ML20310A324

Proposed License Amendment Request, Addition of Analytical Methodology to the Core Operating Limits Report for a Large Break Loss of Coolant Accident (LBLOCA)

ML20310A324
Person / Time
Site:	Millstone
Issue date:	11/05/2020
From:	Mark D. Sartain Dominion Energy Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
20-333
Download: ML20310A324 (51)

Text

Dominion Energy Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 Dominion Dominion Energy.com Energy November 5, 2020 U.S. Nuclear Regulatory Commission Serial No.20-333 Attention: Document Control Desk NSS&L/TFO: RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 PROPOSED LICENSE AMENDMENT REQUEST ADDITION OF ANALYTICAL METHODOLOGY TO THE CORE OPERATING LIMITS REPORT FOR A LARGE BREAK LOSS OF COOLANT ACCIDENT (LBLOCA)

Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENG) 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 reflect Topical Report WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to*the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)." provides DENC's description and assessment of the proposed change. provides the marked-up MPS3 TS page to reflect the proposed amendment.

There is no associated TS Bases change. Attachment 3 provides a technical evaluation of the application of the Full Spectrum Loss of Coolant Accident (FSLOCA) Evaluation Model to MPS3.

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

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

DENG requests approval of this license amendment request by November 30, 2021, with a 90-day implementation period.

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

Serial No.20-333 Docket No. 50-423 Page 2 of 3 If you have any questions or require additional information, please contact Mr. Shayan Sinha at (804) 273-4687.

Sincerely, Mark D. Sartain 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 Mark D. Sartain, who is Vice President - Nuclear Engineering and Fleet Support of Dominion Energy Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this 5'-fhday of iVc>Ve;wit,.er' , 2020.

My Commission Expires: l~f ~r }.,;,o

--.t---+, -------

CRAIG D SLY Notary Public Commonwealth of Virginia Reg.# 7518653 "ln My Commission Expires December 31, 20::

Attachments:

1. Description and Assessment of Proposed Change

2. Marked-up Technical Specification Page

3. Westinghouse Technical Evaluation of the Application of the FSLOCA Evaluation Model to the Millstone Power Station Unit 3 Commitments made in this letter: None

Serial No.20-333 Docket No. 50-423 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 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.20-333 Docket No. 50-423 Attachment 1 DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGE Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.20-333 Docket No. 50-423 Attachment 1, Page 1 of 11 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 reflect Westinghouse (WEC) Topical Report WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)" [Reference 1]. This reference will replace the existing reference for Topical Report WCAP-12945-P-A, "Code Qualification Document for Best Estimate LOCA Analysis," which is a legacy code qualification document that is no longer used at MPS3.

This attachment provides DEN C's description and assessment of the proposed change. provides the marked-up MPS3 TS page to reflect the proposed amendment.

There is no associated TS Bases change. Attachment 3 provides a technical evaluation of the application of the Full Spectrum Loss of Coolant Accident (FSLOCA) Evaluation Model (EM) to MPS3.

2.0 DETAILED DESCRIPTION 2.1. System Design and Operation The design requirements of existing plant systems, structures, and components (SSCs) are used as inputs to the Large Break Loss of Coolant Accident (LBLOCA) analysis, with appropriate technical conservatisms applied.

Therefore, the LBLOCA analysis does not directly impact the existing design or configuration of any plant SSCs.

2.2. Current Technical Specification Requirement TS 6.9.1.6 requires core operating limits to be established for each reload cycle and contains references to the approved analytical methods used to determine the core operating limits. The TS 6.9.1.6.b Core Operating Limits Report References list includes documents that define the methods used to determine the core operating limits for MPS3. The references of concern under TS 6.9.1.6.b include Reference 5 and 6:

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

Serial No.20-333 Docket No. 50-423 Attachment 1, Page 2 of 11

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.)

Reference 5 in the TS 6.9.1.6.b list is a legacy code qualification document for LBLOCA that is no longer in use at MPS3. Reference 6 in the TS 6.9.1.6.b list is the current method listed for LBLOCA.

