Text

License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits and Exemption Request for Use of M5 Cladding
	ML23123A279
Person / Time
Site:	Millstone
Issue date:	05/02/2023
From:	James Holloway; Dominion Energy Nuclear Connecticut
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML23123A276	List: ML23123A279;
References
	23-105
	Download: ML23123A279 (1)
	v • d • e

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 Dom inion En ergy Nuclear Connecti cut, Inc.

5000 Dominion Bou leva rd, Glen All en , VA 23 060 Dominion Energy.com May 2, 2023 U. S. Nuclear Regulatory Commission Serial No.23-105 Attention: Document Control Desk NRA/SS: RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 LICENSE AMENDMENT REQUEST TO USE FRAMATOME SMALL BREAK AND REALISITIC LARGE BREAK LOSS OF COOLANT ACCIDENT EVALUATION METHODOLOGIES FOR ESTABLISHING CORE OPERATING LIMITS AND EXEMPTION REQUEST FOR USE OF M5' CLADDING Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENC) is submitting a license amendment request (LAR) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3). DENC proposes a change to MPS3 TS 6.9.1.6.b to add Framatome (FRM) Topical Reports EMF-2328-P-A, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based ," EMF-2103-P-A, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," and ANP-10349-P-A, "GALILEO Implementation in LOCA Methods" to the list of methodologies for establishing core operating limits. The proposed TS changes are needed to support the transition to FRM GAIA fuel with M5' cladding at MPS3, which requires the application of the FRM Small Break Loss of Coolant Accident (SBLOCA) and Realistic Large Break Loss of Coolant Accident (RLBLOCA) methodologies and the associated use of the GALILEO fuel performance code within the loss of coolant accident (LOCA) methods. DENC and FRM .

have entered into an agreement for batch implementation of the GAIA fuel at MPS3. A full reload batch of GAIA fuel assemblies is planned for initial insertion in Cycle 24. This onload is currently scheduled for the spring 2025 refueling outage.

DENC is also requesting an exemption pursuant to 10 CFR 50.12 to facilitate the use FRM GAIA fuel assemblies containing fuel rods fabricated with M5' cladding material at MPS3. Specifically, DENC is requesting an exemption from certain requirements of 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," and Appendix K of 10 CFR 50, "ECCS Evaluation Models". As written, 10 CFR 50.46 and 10 CFR 50, Appendix K presume the use of zircaloy or ZIRLO' fuel rod cladding. However, M5' cladding has a slightly different composition than either of these alloys. provides DENC's description and assessment of the proposed TS changes. provides the marked-up MPS3 TS pages to reflect the proposed amendment. A site-specific Licensing Report for MPS3 SBLOCA analysis with GAIA fuel is provided, with proprietary and non-proprietary versions included as Attachment 3 and 4, respectively. A site-specific Licensing Report for MPS3 RLBLOCA analysis with GAIA Attachments .3 &5 contain information that is being withheld from public disclosure under 10 CFR 2.390. Upon separation from Attachments 3 & 5, this letter is decontrolled.

Serial No.23-105 Docket No. 50-423 Page 2 of 4 fuel is provided, with proprietary and non-proprietary versions included in Attachment 5 and 6, respectively. Attachment 7 provides the FRM Application for Withholding and Affidavit. Attachment 8 provides the detailed basis and justification for the exemption request, which addresses the exemption requirements of 10 CFR 50.12.

Attachments 3 and 5 contain information proprietary to Framatome Inc. and are supported by an affidavit (Attachment 7) 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 respectfully requested the proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390.

The proposed amendment and proposed exemption do not involve a Significant Hazards Consideration under the standards set forth in 10 CFR 50.92. The basis for this determination is included in Attachment 1 for the proposed amendment and in Attachment 8 for the proposed exemption. DENG 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 and proposed exemption are eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9) and 10 CFR 51.22(c)(25), respectively. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with approval of the proposed amendment or the proposed exemption.

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

DENG requests approval of this LAR and exemption by May 2, 2024, with a 60-day implementation period.

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

Serial No.23-105 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 COMMONWEALTH 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.

. -1 ,vt>

---,--_,, 2023.

Acknowledged before me this _,c._day of MAL/

My Commission Expires:

Notary Public Attachments:

1. Description and Assessment of Proposed Changes

2. Marked-up Technical Specifications Pages

3. ANP-4031 P, Revision 0, Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design, Licensing Report (Proprietary)

4. ANP-4031 NP, Revision 0, Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design, Licensing Report (Non-Proprietary)

5. ANP-4032P, Revision 0, Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design, Licensing Report (Proprietary)

6. ANP-4032NP, Revision 0, Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design, Licensing Report (Non-Proprietary)

7. Framatome Application for Withholding and Affidavit

8. Request for Exemption Related to 10 CFR 50.46 and 10 CFR 50, Appendix K Commitments made in this letter: None

Serial No.23-105 Docket No. 50-423 Page 4 of 4 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.23-105 Docket No. 50-423 Page 1 of 12 Attachment 1 DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGES Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.23-105 Docket No. 50-423 Attachment 1, Page 2 of 12 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENC) is submitting a License Amendment Request (LAR) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3).

This LAR proposes updating the list of approved methodologies for the Core Operating Limits Report (COLR) in MPS3 TS 6.9.1.6.b to include the following Framatome (FRM)

Topical Reports:

1. EMF-2328-P-A, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, [Reference 1],

2. EMF-2103-P-A, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," [Reference 3], and

3. ANP-10349-P-A, GALILEO Implementation in LOCA Methods [Reference 4].

This LAR also proposes updating the list of approved methodologies for the COLR in MPS3 TS 6.9.1.6.b by removing a legacy Westinghouse (WEC) Loss Of Coolant Accident (LOCA) Topical Report:

1. WCAP-16009-P-A, REALISTIC LARGE-BREAK LOCA EVALUATION METHODOLOGY USING THE AUTOMATED STATISTICAL TREATMENT OF UNCERTAINTY METHOD (ASTRUM).

2.0 DETAILED DESCRIPTION 2.1. System Design and Operation The proposed change is relevant to the Emergency Core Cooling System (ECCS) insofar as the method for demonstrating its performance to meet the acceptance criteria set forth in 10 CFR 50.46. However, the ECCS structures, systems, and components (SSCs) are not physically altered by the requested change. Similarly, the way the ECCS is operated is also unaltered by the change.

The design requirements of existing plant SSCs are used as inputs to the Small Break LOCA (SBLOCA) and Realistic Large Break LOCA (RLBLOCA) analyses, with appropriate technical conservatisms applied. Therefore, these analyses do not directly impact the existing design or configuration of any plant SSCs.

Serial No.23-105 Docket No. 50-423 Attachment 1, Page 3 of 12 2.2. Current Technical Specification Requirement TS 6.9.1.6 requires core operating limits 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 for Large Break LOCA (LBLOCA) and SBLOCA are:

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

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

8. WCAP-10054-P-A, WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE, (W Proprietary).

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

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

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

Methodology for Specification:

3.2.2.1 - Heat Flux Hot Channel Factor

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

3.2.2.1 - Heat Flux Hot Channel Factor 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.23-105 Docket No. 50-423 Attachment 1, Page 4 of 12 2.3. Reason for the Proposed Change The proposed TS changes are needed to support the transition to FRM GAIA fuel with M5TM cladding at MPS3, which requires the application of the FRM SBLOCA and RLBLOCA methodologies. DENC and FRM have entered into an agreement for batch implementation of the GAIA fuel at MPS3. A full reload batch of GAIA fuel assemblies is planned for initial insertion in Cycle 24. This onload is currently scheduled for the spring 2025 refueling outage.

Additionally, the proposed changes amend MPS3 TS 6.9.1.6.b by removing a legacy COLR reference no longer used to establish core operating limits.

2.4. Description of Proposed Changes TS 6.9.1.6.b lists methodology documents used to determine the core operating limits for MPS3. 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, the TS 6.9.1.6.b list only identifies the topical report document identification number and title. The proposed revision to this list adds three documents in support of the FRM SBLOCA and RLBLOCA methods, as shown below.

24. EMF-2328-P-A, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, (Framatome Proprietary). Methodology for Specification:

3.2.2.1 Heat Flux Hot Channel Factor

25. EMF-2103-P-A, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," (Framatome Proprietary). Methodology for Specification:

3.2.2.1 Heat Flux Hot Channel Factor

26. ANP-10349-P-A, GALILEO Implementation in LOCA Methods, (Framatome Proprietary). Methodology for Specification:

3.2.2.1 Heat Flux Hot Channel Factor The proposed revision to this list also deletes one methodology no longer used to define core operating limits at MPS3, as shown below.

Document Number 6

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

Serial No.23-105 Docket No. 50-423 Attachment 1, Page 5 of 12 will be replaced as follows:

6. Deleted TS 6.9.1.6.b will retain the remaining WEC LOCA methods to support the current MPS3 fuel product prior to and during the transition to GAIA.

Markups of the proposed TS changes are provided in Attachment 2.

3.0 TECHNICAL EVALUATION

3.1. Addition of FRM SBLOCA and GALILEO Methodology Attachments 3 (proprietary) and 4 (non-proprietary) provide a detailed description of the MPS3 SBLOCA analysis using GAIA fuel with M5TM cladding, with analytical results. This analysis was performed in accordance with NRC-approved FRM Topical Reports EMF-2328-P-A, Revision 0, including Supplement 1-P-A, Revision 0, and ANP-10349-P-A, Revision 0 [References 1, 2, and 4, respectively]. As reported in Section 2.0 of Attachments 3 and 4, the SBLOCA analysis results meet the ECCS performance acceptance criteria in 10 CFR 50.46(b) (1-4). The limiting peak cladding temperature (PCT) is 1622°F for an 8.60-inch diameter cold leg pump discharge break. The high head safety injection line (HHSI) line break produced the limiting maximum local oxidation (MLO) and core wide oxidation (CWO) values. The limiting total MLO and limiting CWO values are 4.28% and 0.08%, respectively.

Section 3.5 of Attachments 3 and 4 includes a discussion of compliance with the limitations contained in the issued NRC Safety Evaluations (SEs) associated with the methodologies of EMF-2328-P-A, Revision 0, including Supplement 1-P-A, Revision 0, and ANP-10349-P-A, Revision 0. DENC concludes that the MPS3 SBLOCA analysis complies with all of the applicable SE limitations.

3.2. Addition of FRM RLBLOCA and GALILEO Methodology Attachments 5 (proprietary) and 6 (non-proprietary) provide a detailed description of the MPS3 RLBLOCA analysis using GAIA fuel with M5TM cladding, with analytical results. This analysis was performed in accordance with NRC-approved FRM Topical Reports EMF-2103-P-A, Revision 3 and ANP-10349-P-A, Revision 0 [Reference 3 and 4, respectively]. As reported in Section 2.0 of Attachments 5 and 6, the RLBLOCA analysis results meet the ECCS performance acceptance criteria in 10 CFR 50.46(b) (1-3). The Upper Tolerance Limit (UTL) results providing 95/95

Serial No.23-105 Docket No. 50-423 Attachment 1, Page 6 of 12 simultaneous coverage from this evaluation meet the 10 CFR 50.46(b) criteria with a PCT of 1835°F, a maximum local oxidation of 6.47% and a total core-wide oxidation of 0.062%. The PCT of 1835°F occurred in a fresh UO2 rod with an assembly burnup of 2.0 GWd/mtU.

Section 3.7 of Attachments 5 and 6 includes a discussion of compliance with the limitations contained in the issued NRC SEs associated with the methodologies of EMF-2103-P-A, Revision 3 and ANP-10349-P-A, Revision 0. DENC concludes that the MPS3 RLBLOCA analysis complies with all applicable SE limitations.

3.3. Deletion of Legacy WEC LOCA Methodology The proposed TS change removes the legacy WEC LOCA Topical Report WCAP-16009-P-A that is no longer in use at MPS3.

According to the intent of NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific, Parameter Limits from Technical Specifications," dated October 4, 1988

[Reference 5], TS 6.9.1.6.b references should be those that describe a methodology that is applicable for use in establishing core operating limits. MPS3 TS 6.9.1.6.b currently contains multiple LOCA methodology references, only some of which are used to establish core operating limits. Therefore, DENC proposes to amend TS 6.9.1.6.b by deleting Reference 6 since this reference is a legacy LOCA methodology that is no longer used to establish MPS3 cycle specific core operating limits. The remaining WEC LOCA methods are currently used and required to support the resident MPS3 fuel product during the transition to GAIA.

4.0 REGULATORY EVALUATION

4.1. Applicable Regulatory Requirements and Criteria The following regulatory requirements are applicable to ECCS functions and associated TS:

General Design Criterion (GDC) 35 - Emergency Core Cooling A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with

Serial No.23-105 Docket No. 50-423 Attachment 1, Page 7 of 12 continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

The ECCS is described in Section 6.3 of the MPS3 Final Safety Analysis Report (FSAR). Conformance with GDC 35 is described in Section 3.1.2.35 of the MPS3 FSAR and is unaffected by this change. The proposed TS change does not require relief from any other regulatory requirements and does not affect conformance with a General Design Criterion differently than described in the MPS3 FSAR.

10 CFR 50.36 - Technical Specifications 10 CFR 50.36(c)(5) requires TS to include Administrative Controls, which are provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

The proposed addition of NRC-approved LOCA methodologies will be included in the Administrative Controls section of the MPS3 TS and these methods would be used to determine a core operating limit. Use of the proposed NRC-approved LOCA methodologies will continue to ensure that the plant is operated in a safe manner.

Therefore, the proposed change is consistent with the Administrative Controls requirement of 10 CFR 50.36(c)(5).

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors and 10 CFR 50 Appendix K - ECCS Evaluation Models 10 CFR 50.46 requires, in part, that each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLOTM cladding must be provided with an ECCS that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in 10 CFR 50.46(b). Appendix K to 10 CFR Part 50 establishes regulations for conservative ECCS evaluation models.

The proposed change requests NRC approval to use the FRM SBLOCA and RLBLOCA methodologies described in EMF-2328-P-A, including its supplement, and EMF-2103-P-A, respectively, using the GALILEO fuel performance code as

Serial No.23-105 Docket No. 50-423 Attachment 1, Page 8 of 12 prescribed in ANP-10349-P-A for establishing core operating limits. The NRC has reviewed these methods and found them acceptable for reference in licensing applications for Westinghouse-designed 4-loop plants with cold leg ECCS injection.

As demonstrated in Attachments 3 through 6, EMF-2328-P-A, Revision 0, including Supplement 1-P-A, Revision 0, EMF-2103-P-A, Revision 3, and ANP-10349-P-A, Revision 0 are applicable to MPS3. The plant-specific application of these methodologies to the LOCA analyses have been performed in accordance with the conditions and limitations of the topical reports and associated NRC SEs. The plant-specific analyses demonstrate the requirements of 10 CFR 50.46 will continue to be met, thus ensuring continued safe plant operation.

To support future use of M5TM fuel rod cladding at MPS3, an exemption to the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50 is required.1 NRC Generic Letter (GL) 88-16 NRC GL 88-16 states that it is acceptable for licensees to control reactor physics parameter limits by specifying an NRC-approved calculation methodology. These parameter limits may be removed from the TS and placed in a cycle specific COLR, which is required to be submitted to the NRC every operating cycle or each time it is revised.

