-D/ORFEO-NMGRID STATEPOINT sTOTAL

% of Rods in DNB Full Core SDL99.9 A

0.1542 0.0989 1.295 B

0.1584 0.0990 1.298 C

0.1549 0.0989 1.297 D

0.1533 0.0989 1.285 E

0.1522 0.0990 1.291 F

0.1549 0.0990 1.290 G

0.1472 0.0990 1.279 H

0.1512 0.0989 1.282 I

0.1520 0.0989 1.285

Serial No.23-251 Docket No. 50-423, Page 16 of 29 3.8 Verification of Nominal Statepoints Condition 1 of the USNRCs SER for VEP-NE-2-A (Reference 11) requires that the Nominal Statepoints be shown to provide a bounding DNBR standard deviation for any set of conditions to which the methodology may potentially be applied.

It is therefore necessary to demonstrate that stotal as calculated herein is maximized for any conceivable set of conditions at which the core may approach the SDL. To do so, a regression analysis is performed using the unrandomized DNBR standard deviations at each Nominal Statepoint as the dependent variable (i.e., the raw MDNBR results obtained from the Monte Carlo simulation). The Nominal Statepoint pressures, inlet temperatures, powers and flow rates are used as the independent variable. If no clear trend appears in the plot it can be concluded that the standard deviation has been maximized. If a clear trend is displayed, the regression function is determined. This regression equation is evaluated to determine the values of the independent variable for which the standard deviation would be maximized, and it is verified that the Nominal Statepoints selected bound those conditions. In addition, the residuals of the regression are plotted again against all the independent variables, and it is verified that no trends are discernible.

Tables 3.8-1 and 3.8-2 show the R2 coefficients obtained for the verification of the nominal statepoints for ORFEO-GAIA and ORFEO-NMGRID, respectively. The largest linear curve fit R2 coefficient is 51.79% for ORFEO-GAIA and 74.31% for ORFEO-NMGRID. An evaluation has been performed considering all the data, linear fits, and R2 coefficients. It was concluded that sTOTAL had been maximized for any conceivable set of conditions at which the core may approach the SDL and that the selected Nominal Statepoints provide a bounding standard deviation for any set of conditions to which the methodology may potentially be applied.

Figure 3.8-1 and Figure 3.8-2 display a sample regression plots for ORFEO-GAIA and ORFEO-NMGRID, respectively, and clearly shows the trends discussed above.

Table 3.8-1: R2 Coefficients for the Verification of the Nominal Statepoints for Framatome 17x17 GAIA Fuel at MPS3 with VIPRE-D/ORFEO-GAIA R2 - Linear Regression Pressure 15.55%

Temperature 33.46%

Flow Rate 14.07%

Power 51.79%

Serial No.23-251 Docket No. 50-423, Page 17 of 29 Table 3.8-2: R2 Coefficients for the Verification of the Nominal Statepoints for Framatome 17x17 GAIA Fuel at MPS3 with VIPRE-D/ORFEO-NMGRID R2 - Linear Regression Pressure 29.78%

Temperature 74.31%

Flow Rate 7.77%

Power 43.02%

Serial No.23-251 Docket No. 50-423, Page 18 of 29 Figure 3.8-1: Variation of the Unrandomized Standard Deviation with Power for the ORFEO-GAIA CHF Correlation y=0.0005x+0.0387 R²=0.5179 0.03 0.04 0.05 0.06 0.07 0.08 0.09 0.1 0.11 60 70 80 90 100 110 120 130 SIGMADNBR Power[%]

Serial No.23-251 Docket No. 50-423, Page 19 of 29 Figure 3.8-2: Variation of the Unrandomized Standard Deviation with Temperature for the ORFEO-NMGRID CHF Correlation 3.9 Scope of Applicability This section is included herein to satisfy Item 4 of the plant specific application list in Section 2.1 of DOM-NAF-2-P-A (Reference 2).

The Statistical DNBR Evaluation Methodology may be applied to all Condition I and II DNB events (except Rod Withdrawal from Subcritical (RWSC) which is initiated from zero power), and to the Complete Loss of Flow and the Locked Rotor Accident. The accidents to which the methodology is applicable are listed in Table 3.9-1. This table corresponds to Table 2.1-1 in Reference 2. The range of application is consistent with previous applications of Dominion Energys Statistical DNBR Evaluation Methodology at MPS3. This methodology will not be applied to accidents that are initiated from zero power where the parameter uncertainties are higher.