2.3. Reason for the Proposed Change In a letter dated November 29, 2012 [Reference 2], DENC stated MPS3 will adopt a Nuclear Regulatory Commission (NRC)-approved LBLOCA analysis method which includes the effects of fuel pellet thermal conductivity degradation (TCD). In this letter, DENC proposed submitting this analysis for NRC review and approval by November 30, 2017. In a subsequent letter dated November 6, 2017 [Reference 3], the planned submittal date was revised to November 30, 2020. The WEC FSLOCA EM provided in WCAP-16996-P-A has been generically approved by the NRC and includes the effects of TCD.

Attachment 3 provides a detailed Westinghouse technical analysis which supports the application of WCAP 16996-P-A to MPS3. This LAR proposes to incorporate this reference in TS 6.9.1.6.b Core Operating Limits Report References, to address the action specified in the November 29, 2012 letter.

2.4. Description of Proposed Changes The TS 6.9.1.6.b Core Operating Limits Report References list methodology documents used to determine the core operating limits for MPS3. The proposed revision to this list of references would replace the current Reference 5:

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

with WCAP-16996-P-A, as follows:

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.)

Serial No.20-333 Docket No. 50-423 Attachment 1, Page 3 of 11 Note that the current Reference 5 in the TS 6.9.1.6.b list is a legacy code qualification document no longer in use at MPS3.

The Loss of Coolant Accident (LOCA) methodologies of Topical Report WCAP-16009-P-A will no longer be used following the implementation of Topical Report WCAP-16996-P-A. WCAP-16009-P-A is being retained as Reference 6 in the TS 6.9.1.6.b reference list, to allow for an orderly transition to FSLOCA with PAD5 in a subsequent reload cycle.

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

3.0 TECHNICAL EVALUATION

3.1 Background

On November 29, 2012, DENC submitted a 10 CFR 50.46 30-day report of Emergency Core Cooling System (ECCS) model changes regarding an evaluation of fuel pellet thermal conductivity with fuel burnup in the Westinghouse Best-Estimate LBLOCA analysis methodology for MPS3 and its effect on peak cladding temperature (PCT) [Reference 2]. Fuel pellet TCD and peaking factor burndown were not explicitly considered in the MPS3 Best Estimate Large Break Loss-of-Coolant Accident (BE LBLOCA) Analysis of Record (AOR). DENC performed an evaluation of PCT for comparison to 10 CFR 50.46 requirements and determined that requirements of 10 CFR 50.46(b)(1) were met but the change was significant and required reporting to the NRC per 10 CFR 50.46(a)(3)(i).

In Reference 2, it was noted that DENC would submit for NRC review and approval a LBLOCA analysis that applies NRG-approved methods that include the effects of fuel TCD. Per Reference 2, this would occur prior to November 30, 2017, but the submittal date was contingent upon receiving:

I

1. Prior NRC approval of a fuel performance analysis methodology that includes the effects of TCD which would replace the current licensing basis methodology in WCAP-15063-P-A, Revision 1, which is referenced in Section 4.2.3.3 of the MPS3 Updated Final Safety Analysis Report (UFSAR) to develop inputs to the LBLOCA EM.

2. Prior NRC approval of a LBLOCA EM that includes the effects of TCD and accommodates the ongoing 10 CFR 50.46(c) rulemaking process and which

Serial No.20-333 Docket No. 50-423 Attachment 1, Page 4 of 11 would replace the current licensing basis methodology in WCAP-16009-P-A which is referenced in Section 15.6.5.2 of the MPS3 UFSAR.

DENG determined that WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)", would be used for the MPS3 LBLOCA analysis. In a letter dated November 6, 2017 [Reference 3], the original planned submittal date of November 30, 2017 was revised to November 30, 2020 as a result of the NRC schedule for WCAP-16996 approval.