TS 6.9.1.6.b identifies the NRC-approved analytical methodologies that are used to determine the core operating limits for MPS3. The guidance in NRC GL 88-16 continues to be met since the proposed changes will continue to specify NRC-approved methodologies used to determine the core operating limits.

Summary As described above, the addition of the proposed LOCA methodologies and removal of a legacy LOCA methodology at MPS3 continues to satisfy the requirements of 10 CFR 50.36, 10 CFR 50.46, 10 CFR 50 Appendix K, and NRC GL 88-16. The proposed changes meet the current regulatory requirements and do not affect conformance with GDC 35 as described in the MPS3 FSAR.

1 Attachment 8 to this submittal letter provides the MPS3 M5TM exemption request in accordance with 10 CFR 50.12.

Serial No.23-105 Docket No. 50-423 Attachment 1, Page 9 of 12 4.2. Precedents The proposed change to MPS3 TS 6.9.1.6.b adds FRM Topical Reports EMF-2328-P-A, EMF-2103-P-A, and ANP-10349-P-A to the list of approved methodologies for determining core operating limits. Additionally, one legacy COLR methodology is removed. Numerous previous requests have been approved for methodology changes in plant-specific TS COLR reference lists. Examples include:

EMF-2328-P-A o Harris Unit 1 (ML030910469, dated March 28, 2003).

o Robinson Unit 2 (ML11342A165, dated December 29, 2011) o Millstone Unit 2 (ML16249A001, dated September 30, 2016) o Palo Verde Units 1, 2, and 3 (ML20031C947, dated March 4, 2020)

EMF-2103-P-A o Robinson Unit 2 (ML062330018, dated September 20, 2006) o Harris Unit 1 (ML12076A103, dated May 30, 2012) o Millstone Unit 2 (ML17025A218, dated January 24, 2017) o Palo Verde Units 1, 2, and 3 (ML20031C947, dated March 4, 2020)

Precedents supporting the removal of legacy COLR methodologies include:

Millstone Unit 3 (ML16131A728, dated July 28, 2016)

Robinson Unit 2 (ML11342A165, dated December 29, 2011).

Due to the recent issuance of ANP-10349-P-A, no previous precedents exist for adding this topical report to the TS COLR reference list.

4.3. No Significant Hazards Consideration Dominion Energy Nuclear Connecticut, Inc. (DENC) proposes a change to Millstone Power Station Unit 3 (MPS3) Technical Specifications (TS) 6.9.1.6.b to add Framatome (FRM) Topical Reports EMF-2328-P-A, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, EMF-2103-P-A, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," and ANP-10349-P-A, GALILEO Implementation in LOCA Methods to the list of methodologies approved for reference in the Core Operating Limits Report (COLR). The added references identify

Serial No.23-105 Docket No. 50-423 Attachment 1, Page 10 of 12 the analytical methods used to determine core operating limits for the Small Break Loss Of Coolant Accident (SBLOCA) and Realistic Large Break Loss Of Coolant Accident (RLBLOCA) events described in the MPS3 Final Safety Analysis Report (FSAR), Section 15.6.5.3 and Section 15.6.5.2, respectively. Additionally, the proposed change amends MPS3 TS 6.9.1.6.b by removing a legacy COLR reference no longer used to establish core operating limits.

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 change to TS 6.9.1.6.b permits the use of NRC-approved methodologies for analysis of the SBLOCA and RLBLOCA to determine if MPS3 continues to meet the applicable design and safety analysis acceptance criteria. The proposed change to the list of NRC-approved methodologies in TS 6.9.1.6.b has no direct impact upon plant operation or configuration and does not impact either the initiation of an accident or the mitigation of its consequences.

The results of the SBLOCA and RLBLOCA analysis demonstrate that MPS3 continues to satisfy the 10 CFR 50.46 Emergency Core Cooling System (ECCS) performance acceptance criteria using an NRC-approved evaluation model.

Therefore, the proposed change does 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 change will not create the possibility of a new or different accident due to credible new failure mechanisms, malfunctions, or accident initiators not previously considered. There is no change to the parameters within which the plant is normally operated, and no physical plant modifications are being made; thus, the possibility of a new or different type of accident is not created.

Serial No.23-105 Docket No. 50-423 Attachment 1, Page 11 of 12 Therefore, the proposed change does not create the possibility of a new or different kind of accident or malfunction from those previously evaluated within the FSAR.

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

Response: No.

No design basis or safety limits are exceeded or altered by this change.

Approved methodologies will be used to ensure the plant continues to meet applicable design criteria and safety analysis acceptance criteria.

The results of the SBLOCA and RLBLOCA analyses demonstrate that MPS3 continues to satisfy the 10 CFR 50.46 ECCS performance acceptance criteria using an NRC-approved evaluation model.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above information, DENC concludes that the proposed change does 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. Conclusion Based on the considerations presented above, there is reasonable assurance that:

(1) 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

Serial No.23-105 Docket No. 50-423 Attachment 1, Page 12 of 12 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.

6.0 References

1. FRM Topical Report EMF-2328-P-A, Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2001.

2. FRM Topical Report EMF-2328-P-A, Revision 0, Supplement 1-P-A, Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2012.

3. FRM Topical Report EMF-2103-P-A, Revision 3, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors, June 2016.

4. FRM Topical Report ANP-10349-P-A, Revision 0, GALILEO Implementation in LOCA Methods, November 2021.

5. US NRC Generic Letter 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, October 4, 1988.

Serial No.23-105 Docket No. 50-423 Page 1 of 4 Attachment 2 MARKED-UP TECHNICAL SPECIFICATIONS PAGES Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.23-105 Docket No. 50-423 , Page 2 of 4

Serial No.23-105 Docket No. 50-423 , Page 3 of 4

Serial No.23-105 Docket No. 50-423 Attachment 2, Page 4 of 4 INSERT A

24. EMF-2328-P-A, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, (Framatome Proprietary). Methodology for Specification:

3.2.2.1 Heat Flux Hot Channel Factor

25. EMF-2103-P-A, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," (Framatome Proprietary). Methodology for Specification:

3.2.2.1 Heat Flux Hot Channel Factor

26. ANP-10349-P-A, GALILEO Implementation in LOCA Methods, (Framatome Proprietary). Methodology for Specification:

3.2.2.1 Heat Flux Hot Channel Factor

Serial No.23-105 Docket No. 50-423 Page 1 of 62 Attachment 4 ANP-4031NP, REVISION 0, MILLSTONE UNIT 3 SMALL BREAK LOCA ANALYSIS WITH GAIA FUEL DESIGN, LICENSING REPORT (NON-PROPRIETARY)

Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.23-105 Docket No. 50-423 Attachment 4, Page 2 of 62

Serial No.23-105 Docket No. 50-423 Attachment 4, Page 3 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Initial Issue Serial No.23-105 Docket No. 50-423 Attachment 4, Page 4 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1 2.0

SUMMARY

OF RESULTS ................................................................................. 2-1

3.0 DESCRIPTION

OF ANALYSIS.......................................................................... 3-1 3.1 Acceptance Criteria ................................................................................ 3-1 3.2 Description of SBLOCA Event ................................................................ 3-1 3.3 Description of Analytical Methods ........................................................... 3-4 3.4 Plant Description and Summary of Analysis Parameters ........................ 3-9 3.5 Safety Evaluation Compliance .............................................................. 3-11 4.0 SBLOCA ANALYSIS ......................................................................................... 4-1 4.1 Cold Leg Pump Discharge Break Spectrum Results .............................. 4-1 4.2 Discussion of Transient for Limiting PCT Break...................................... 4-2 4.3 Attached Piping Break Results ............................................................... 4-3 4.4 Delayed RCP Trip Study ......................................................................... 4-4 4.5 ECCS Temperature Sensitivity Study ..................................................... 4-5 4.6 RWST Drain Down Evaluation ................................................................ 4-6

5.0 REFERENCES

.................................................................................................. 5-1 Serial No.23-105 Docket No. 50-423 Attachment 4, Page 5 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page iii List of Tables Table 3-1 System Parameters and Initial Conditions ............................................... 3-12 Table 3-2 High Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum ....................................................................................... 3-13 Table 3-3 Intermediate Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum ...................................................................... 3-14 Table 3-4 Low Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum ....................................................................................... 3-15 Table 3-5 ANP-10349P-A, Revision 0, Limitations and Conditions on GALILEO Applications ............................................................................................. 3-17 Table 4-1 Summary of Cold Leg Pump Discharge Break Spectrum Results ............. 4-7 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum ........ 4-8 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd) ...................................................................................................... 4-9 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd) .................................................................................................... 4-10 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd) .................................................................................................... 4-11 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd) .................................................................................................... 4-12 Serial No.23-105 Docket No. 50-423 Attachment 4, Page 6 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page iv List of Figures Figure 3-1 S-RELAP5 SBLOCA Reactor Coolant System Nodalization .................... 3-6 Figure 3-2 S-RELAP5 SBLOCA Secondary System Nodalization ............................. 3-7 Figure 3-3 S-RELAP5 SBLOCA Reactor Vessel Nodalization .................................. 3-8 Figure 3-4 Axial Power Distribution Comparison ..................................................... 3-16 Figure 4-1 Cold Leg Pump Discharge Break Spectrum Peak Cladding Temperature versus Break Size ............................................................. 4-13 Figure 4-2 Reactor Power - 8.60 inch Break........................................................... 4-14 Figure 4-3 Primary and Secondary System Pressures - 8.60 inch Break ............... 4-15 Figure 4-4 Break Mass Flow Rate - 8.60 inch Break .............................................. 4-16 Figure 4-5 Break Vapor Void Fraction - 8.60 inch Break ........................................ 4-17 Figure 4-6 Loop Seal Upside Collapsed Levels - 8.60 inch Break .......................... 4-18 Figure 4-7 Downcomer Collapsed Liquid Level - 8.60 inch Break .......................... 4-19 Figure 4-8 Primary System Masses - 8.60 inch Break ............................................ 4-20 Figure 4-9 RCS Loop Mass Flow Rates - 8.60 inch Break ..................................... 4-21 Figure 4-10 Steam Generator Main Feedwater Flow Mass Rates - 8.60 inch Break ................................................................................................... 4-22 Figure 4-11 Steam Generator MSSV Mass Flow Rates - 8.60 inch Break ............... 4-23 Figure 4-12 Steam Generator Auxiliary Feedwater Flow Rate - 8.60 inch Break ..... 4-24 Figure 4-13 Steam Generator Total Secondary Side Mass - 8.60 inch Break .......... 4-25 Figure 4-14 Steam Generator Narrow Range Level - 8.60 inch Break ..................... 4-26 Figure 4-15 High Head Safety Injection Mass Flow Rates - 8.60 inch Break............ 4-27 Figure 4-16 Intermediate Head Safety Injection Mass Flow Rates - 8.60 inch Break ................................................................................................... 4-28 Figure 4-17 Low Head Safety Injection Mass Flow Rates - 8.60 inch Break ............ 4-29 Figure 4-18 Accumulator Mass Flow Rates - 8.60 inch Break .................................. 4-30 Figure 4-19 Total ECCS and Break Mass Flow Rates - 8.60 inch Break .................. 4-31 Figure 4-20 Hot Assembly Collapsed Liquid Level - 8.60 inch Break ....................... 4-32 Figure 4-21 Cladding Temperature at PCT Node - 8.60 inch Break ......................... 4-33 Serial No.23-105 Docket No. 50-423 Attachment 4, Page 7 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page v Nomenclature Acronym Definition AFW Auxiliary Feedwater BOC Beginning-of-Cycle CFR Code of Federal Regulations CWO Core Wide Oxidation DC Downcomer ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EOC End-of-Cycle FH Nuclear Enthalpy Rise Factor/Radial Peaking Factor FQ Total Peaking Factor Framatome Framatome Inc.

HEM Homogeneous Equilibrium Model HHSI High Head Safety Injection HMP High Mechanical Performance IGM Intermediate GAIA Mixing Grid IHSI Intermediate Head Safety Injection k(z) Axial-Dependent Peaking Factor LEU Low-Enriched Uranium LHSI Low Head Safety Injection LOCA Loss-of-Coolant Accident LOOP Loss-of-Offsite Power LS Loop Seal LSC Loop Seal Clearing Serial No.23-105 Docket No. 50-423 Attachment 4, Page 8 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page vi Acronym Definition MFW Main Feedwater MLO Maximum Local Oxidation MSSV Main Steam Safety Valve NRC U.S. Nuclear Regulatory Commission RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RPS Reactor Protection System RT Reactor Trip RV Reactor Vessel RWST Refueling Water Storage Tank PCT Peak Cladding Temperature PWR Pressurized Water Reactor PZR Pressurizer SBLOCA Small Break Loss-of-Coolant Accident SE Safety Evaluation SG Steam Generator SI Safety Injection SIAS Safety Injection Actuation Signal Tavg RCS Average Temperature TT Turbine Trip Serial No.23-105 Docket No. 50-423 Attachment 4, Page 9 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 1-1

1.0 INTRODUCTION

This report summarizes the small break loss-of-coolant accident (SBLOCA) analysis for Millstone Power Station Unit 3. The purpose of the SBLOCA analysis is to support the fuel transition at Millstone Unit 3 to the Framatome GAIA fuel design. This analysis was performed in accordance with the U.S. Nuclear Regulatory Commission (NRC)-approved S-RELAP5-based methodology described in Reference 1 as supplemented by Reference 2 and Reference 3.

Millstone Unit 3 is a four-loop, Westinghouse-designed Pressurized Water Reactor (PWR). The Framatome GAIA fuel design with M5Framatome cladding for Millstone Unit 3 consists of a 17x17 array with GAIA and intermediate GAIA Mixing (IGM) grids, a lower high mechanical performance (HMP) grid and an upper HMP grid. The fuel assembly includes a MONOBLOC guide tube design, M5Framatome fuel rod design and a GRIP lower nozzle.

The analysis supports plant operation at a core power level of 3723 MWt (includes measurement uncertainty), a maximum-allowed total peaking factor (FQ) of 2.6 (represents total peaking with uncertainties applied and an axial-dependent factor k(z) set to 1.0), a radial peaking factor of (FH) of 1.70 (includes measurement uncertainty),

and up to 10% steam generator (SG) tube plugging per SG.

A complete spectrum including cold leg pump discharge break sizes ranging from 1.00 inch diameter to 8.70 inch diameter and breaks in attached piping were considered.

Other supporting analyses prescribed by the methodology to assess a delayed reactor coolant pump (RCP) trip and the sensitivity to Emergency Core Cooling System (ECCS) fluid temperature were performed. Beyond the supporting analyses prescribed by the methodology, a plant specific request for a refueling water storage tank (RWST) drain down evaluation and sensitivity for delayed RCP trip were performed.