The Statistical DNBR Evaluation Methodology provides analytical margin by permitting transient analyses to be initiated from nominal operating conditions, and by allowing core thermal limits to be generated without the y=4E05x+0.0242 R²=0.7431 0.03 0.04 0.05 0.06 0.07 0.08 0.09 0.1 0.11 400 420 440 460 480 500 520 540 560 SIGMADNBR Temperature[°F]

Serial No.23-251 Docket No. 50-423, Page 20 of 29 application of the bypass flow, FHN (measurement component) and hot channel uncertainties. These uncertainties are convoluted statistically into the DNBR limit.

Table 3.9-1: UFSAR Transients Analyzed with VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID for MPS3 ACCIDENT MPS3 FSAR SECTION APPLICATION Feedwater system malfunctions that result in a decrease in feedwater temperature (Excessive heat removal due to feedwater system malfunctions) 15.1.1 STAT-DNB Feedwater system malfunctions that result in an increase in feedwater flow (Excessive heat removal due to feedwater system malfunctions) 15.1.2 STAT-DNB Excessive increase in secondary steam flow (Excessive load increase) 15.1.3 STAT-DNB Inadvertent opening of a steam generator relief or safety valve (Accidental depressurization of the main steam system) 15.1.4 DET-DNB Steam system piping failure (Rupture of a main steam pipe) 15.1.5 DET-DNB Loss of external electrical load (Loss of external electrical load and/or turbine trip) 15.2.2 STAT-DNB Turbine trip (Loss of external electrical load and/or turbine trip) 15.2.3 STAT-DNB Inadvertent closure of a main steam isolation valve (Loss of external electrical load and/or turbine trip) 15.2.4 STAT-DNB Loss of condenser vacuum and other events resulting in turbine trip (Loss of external electrical load and/or turbine trip) 15.2.5 STAT-DNB Loss of nonemergency AC power to the station auxiliaries (Loss of external electrical load and/or turbine trip) 15.2.6 STAT-DNB Loss of normal feedwater flow (Loss of normal feedwater flow) 15.2.7 STAT-DNB Feedwater system pipe break (Major rupture of a main feed water pipe) 15.2.8 STAT-DNB Partial loss of forced reactor coolant flow (Loss of forced reactor coolant flow) 15.3.1 STAT-DNB

Serial No.23-251 Docket No. 50-423, Page 21 of 29 ACCIDENT MPS3 FSAR SECTION APPLICATION Complete loss of forced reactor coolant flow(Loss of forced reactor coolant flow) 15.3.2 STAT-DNB Reactor coolant pump shaft seizure (Locked reactor coolant pump rotor or shaft break) 15.3.3 STAT-DNB Reactor coolant pump shaft break (Locked reactor coolant pump rotor or shaft break) 15.3.4 STAT-DNB Uncontrolled rod cluster control assembly bank withdrawal from a subcritical or low power startup condition (Rod cluster control assembly bank withdrawal from subcritical) 15.4.1 DET-DNB Uncontrolled rod cluster control assembly bank withdrawal at power (Rod cluster control assembly bank withdrawal at power) 15.4.2 STAT-DNB Rod cluster control assembly misalignment (Single rod cluster control assembly withdrawal at full power) 15.4.3 STAT-DNB Chemical and volume control system malfunction that results in a decrease in the boron concentration in the reactor coolant (Uncontrolled boron dilution) 15.4.6 STAT-DNB Inadvertent operation of the emergency core cooling system during power operation (Inadvertent operation of the emergency core cooling system during power operation) 15.5.1 STAT-DNB Inadvertent opening of a pressurizer safety or relief valve (Accidental depressurization of the reactor coolant system) 15.6.1 STAT-DNB 3.10 Summary of Analysis The steps of the SDL derivation analysis may be summarized as follows:

In accordance with the Statistical DNBR Evaluation Methodology, 2,000 random statepoints are generated about each nominal statepoint and VIPRE-D is then executed to obtain MDNBRs. The standard deviation for the distribution of 2,000 MDNBRs is referred to as the unrandomized standard deviation. At the limiting Nominal Statepoint, the standard deviation of the randomized DNBR distributions, which is the unrandomized standard deviation corrected for CHF correlation uncertainty. This value was then combined Root Sum Square with code and model uncertainty standard deviations to obtain a total DNBR standard deviation, listed in Tables 3.6-3 and 3.6-4. The use of total DNBR standard deviation in Equation 3.1 yields a peak pin DNBR

Serial No.23-251 Docket No. 50-423, Page 22 of 29 limit of 1.243 for VIPRE-D/ORFEO-GAIA and 1.261 for VIPRE-D/ORFEO-NMGRID with at least 95%

probability at a 95% confidence level. The total DNBR standard deviation was then used to obtain 99.9% DNB protection in the full core DNBR limit of 1.251 for VIPRE-D/ORFEO-GAIA and 1.298 for VIPRE-D/ORFEO-NMGRID.

Therefore, the VIPRE-D/ORFEO-GAIA code/correlation pair SDL for MPS3 Framatome 17x17 GAIA Fuel is set to 1.26 and the VIPRE-D/ORFEO-NMGRID code/correlation pair SDL for MPS3 Framatome 17x17 GAIA Fuel is set to 1.31.

Serial No.23-251 Docket No. 50-423, Page 23 of 29 4 Application of VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID to MPS3 The VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID CHF code/correlation pairs are used for the calculation of DNBRs in the Framatome 17x17 GAIA fuel product within the restrictions of Reference 3, as supplemented by References 4 and 5.

VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs together with the Statistical DNBR Evaluation Methodology will be applied to all Condition I and II DNB events (except Rod Withdrawal from Subcritical, RWSC), and to the Complete Loss of Flow events and the Locked Rotor Accident. The Statistical DNBR Evaluation Methodology provides analytical margin by permitting transient analyses to be initiated from nominal operating conditions, and by allowing core thermal limits to be generated without the application of the bypass flow, FHN (measurement component) and FHE (engineering component) uncertainties. These uncertainties are convoluted statistically into the DNBR limit.

In addition, there are a few events that will be evaluated with deterministic models because they do not meet the applicability requirements of the Statistical DNBR Evaluation Methodology (see the events in Table 3.9-1 labeled DET-DNB). These events are initiated from bounding operating conditions considering the nominal value and the appropriate uncertainty value, and require the application of the bypass flow, FHN (measurement component) and FHE uncertainties. The events modeled deterministically are limited by the DDLs stated in DOM-NAF-2-P-A (Reference 3, as supplemented by References 4 and 5).

4.1 VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID SDLs for MPS3 The SDLs for MPS3 cores containing Framatome 17x17 GAIA fuel being analyzed with the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs were derived in Section 3 of this report. The SDL for the VIPRE-D/ORFEO-GAIA code/correlation pair is determined to be 1.26. The SDL for the VIPRE-D/ORFEO-NMGRID code/correlation pair is determined to be 1.31. The SDL limit provides a peak fuel rod DNB protection with at least 95% probability at a 95% confidence level and a 99.9% DNB protection for the full core. This SDL is plant specific as it includes MPS3 specific uncertainties for the key parameters accounted for in the application of the Statistical DNBR Evaluation Methodology. Therefore, these limits are applicable to the analysis of statistical DNB events of Framatome 17x17 GAIA fuel in MPS3 cores with the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs.

4.2 Safety Analysis Limits (SAL)

In the performance of in-house DNB thermal-hydraulic evaluations, design limits and safety analysis limits are used to define the available retained DNBR margin for each application. The difference between the safety analysis (self-imposed) limit and the design limit is the available retained DNBR margin.

Serial No.23-251 Docket No. 50-423, Page 24 of 29 For deterministic DNB analyses, the design DNBR limit is set equal to the applicable code/correlation limit and it is termed the DDL. For statistical DNB analyses, the design DNBR limit is set equal to the applicable SDL.

These design limits are two of the Design Basis Limits for Fission Product Barriers (DBLFPB) described in Reference 13. The DDLs and SDLs are fixed and any changes to their value require USNRC review and approval. However, the safety analysis limits for deterministic and statistical DNB analyses (SALDET and SALSTAT, respectively) may be changed without prior USNRC review and approval, provided the changes meet the criteria established in Reference

13.