To address Items 1 and 2, the WEC LBLOCA EM WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)" has been generically approved by the NRC for WEC 3-loop and 4-loop plants with cold leg ECCS injection [NRC SE in Reference 1]. Since MPS3 is a WEC-designed 4-loop plant with cold leg ECCS injection, the approved method is applicable. The Compliance discussion related to Limitation and Condition 6 contained in Attachment 3 states that the PADS fuel performance code which accounts for the effects of TCD has been used to develop inputs to the MPS3 LBLOCA analysis using the FSLOCA EM. The FSLOCA analysis with inputs developed using PADS accounts for TCD and positions MPS3 to be able to accommodate the ongoing 10 CFR 50.46(c) rulemaking process.

TS 6.9.1.6 requires core operating limits to be established for each reload cycle and contains references to the approved analytical methods used to determine the core operating limits. The TS 6.9.1.6.b Core Operating Limits Report References list includes methodology documents used to determine the core operating limits for MPS3. DENG proposes updating the list of approved methodologies for the COLR in M PS3 TS 6.9 .1 .6. b to reflect WEC Topical Report WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)" [Reference 1], to satisfy the action specified in the November 29, 2012 letter [Reference 2].

The MPS3 COLR is contained in the Technical Requirements Manual '(TRM) as Appendix 8.1. The added reference identifies the analytical methods used to determine core operating limits for the BE LBLOCA event described in the MPS3 UFSAR, Section 15.6.5.2. As described in MPS3 UFSAR Section 15.6.5.2, it must be demonstrated that there is a high level of probability that the limits set forth in 10 CFR 50.46 are met for the event.

Serial No.20-333 Docket No. 50-423 Attachment 1, Page 5 of 11 3.2 Technical Analysis Attachment 3 provides the WEC technical evaluation for the application of the FSLOCA EM to MPS3. This analysis was performed in accordance with the NRC-approved FSLOCA EM in Westinghouse Topical Report WCAP-16996-P-A. The application of the Topical Report to MPS3 only involves the analysis for Region II (LBLOCA), as discussed in Section 1.0 of Attachment 3. This application is a replacement for the existing ASTRUM LBLOCA analysis, which addresses breaks from 1.0 ft2 to two times the area of the RCS cold leg piping.

An LBLOCA analysis (Attachment 3) has been completed for MPS3 with the FSLOCA Methodology, [Reference 1]. The analysis was performed in compliance with the conditions and limitations included in the NRC Safety Evaluation (SE) in WCAP-16996-P-A, Revision 1. The analysis results confirm that MPS3 continues to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.

The small break LOCA (SBLOCA) portion of the break spectrum is currently addressed for MPS3 with the NOTRUMP EM (References 8, 9, and 17 in the TS 6.9.1.6.b list).

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 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

Serial No.20-333 Docket No. 50-423 Attachment 1, Page 6 of 11 available) the system safety function can be accomplished, assuming a single failure.

The ECCS is described in Section 6.3 of the MPS3 UFSAR. Conformance with GDC 35 is described in Section 3.1.2.35 of the MPS3 UFSAR and is unaffected by this change.

10 CFR 50.36 - Technical Specifications 10 CFR 50.36(c)(5) requires Technical Specifications 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 replacement for NRC-approved LOCA methodology will be included in the Administrative Controls section of the MPS3 TS and would be used to determine a core operating limit. The use of the proposed NRC-approved LOCA methodology 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 10 CFR 50.46 includes requirements and acceptance criteria pertaining to the evaluation of post-accident ECCS performance. This regulation includes the requirement that " ... uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated-results can be estimated. This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria ... there is a high level of probability that the criteria would not be exceeded."

The proposed change requests NRC approval to use the FSLOCA methodology described in WCAP-16996-P-A, Revision 1. This methodology is used for performance of the MPS3 LBLOCA analyses, including treatment of uncertainties in the inputs used for the analysis. No change is proposed to the analysis acceptance criteria specified in 10 CFR 50.46. The NRC has reviewed WCAP-16996-P-A, Revision 1 and found it acceptable for referencing in licensing applications for WEC-designed 4-loop plant with cold leg ECCS injection. As demonstrated in Attachment 3, WCAP-16996-P-A, Revision 1 is applicable to MPS3, and the plant-specific application of the FSLOCA

Serial No.20-333 Docket No. 50-423 Attachment 1, Page 7 of 11 methodology to the LOCA analyses have been performed in accordance with the conditions and limitations of the topical report and associated NRC SE. The plant-specific analyses demonstrate that the requirements of 10 CFR 50.46(b) paragraphs (1) through (4) will continue to be met, thus ensuring continued safe plant operation.