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 2-1 2.0

SUMMARY

OF RESULTS The SBLOCA analysis results demonstrate the adequacy of the ECCS to satisfy the 10 CFR 50.46(b) (1-4) criteria (Reference 5) for Millstone Unit 3 operating with Framatome supplied GAIA fuel design with M5Framatome cladding. The limiting peak cladding temperature (PCT) is 1622°F for an 8.60-inch diameter cold leg pump discharge break. The high head safety injection line (HHSI) line break produced the limiting maximum local oxidation (MLO) and core wide oxidation (CWO) values. The limiting total MLO and limiting CWO values are 4.28% and 0.08%, respectively. The total MLO value includes [ ]

In addition to the analysis of the cold leg pump discharge breaks, breaks in the attached piping were considered. The attached piping break analyses considered breaks in the accumulator line and HHSI line. The HHSI line break established the limiting MLO and CWO values, but both attached piping break PCTs are bounded by the limiting PCT from the cold leg pump discharge break spectrum. Two supporting analyses prescribed by the methodology were performed to investigate a delayed RCP trip and a different ECCS temperature. The results of the delayed RCP trip study demonstrated that there is at least 5 minutes for operators to trip all four RCPs after the specified trip criteria being met. The results of the ECCS temperature sensitivity study support the applicability of the ECCS temperature used in the SBLOCA analysis.

Additionally, a plant specific RWST drain down evaluation was performed to evaluate the sensitivity to sump switchover. The results of the evaluation determined that the SBLOCA analysis results are not impacted by a higher ECCS temperature following the switchover.

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-1

3.0 DESCRIPTION

OF ANALYSIS 3.1 Acceptance Criteria The purpose of the analysis is to verify the adequacy of the Millstone Unit 3 ECCS by demonstrating compliance with the following 10 CFR 50.46(b) criteria (Reference 5):

1. The calculated maximum fuel element cladding temperature shall not exceed 2200°F.

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

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

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

3.2 Description of SBLOCA Event The postulated SBLOCA is defined as a break in the Reactor Coolant System (RCS) pressure boundary with an area less than or equal to 10% of the cold leg pipe area.

The most limiting break location is typically in the cold leg pipe on the discharge side of the RCP. This break location results in the largest amount of RCS inventory loss, the largest fraction of ECCS fluid discharged out the break, and the largest pressure difference between the core exit and the top of the downcomer (DC). This typically produces the greatest degree of core uncovery, the longest fuel rod heatup time and, consequently, the greatest challenge to the 10 CFR 50.46(b)(1-4) criteria (Reference 5).

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-2 The SBLOCA event progression develops in the following distinct phases: (1) subcooled depressurization (also known as blowdown), (2) natural circulation, (3) loop seal clearing, (4) core boil-off, (5) core recovery and long-term cooling. The duration of each of these phases is break size and system dependent.

Following the break, the RCS rapidly depressurizes to the saturation pressure of the hot leg fluid. During the initial depressurization phase, a reactor trip signal is generated on low pressurizer pressure, and the turbine is tripped on the reactor trip. The assumption of a loss-of-offsite power (LOOP) concurrent with the reactor scram results in RCP trip.

In the second phase of the transient, the RCS transitions to a quasi-equilibrium condition in which the core decay heat, leak flow, SG heat removal, and system hydrostatic head balance combine to control the core inventory. During this period, the RCPs are coasting down and the system drains from the top of the RCS with the first voiding occurring at the top of the SG tubes, in the reactor vessel (RV) upper head, and at the top of the RV upper plenum region. Also, the loop seals remain plugged during this phase, trapping vapor generated by the core in the RCS, resulting in a low-quality flow at the break.

The third phase in the transient is characterized by loop seal clearing (LSC). During this phase the loop seal (i.e., liquid trapped in the RCP suction piping) can prevent steam from venting to the break. The maximum pressure difference between the RV upper head and DC is reached when the liquid level on the downside of the SG is depressed to the elevation of the horizontal loop seal piping. When this point is reached, the loop seal clears, and the trapped steam can be vented to the break. For some break sizes, the transient develops slowly, and the core can become temporarily uncovered before the loop seal clears. Following LSC, the break flow transitions to primarily steam and the core recovers to approximately the cold leg elevation as pressure imbalances throughout the RCS are relieved.

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-3 The fourth phase is characterized as core boil-off. With the loop seal cleared, the venting of steam through the break causes a rapid RCS depressurization below the secondary pressure. As boiling increases in the core, the core mixture level decreases.

The core mixture level will reach a minimum, in some cases resulting in deep core uncovering. The transient boil-off period ends when the core liquid level reaches this minimum. At this time, the RCS has depressurized to the point where ECCS flow into the RV matches the rate of boil-off from the core.

The last phase of the transient is characterized as core recovery. The core recovery period extends from the time at which the core mixture level reaches a minimum in the core boil-off phase until all parts of the core are quenched and covered by a low-quality mixture. Core recovery is provided by pumped injection and passive accumulator injection when the RCS pressure decreases below the accumulator pressure.

Generally, PCT occurs at the beginning of the core recovery phase before the mixture level has increased enough to provide enhanced cooling to the PCT location on the hot rod.

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-4 The SBLOCA transient progression is dependent on the size of the break and is typically broken into three different break size ranges. For break sizes towards the larger end of the break spectrum, significant RCS inventory loss results in more rapid RCS depressurization to the accumulator actuation pressure. Accumulator flow provides sufficient inventory early in the transient to limit the core uncovery and hot rod heatup. For break sizes in the middle of the spectrum, the rate of inventory loss from the RCS is such that the HHSI pumps typically cannot preclude significant core uncovery. The RCS depressurization rate is slow, extending the time required to reach the accumulator injection pressure, if reached at all. Break sizes in this range, will either exhibit core recovery with the HHSI pumped injection alone while the RCS pressure remains barely above the accumulator injection setpoint, or exhibit core recovery from accumulator injection. For break sizes at the low end of the spectrum, the RCS pressure does not reach the accumulator injection pressure. However, RCS inventory loss is not significant and typically within the means of HHSI makeup capacity such that core uncovery is minimal if not precluded.

3.3 Description of Analytical Methods This analysis was performed in accordance with the NRC-approved S-RELAP5-based methodology described in Reference 1 as supplemented by Reference 2 and Reference

3. The Framatome S-RELAP5 SBLOCA evaluation model for event response of the primary and secondary systems and the hot fuel rod is based on the use of two computer codes. The appropriate conservatisms, as prescribed by Appendix K of 10 CFR 50 (Reference 6), are incorporated.

The two Framatome computer codes used in this analysis are:

1. The GALILEO code was used to determine the burnup dependent initial fuel rod conditions for the system calculations.

2. The S-RELAP5 code was used to predict the primary and secondary system thermal-hydraulic and hot rod transient response.

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-5 Representative system nodalization figures for a Westinghouse four-loop plant are shown in Figure 3-1 through Figure 3-3. See Section 3.4 for a description of the Millstone Unit 3 analysis model.

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-6 Figure 3-1 S-RELAP5 SBLOCA Reactor Coolant System Nodalization Serial No.23-105 Docket No. 50-423 Attachment 4, Page 17 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-7 Figure 3-2 S-RELAP5 SBLOCA Secondary System Nodalization Serial No.23-105 Docket No. 50-423 Attachment 4, Page 18 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-8 Figure 3-3 S-RELAP5 SBLOCA Reactor Vessel Nodalization Serial No.23-105 Docket No. 50-423 Attachment 4, Page 19 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-9 3.4 Plant Description and Summary of Analysis Parameters The plant analyzed is the Millstone Unit 3, Westinghouse-designed PWR, which has four loops, each with a hot leg, a U-tube steam generator, and a cold leg with an RCP.

The RCS includes one pressurizer connected to a hot leg. Main feedwater (MFW) is injected into the downcomer of each SG. The auxiliary feedwater (AFW) system provides flow to the four SGs when normal feedwater is not available. The ECCS provides injection to each of the four loops via the centrifugal charging/HHSI system, SI/intermediate head safety injection (IHSI) system, residual heat removal (RHR)/low head safety injection (LHSI) system, and accumulators. For the purpose of this report, the centrifugal charging/HHSI, SI/IHSI, and RHR/LHSI systems are referred to as the HHSI, IHSI, and LHSI systems, respectively.

The RCS, SG, reactor vessel, pressurizer, and ECCS are explicitly modeled in the S-RELAP5 model to provide an accurate representation of the plant. The model includes four accumulators, a pressurizer, and four SGs with both primary and secondary sides modeled. For the secondary side, the model includes the main steam lines between their respective SGs and the turbine control valve, including the connected main steam safety valve (MSSV) inlet piping.

For each RCS loop, the ECCS model includes an injection connection to the cold leg for the accumulator and another connection for HHSI. IHSI and LHSI are modeled with separate injection connections to each of the four accumulator lines. The accumulator and HHSI injection connections to the cold leg pipe are downstream of the RCP discharge. The ECCS pumped injection is modeled as a table of flow versus RCS backpressure.

Important system parameters and initial conditions used in the analysis are given in Table 3-1. The heat generation rate in the S-RELAP5 reactor core model is determined from reactor kinetics equations with actinide and decay heating as prescribed by 10 CFR 50 Appendix K (Reference 6).

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-10 The break spectrum analysis assumes a LOOP concurrent with reactor scram, which is based on the reactor protection system (RPS) low pressurizer pressure reactor trip and includes delays as stated in Table 3-1. The assumption of LOOP concurrent with reactor scram results in an RCP trip.

The RCPs are tripped at the time of reactor scram, instead of the opening of the break (time zero). This is considered to be conservative, since continued RCP operation will delay LSC. This delay in LSC will result in additional RCS inventory loss since the break flow is mostly liquid until the time of LSC. After LSC, a path for steam venting is established and the break flow transitions from liquid to steam, lowering the break mass flow rate.

The single failure criterion required by 10 CFR 50 Appendix K (Reference 6) is satisfied by assuming the loss of one emergency diesel generator (EDG). The loss of an EDG disables one of the two ECCS trains. Thus, one motor-driven AFW pump, one HHSI pump, one IHSI pump, and one LHSI pump are assumed unavailable.

Following the safety injection actuation system (SIAS) activation on low pressurizer pressure, actuation of the HHSI, IHSI, and LHSI systems are delayed by 45 seconds.

Table 3-2, Table 3-3, and Table 3-4 show the minimum ECCS flow rates with one EDG failure for HHSI, IHSI, and LHSI, respectively, for a break in the RCS loop. The pumped safety injection flow to the intact loops is modeled to be distributed equally among the three intact loops.

With one of the two motor-driven AFW pumps assumed unavailable for the single failure criterion, AFW flow is supplied via the remaining motor-driven AFW pump. AFW injection is delayed by 60 seconds beyond the time of AFW system initiation on low-low SG level. Only two of the SGs are credited with receiving AFW flow. A SG tube plugging level of 10% is modeled in each SG. The MSSVs are set to open at their nominal setpoints plus 3% accumulation.

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-11 The axial power shapes for this analysis are shown in Figure 3-4. The figure shows the input axial power shape and the axial power shape after being adjusted so that it is consistent with the Technical Specification total and radial peaking factors.

3.5 Safety Evaluation Compliance The SBLOCA analysis for Millstone Unit 3 presented herein is consistent with the submitted SBLOCA methodology documented in EMF-2328, Revision 0 (Reference 1) as modified by EMF-2328, Revision 0, Supplement 1, Revision 0 (Reference 2) and supplemented by ANP-10349P-A, Revision 0 (Reference 3). The limitations and conditions from the NRC Safety Evaluation (SE) for EMF-2328, Revision 0, are addressed below. There are no limitations and conditions from the NRC for EMF-2328, Revision 0, Supplement 1. The limitations and conditions from Reference 4 on GALILEO applications for ANP-10349P-A, Revision 0 (Reference 3), are addressed in Table 3-5. From the disposition provided below and the responses given in Table 3-5, the Millstone Unit 3 SBLOCA analysis documented herein are compliant with all requirements of the applicable topical reports.

There is one SE limitation and condition for the application of the SBLOCA evaluation model EMF-2328, Revision 0 (Reference 1), that S-RELAP5 is acceptable for modeling transients where the break flow is less than or equal to 10% of the cold leg flow area. A spectrum of cold leg break sizes from 0.00545 ft2 (1.00-inch diameter) to 0.41282 ft2 (8.70-inch diameter, 10% of cold leg pipe area) are analyzed. This satisfies the limitation placed on EMF-2328, Revision 0 for the cold leg break spectrum.

The attached pipe break in the accumulator lines is greater than 10% of the cold leg flow area with a break area of 0.4176 ft2 (8.75-inch accumulator line inside diameter).

However, the flow area limitation on EMF-2328, Revision 0 (Reference 1) is addressed for this break location in Supplement 1 to EMF-2328 (Reference 2) which is applied for the Millstone Unit 3 SBLOCA analyses. Therefore, no further actions are required to address the limitation for the accumulator line break area.

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-12 Table 3-1 System Parameters and Initial Conditions Parameter Value Reactor Power (MWt) 3723(1)

Axial Power Shape Figure 3-4 Radial Peaking Factor (FH) 1.70(1)

Maximum-Allowed Total Power Peaking Factor (FQ) 2.6(2)

Total RCS Flow Rate (gpm) 363,200 Pressurizer Pressure (psia) 2250 RCS Operating Temperature, Tavg (°F) 594.5 SG Tube Plugging per SG (%) 10 SG Secondary Pressure (psia) 937.4 MFW Temperature (°F) 447.9 RPS Low Pressurizer Pressure for Reactor Trip (psia) 1859.7 RPS Low Pressurizer Pressure Trip Delay(3) (sec) 2 RPS Scram Delay (sec) 0 SIAS Low Pressurizer Pressure Activation Setpoint (psia) 1700 Accumulator Pressure (psia) 626 Accumulator Fluid Temperature (°F) 125 Accumulator Water Volume per Accumulator (ft3) 912 AFW Temperature (°F) 80 Total AFW Flow Rate (gpm) 246.5(4)

AFW Initiation on Low-Low SG Narrow Range Level 0

Setpoint (% Narrow Range Span)

AFW Injection Delay (sec) 60(5)

ECCS Pumped Injection Temperature (°F) 100 HHSI, IHSI, and LHSI Injection Delay Time on SIAS (sec) 45(5)

MSSV Lift Pressure and Accumulation Nominal + 3% Accumulation 1

Includes measurement uncertainty.