DNBR limits for the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs are listed in Table 4.2-1.

Table 4.2-1: DNBR Limits for ORFEO-GAIA and ORFEO-NMGRID ORFEO-GAIA DDL 1.13 SDL 1.26 SALDET 1.45 SALSTAT 1.45 ORFEO-NMGRID DDL 1.21 SDL 1.31 SALDET 1.45 SALSTAT 1.45

Serial No.23-251 Docket No. 50-423, Page 25 of 29 4.3 Retained DNBR Margin The difference between the safety analysis (self-imposed) limit and the design limit (the DBLFPB) is the available retained DNBR margin:

%100

The resulting available retained DNBR margins are listed in Tables 4.3-1 and 4.3-2.

Table 4.3-1: DNBR Limits and Retained DNBR Margin for Deterministic DNB Applications DETERMINISTIC DNB APPLICATIONS DNB CORRELATION DDL SALDET RETAINED DNBR MARGIN [%]

ORFEO-GAIA 1.13 1.45 22.0 ORFEO-NMGRID 1.21 1.45 16.5 Table 4.3-2: DNBR Limits and Retained DNBR Margin for Statistical DNB Applications STATISTICAL DNB APPLICATIONS DNB CORRELATION SDL SALSTAT RETAINED DNBR MARGIN [%]

ORFEO-GAIA 1.26 1.45 13.1 ORFEO-NMGRID 1.31 1.45 9.6 This method of defining retained DNBR margin allows all of the DNBR margin to be found in a single, clearly defined location. The retained DNBR margin may be used to offset generic DNBR penalties, such as a rod bow penalty and a transition core penalty for mixed cores.

The reload thermal-hydraulics evaluation prepared as part of the reload safety analysis process (Reference

13) presents tables and descriptions of retained DNBR margin and applicable penalties. Retained DNBR margin is tracked separately for each CHF correlation and for statistical and deterministic analyses.

Serial No.23-251 Docket No. 50-423, Page 26 of 29 4.4 Verification of Existing Reactor Core Safety Limits, Protection Setpoints and MPS3 UFSAR Chapter 15 Events This section is included herein to satisfy Item 3 of the plant specific application list in Section 2.1 of DOM-NAF-2-P-A (Reference 2).

To demonstrate that the DNB performance of the Framatome 17x17 GAIA fuel is acceptable, Dominion Energy will perform calculations for full-core configurations of Framatome 17x17 GAIA fuel. The calculations will be performed using the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs and selected statepoints including: the reactor core safety limits (RCSL), various axial flux shapes, rod withdrawal from subcritical (RWSC), rod withdrawal at power (RWAP), loss of flow (LOFA), locked rotor events (LOCROT), hot zero power steam line break (MSLB), dropped rod limit line (DRLL), and static rod misalignment (SRM). These various statepoints provide sensitivity of DNB performance to the following: (a) power level (including the impact of the part-power multiplier on the allowable hot rod power FH), pressure and temperature (RCSL); (b) various axial flux shapes at several axial offsets (AO); and (c) low flow (LOFA and LOCROT). The statepoints for the RWSC and MSLB will be evaluated with deterministic DNB methods.

The remaining statepoints will be evaluated using statistical DNB methods. The evaluation criterion for these analyses is that the minimum DNBR must be equal to or greater than the applicable safety analysis limit (SAL) listed in Table 4.2-1.

The results of the calculations demonstrate that changes to reactor protection setpoints are required to ensure minimum DNBR values are equal to or greater than the applicable safety analysis limit for the analyses performed to address statepoints of the Reactor Core Safety Limits (RCSLs), the OTT, and Permissive 8 (P-

8) trip setpoints. The RCSL and OTT analyses were performed with an FH of 1.587 and a Rated Thermal Power (RTP) of 3712 MW, which is equal to the current FH limit and bounds the licensed RTP, respectively.