NRC Generic Letter (GL) 88-16 NRC GL 88-16, "Removal of Cycle-Specific, Parameter Limits from Technical Specifications," dated October 4, 1988 [Reference 1O], states that it is acceptable for licensees to control reactor physics parameter limits by specifying an NRG-approved calculation methodology. These parameter limits may be removed from the TS and placed in a cycle-specific COLR that 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 NRG-approved analytical methodologies that are used to determine the core operating limits for MPS3. Upon approval of the proposed change, the guidance in the GL continues to be met since the proposed change will continue to specify the NRC approved methodologies used to determine the core operating limits.

Therefore, the application of the FSLOCA EM in WCAP-16996-P-A to MPS3 continues to satisfy the requirements of 10 CFR 50.36, 10 CFR 50.46(b) paragraphs (1) through (4), and NRC GL 88-16. The proposed change meets the current regulatory requirements and does not affect conformance with GDC 35 as described in the MPS3 UFSAR.

4.2. Precedents The proposed change to TS 6.9.1.6.b adds WEC Topical Report WCAP-16996-P-A to the list of approved methodologies for determining core operating limits at MPS3. Numerous previous requests have been approved for methodology reference changes in plant-specific TS COLR reference lists. The NRC has approved the addition of WCAP-16996-P-A to the list of approved methodologies for determining core operating limits for Diablo Canyon

[Reference 4].

LARs requesting a similar change have been approved for Surry Power Station

• Units 1 & 2 [Reference 5] and North Anna Power Station Units 1 & 2

[References 6]. The MPS3 LAR follows a similar format and contains similar content as these submittals. Additionally, the NRC is currently reviewing similar

Serial No.20-333 Docket No. 50-423 Attachment 1, Page 8 of 11 submittals to include the FSLOCA methodology for Watts Bar [Reference 7],

Byron and Braidwood [Reference 8], and V. C. Summer [Reference 9].

4.3. No Significant Hazards Consideration Dominion Energy Nuclear Connecticut, Inc. (DENC) proposes a change to Millstone Power Station Unit 3 (MPS3) Technical Specification (TS) 6.9.1.6.b to add Westinghouse Topical Report WCAP-16996-P-A, "Realistic LOCA [Loss of Coolant Accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," in the list of methodologies approved for reference in the Core Operating Limits Report (COLR), and delete Topical Reports WCAP-12945-P-A, "Code Qualification Document for Best Estimate LOCA Analysis." The added reference identifies the analytical methods used to determine core operating limits for the Large Break Loss of Coolant Accident (LBLOCA) event described in the MPS3 Updated Final Safety Analysis Report (UFSAR), Section 15.6.5.2.

DENC has evaluated whether a significant hazards consideration is involved with the proposed amendment, 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 a Nuclear Regulatory Commission (NRC)-approved methodology for analysis of the LBLOCA to determine if MPS3 continues to meet the applicable design and safety analysis acceptance criteria. The proposed change to the list of NRG-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 LBLOCA analysis demonstrate that MPS3 continues to satisfy the 10 CFR 50.46(b)(1-4), Emergency Core Cooling System (ECCS) performance acceptance criteria using an NRG-approved evaluation model.

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

Serial No.20-333 Docket No. 50-423 Attachment 1, Page 9 of 11

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.

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 UFSAR.

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.

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

Implementation of the proposed license amendment is safe and will have no effect on plant operation. The proposed change will make no physical modifications to equipment or how equipment is operated or maintained.