2 Includes uncertainties and k(z) set to 1.0.

3 Includes scram delay.

4 Total flow is split evenly among the four SGs. Only two SGs are credited with receiving AFW flow.

5 Delay applied to all analyses regardless of offsite power availability.

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-13 Table 3-2 High Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum RCS Total Intact Broken Pressure Loops Flow Loop Flow (psia) (gpm) (gpm) 14.7 281.6 102.9 114.7 274.7 100.4 214.7 267.9 97.9 314.7 260.2 95.1 414.7 252.5 92.3 514.7 244.8 89.5 614.7 237.1 86.6 714.7 228.9 83.7 814.7 220.6 80.7 914.7 212.3 77.6 1014.7 203.8 74.4 1114.7 194.1 71.0 1214.7 184.4 67.4 1314.7 174.7 63.8 1414.7 164.7 60.2 1514.7 154.8 56.5 1614.7 144.5 52.8 1714.7 132.0 48.3 1814.7 118.8 43.4 1914.7 105.5 38.6 2014.7 92.1 33.6 2114.7 79.8 29.2 2214.7 70.7 25.8 2314.7 61.6 22.5 2482.7 0.0 0.0 Serial No.23-105 Docket No. 50-423 Attachment 4, Page 24 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-14 Table 3-3 Intermediate Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum RCS Total Intact Broken Pressure Loops Flow Loop Flow (psia) (gpm) (gpm) 14.7 402.2 143.2 114.7 388.1 138.1 214.7 373.4 132.9 314.7 357.7 127.3 414.7 341.2 121.4 514.7 324.1 115.3 614.7 305.2 108.7 714.7 285.5 101.7 814.7 263.8 93.9 914.7 239.9 85.4 1014.7 213.0 75.8 1114.7 180.1 64.1 1214.7 133.0 47.3 1314.7 61.7 21.9 1414.7 0.0 0.0 Serial No.23-105 Docket No. 50-423 Attachment 4, Page 25 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-15 Table 3-4 Low Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum RCS Total Intact Broken Pressure Loops Flow Loop Flow (psia) (gpm) (gpm) 14.7 1947.6 692.7 114.7 89.5 31.8 214.7 0.0 0.0 Serial No.23-105 Docket No. 50-423 Attachment 4, Page 26 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-16 Figure 3-4 Axial Power Distribution Comparison Serial No.23-105 Docket No. 50-423 Attachment 4, Page 27 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-17 Table 3-5 ANP-10349P-A, Revision 0, Limitations and Conditions on GALILEO Applications Ranges of Applicability

Response

(Section 1.2 in Reference 4)

Pressurized water reactor designs using This analysis was performed for the Millstone Low-Enriched Uranium (LEU) fuel loading Unit 3 plant, which is a PWR, using LEU fuel.

The fuel burnups applied in this analysis do Rod average burnups up to [ ] gigawatt-days not exceed the rod average burnup of per metric ton of uranium (GWd/MTU) for Zircaloy-4 and up to [ ] GWd/MTU for M5 cladding [ ]

Zircaloy-4 and M5 cladding The analysis supports operation with M5Framatome cladding.

This analysis was performed using fuel with a Rod diameter between [ ] mm and [ ] rod outside diameter of 9.5 mm.

mm Uranium 235U enrichments up to 5 weight percent The 235U enrichments applied in this analysis (wt%) do not exceed 5 weight percent.

Gadolinia concentrations up to 10 wt% Gadolinia fuel is not analyzed as part of the SBLOCA methodology. Therefore, this parameter is not subject to the limitation for this loss-of-coolant accident (LOCA) analysis.

Nominal true pellet density ranging from [ The initial pellet density is [ ] percent of the theoretical density of UO2.

] percent of the theoretical density of UO2 This analysis was performed using fuel Fuel grain sizes ranging from [ ]

microns (mean linear intercept) pellets with a grain size of [ ]

Pellets manufactured by dry conversion and The fuel pellet manufacturing process for the ammonium diuranate fuel design considered in this analysis is dry conversion and ammonium diuranate.

Fuel temperature up to the melting point to the This is related to thermo-mechanical methods approved burnup range and is not subject to the limitation for this LOCA analysis.

Cladding strain up to the approved transient clad This is related to thermo-mechanical methods strain limit and is not subject to the limitation for this LOCA analysis.

Internal rod pressure up to pressures that protect This is related to thermo-mechanical methods from clad lift-off and hydride reorientation and is not subject to the limitation for this LOCA analysis.

Fuel rod power not to exceed levels as limited by This is related to thermo-mechanical methods fuel melt, cladding strain, and rod pressure criteria and is not subject to the limitation for this LOCA analysis.

Serial No.23-105 Docket No. 50-423 Attachment 4, Page 28 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-1 4.0 SBLOCA ANALYSIS The analysis results demonstrate the adequacy of the ECCS to satisfy the criteria given in 10 CFR 50.46(b)(1-4) for Millstone Unit 3 operating with Framatome supplied GAIA fuel design with M5Framatome cladding.

4.1 Cold Leg Pump Discharge Break Spectrum Results The Millstone Unit 3 break spectrum analysis for SBLOCA includes breaks of varying diameter up to 10% of the flow area for the cold leg pump discharge. The spectrum includes a break size range from 1.00 to 8.70 inches in diameter, where the break size interval is sufficient to establish a PCT trend. Additional break sizes are analyzed with a smaller break interval once the potential limiting break size is determined to confirm the limiting break size, which is the case with the highest PCT. Figure 4-1 displays the PCT results as a function of break size. For the break spectrum analysis, RCP trip is assumed to occur on reactor scram.

The results of the cold leg pump discharge SBLOCA break spectrum analysis are presented in Table 4-1. The predicted event times for the break spectrum are provided in Table 4-2. The limiting PCT break size is determined to be 8.60 inches in diameter (0.40338 ft2), resulting in a PCT of 1622°F. The highest transient MLO and CWO values from the cold leg pump discharge break spectrum results are less than those from the HHSI line attached piping break discussed in Section 4.3, and therefore, are not limiting.

Serial No.23-105 Docket No. 50-423 Attachment 4, Page 29 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-2 4.2 Discussion of Transient for Limiting PCT Break The limiting PCT break from the cold leg pump discharge break spectrum is the 8.60-inch diameter break with a PCT of 1622°F. The transient progression is shown in Figure 4-2 through Figure 4-21. The cladding temperature at the PCT location is shown in Figure 4-21. Key transient results are presented in Table 4-1. The sequence of events is provided in Table 4-2.

The break opens at t=0 seconds and initiates a subcooled depressurization of the RCS.

The RPS low pressurizer pressure trip setpoint is reached at 7.03 seconds and at 9.03 seconds the reactor is scrammed, coincident with the RCP and turbine trips (Figure 4-2 and Figure 4-9). MFW is also isolated on reactor scram (Figure 4-10). The pressure in the secondary side begins to increase but does not reach the MSSV lift setpoint, and the MSSVs remain closed for the duration of the transient (Figure 4-11).

The SIAS is issued at 9.73 seconds. Following the EDG loading delay, HHSI and IHSI begin to inject at 55 seconds (Figure 4-15 and Figure 4-16). Prior to this, the core began to uncover at 48 seconds with effective cooling lost to most of the hot assembly in a short period of time (Figure 4-20). HHSI and IHSI do not provide sufficient inventory to offset the large amount of coolant mass lost out the break at this time (Figure 4-19).

All four loop seals clear before the time of PCT, with the broken loop clearing first after 73 seconds, followed closely by the other three loops clearing between 75 and 96 seconds (Figure 4-6). The clearing of the loop seals produces a temporary increase in core level at approximately 90 seconds (Figure 4-20). However, the mixture level remains near the bottom of the active core, resulting in continuation of the clad temperature excursion (Figure 4-21).

Serial No.23-105 Docket No. 50-423 Attachment 4, Page 30 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-3 The accumulators begin injecting at 152 seconds (Figure 4-18). The minimum RV mass occurs around 170 seconds (Figure 4-8). There is a time delay from the time when accumulator injection begins to the mixture level reaching sufficient levels to cool the upper locations in the core. The delay results in a rupture of the hot rod at 170 seconds. The rupture allows for interior metal-water reaction, increasing the local oxidation at the rupture node.

The cladding temperature excursion is terminated at 181 seconds with a PCT of 1622°F (Figure 4-21). The core is quenched at approximately 220 seconds with the initial accumulator injection ending around the same time (Figure 4-18). LHSI initially begins injecting at 179 seconds, just prior to the time of PCT. However, the LHSI flow rate is significantly less than that of the accumulators and any impact on transient mitigation is considered minimal (Figure 4-17 and Figure 4-18). By the time of core quench, enough decay heat is being removed and an adequate mixture level is sustained primarily by pumped ECCS injection (Figure 4-15, Figure 4-16, and Figure 4-17).

4.3 Attached Piping Break Results The ECCS must cope with ruptures of the main RCS piping and breaks in attached piping. To demonstrate this, as prescribed by the NRC-approved supplement to EMF-2328 (Reference 2), an analysis of the ruptures in attached piping that compromise the ability to inject emergency coolant into the RCS is performed. The size of the rupture and the portion of ECCS lost directly to containment are dependent on the plant design.

Serial No.23-105 Docket No. 50-423 Attachment 4, Page 31 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-4 The Millstone Unit 3 plant design injects IHSI and LHSI to the accumulator injection line which is connected to each cold leg while HHSI is injected through a separate line that is connected to each cold leg. Therefore, two break locations are analyzed, accumulator line and HHSI line. The break areas analyzed represents a double-ended guillotine of the accumulator line and HHSI line. The accumulator line break analysis results are less limiting than those of the break spectrum analysis. The HHSI line break PCT is less limiting than that of the break spectrum analysis, however, the MLO and CWO values are not. The HHSI line break analysis produced the limiting total MLO and limiting CWO values of 4.28% and 0.08%, respectively.

4.4 Delayed RCP Trip Study The delayed RCP trip study is performed in accordance with the NRC-approved supplement to the EMF-2328 methodology (Reference 2). For plants such as Millstone Unit 3 that do not have an automatic RCP trip, a delayed RCP trip can potentially result in a more limiting condition than tripping the RCPs at reactor scram. Continued operation of the RCPs can result in more overall inventory loss out the break. It has been postulated that tripping the pumps when the minimum RCS inventory occurs could cause a collapse of voids in the core, thus depressing the core level and provoking a deeper core uncovering, and a potentially higher PCT. Therefore, the methodology prescribes an RCP trip study for both the cold and hot leg breaks consistent with the plant licensing basis and Emergency Operating Procedures.

For Millstone Unit 3, a delayed RCP trip time for operator action of 5 minutes following the RCP trip criteria is analyzed to evaluate the adequacy of the trip criteria and delay time by demonstrating compliance to 10 CFR 50.46(b)(1-4) criteria (Reference 5). The RCP trip criteria, which are based on plant-specific procedures, are modeled as the occurrence of the RCS pressure trip and at least one charging pump running or one safety injection (SI) pump capable of injecting. The spectrum of cold and hot leg breaks in this study includes break sizes from 1.00 to 8.70 inches.

Serial No.23-105 Docket No. 50-423 Attachment 4, Page 32 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-5 The results of the delayed RCP trip cases indicate that there is at least 5 minutes for operators to trip all four RCPs after the specified trip criteria is met with considerable margin to the 10 CFR 50.46(b)(1-4) criteria. [

]

4.5 ECCS Temperature Sensitivity Study Serial No.23-105 Docket No. 50-423 Attachment 4, Page 33 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-6 4.6 RWST Drain Down Evaluation When the RWST drains down to the point at which the ECCS suction is switched from the RWST to the containment sump, the ECCS fluid temperature is expected to increase. [

] an evaluation was performed for Millstone Unit 3 to determine the potential impact of a higher pumped ECCS injection source temperature following sump switchover on the SBLOCA results.

[

] Therefore, the Millstone Unit 3 SBLOCA analysis results are not impacted by a higher ECCS temperature from the switchover.

Serial No.23-105 Docket No. 50-423 Attachment 4, Page 34 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-7 Table 4-1 Summary of Cold Leg Pump Discharge Break Spectrum Results 6

No clad heat-up experienced, therefore, reported value is the initial clad temperature.

7

[

]

8

[ ]

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-8 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum 9

No clad heat-up experienced, therefore, reported value is the initial clad temperature.

10

[

]

11

[ ]

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-9 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd)

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-10 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd)

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-11 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd)

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-12 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd)

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Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-13 Figure 4-1 Cold Leg Pump Discharge Break Spectrum Peak Cladding Temperature versus Break Size Serial No.23-105 Docket No. 50-423 Attachment 4, Page 41 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-14 Figure 4-2 Reactor Power - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 42 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-15 Figure 4-3 Primary and Secondary System Pressures - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 43 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-16 Figure 4-4 Break Mass Flow Rate - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 44 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-17 Figure 4-5 Break Vapor Void Fraction - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 45 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-18 Figure 4-6 Loop Seal Upside Collapsed Levels - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 46 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-19 Figure 4-7 Downcomer Collapsed Liquid Level - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 47 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-20 Figure 4-8 Primary System Masses - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 48 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-21 Figure 4-9 RCS Loop Mass Flow Rates - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 49 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-22 Figure 4-10 Steam Generator Main Feedwater Flow Mass Rates - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 50 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-23 Figure 4-11 Steam Generator MSSV Mass Flow Rates - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 51 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-24 Figure 4-12 Steam Generator Auxiliary Feedwater Flow Rate - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 52 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-25 Figure 4-13 Steam Generator Total Secondary Side Mass - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 53 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-26 Figure 4-14 Steam Generator Narrow Range Level - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 54 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-27 Figure 4-15 High Head Safety Injection Mass Flow Rates - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 55 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-28 Figure 4-16 Intermediate Head Safety Injection Mass Flow Rates - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 56 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-29 Figure 4-17 Low Head Safety Injection Mass Flow Rates - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 57 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-30 Figure 4-18 Accumulator Mass Flow Rates - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 58 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-31 Figure 4-19 Total ECCS and Break Mass Flow Rates - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 59 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-32 Figure 4-20 Hot Assembly Collapsed Liquid Level - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 60 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-33 Figure 4-21 Cladding Temperature at PCT Node - 8.60 inch Break Serial No.23-105 Docket No. 50-423 Attachment 4, Page 61 of 62

Framatome Inc. ANP-4031NP Revision 0 Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 5-1

5.0 REFERENCES

1. EMF-2328(P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2001.

2. EMF-2328(P)(A), Revision 0; Supplement 1(P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2012.

3. ANP-10349P-A, Revision 0, GALILEO Implementation in LOCA Methods, November 2021.

4. ANP-10323P-A, Revision 1, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors, November 2020.

5. Code of Federal Regulations, Title 10, Part 50, Section 46, Acceptance Criteria For Emergency Core Cooling Systems For Light-Water Nuclear Power Reactors, August 2007.