The results of the calculations also demonstrate that the minimum DNBR values are equal to or greater than the applicable safety analysis limit for the analyses that are performed to address statepoints of the OPT trip setpoints, as well as evaluated Chapter 15 events. The P-8 and evaluated Chapter 15 analyses were performed with an FH of 1.635 and a RTP of 3712 MW, which bound the current FH limit and licensed RTP, respectively. Dominion Energy will evaluate analyses for the FI trip reset function and remaining Chapter 15 events supporting the GAIA transition and submit for USNRC review any items for which such review is deemed necessary per the criteria of 10 CFR 50.59. Note, where used, an FH of 1.635 supports a potential future increase in the FH limit for Millstone Unit 3. Dominion Energy will evaluate any potential future increase in the FH limit and submit for USNRC review any items for which such review is deemed necessary per the criteria of 10 CFR 50.59.

Serial No.23-251 Docket No. 50-423, Page 27 of 29 5 Conclusions Dominion Energys Statistical DNBR Evaluation Methodology has been used to derive Statistical Design Limits (SDLs). This application employs the VIPRE-D code with the Framatome ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux (CHF) correlations (VIPRE-D/ORFEO-GAIA and VIPRE-D/

ORFEO-NMGRID code/correlation pairs) for the thermal-hydraulic analysis of Framatome 17x17 GAIA fuel assemblies at MPS3. In particular, Dominion Energy seeks the review and approval of the Statistical Design Limits (SDLs) of 1.26 and 1.31 for the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs, respectively, documented herein as per 10 CFR 50.59(c)(2)(vii) the SDL constitutes a Design Basis Limit for Fission Products Barrier (DBLFPB).

Dominion Energy is also seeking the approval for the inclusion of Appendix F of Fleet Report DOM-NAF-2-P-A (submitted in Reference 3, as supplemented by References 4 and 5), to the Technical Specification 6.9.1.6.b list of USNRC approved methodologies used to determine core operating limits (i.e., the reference list of the Millstone Unit 3 COLR). This would allow Dominion Energy the use of the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs to perform licensing calculations for Framatome 17x17 GAIA fuel assemblies in MPS3 cores, using the deterministic design limits (DDLs) qualified in Appendix F of Fleet Report DOM-NAF-2-P-A, and the SDLs documented herein.

Serial No.23-251 Docket No. 50-423, Page 28 of 29 6 References

1.

Dominion Energy Topical Report, VEP-NE-2-A, Statistical DNBR Evaluation Methodology, June 1987.

2.

Dominion Energy Fleet Report, DOM-NAF-2-P-A, Revision 0, Minor Revision 4, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code, with Appendices A, B, C, D, and E, March 2023.

3.

Letter from J. Holloway (Dominion Energy) to USNRC, Virginia Electric and Power Company (Dominion Energy Virginia), North Anna and Surry Power Stations Units 1 and 2, Dominion Energy Nuclear Connecticut, Inc. (DENC), Millstone Power Station Units 2 and 3, Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code, December 19, 2022. Serial No.22-362; ML22353A619 (Proprietary, Non-Public), ML22353A620 (Public).

4.

Letter from J. Holloway (Dominion Energy) to USNRC, Virginia Electric and Power Company (Dominion Energy Virginia), North Anna and Surry Power Stations Units 1 and 2, Dominion Energy Nuclear Connecticut, Inc. (DENC), Millstone Power Station Units 2 and 3, Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code Response to Request for Additional Information, April 6, 2023. Serial No. 22-362B; ML23096A297 (Proprietary, Non-Public), ML23096A298 (Public).

5.

Letter from J. Holloway (Dominion Energy) to USNRC, Virginia Electric and Power Company (Dominion Energy Virginia), North Anna and Surry Power Stations Units 1 and 2, Dominion Energy Nuclear Connecticut, Inc. (DENC), Millstone Power Station Units 2 and 3, Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code Response to Second Request for Additional Information, July 26, 2023. Serial No. 22-362D; ML23208A092 (Public).

6.

Letter from M. D. Sartain (DNC) to USNRC, Dominion Nuclear Connecticut, Inc., Millstone Power Station Unit 3, License Amendment Request to Adopt Dominion Core Design and Safety Analysis Methods and to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1, NSAL-15-1, and 06-IC-03., Dominion Serial No.15-159, dated May 8, 2015 (USNRC ADAMS Accession No. ML15134A244).

7.