Based on the above information, DENC concludes that the proposed change does not involve a significant hazards cons.ideration 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 the demonstration that MPS3 continues to meet applicable design criteria and safety analysis acceptance criteria, (2) such activities will be conducted in

Serial No.20-333 Docket No. 50-423 Attachment 1, Page 10 of 11 compliance with the Commission's regulations, and (3) the issuance of the requested license amendment will not be inimical to the common defense and security or to the health and safety of the public.

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

6.0 References

1. Westinghouse Topical Report WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.

2. Letter from J. Alan Price (Dominion Nuclear Connecticut) to USNRC (Serial No.12-705), "Dominion Nuclear Connecticut, Inc., Millstone Power Station Unit 3, 30-Day Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the Requirements of 10 CFR 50.46," November 29, 2012. (ADAMS Accession Number ML12340A010).

3. Letter from M. D. Sartc='in (DENC) to USNRC (Serial No.17-425), "Dominion Nuclear Connecticut, Inc., Millstone Power Station Unit 3, Emergency Core Cooling System (ECCS) Model Change Pursuant to the Requirements of 10 CFR 50.46 Submittal Schedule Date Change," November 6, 2017 (ADAMS Accession Number ML17318A084).

4. Letter from USNRC to James Welsch (Pacific Gas and Electric Company),

"Diablo Canyon Nuclear Plant Units 1 and 2 - Issuance of Amendment Nos. 234 and 236 to Revise Technical Specification 5.6.5b, 'Core Operating Limits Report (COLR),' for Full Spectrum Loss of Coolant Accident Methodology (EPID L-2018-LLA-0730)," January 9, 2020 (ADAMS Accession Number ML19316A109).

Serial No.20-333 Docket No. 50-423 Attachment 1, Page 11 of 11

5. Letter from USNRC to D. G. Stoddard (Virginia Electric and Power Company),

"Surry Power Station Units Nos. 1 and 2, Issuance of Amendment Nos. 300 and 300 to Revise Technical Specifications 6.2.C, "Core Operating Limits Report,"

(EPID L-2019-LLA-0243)," October 28, 2020, (ADAMS Accession Number ML20301A452).

6. Letter from USNRC to D. G. Stoddard (Virginia Electric and Power Company),

"North Anna Power Station Units Nos. 1 and 2 - Issuance of Amendment Nos.

286 and 269 to Revise Technical Specifications to Allow Usage of a Full Spectrum Loss-of-Coolant-Accident (LOCA) Methodology (EPID L-2019-LLA-0236)," October 29, 2020, (ADAMS Accession Number ML20302A179).

7. Letter from James T. Polickoski (Tennessee Valley Authority) to USNRC, "Watts Bar Nuclear Plant, Units 1 and 2, Facility Operating License Nos. NFP-90 and NFP-96, NRC Docket Nos. 50-390 and 50-391, Application to Implement FULL SPECTRUM LOCA (FSLOCA) Methodology for Loss of Coolant Accident (LOCA) Analysis and New LOCA-specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology (WBN-TS-19-04)," January 17, 2020 (ADAMS Accession Number ML20017A338).

8. Letter from Dwi Murray (Exelon Generation Cqmpany) to USNRC, "Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, Application to Revise Technical Specifications 5.6., 'Core Operating Limits Report (COLR),"' February 28, 2020 (ADAMS Accession Number ML20063L282 and ML20063L295).

9. Letter from M. D. Sartain (Dominion Energy South Carolina) to USNRC (Serial No.20-176), "Dominion Energy South Carolina (DESC), Virgil C. Summer Nuclear Station (VCSNS) Unit 1, License Amendment Request, Update of Analytical Method to the Core Operating Limits Report with the Full Spectrum Loss of Coolant Accident Approach," June 4, 2020 (ADAMS Accession Number ML20156A303).