6. Code of Federal Regulations, Title 10, Part 50, Appendix K, ECCS Evaluation Models, June 2000.

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Serial No.23-105 Docket No. 50-423 Page 1 of 106 Attachment 6 ANP-4032NP, REVISION 0, MILLSTONE UNIT 3 REALISTIC LARGE BREAK LOCA ANALYSIS WITH GAIA FUEL DESIGN, LICENSING REPORT (NON-PROPRIETARY)

Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 2 of 106

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 3 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Initial Issue Serial No.23-105 Docket No. 50-423 Attachment 6, Page 4 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1 2.0

SUMMARY

OF RESULTS ................................................................................. 2-1

3.0 DESCRIPTION

OF ANALYSIS.......................................................................... 3-1 3.1 Acceptance Criteria ................................................................................ 3-1 3.2 Description of LBLOCA Event ................................................................. 3-1 3.3 Description of Analytical Models ............................................................. 3-3 3.4 GDC-35 Limiting Condition Determination .............................................. 3-6 3.5 Overall Statistical Compliance to Criteria ................................................ 3-7 3.6 Plant Description ..................................................................................... 3-7 3.7 Safety Evaluation Limitations .................................................................. 3-9 4.0 RLBLOCA ANALYSIS ....................................................................................... 4-1 4.1 RLBLOCA Results .................................................................................. 4-1 4.2 Conclusions ............................................................................................ 4-3

5.0 REFERENCES

.................................................................................................. 5-1 APPENDIX A [ ]

SUMMARY

OF KEY INPUT AND OUTPUT PARAMETERS.......................................................................... A-1 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 5 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page iii List of Tables Page Table 3-1 EMF-2103(P)(A), Revision 3, SE Limitations Evaluation ......................... 3-10 Table 3-2 ANP-10349P-A, Revision 0, Limitations and Conditions on GALILEO Applications ............................................................................................. 3-13 Table 4-1 RLBLOCA Analysis - Plant Parameter Values and Ranges ...................... 4-4 Table 4-2 Statistical Distribution Used for Process Parameters ................................ 4-8 Table 4-3 Passive Heat Sinks and Material Properties in Containment Geometry ................................................................................................... 4-9 Table 4-4 Compliance with 10 CFR 50.46(b)........................................................... 4-10 Table 4-5 Summary of Major Parameters for the Demonstration Case ................... 4-11 Table 4-6 Calculated Event Times for the Demonstration Case .............................. 4-12 Table 4-7 Heat Transfer Parameters for the Demonstration Case .......................... 4-13 Table 4-8 Fuel Rod Rupture Ranges of Parameters ................................................. 4-2 Table A-1 Summary of Key Input and Output Parameters, Part 1 ............................. A-1 Table A-2 Summary of Key Input and Output Parameters, Part 2 ........................... A-16 Table A-3 Summary of Key Input and Output Parameters, Part 3 ........................... A-30 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 6 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page iv List of Figures Page Figure 3-1 Representative Primary System Noding ............................................... 3-14 Figure 3-2 Representative Secondary System Noding .......................................... 3-15 Figure 3-3 Representative Reactor Vessel Noding ................................................ 3-16 Figure 4-1 Scatter Plot Key Parameters................................................................... 4-3 Figure 4-1 Scatter Plot Key Parameters (continued) ................................................ 4-4 Figure 4-2 PCT versus PCT Time Scatter Plot ........................................................ 4-5 Figure 4-3 PCT versus Break Size Scatter Plot ....................................................... 4-6 Figure 4-4 Maximum Local Oxidation versus PCT Scatter Plot ............................... 4-7 Figure 4-5 Total Core Wide Oxidation versus PCT Scatter Plot ............................... 4-8 Figure 4-6 Demonstration Case - Peak Cladding Temperature (Independent of Elevation) ................................................................................................ 4-9 Figure 4-7 Demonstration Case - Break Flow ........................................................ 4-10 Figure 4-8 Demonstration Case - Core Inlet Mass Flux ......................................... 4-11 Figure 4-9 Demonstration Case - Core Outlet Mass Flux ...................................... 4-12 Figure 4-10 Demonstration Case - Void Fraction at RCS Pumps ............................ 4-13 Figure 4-11 Demonstration Case - ECCS Flows (Includes Accumulator, HHSI, IHSI and LHSI) ...................................................................................... 4-14 Figure 4-12 Demonstration Case - Upper Plenum Pressure .................................... 4-15 Figure 4-13 Demonstration Case - Collapsed Liquid Level in the Downcomer ........ 4-16 Figure 4-14 Demonstration Case - Collapsed Liquid Level in the Lower Plenum .... 4-17 Figure 4-15 Demonstration Case - Core Collapsed Liquid Level ............................ 4-18 Figure 4-16 Demonstration Case - Containment and Loop Pressures ..................... 4-19 Figure 4-17 Demonstration Case - Pressure Differences between Upper Plenum and Downcomer....................................................................... 4-20 Figure 4-18 [ ] .................................... 4-21 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 7 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page v Nomenclature Acronym Definition AO Axial Offset CFR Code of Federal Regulations CHF Critical Heat Flux CSAU Code Scaling, Applicability and Uncertainty CWO Core-Wide Oxidation ECCS Emergency Core Cooling System ECR Equivalent Cladding Reacted EM Evaluation Model EMDAP Evaluation Model Development and Assessment Process FH Nuclear Enthalpy Rise Factor/Radial Peaking Factor FQ Total Peaking Factor/Global Peaking Factor Framatome Framatome Inc.

FSRR Fuel Swell Rupture and Relocation Gd2O3 Gadolinia or Gad GDC General Design Criteria HHSI High Head Safety Injection HMP High Mechanical Performance HTC Heat Transfer Coefficient IGM Intermediate GAIA Mixing Grid IHSI Intermediate Head Safety Injection k(z) Axial-Dependent Peaking Factor LBLOCA Large Break Loss-of-Coolant Accident LCO Limiting Condition of Operation LEU Low-Enriched Uranium LHGR Linear Heat Generation Rate Serial No.23-105 Docket No. 50-423 Attachment 6, Page 8 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page vi Acronym Definition LHSI Low Head Safety Injection LOCA Loss-of-Coolant Accident LOOP Loss-of-Offsite Power MLO Maximum Local Oxidation No-LOOP No Loss of Offsite Power NRC U.S. Nuclear Regulatory Commission NSSS Nuclear Steam Supply System PCT Peak Clad Temperature PIRT Phenomena Identification and Ranking Table PWR Pressurized Water Reactor RAI Request for Additional Information RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RLBLOCA Realistic Large Break Loss of Coolant Accident SE Safety Evaluation SG Steam Generator SI Safety Injection SIAS Safety Injection Actuation Signal TS Technical Specification UTL Upper Tolerance Limit Serial No.23-105 Docket No. 50-423 Attachment 6, Page 9 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 1-1

1.0 INTRODUCTION

This report summarizes the Realistic Large Break Loss-of-Coolant Accident (RLBLOCA) analysis for Millstone Power Station Unit 3. The purpose of the RLBLOCA analysis is to support the fuel transition at Millstone Unit 3 to the Framatome GAIA fuel design. This analysis was performed in accordance with the U.S. Nuclear Regulatory Commission (NRC)-approved S-RELAP5-based methodology described in Reference 1 and supplemented by Reference 2.

Millstone Unit 3 is a 4-loop, Westinghouse-designed Pressurized Water Reactor (PWR).

The Framatome GAIA fuel design with M5Framatome cladding for Millstone Unit 3 consists of a 17x17 array with GAIA and Intermediate GAIA Mixing (IGM) grids, a lower High Mechanical Performance (HMP) grid and an upper HMP grid. The fuel assembly includes a MONOBLOC guide tube design, M5Framatome fuel rod design and a GRIP lower nozzle. The fuel is standard UO2 fuel with 2, 4, 6, and 8 weight-percent Gadolinia (Gd2O3) rods included.

The analysis assumes full-power operation at a core power level of 3723 MWt (includes measurement uncertainty), a maximum-allowed total peaking factor (FQ) of 2.6 (represents total peaking with an axial-dependent factor k(z) set to 1.0), a radial peaking factor (FH) of 1.70 (includes uncertainty), and up to 10% steam generator (SG) tube plugging per SG. This analysis also addresses typical operational ranges or technical specification (TS) limits (whichever is applicable) with regard to [

] The analysis explicitly analyzes fresh and once-burned fuel assemblies.

The plant parameter specification for this analysis is provided in Table 4-1. The analysis uses the Fuel Swelling, Rupture, and Relocation (FSRR) model to determine if cladding rupture occurs and evaluate the consequences of FSRR on the transient response.

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 1-2 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 11 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 2-1 2.0

SUMMARY

OF RESULTS The UTL results providing 95/95 simultaneous coverage from this evaluation meet the 10 CFR 50.46(b) criteria with a PCT of 1835°F, a maximum local oxidation of 6.47 percent and a total core-wide oxidation of 0.062 percent. The PCT of 1835°F occurred in a fresh UO2 rod with an assembly burnup of 2.0 GWd/mtU. The results of the analysis demonstrate the adequacy of the ECCS to support the 10 CFR 50.46(b) (1-3) criteria (Reference 4).

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-1

3.0 DESCRIPTION

OF ANALYSIS 3.1 Acceptance Criteria The purpose of the analysis is to verify the adequacy of the Millstone Unit 3 ECCS by demonstrating compliance with the following 10 CFR 50.46(b) criteria (Reference 4):

1. The calculated maximum fuel element cladding temperature shall not exceed 2200°F.

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

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

The final two criteria, coolable geometry and long-term cooling, are treated in separate plant-specific evaluations.

Note: The original 17% value in the second acceptance criterion for MLO was based on the usage of the Baker-Just correlation. For present reviews on ECCS Evaluation Model (EM) applications, the NRC staff imposed a limitation specifying that the equivalent cladding reacted (ECR) results calculated using the Cathcart-Pawel correlation are considered acceptable in conformance with 10 CFR 50.46(b)(2) if the ECR value is less than 13% (Section 3.3.3, NRC Final Safety Evaluation (SE) for EMF-2103(P) Revision 3, See Reference 1). The limitation is addressed in Table 3-1.

3.2 Description of LBLOCA Event A Large Break Loss-of-Coolant Accident (LBLOCA) is initiated by a postulated rupture of the Reactor Coolant System (RCS) primary piping. The most challenging break Serial No.23-105 Docket No. 50-423 Attachment 6, Page 13 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-2 location is in the cold leg piping between the reactor coolant pump and the reactor. The plant is assumed to be operating normally at full power prior to the accident and the break is assumed to open instantaneously. A worst case single-failure is also assumed to occur during the accident. The single-failure for this analysis, as defined in the EM, is the loss of one ECCS pumped injection train without the loss of containment spray.

The LBLOCA event is typically described in three phases: blowdown, refill, and reflood.

Following the initiation of the break, the blowdown phase is characterized by a sudden depressurization from operating pressure down to the saturation pressure of the hot leg fluid. For larger cold leg breaks, an immediate flow reversal and stagnation occurs in the core due to flow out the break, which causes the fuel rods to pass through critical heat flux (CHF), usually within one second following the break. Following this initial rapid depressurization, the RCS depressurizes at a more gradual rate. Reactor trip and emergency injection signals occur when either the low pressure setpoint or the containment high-pressure setpoint are reached. However, for LBLOCA, reactor trip and scram are not modeled and reactor shutdown is accomplished by moderator reactivity feedback. During blowdown, core cooling is supported by the natural evolution of the RCS flow pattern as driven by the break flow.

When the system pressure falls below the accumulator pressure, flow from the accumulator is injected into the cold legs ending the blowdown period and initiating the refill period. Once the system pressure falls below the respective shutoff heads of the safety injection systems and the system startup time delays are met, flow from the pumped safety injection systems is injected into the RCS. While some of the ECCS flow bypasses the core and goes directly out of the break, the downcomer and lower plenum gradually refill until the mixture in the lower head and lower plenum regions reaches the bottom of the active core and the reflood period begins. Core cooling is supported by the natural evolution of the RCS flow pattern as driven by the break flow and condensation in the RCS promoted by safety injection. Towards the end of the refill period, heat transfer from the fuel rods is relatively low, steam cooling and rod-to-rod radiation being the primary mechanisms of core heat removal.

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-3 Once the lower plenum is refilled to the bottom of the fuel rod heated length, refill ends and the reflood phase begins. Substantial ECCS fluid is retained in the downcomer during refill. This provides the driving head to move coolant into the core. As the mixture level moves up the core, steam is generated and liquid is entrained, providing cooling in the upper core regions. The two-phase mixture extends into the upper plenum and some liquid may de-entrain and flow downward back into the cooler core regions. The remaining entrained liquid passes into the steam generators where it vaporizes, adding to the steam that must be discharged through the break and out of the system. The difficulty of venting steam is, in general, referred to as steam binding.

It acts to impede core reflood rates. With the initiation of reflood, a quench front starts to progress up the core. With the advancement of the quench front, the cooling in the upper regions of the core increases, eventually arresting the rise in fuel rod surface temperatures. Later the core is quenched and a pool cooling process is established that can maintain the cladding temperature near saturation, so long as the ECCS makes up for the core boil off.

3.3 Description of Analytical Models The NRC-approved RLBLOCA methodology is documented in EMF-2103(P)(A)

Realistic Large Break LOCA Methodology for Pressurized Water Reactors (Reference 1). The methodology follows the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology (Reference 5) and the requirements of the Evaluation Model Development and Assessment Process (EMDAP) documented in Reference 6. The CSAU method outlines an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifies the uncertainties in a Loss-of-Coolant Accident (LOCA) analysis.

The Framatome RLBLOCA methodology evaluation model for event response of the primary and secondary systems and the hot fuel rod used in this analysis is based on the use of two computer codes.

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-4

GALILEO for computation of the initial fuel stored energy, fission gas release, and the transient fuel-cladding gap conductance.

S-RELAP5 for the thermal-hydraulic system calculations (includes ICECON for containment response).

The methodology (Reference 1) has been reviewed and approved by the NRC to perform LBLOCA analyses. The governing two-fluid (plus non-condensable) model with conservation equations for mass, energy, and momentum transfer is used. The reactor core is modeled in S-RELAP5 with heat generation rates determined from reactor kinetics equations (point kinetics) with reactivity feedback, and with actinide and fission product decay heat.

The two-fluid formulation uses a separate set of conservation equations and constitutive relations for each phase. The effects of one phase on the other are accounted for by interfacial friction and heat and mass transfer interaction terms in the equations. The conservation equations have the same form for each phase; only the constitutive relations and physical properties differ.

The modeling of plant components is performed by following guidelines developed to ensure accurate accounting for physical dimensions and that the dominant phenomena expected during the LBLOCA event are captured. The basic building blocks for modeling are hydraulic volumes for fluid paths and heat structures for heat transfer. In addition, special purpose components exist to represent specific components such as the Reactor Coolant Pumps (RCPs) or the SG separators. All geometries are modeled at the resolution necessary to best resolve the flow field and the phenomena being modeled within practical computational limitations.

The analysis considers blockage effects due to clad swelling and rupture as well as increased heat load due to fuel relocation in the ballooned region of the cladding in the prediction of the hot fuel rod PCT.

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-5 A typical calculation using S-RELAP5 begins with the establishment of a steady-state initial condition with all loops intact. The input parameters and initial conditions for this steady-state calculation are chosen to reflect plant technical specifications or to match measured data. Specific parameters are discussed in Section 3.6. Additionally, the GALILEO code provides initial conditions for the S-RELAP5 fuel models.

Following the establishment of an acceptable steady-state condition, the transient calculation is initiated by introducing a break into one of the loops. The evolution of the transient through blowdown, refill, and reflood is computed continuously using S-RELAP5. Containment pressure is calculated by the ICECON module within S-RELAP5.

A detailed assessment of the S-RELAP5 computer code was made through comparisons to experimental data. These assessments were used to develop quantitative estimates of the ability of the code to predict key physical phenomena in a PWR LBLOCA. The final step of the best-estimate methodology is to combine all the uncertainties related to the code and plant parameters and estimate values for the first three criteria of 10 CFR 50.46(b) with a probability of at least 95 percent with 95 percent confidence. The steps taken to derive the uncertainty estimate are summarized below:

1. Base Plant Input File Development First, base GALILEO and S-RELAP5 input files for the plant (including the containment input file) are developed. The code input development guidelines documented in Appendix A of Reference 1 are applied to ensure that model nodalization is consistent with the model nodalization used in the code validation.

2. Sampled Case Development The statistical approach requires that many sampled cases be created and processed. For every set of input created, each key LOCA parameter is randomly sampled over a range established through code uncertainty assessment or expected operating limits (provided by plant technical Serial No.23-105 Docket No. 50-423 Attachment 6, Page 17 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-6 specifications or data). Those parameters considered "key LOCA parameters" are listed in Table A-6 of Reference 1. This list includes both parameters related to LOCA phenomena, based on the PIRT provided in Reference 1, and to plant operating parameters. The uncertainty ranges associated with each of the model parameters are provided in Table A-7 of Reference 1.