Letter to D. A. Heacock (Dominion Nuclear) from R. V. Guzman (USNRC), "Millstone Power Station, Unit No. 3 - Issuance of Amendment Adopting Dominion Core Design and Safety Analysis Methods and Addressing the Issues Identified in Three Westinghouse Communication Documents (CAC No. MF6251),"

Dominion Serial No.16-317, dated July 28, 2016. (USNRC ADAMS Accession No. ML16131A728).

8.

Letter from M. D. Sartain (DNC) to USNRC, "Dominion Nuclear Connecticut, Inc., Millstone Power Station Unit 3, Response to Request For Additional Information Regarding License Amendment Request to Adopt Dominion Core Design and Safety Analysis Methods and to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1, NSAL-15-1, and 06-IC-03 (CAC No. MF6251), Dominion Serial No.16-011, dated January 28, 2016 (USNRC ADAMS Accession No. ML16034A216).

Serial No.23-251 Docket No. 50-423, Page 29 of 29

9.

Letter from C. E. Rossi (USNRC) to J. A. Blaisdell (UGRA Executive Committee), Acceptance for Referencing of Licensing Topical Report, EPRI NP-2511-CCM, VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores, Volumes 1, 2, 3 and 4, May 1, 1986.

10. Letter from A. C. Thadani (USNRC) to Y. Y. Yung (VIPRE-01 Maintenance Group), Acceptance for Referencing of the Modified Licensing Topical Report, EPRI NP-2511-CCM, Revision 3, VIPRE-01: A Thermal Hydraulic Analysis Code for Reactor Cores, (TAC No. M79498), October 30, 1993.

11. Letter from L. B. Engle (USNRC) to W. L. Stewart (Virginia Power), Statistical DNBR Evaluation Methodology, VEP-NE-2, Surry Power Station, Units No. 1 & No. 2 (Surry-1&2) and North Anna Power Station, Units No. 1 & No. 2 (NA-1&2), Serial No.87-335, May 28, 1987.

12. USNRC Information Notice, IN-2014-01, Fuel Safety Limit Calculation Inputs were Inconsistent with NRC-Approved Correlation Limit Values, February 21, 2014.

13. Technical Report, NEI 96-07, Revision 1, Guidelines for 10 CFR 50.59 Implementation, Nuclear Energy Institute, November 2000.

14. Dominion Energy Topical Report, VEP-FRD-42-A, Revision 2, Minor Revision 2, Reload Nuclear Design Methodology, October 2017.

Serial No.23-251 Docket No. 50-423 FRAMATOME APPLICATION FOR WITHHOLDING AND AFFIDAVIT Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

A F F I D A V I T 1.

My name is Morris Byram. I am Product Manager, Licensing & Regulatory Affairs for Framatome Inc. (Framatome) and as such I am authorized to execute this Affidavit.

2.

I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.

3.

I am familiar with the Framatome information contained in Attachment 4 to Dominion Energy letter Serial No. 23-251R(O), entitled Dominion Energy Nuclear Connecticut, Inc. Millstone Power Station Unit 3 Proposed Amendment to Support Implementation of Framatome GAIA Fuel, and referred to herein as Document. Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.

4.

This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5.

This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is Serial No.23-251 Docket No. 50-423, Page 1 of 3

requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.

6.

The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:

(a)

The information reveals details of Framatomes research and development plans and programs or their results.

(b)

Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c)

The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome.

(d)

The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability.

(e)

The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome.

The information in this Document is considered proprietary for the reasons set forth in paragraph 6(c) and 6(d) above.

7.

In accordance with Framatomes policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8.

Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

Serial No.23-251 Docket No. 50-423, Page 2 of 3

9.

The foregoing statements are true and correct to the best of my knowledge, information, and belief.

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

Executed on: (10/06/2023)

(NAME)

`

morris.byram@framatome.com BYRAM Morris Digitally signed by BYRAM Morris Date: 2023.10.06 10:19:51 -07'00' Serial No.23-251 Docket No. 50-423, Page 3 of 3

v t e Letter Sequence Request
EPID:L-2023-LLA-0074, (Approved, Closed)
Initiation Request Acceptance Administration Withholding Request Acceptance Questions 14 November 2023 Answers 21 December 2023 Results Approval Other: ML23145A195
MONTHYEAR2023-05-23 [Table View]

ML23304A047

Text

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

SUMMARY

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