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

Serial No.20-333 Docket No. 50-423 Attachment 2 MARKED-UP TECHNICAL SPECIFICATIONS PAGE Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.20-333 Docket No. 50-423 Attachment 2, Page 1 of 1 Jtily 28, 2916 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)

6.9.1.6.b TI1e analytical methods used to detennine the core operating limits in Specification 6.9.1.6.a shall be those previously reviewed and approved by the NRC and identified below. The CORE OPERATING LIMITS REPORT will contain the complete identification for each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT (i.e.,

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

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

• 2.1.1 Reactor Core Safety Lin1its

• 3.1.1.1.1 SHUTDOWN MARGIN -MODE 1 and 2

• 3.1.1.1.2 SHUTDOWN MARGIN - MODES 3, 4 and 5 Loops Filled

• 3.1.1.2 SHUTDOWN MARGIN - Cold Shutdown - Loops Not Filled

• 3.1.1.3 Moderator Temperature Coefficient

• 3.1.3.5 Shutdown Rod Insertion Limit

• 3.1.3.6 Control Rod Insertion Limits

• 3.2.1.1 AXIAL FLUX DIFFERENCE r:-:W:-:-:C:-:A-::P,---1,....,,6"""9-=-96=---"'"p*....,.A-,- - - - ,

• 3 .2 .2 . 1 Heat Flux Hot Clianne 1 F actor "REALISTIC LOCA

• 3.2.3.1 RCS Total Flow Rate, Nuclear Enthalpy Rise Hot Challllel Factor EVALUATION METHODOLOGY APPLIED

• 3.9.1.1 REFUELING Boron Concentration TO THE FULL SPECTRUM

• 3.2.5 DNB Parameters OF BREAK SIZES (FULL

• 3.3.5 Shutdown Margin Monitor SPECTRUM LOCA METHODOLOGY)," 0flJ_

Proprietary) (Methodology for Specification 3.2.2.1--

Heat Flux Hot Channel 10216-P-A-RlA, "RELAXATION OF CONSTANT AXIAL OFFSET

,_F_a_ct_or.....)'---_ _ _ _ _ _......OL FQ SURVEILLANCE TECHNICAL SPECIFICATION,"

(Ji. Proprietary). (Methodology for Specifications 3.2.1.1--AXIAL FLUX DIFFERENCE and 3.2.2.1--Heat Flux Hot Channel Factor)

5. WGAP 1294§ P A, "GODE QUALIFIGt'.TI~I DOGUME~ff FOR BEST ESTIMATE LOGA ML'\IrYSIS," (R PFeprietary). (Methetlelegy fep Speeifiemieti 3.2.2.l Hem Flun Het Ghlttltlel Faeter.) ~

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

MILLSTONE - UNIT 3 6-20 Amendment No. ,1;4, 3-7-, 60, 69, 8-1-,

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Serial No.20-333 Docket No. 50-423 Attachment 3 WESTINGHOUSE TECHNICAL EVALUATION OF THE APPLICATION OF THE FSLOCA EVALUATION MODEL TO THE MILLSTONE POWER STATION UNIT 3 Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 1 of 33 Westinghouse Non-Proprietary Class 3 APPLICATION OF WESTINGHOUSE FULL SPECTRUM LOCA EVALUATION MODEL TO THE MILLSTONE POWER STATION UNIT 3

1.0 INTRODUCTION

An analysis with the FULL SPECTRUMTM loss-of-coolant accident (FSLOCA'11\1) evaluation model (EM) has been completed for :tvfillstone Power Station Unit 3. This license amendment request (LAR) for

tvfillstone Unit 3 requests approval to apply the Westinghouse FSLOCA EMfor the large-break loss-of-coolant accident (LBLOCA) analysis.

The FSLOCA EM (Reference 1) was developed to address the full spectrum ofloss-of-coolant accidents (LOCAs) which result from a postulated break in the reactor coolant system (RCS) of a pressurized water reactor (PWR). The break sizes considered in the Westinghouse FSLOCA EM include any break size in which break flow is beyond the capacity of the normal charging pumps, up to and including a double-ended guillotine (DEG) rupture of an RCS cold leg with a break flow area equal to two times the pipe area, including what traditionally are defined as Small and Large Break LOCAs.