3. Determination of Adequacy of ECCS The RLBLOCA methodology uses a non-parametric statistical approach to determine that the first three criteria of 10 CFR 50.46(b) are met with a probability higher than 95 percent with 95 percent confidence.

3.4 GDC-35 Limiting Condition Determination General Design Criteria (GDC)-35 requires that a system be designed to provide abundant core cooling with suitable redundancy such that the capability is maintained in either the LOOP or No-LOOP conditions. [

]

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-7 3.5 Overall Statistical Compliance to Criteria 3.6 Plant Description The plant analyzed is the Millstone Unit 3, Westinghouse-designed PWR, which has four loops, each with a hot leg, a U-tube steam generator, and a cold leg with an RCP.

The Reactor Coolant System (RCS) includes one pressurizer connected to a hot leg.

The ECCS provides injection to each of the four loops via the centrifugal charging/high head safety injection (HHSI) system, SI/intermediate head safety injection (IHSI) system, residual heat removal (RHR)/low head safety injection (LHSI) system, and accumulators. For the purpose of this report, the centrifugal charging/HHSI, SI/IHSI, and RHR/LHSI systems are referred to as the HHSI, IHSI, and LHSI systems, respectively. The RLBLOCA transients are of sufficiently short duration that the switchover to sump cooling water for ECCS pumped injection does not need to be considered.

The RCS, reactor vessel, pressurizer, and ECCS are explicitly modeled in the S-RELAP5 model. For each RCS loop, the LOCA ECCS model includes an injection connection to the cold leg for the accumulator and another connection for HHSI. IHSI and LHSI are modeled with separate injection connections to each of the four accumulator lines. The accumulator and HHSI injection connections to the cold leg pipe are downstream of the RCP discharge. The ECCS pumped injection is modeled as a table of flow versus backpressure. Also modeled is the secondary-side steam generator that is instantaneously isolated (closed main steam isolation valve and feedwater trip) at the time of the break. The primary and secondary coolant systems for Millstone Unit 3 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 19 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-8 were nodalized to be consistent with code input guidelines in Appendix A of Reference 1. Representative system nodalization details are shown in Figure 3-1 through Figure 3-3.

The results used to demonstrate compliance with the 10 CFR 50.46(b) criteria are only applicable to the Framatome fuel product. However, the analysis includes considerations for the mixed core scenario. [

]

As described in Section 3.3, many parameters associated with LBLOCA phenomenological uncertainties and plant operation ranges are sampled. Values for process or operational parameters, including ranges of sampled process parameters, and fuel design parameters used in this analysis are given in Table 4-1. Table 4-2 presents a summary of the uncertainties used in the analysis. [

] The passive heat sinks and material properties used in the containment input model are provided in Table 4-3.

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-9 3.7 Safety Evaluation Limitations The RLBLOCA analysis for Millstone Unit 3 presented herein is consistent with the submitted RLBLOCA methodology documented in EMF-2103(P)(A), Revision 3 (Reference 1) and supplemented by ANP-10349P-A, Revision 0 (Reference 2). The limitations and conditions from the NRC SE for EMF-2103(P)(A), Revision 3 (Reference 1), are addressed in Table 3-1. The limitations and conditions from Reference 3 on GALILEO applications for ANP-10349P-A, Revision 0 (Reference 2), are addressed in Table 3-2.

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 21 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-10 Table 3-1 EMF-2103(P)(A), Revision 3, SE Limitations Evaluation Limitations

Response

(Sub-sections of Section 4.0 in Reference 1) 1 This EM was specifically reviewed in This analysis applies only to the accordance with statements in EMF-2103, acceptance criteria set forth in 10 Revision 3. The NRC staff determined that the CFR 50.46(b), paragraphs (1)

EM is acceptable for determining whether plant- through (3).

specific results comply with the acceptance criteria set forth in 10 CFR 50.46(b), paragraphs (1) through (3). AREVA did not request, and the NRC staff did not consider, whether this EM would be considered applicable if used to determine whether the requirements of 10 CFR 50.46(b)(4), regarding coolable geometry, or (b)(5), regarding long-term core cooling, are satisfied. Thus, this approval does not apply to the use of SRELAP5-based methods of evaluating the effects of grid deformation due to seismic of LOCA blowdown loads, or for evaluating the effects of reactor coolant system boric acid transport. Such evaluations would be considered separate methods.

2 EMF-2103, Revision 3, approval is limited to Millstone Unit 3 is a 4-loop application for 3-loop and 4-loop Westinghouse- Westinghouse-designed NSSS with designed nuclear steam supply systems cold leg ECCS injection.

(NSSSs), and to Combustion Engineering-designed NSSSs with cold leg ECCS injection, only. The NRC staff did not consider model applicability to other NSSS designs in its review.

3 The EM is approved based on models that are The analysis supports operation with specific to AREVA proprietary M5 fuel cladding. M5Framatome cladding.

The application of the model to other cladding types has not been reviewed.

4 Plant-specific applications will generally be The modeling guidelines contained in considered acceptable if they follow the Appendix A of EMF-2103(P)(A),

modeling guidelines contained in Appendix A to Revision 3 (Reference 1) were EMF 2103, Revision 3. Plant-specific licensing followed completely for the analysis actions referencing EMF 2103, Revision 3, described in this report.

analyses should include a statement summarizing the extent to which the guidelines were followed, and justification for any departures.

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-11 Limitations

Response

(Sub-sections of Section 4.0 in Reference 1) 5 The response to RAI 15 indicates that the fuel The analysis burnups applied in this pellet relocation packing factor is derived from analysis do not exceed the rod data that extend to currently licensed fuel average burnup of burnup limits (i.e., rod average burnup of [ ]

[ ]). Thus, the approval of this method is limited to fuel burnup below this value. Extension beyond rod average burnup of

[ ] would require a revision or supplement to EMF-2103, Revision 3, or plant-specific justification.

6 The response to RAI 15 indicates that the fuel The analysis uses the approved pellet relocation packing factor is derived from EMF- 2103(P)(A), Revision 3 currently available data. Should new data (Reference 1) relocation packing become available to suggest that fuel pellet factor application. [

fragmentation behavior is other than that suggested by the currently available database, the NRC may request AREVA to update its ]

model to reflect such new data.

7 The regulatory limit contained in 10 CFR The MLO UTL is less than 13%

50.46(b)(2), requiring cladding oxidation not to (Table 4-4).

exceed 17 percent of the initial cladding thickness prior to oxidation, is based on the use of the Baker-Just oxidation correlation. To account for the use of the Cathcart-Pawel correlation, this limit shall be reduced to 13 percent, inclusive of pre-transient oxide layer thickness.

8 In conjunction with Limitation 7 above, Cathcart- All second cycle fuel rod [

Pawel oxidation results will be considered

]

acceptable, provided plant-specific [

] If second-cycle fuel is identified in a plant-specific analysis, whose

[ ] the NRC staff reviewing the plant-specific analysis may request technical justification or quantitative assessment, demonstrating that

[

]

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-12 Limitations

Response

(Sub-sections of Section 4.0 in Reference 1) 9 The response to RAI 13 states that all operating ranges used in a plant-specific analysis are

[

supplied for review by the NRC in a table like Table B-8 of EMF-2103, Revision 3. In plant-specific reviews, the uncertainty treatment for plant parameters will be considered acceptable if plant parameters are [

]

] as appropriate.

Alternative approaches may be used, provided they are supported with appropriate justification.

10 [ [ ] were not used in this analysis.

]

11 Any plant submittal to the NRC using The present analysis is the first EMF-2103, Revision 3, which is not based on statistical application of EMF-2103, the first statistical calculation intended to be the Revision 3 for this plant.

analysis of record must state that a re-analysis has been performed and must identify the changes that were made to the evaluation model and/or input in order to obtain the results in the submitted analysis.

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 24 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-13 Table 3-2 ANP-10349P-A, Revision 0, Limitations and Conditions on GALILEO Applications Ranges of Applicability

Response

(Section 1.2 in Reference 3)

Pressurized water reactor designs using Low- This analysis was performed for the Enriched Uranium (LEU) fuel loading Millstone Unit 3 plant, which is a PWR, using LEU fuel.

The fuel burnups applied in this Rod average burnups up to [ ] gigawatt-days analysis do not exceed the rod average per metric ton of uranium (GWd/MTU) for Zircaloy-4 and up to [ ] GWd/MTU for M5 cladding burnup of [ ]

Zircaloy-4 and M5 cladding The analysis supports operation with M5Framatome cladding.

This analysis was performed using fuel Rod diameter between [ ] mm and [ ] mm with a rod outside diameter of 9.5 mm.

Uranium 235U enrichments up to 5 weight percent The 235U enrichments applied in this (wt%) analysis do not exceed 5 weight percent.

Gadolinia concentrations up to 10 wt% The Gadolinia concentration analyzed does not exceed 10 wt%.

Nominal true pellet density ranging from [ The initial pellet density is [ ]

percent of the theoretical density of

] percent of the theoretical density of UO2 UO2.

This analysis was performed using fuel Fuel grain sizes ranging from [ ] pellets with a grain size of microns (mean linear intercept)

[ ]

Pellets manufactured by dry conversion and The fuel pellet manufacturing process ammonium diuranate for the fuel design considered in this analysis is dry conversion and ammonium diuranate.

Fuel temperature up to the melting point to the This is related to thermo-mechanical approved burnup range methods and is not subject to the limitation for this LOCA analysis.

Cladding strain up to the approved transient clad This is related to thermo-mechanical strain limit methods and is not subject to the limitation for this LOCA analysis.

Internal rod pressure up to pressures that protect This is related to thermo-mechanical from clad lift-off and hydride reorientation methods and is not subject to the limitation for this LOCA analysis.

Fuel rod power not to exceed levels as limited by This is related to thermo-mechanical fuel melt, cladding strain, and rod pressure criteria methods and is not subject to the limitation for this LOCA analysis.

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-14 Figure 3-1 Representative Primary System Noding Serial No.23-105 Docket No. 50-423 Attachment 6, Page 26 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-15 Figure 3-2 Representative Secondary System Noding Serial No.23-105 Docket No. 50-423 Attachment 6, Page 27 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 3-16 Figure 3-3 Representative Reactor Vessel Noding Serial No.23-105 Docket No. 50-423 Attachment 6, Page 28 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-1 4.0 RLBLOCA ANALYSIS 4.1 RLBLOCA Results For a simultaneous coverage/confidence level of 95/95, the UTL values, [

] are a PCT of 1835°F, an MLO of 6.47 percent, and a CWO of 0.062 percent. The fraction of total hydrogen generated was not directly calculated; however, it is conservatively bounded by the calculated total core wide percent oxidation, which is well below the 1 percent limit.

A summary of the major input parameters for the demonstration case is provided in Table 4-5. The sequence of event times for the demonstration case is provided in Table 4-6. The heat transfer parameter ranges for the demonstration case are provided in Table 4-7. Table 4-8 [

]

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 29 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-2 The analysis scatter plots for the case set are shown in Figure 4-1 through Figure 4-5.

Figure 4-1 shows linear scatter plots of the key parameters sampled for all cases.

These figures illustrate the parameter ranges used in the analysis. Visual examination of the linear scatter plots demonstrates that the spread and coverage of all the values used is appropriate and within the uncertainty ranges listed in Table 4-2. APPENDIX A provides a listing of all the sampled input values for each case. Key results such as the PCT and event timings are also listed for the case set.

Figure 4-2 and Figure 4-3 show PCT scatter plots versus the time of PCT and versus break size, respectively. The scatter plots for the maximum local oxidation and total core-wide oxidation are shown in Figure 4-4 and Figure 4-5, respectively.

Figure 4-2 shows about 7% of cases have PCT during the blowdown phase (PCT time less than approximately 30 seconds). The next cluster of PCTs occurs during the early to late reflood period. Blowdown PCT cases are dominated by rapid RCS depressurization and stored energy content. Early reflood PCT cases are dominated by decay heat removal capacity. In general, plants with high pressure accumulators inject early in the transient when the break flow is still high. The high pressure and high break flow drive some of this fluid to bypass the core, delaying the progression of the core reflood. This results in cases with PCTs in the early reflood phase of the transient.

The high PCT cases in the upper part of Figure 4-2 are mainly influenced by the area of the break. This is demonstrated in Figure 4-3 which shows a general increasing trend in PCT with break size. And similarly smaller break sizes result in lower PCTs. From all sampled parameters, the break size is a dominant effect on PCT because of its influence on the rate of primary depressurization.

Figure 4-4 shows general trend in increasing oxidation results with increasing PCT.

Since the MLO includes the pre-transient oxidation, the MLO is not only a function of cladding temperature but also of time in cycle (burnup). The CWO also shows a strong correlation to PCT as demonstrated in Figure 4-5, as higher PCT cases would have higher oxidation throughout the core.

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-3 The demonstration case is a reflood peak case with a PCT timing of 137 seconds.

Figure 4-6 through Figure 4-17 show key parameters from the S-RELAP5 calculations for the demonstration case. The transient progression for the demonstration case follows that described in Section 3.2.

4.2 Conclusions This report describes and provides results from the RLBLOCA analysis for the Millstone Unit 3 with the Framatome GAIA fuel design. The application of the Framatome RLBLOCA methodology involves developing input decks, executing the simulations that comprise the uncertainty analysis, retrieving PCT, MLO, and CWO information and determining the simultaneous UTL results for the criteria. [

] The UTL results providing a 95/95 simultaneous coverage/confidence level from this evaluation meet the 10 CFR 50.46(b) criteria with a PCT of 1835°F, a MLO of 6.47 percent and a CWO of 0.062 percent.

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-4 Table 4-1 RLBLOCA Analysis - Plant Parameter Values and Ranges Plant Parameter Parameter Value 1.0 Plant Physical Description 1.1 Fuel a) Cladding outside diameter 0.374 in.

b) Cladding inside diameter 0.329 in.

c) Pellet outside diameter 0.3225 in.

d) Initial Pellet density [ ]

e) Active fuel length 144 in.

f) Gd2O3 concentrations 2, 4, 6, 8 weight-percent 1.2 RCS a) Flow resistance Analysis b) Pressurizer location

[

]

c) Hot assembly location Anywhere in core d) Hot assembly type 17x17 e) SG tube plugging 10 percent 2.0 Plant Initial Operating Conditions 2.1 Reactor Power a) Analyzed reactor power 3723(1) MWt b) FQ 2.6(1) c) FH 1.70(1) 2.2 Fluid Conditions a) Total Loop flow 134.8 Mlbm/hr M 156.8 Mlbm/hr b) RCS average temperature 576.5°F T 594.5°F c) Upper head temperature ~RCS Cold Leg Temperature(2) d) Pressurizer pressure 2200 psia P 2300 psia e) Pressurizer liquid level 49.9 percent L 71.5 percent f) Accumulator pressure 626 psia P 704 psia g) Accumulator liquid volume 884.5 ft3 V 939.5 ft3 h) Accumulator temperature 75°F T 125°F(3) i) Accumulator resistance fL/D As-built piping configuration j) Accumulator boron 2550 ppm 1

Includes measurement uncertainty.