The break size spectrum is divided into two regions. Region I includes breaks that are typically defined as Small Break LOCAs (SBLOCAs). Region II includes break sizes that are typically defined as LBLOCAs.

Only the Region II (LBLOCA) analysis was performed for this application of the FSLOCA EM. A Region I (SBLOCA) analysis was not pe1f01med.

The FSLOCA EM explicitly considers the effects of fuel pellet the1mal conductivity degradation (TCD) and other burnup-related effects by initializing to fuel rod petf01mance data input generated by the PADS code (Reference 2), which explicitly models TCD and is benchmarked to high burnup data in Reference

2. The fuel pellet thermal conductivity model in the WCOBRNTRAC-TF2 code used in the FSLOCA EM explicitly accounts for pellet TCD.

Tlu-ee of the Title 10 of the Code ofFederal Regulations (CFR) 50.46 criteria (peak cladding temperature (PCT), maximum local oxidation (:tvlLO), and core-wide oxidation (CWO)) are considered directly in the FSLOCA EM. A high probability statement is developed for the PCT, :tvlLO, and CWO that is needed to demonstrate compliance with 10 CFR 50.46 acceptance criteria (b)(l), (b)(2), and (b)(3) (Reference 3) via statistical methods. The lVILO is defined as the sum of pre-transient corrosion and transient oxidation consistent with tl1e position in Information Notice 98-29 (Reference 4). The coolable geometry acceptance criterion, 10 CFR 50.46 (b)(4), is assured by compliance with acceptance criteria (b)(l),

(b)(2), and (b)(3), and demonstrating that fuel assembly grid deformation due to combined seismic and LOCA loads does not extend to the in-board fuel assemblies such that a coolable geometry is maintained.

The FSLOCA EM has been generically approved by the Nuclear Regulato1y Commission (NRC) for Westinghouse 3-loop and 4-loop plants with cold leg Emergency Core Cooling System (ECCS) injection (Reference 1). Since Millstone Unit 3 is a Westinghouse designed 4-loop plant with cold leg ECCS injection, the approved method is applicable. Infonnation required to address Lintltations and Conditions 9 and 10 of the NRC 's Safety Evaluation Report (SER) for Reference 1 was docketed in Reference 12 in suppo1t of application of the FSLOCA EM to Westinghouse 4-loop plants.

FULL SPECTRUlvl, FSLOCA, and OptimizedZIRLO are trademarks of\Vestinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 2 of 33 Westinghouse Non-Proprietary Class 3 This report summarizes the application of the Westinghouse FSLOCA EM to lvlillstone Unit 3. The application of the FSLOCA EM to lvlillstone Unit 3 is consistent with the NRC-approved methodology (Reference 1), with exceptions identified under Llmitation and Condition Number 2 in Section 2.3. A consistency review for the application of the FSLOCA EM to lvlillstone Unit 3 with the conditions and limitations as identified in the NRC's SER for Reference 1 is documented in Section 2.3.

Both Dominion Energy and its analysis vendor (Westinghouse) have interface processes which identify plant configuration changes potentially impacting safety analyses. TI1ese inte1face processes, along with Westinghouse internal processes for assessing EM changes and errors, are used to identify the need for LOCA analysis impact assessments.

The major plant parameter and analysis assumptions used in the lvlillstone Unit 3 analysis with the FSLOCA EM are provided in Tables 1 through 4.

2.0 METHOD OF ANALYSIS 2.1 FULL SPECTRUM LOCA Evaluation Model Development In 1988, the NRC Staff amended the requirements of 10 CFR 50.46 (Reference 3 and Reference 6) and Appendix K, "ECCS Evaluation Models," to permit the use of a realistic EM to analyze the performance of the ECCS during a hypothetical LOCA. Westinghouse's previously approved best-estimate LBLOCA EM is discussed in Reference 7. TI1e EM is referred to as the Automated Statistical Treatment of Unce1iainty Method (ASTRUlvl), and was developed following Regulatory Guide (RG) 1.157 (Reference 8).