2 Upper head temperature will change based on sampling of RCS temperature.

3 Coupled with containment temperature.

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-5 Table 4-1 RLBLOCA Analysis - Plant Parameter Values and Ranges (Continued)

Plant Parameter Parameter Value 3.0 Accident Boundary Conditions a) Break location Cold leg pump discharge b) Break type Double-ended guillotine or split c) Break size (each side, relative to cold leg pipe area) d) ECCS pumped injection 100°F temperature e) HHSI pump delay 45 s (No-LOOP and LOOP) f) IHSI pump delay 45 s (No-LOOP and LOOP) g) LHSI pump delay 45 s (No-LOOP and LOOP) h) Initial containment pressure 12.3 psia i) Initial containment temperature 75°F T 125°F j) Containment sprays delay 0s k) Containment spray water 40°F temperature l) LHSI Flow RCS Cold Leg Broken Loop Total Intact Loops Pressure (psia) Flow (gpm) Flow (gpm) 14.7 1027.50 2596.89 34.7 962.10 2429.61 54.7 890.50 2246.79 74.7 813.10 2049.09 94.7 543.70 1359.39 114.7 289.40 723.39 134.7 52.20 130.41 134.8 0.00 0.00 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 33 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-6 Table 4-1 RLBLOCA Analysis - Plant Parameter Values and Ranges (Continued)

Plant Parameter Parameter Value m) IHSI Flow RCS Cold Leg Broken Loop Total Intact Loops Pressure (psia) Flow (gpm) Flow (gpm) 14.7 143.20 402.21 114.7 138.10 388.11 214.7 132.90 373.41 314.7 127.30 357.69 414.7 121.40 341.19 514.7 115.30 324.09 614.7 108.70 305.19 714.7 101.70 285.51 814.7 93.90 263.79 914.7 85.40 239.91 1014.7 75.80 213.00 1114.7 64.10 180.09 1214.7 47.30 132.99 1314.7 21.90 61.71 1414.7 0.00 0.00 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 34 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-7 Table 4-1 RLBLOCA Analysis - Plant Parameter Values and Ranges (Continued)

Plant Parameter Parameter Value n) HHSI Flow RCS Cold Leg Broken Loop Total Intact Loops Pressure (psia) Flow (gpm) Flow (gpm) 14.7 102.90 281.61 114.7 100.40 274.71 214.7 97.90 267.90 314.7 95.10 260.19 414.7 92.30 252.51 514.7 89.50 244.80 614.7 86.60 237.09 714.7 83.70 228.90 814.7 80.70 220.59 914.7 77.60 212.31 1014.7 74.40 203.79 1114.7 71.00 194.10 1214.7 67.40 184.41 1314.7 63.80 174.69 1414.7 60.20 164.70 1514.7 56.50 154.80 1614.7 52.80 144.51 1714.7 48.30 132.00 1814.7 43.40 118.80 1914.7 38.60 105.51 2014.7 33.60 92.10 2114.7 29.20 79.80 2214.7 25.80 70.71 2314.7 22.50 61.59 2482.7 0.00 0.00 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 35 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-8 Table 4-2 Statistical Distribution Used for Process Parameters Lower Upper Value Value 2200 2300 49.9 71.5 884.5 939.5 626.0 704.0 75 125 2.26 2.35 134.8 156.8 576.5 594.5 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 36 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-9 Table 4-3 Passive Heat Sinks and Material Properties in Containment Geometry Heat Sink Surface Area, ft2 Thickness, ft Material 0.0375 Stainless Steel 1 (Stainless-Steel/Concrete-2.16 ft) 866.0 4.32 Concrete 0.0375 Stainless Steel 2 (Stainless Steel/Concrete-1.5 ft) 7674.0 1.5 Concrete 3 (Concrete 1.5 ft) 157766.0 1.52 Concrete 4 (Concrete - 1.75 ft) 2007.0 1.75 Concrete 2.0 Concrete 5 (Concrete/Carbon 12269.0 0.25 Carbon Steel Steel/Concrete) 10.0 Concrete 0.0428 Carbon Steel 6 (Carbon Steel/Concrete-4.5 ft) 63168.0 4.5 Concrete 0.0462 Carbon Steel 7 (Carbon Steel/Concrete-2.56 ft) 34100.0 2.56 Concrete 8 (Stainless Steel-0.1075 ft) 1722.0 0.1075 Stainless Steel 9 (Carbon Steel-0.0592 ft) 552.0 0.0592 Carbon Steel 10 (Stainless Steel-0.2 ft) 13230.0 0.02 Stainless Steel 11 (Stainless Steel-0.0548 ft) 2063.0 0.0548 Stainless Steel 12 (Carbon Steel-0.0231 ft) 8966.0 0.0231 Carbon Steel 13 (Carbon Steel-0.0825 ft) 1282.0 0.0825 Carbon Steel 14 (Carbon Steel-0.0182 ft) 514279.0 0.0182 Carbon Steel 15 (Carbon Steel-0.00925 ft) 182517.0 0.00925 Carbon Steel 16 (Stainless Steel-0.0304 ft) 11033.0 0.0304 Stainless Steel 17 (Carbon Steel-0.0651 ft) 37068.0 0.0651 Carbon Steel 18 (Stainless Steel-0.0119 ft) 21000.0 0.0119 Stainless Steel 19 (Refueling Cavity Floor-2.16 ft) 866.0 2.16 Concrete 20 (Refueling Cavity Floor-1.5 ft) 7674.0 1.5 Concrete Thermal Volumetric Heat Capacity Heat Sink Material Conductivity Btu/ft3-°F Btu/hr-ft-°F Concrete 1.0 22.152 Carbon Steel 27.0 49.0 Stainless Steel 9.4 60.12 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 37 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-10 Table 4-4 Compliance with 10 CFR 50.46(b)

UTL for 95/95 Simultaneous Coverage/Confidence Parameter Value Case Number PCT (°F) 1835 [ ]

MLO (%) 6.47 [ ]

CWO (%) 0.062 [ ]

Characteristics of Case Setting the PCT UTL PCT (°F) 1835 PCT Rod Type Fresh UO2 Rod Time of PCT (s) 137.67 Elevation within Core (ft) 9.82 Local Maximum Oxidation (%) 5.21 Total Core-Wide Oxidation (%) 0.062 PCT Rod Rupture Time (s) 25.21 Rod Rupture Elevation within Core (ft) 7.89 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 38 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-11 Table 4-5 Summary of Major Parameters for the Demonstration Case Parameter Value Core Power (MWt) 3723 Fresh Fuel Time in Cycle (hrs) 877 Burned Fuel Time in Cycle (hrs) 13737 Fresh Fuel Assembly Avg. Burnup (GWd/mtU) 2.0 Burned Fuel Assembly Avg. Burnup (GWd/mtU) 23.5 Core Peaking Factor, FQ 2.35 Radial Peaking Factor, FH 1.70 Fresh Fuel Axial Offset 0.218 Burned Fuel Axial Offset 0.220 Break Type Guillotine Break Size (ft2/side) 4.105

[ ] [ ]

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 39 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-12 Table 4-6 Calculated Event Times for the Demonstration Case Event Time (sec)

Break Opens 0.0 RCP Trip 0.0 SIAS Issued 1.2 Start of Broken Loop Accumulator Injection 4.7 Start of Intact Loop Accumulator Injection 12.9, 12.9 and 12.9

[ ]

Beginning of Core Recovery (Beginning of Reflood) 30.7 Broken Loop Accumulator Emptied 44.2 Intact Loop Accumulator Emptied 46.3, 45.9 and 46.3

[ ]

Charging/IHSI/LHSI Available 46.2 Broken Loop Charging Delivery Began 46.2 Intact Loop Charging Delivery Began 46.2, 46.2 and 46.2

[ ]

Broken Loop IHSI Delivery Began 46.2 Intact Loop IHSI Delivery Began 46.2, 46.2 and 46.2

[ ]

Broken Loop LHSI Delivery Began 46.2 Intact Loop LHSI Delivery Began 46.2, 46.2 and 46.2

[ ]

PCT Occurred 137.7 Transient Calculation Terminated 1058.7 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 40 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-13 Table 4-7 Heat Transfer Parameters for the Demonstration Case Time (s)

Early Long Term LOCA Phase Blowdown1 Refill Reflood Quench Blowdown Cooling2 Heat Transfer Mode Heat Transfer Correlations Maximum LHGR (kW/ft)

Pressure (psia)

Core Inlet Mass Flux (lbm/s-ft2)

Vapor4 Reynolds Number Liquid Reynolds Number Vapor Prandtl Number Liquid Prandtl Number Vapor5 Superheat

(°F) 1 End of blowdown considered as beginning of refill.

2 Quench to End of Transient.

3

[ ]

4 Not important in pre-CHF heat transfer.

5 Vapor superheat is meaningless during blowdown and system depressurization.

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 41 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-2 Table 4-8 Fuel Rod Rupture Ranges of Parameters Serial No.23-105 Docket No. 50-423 Attachment 6, Page 42 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-3 Figure 4-1 Scatter Plot Key Parameters Serial No.23-105 Docket No. 50-423 Attachment 6, Page 43 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-4 Figure 4-1 Scatter Plot Key Parameters (continued)

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 44 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-5 Figure 4-2 PCT versus PCT Time Scatter Plot Serial No.23-105 Docket No. 50-423 Attachment 6, Page 45 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-6 Figure 4-3 PCT versus Break Size Scatter Plot Serial No.23-105 Docket No. 50-423 Attachment 6, Page 46 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-7 Figure 4-4 Maximum Local Oxidation versus PCT Scatter Plot Serial No.23-105 Docket No. 50-423 Attachment 6, Page 47 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-8 Figure 4-5 Total Core Wide Oxidation versus PCT Scatter Plot Serial No.23-105 Docket No. 50-423 Attachment 6, Page 48 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-9 Figure 4-6 Demonstration Case - Peak Cladding Temperature (Independent of Elevation)

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 49 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-10 Figure 4-7 Demonstration Case - Break Flow Serial No.23-105 Docket No. 50-423 Attachment 6, Page 50 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-11 Figure 4-8 Demonstration Case - Core Inlet Mass Flux Serial No.23-105 Docket No. 50-423 Attachment 6, Page 51 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-12 Figure 4-9 Demonstration Case - Core Outlet Mass Flux Serial No.23-105 Docket No. 50-423 Attachment 6, Page 52 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-13 Figure 4-10 Demonstration Case - Void Fraction at RCS Pumps Serial No.23-105 Docket No. 50-423 Attachment 6, Page 53 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-14 Figure 4-11 Demonstration Case - ECCS Flows (Includes Accumulator, HHSI, IHSI and LHSI)

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 54 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-15 Figure 4-12 Demonstration Case - Upper Plenum Pressure Serial No.23-105 Docket No. 50-423 Attachment 6, Page 55 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-16 Figure 4-13 Demonstration Case - Collapsed Liquid Level in the Downcomer Serial No.23-105 Docket No. 50-423 Attachment 6, Page 56 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-17 Figure 4-14 Demonstration Case - Collapsed Liquid Level in the Lower Plenum Serial No.23-105 Docket No. 50-423 Attachment 6, Page 57 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-18 Figure 4-15 Demonstration Case - Core Collapsed Liquid Level Serial No.23-105 Docket No. 50-423 Attachment 6, Page 58 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-19 Figure 4-16 Demonstration Case - Containment and Loop Pressures Serial No.23-105 Docket No. 50-423 Attachment 6, Page 59 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-20 Figure 4-17 Demonstration Case - Pressure Differences between Upper Plenum and Downcomer Serial No.23-105 Docket No. 50-423 Attachment 6, Page 60 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 4-21 Figure 4-18

[ ]

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 61 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page 5-1

5.0 REFERENCES

1. EMF-2103(P)(A) Revision 3, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, Framatome, June 2016.

2. ANP-10349P-A, Revision 0, GALILEO Implementation in LOCA Methods, November 2021.

3. ANP-10323P-A, Revision 1, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors, November 2020.

4. Code of Federal Regulations, Title 10, Part 50, Section 46, Acceptance Criteria For Emergency Core Cooling Systems For Light-Water Nuclear Power Reactors, August 2007.

5. NUREG/CR-5249, Quantifying Reactor Safety Margins, Application of Code Scaling, Applicability, and Uncertainty Evaluation Methodology to a Large Break, Loss-of-Coolant Accident, U.S. NRC, December 1989.

6. Regulatory Guide 1.203, Transient and Accident Analysis Methods, U.S.

NRC, December 2005.

Serial No.23-105 Docket No. 50-423 Attachment 6, Page 62 of 106

Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page A-1 APPENDIX A [ ]

SUMMARY

OF KEY INPUT AND OUTPUT PARAMETERS The following tables contain the sampled input values for all the cases analyzed. Key results are also included in columns 2 through 6 in Table A-1 for the case set. In all cases, the core power is 3723 MWt (includes uncertainty).

Table A-1 Summary of Key Input and Output Parameters, Part 1 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 63 of 106

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page A-16 Table A-2 Summary of Key Input and Output Parameters, Part 2 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 78 of 106

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Framatome Inc. ANP-4032NP Revision 0 Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design Licensing Report Page A-30 Table A-3 Summary of Key Input and Output Parameters, Part 3 Serial No.23-105 Docket No. 50-423 Attachment 6, Page 92 of 106

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Serial No.23-105 Docket No. 50-423 Page 1 of 4 Attachment 7 FRAMATOME APPLICATION FOR WITHHOLDING AND AFFIDAVIT Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

AFFIDAVIT

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 documents ANP-4031P-000, Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design, and ANP-4032P-000, Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design, and referred to herein as Documents. Information contained in these Documents have 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. These Documents contain 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 these Documents as proprietary and confidential.

5. These Documents have been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in these Documents 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 requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.

Serial No.23-105 Docket No. 50-423 Attachment 7, Page 2 of 4

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 these Documents 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 these Documents have 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.

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

Serial No.23-105 Docket No. 50-423 Attachment 7, Page 3 of 4

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

Executed on: (3/02/2023)

(NAME)

` morris.byram@framatome.com Serial No.23-105 Docket No. 50-423 Attachment 7, Page 4 of 4

Serial No.23-105 Docket No. 50-423 Page 1 of 10 Attachment 8 REQUEST FOR EXEMPTION RELATED TO 10 CFR 50.46 AND 10 CFR 50, APPENDIX K Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.23-105 Docket No. 50-423 Attachment 8, Page 2 of 10

1.

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.12, Dominion Energy Nuclear Connecticut (DENC) requests an exemption from certain requirements of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and Appendix K of 10 CFR 50, ECCS Evaluation Models. The proposed exemption will facilitate the use of Framatome (FRM) GAIA fuel assemblies containing fuel rods fabricated with M5TM cladding material at Millstone Power Station Unit 3 (MPS3). The requested exemption is specific to the types of cladding material specified in these regulations for use in light water reactors.

As written, 10 CFR 50.46 and 10 CFR 50, Appendix K presume the use of zircaloy or ZIRLO' fuel rod cladding. M5TM cladding has a slightly different composition than either of these alloys. Therefore, in order to use M5TM fuel rod cladding, an exemption to these regulations is needed.

2.0 DETAILED DESCRIPTION 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, states Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLOTM cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section. As 10 CFR 50.46 specifically refers to fuel with zircaloy or ZIRLOTM cladding, the use of fuel with M5TM cladding would, in effect, result in noncompliance with this section of the Code.

Also, paragraph I.A.5 of Appendix K to 10 CFR Part 50 states that the rates of energy release, hydrogen generation, and cladding oxidation from the metal-water reaction shall be calculated using the Baker-Just equation. The Baker-Just equation presumes the use of a zirconium alloy different than M5TM. Therefore, use of fuel with M5TM cladding would also require exemption from paragraph I.A.5 of 10 CFR 50, Appendix K.

3.0 DISCUSSION DENC and FRM have entered into agreement for batch implementation of the GAIA fuel at MPS3. A full reload batch of GAIA fuel assemblies is planned for initial insertion in Cycle 24. This onload is currently scheduled for the spring 2025 refueling outage.

Serial No.23-105 Docket No. 50-423 Attachment 8, Page 3 of 10 The GAIA design was generically approved by the NRC in ANP-10342-P-A [Reference 1]. The GAIA design is similar to the fuel assemblies currently used at MPS3. One exception is that the GAIA assemblies use M5TM cladding. M5TM is a fuel rod cladding composed of primarily zirconium (approximately 99%) and niobium (approximately 1%).

It provides improvements in corrosion resistance, hydrogen pickup, axial growth, and diametral creep relative to zircaloy. Use of M5TM as a cladding material does not meet 10 CFR 50.46 and Appendix K of 10 CFR Part 50 requirements, thus an exemption request is needed.

This exemption request supports the use of M5TM cladding in the MPS3 GAIA Small Break and Realistic Large Break Loss of Coolant Accident (SBLOCA and RLBLOCA) evaluations under FRM methods being submitted concurrently with this exemption request1. Note that DENC will separately submit a License Amendment Request (LAR) for the incorporation of M5TM into the MPS3 Technical Specifications (TS), TS 5.3.1, for NRC review and approval to support future use of GAIA fuel at MPS3.

4.0 REGULATORY EVALUATION

4.1. Justification for Exemption Per 10 CFR 50.12, the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that: 1) the exemption is authorized by law, 2) the exemption will not present an undue risk to public health and safety, 3) the exemption is consistent with the common defense and security, and 4) special circumstances, as defined in 10 CFR 50.12(a)(2), are present. The requested exemption from 10 CFR 50.46 and 10 CFR 50, Appendix K, supporting the use of M5TM cladding in loss of coolant accident (LOCA) analysis at MPS3, satisfies these requirements as described below.

1. The requested exemption is authorized by law.

This exemption would allow the use of M5TM fuel rod cladding at MPS3 for GAIA SBLOCA and RLBLOCA analyses performed under FRM methods, instead of zircaloy or ZIRLOTM. The NRC has the authority under 10 CFR 50.12 to grant exemptions from the requirements of 10 CFR Part 50 with proper justification. By submitting this exemption request, MPS3 does not seek an exemption from the acceptance and analytical criteria of 10 CFR 1

Attachments 1-7 to this submittal letter provide the SBLOCA and RLBLOCA TS change request in accordance with 10 CFR 50.90.

Serial No.23-105 Docket No. 50-423 Attachment 8, Page 4 of 10 50.46 and 10 CFR Part 50, Appendix K. Rather, the intent of the request is solely to allow the use of criteria set forth in these regulations for application to the M5TM fuel rod cladding material. Therefore, the exemption is authorized by law.

2. The requested exemption does not present an undue risk to the public health and safety.

The underlying purpose of 10 CFR 50.46 is to ensure nuclear power facilities have adequate acceptance criteria for their ECCS. In the NRC approved Topical Report BAW-10227-P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel" [Reference 2],

FRM demonstrated the effectiveness of the ECCS will not be affected by a change from zircaloy fuel rod cladding to M5TM fuel rod cladding. The analysis described in the Topical Report also demonstrated that ECCS acceptance criteria applied to reactors fueled with zircaloy clad fuel are also applicable to reactors fueled with M5TM fuel rod cladding.

Appendix K, paragraph I.A.5, of 10 CFR 50 ensures cladding oxidation and hydrogen generation are appropriately limited during a LOCA, and conservatively accounted for in the ECCS evaluation model. Appendix K requires the Baker-Just equation be used in the ECCS evaluation model to determine the rate of energy release, cladding oxidation, and hydrogen generation. In the approved Topical Report BAW-10227-P-A, Revision 1, FRM demonstrated that the Baker-Just model is conservative in the evaluated post-LOCA scenarios with respect to the use of the M5TM alloy as a fuel rod cladding material; therefore, the amount of energy release, cladding oxidation, and hydrogen generated in an M5TM clad core during a LOCA remains conservative for MPS3.

Based on the above, no new accident precursors are created by the use of M5TM fuel cladding at MPS3; thus, the probability of postulated accidents is not increased. Also, the consequences of postulated accidents are not increased. Therefore, there is no undue risk to public health and safety.

3. The requested exemption will not endanger the common defense and security.

The M5TM fuel rod cladding is similar in design to the current cladding material used at MPS3. This change in cladding material will not result in any changes to the security aspects associated with the control of special nuclear

Serial No.23-105 Docket No. 50-423 Attachment 8, Page 5 of 10 material. The change in cladding material is unrelated to other nuclear plant security considerations. Therefore, the common defense and security are not impacted by this exemption.

4. Special circumstances are present which necessitate the request of an exemption to the regulations of 10 CFR 50.46 and 10 CFR 50 Appendix K.

Special circumstances, in accordance with 10 CFR 50.12, paragraph (a)(2)(ii), are present whenever application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.

The underlying purpose of 10 CFR 50.46 is to ensure nuclear power facilities have demonstrated the cooling performance of their ECCS. As discussed above, Topical Report BAW-10227-P-A, Revision 1 concluded the M5TM fuel rod cladding does not alter the effectiveness of the ECCS and also demonstrated the ECCS acceptance criteria applied to reactors fueled with zircaloy clad fuel are also applicable to reactors fueled with M5TM fuel rod cladding.

As currently written, the criteria of 10 CFR 50.46 are not applicable to M5TM, even though analysis shows applying the zircaloy or ZIRLOTM criteria to M5TM fuel yields acceptable ECCS performance results. Application of the regulation in this instance is not necessary to achieve the underlying purpose of the rule; therefore, special circumstances exist.

The underlying purpose of 10 CFR 50, Appendix K, paragraph I.A.5 is to ensure that cladding oxidation and hydrogen generation are appropriately limited during a LOCA and conservatively accounted for in the ECCS evaluation model. Specifically, Appendix K requires the Baker-Just equation be used in the ECCS evaluation model to determine the rate of energy release, cladding oxidation, and hydrogen generation. Topical Report BAW-10227-P-A, Revision 1 demonstrated that the Baker-Just model is conservative in the evaluated post-LOCA scenarios with respect to the use of the M5TM alloy as a fuel rod cladding material. Therefore, the amount of energy release, cladding oxidation, and hydrogen generation for an M5TM alloy reactor core during a LOCA remains conservative for MPS3.

As currently written, the criteria of 10 CFR 50, Appendix K, paragraph I.A.5 are not applicable to M5TM, even though analysis shows applying the zircaloy or ZIRLOTM criteria to M5TM fuel yields acceptable results. Application of the regulation in this instance is not necessary to achieve the underlying purpose of the rule; therefore, special circumstances exist.

Serial No.23-105 Docket No. 50-423 Attachment 8, Page 6 of 10 Based on the above, the underlying purpose of 10 CFR 50.46 and 10 CFR 50, Appendix K will continue to be satisfied for planned MPS3 operation using fuel assemblies with M5TM fuel rod cladding. Therefore, required special circumstances exist, which necessitate the exemption request.

Issuance of an exemption from the specified regulations for the use of M5TM fuel rod cladding material at MPS3 will not compromise the safe operation of the reactor.

Because the underlying purposes of the NRC regulations have been preserved, it is concluded that the proposed exemptions do not present an undue risk to the public health and safety and are consistent with the common defense and security.

4.2. Precedents Similar exemptions have been issued for other licensed reactors, including, but not limited to: Calvert Cliffs Unit 2 (ML030640137, dated April 11, 2003), North Anna Units 1 and 2 (ML032590881, dated September 23, 2003), Crystal River Unit 3 (ML032380538, dated September 26, 2003), Arkansas Nuclear One Unit 1 (ML051790417, dated July 25, 2005), H. B. Robinson Unit 2 (ML11297A103, dated October 31, 2011), Shearon Harris Unit 1 (ML12025A162, dated February 24, 2012),

Millstone Unit 2 (ML15093A497, dated May 12, 2015), and Surry Units 1 and 2 (ML16195A516, dated July 27, 2016).

4.3. No Significant Hazards Consideration Dominion Energy Nuclear Connecticut (DENC) is requesting an exemption from the requirements of 10 CFR 50.46, and 10 CFR 50, Appendix K to support refueling batch use of GAIA fuel assemblies with M5TM fuel cladding material at Millstone Unit 3 (MPS3). The requested exemption is required since the types of cladding material specified in the cited regulations for use in light water reactors do not include M5TM cladding material.

The NRC has provided standards for determining whether a significant hazards consideration exists in 10 CFR 50.92(c). A determination that a proposed exemption involves no significant hazards consideration may be made if operation of the facility in accordance with the proposed exemption would not: 1) involve a significant increase in the probability or consequences of an accident previously evaluated, or

2) create the possibility of a new or different kind of accident from any accident previously evaluated, or 3) involve a significant reduction in a margin of safety. DENC

Serial No.23-105 Docket No. 50-423 Attachment 8, Page 7 of 10 has evaluated if a significant hazards consideration is involved with the proposed exemption request. A discussion of these standards as they relate to this exemption request is provided below:

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

Response: No.

The proposed exemption from the requirements of 10 CFR 50.46, and 10 CFR 50, Appendix K does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structures, or components, or their roles in accident prevention or mitigation. M5TM fuel cladding material is not an accident initiator. The response of the fuel to an accident is analyzed using conservative techniques, and the results are compared to approved acceptance criteria. Reload specific analyses conducted by DENC and Framatome (FRM) shall demonstrate the design limits of the fuel cladding are met.

Station operation and analysis at MPS3 will continue to be compliant with NRC regulations. In the NRC approved Topical Report BAW-10227-P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel", FRM demonstrated the effectiveness of the Emergency Core Cooling System (ECCS) performance will not be affected by a change from zircaloy fuel rod cladding to M5TM fuel rod cladding. The analysis described in the topical report also demonstrated that ECCS acceptance criteria applied to reactors fueled with zircaloy clad fuel are also applicable to reactors fueled with M5TM fuel rod cladding. Thus, the plant will continue to meet applicable design criteria and safety analysis acceptance criteria.

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

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

Response: No.

The proposed exemption from the requirements of 10 CFR 50.46, and 10 CFR 50, Appendix K has no impact on any plant configuration or system performance. The proposed exemption does not modify any interfaces with existing equipment, change the equipments function, or change the method of operating the equipment.

Serial No.23-105 Docket No. 50-423 Attachment 8, Page 8 of 10 The proposed exemption does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structures, or components, or their roles in accident prevention or mitigation. In the NRC approved Topical Report BAW-10227-P-A, Revision 1, FRM demonstrated ECCS performance will not be affected by a change from zircaloy fuel rod cladding to M5TM fuel rod cladding. The analysis described in the topical report also demonstrated that ECCS acceptance criteria applied to reactors fueled with zircaloy clad fuel are also applicable to reactors fueled with M5TM fuel rod cladding. Using approved methods, DENC and FRM shall demonstrate on a cycle specific basis that the fuel assemblies perform within the fuel design limits for MPS3.

The proposed exemption assures there is adequate margin available to meet safety analysis criteria and does not introduce any failure modes, accident initiators, or equipment malfunctions that would cause a new or different kind of accident. Therefore, the proposed exemption does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed exemption from the requirements of 10 CFR 50.46, and 10 CFR 50, Appendix K has no impact on plant configuration or system performance and does not adversely affect any fission product barrier.

Approved methodologies will be used to ensure the plant continues to meet applicable design criteria and safety analysis acceptance criteria. The proposed exemption does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, and the dose analysis acceptance criteria are not affected. The proposed exemption does not result in plant operation in a configuration outside the analyses or design basis and does not adversely affect systems that respond to safely shutdown the plant and maintain the plant in a safe shutdown condition.

In the NRC approved Topical Report BAW-10227-P-A, Revision 1, FRM demonstrated the effectiveness of the ECCS will not be affected by a change from zircaloy fuel rod cladding to M5TM fuel rod cladding. The analysis described in the topical report also demonstrated that ECCS acceptance criteria applied to reactors fueled with zircaloy clad fuel are also applicable

Serial No.23-105 Docket No. 50-423 Attachment 8, Page 9 of 10 to reactors fueled with M5TM fuel rod cladding. Using approved methods, DENC and FRM shall demonstrate on a cycle specific basis that the fuel assemblies perform within the fuel design limits.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, DENC concludes that the proposed change does 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. Conclusion The acceptance criteria and requirements of 10 CFR 50.46 and 10 CFR Part 50, Appendix K currently are limited in applicability to the use of fuel rods with zircaloy or ZIRLOTM cladding. 10 CFR 50.46 and 10 CFR Part 50, Appendix K do not apply to the proposed use of M5TM fuel rod cladding material since M5TM cladding has a slightly different composition than either of the currently specified alloys. With the approval of this exemption request, M5TM cladding may be used for LOCA analyses at MPS3.

In order to transition to GAIA fuel assemblies at MPS3, an exemption from the requirements of 10 CFR 50.46, and 10 CFR 50, Appendix K is requested. As required by 10 CFR 50.12, the requested exemption is authorized by law, does not present undue risk to public health and safety, and is consistent with common defense and security. Approval of this exemption request does not violate the underlying purpose of the rule, and special circumstances exist to justify the approval of an exemption from the subject requirements.

5.0 Environmental Considerations DENC has determined that the requested exemption meets the categorical exclusion provision in 10 CFR 51.22(c)(25), as the requested licensing action is an exemption from the requirements of the Commission's regulations and (i) there is no significant hazards consideration; (ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite; (iii) there is no significant increase in individual or cumulative public or occupational radiation exposure; (iv) there is no significant construction impact; (v) there is no significant

Serial No.23-105 Docket No. 50-423 Attachment 8, Page 10 of 10 increase in the potential for or consequences from radiological accidents; and (vi) the intent of the exemption request is to allow application of 10 CFR 50.46 and 10 CFR 50, Appendix K criteria and requirements to the M5TM fuel rod cladding at MPS3.

Therefore, in accordance with 10 CFR 51.22(b), no environmental assessment or environmental impact statement needs to be prepared in connection with the proposed exemption request.

6.0 References

1. ANP-10342-P-A, Revision 0, GAIA Fuel Assembly Mechanical Design, September 2019.

2. BAW-10227-P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," June 2003.

v t e Letter Sequence Request
EPID:L-2023-LLA-0065, License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits and Exemption Request for Use of M5 Cladding (Open)
Initiation Request Acceptance Administration Withholding Request Acceptance Questions Answers Results Approval Other:
MONTHYEAR2023-05-25 [Table View]

ML23123A279

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