When the FSLOCA EM was being developed, the NRC issued RG 1.203 (Reference 9) which expands on the principles ofRG 1.157, while providing a more systematic approach to the development and assessment process of a PWR accident and safety analysis EM. Therefore, the development of the FSLOCA EM followed the Evaluation Model Development and Assessment Process (EMDAP), which is documented in RG 1.203. While RG 1.203 expands upon RG 1.157, there are certain aspects ofRG 1.157 which are more detailed than RG 1.203; therefore, both RGs were used for the development of the FSLOCAEM.

2.2 WCOBRA/TRAC-TF2 Computer Code The FSLOCA EM (Reference 1) uses the WCOBRA/TRAC-TF2 code to analyze the system thennal-hydraulic response for the full spectrnm of break sizes. WCOBRAITRAC-TF2 was created by combining a lD module (TRAC-P) with a 3D module (based on Westinghouse modified COBRA-TF). TI1e lD and 3D modules include an explicit non-condensable gas transport equation. TI1e use of TRAC-P allows for the extension of a two-fluid, six-equation formulation of the two-phase flow to the lD loop components.

Titls new code is WCOBRA/TRAC-TF2, where "TF2" is an identifier that reflects the use of a three-field (TF) formulation of the 3D module derived by COBRA-TF and a two-fluid (TF) formulation of the lD module based on TRAC-P.

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 3 of 33 Westinghouse Non-Proprietary Class 3 This best-estimate computer code contains the following features:

1. Ability to model transient three-dimensional flows in different geometries inside the reactor vessel

2. Ability to model thermal and mechanical non-equilib1ium between phases

3. Ability to mechanistically represent interfacial heat, mass, and momentum transfer in different flow regimes

4. Ability to represent impottant reactor and plant components such as fuel rods, steam generators (SGs ), reactor coolant pumps (RCPs ), etc.

A detailed assessment of the computer code WCOBRA/TRAC-1F2 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 for a LOCA. Modeling of a LOCA introduces additional unce1tainties which are identified and quantified in the plant-specific analysis. The reactor vessel and loop noding scheme used in the FSLOCA EM is consistent with the noding scheme used for the experiment simulations that form the validation basis for the physical models in the code. Such noding choices have been justified by assessing the model against large and full scale experiments.

2.3 Compliance witlt FSLOCA EM Limitations and Conditions The NRC's SER for Reference 1 contains 15 limitations and conditions on the NRC-approved FSLOCA EM. A summary of each limitation and condition and how it was met is provided below.

Limitation and Condition Number 1 Summary The FSLOCA EM is not approved to demonstrate compliance with 10 CFR 50.46 acceptance criterion (b)(5) related to the long-te1m cooling.

Compliance The analysis for Millstone Unit 3 with the FSLOCA EM is only being used to demonstrate compliance with 10 CFR 50.46 (b)(l) through (b)(4).

Limitation and Condition Number 2 Smmnary The FSLOCA EM is approved for the analysis of Westinghouse-designed 3-loop and 4-loop PWRs with cold-side injection. Analyses should be executed consistent with the approved method, or any deviations from the approved method should be described and justified.

Compliance JVfillstone Unit 3 is a Westinghouse-designed 4-loop PWR with cold-side injection, so it is within the NRC-approved methodology. The analysis for Millstone Unit 3 utilized the NRC-approved FSLOCA methodology, except for the changes which were previously transmitted to the NRC pursuant to 10 CFR 50.46 in LTR-NRC-18-30 (Reference 5) and LTR-NRC-19-6 (Reference 13),justified therein, and with the exception of only including an analysis for Region II, which is justified since the Region I and Region Il analyses are pe1fo1med independently such that the Region I and Region Il analyses are separable.

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 4 of 33 Westinghouse Non-Proprietruy Class 3 Limitation and Condition Number 3 Summary For Region II, the containment pressure calculation will be executed in a manner consistent with the approved methodology (i.e., the COCO or LOTIC2 model will be based on appropriate plant-specific design parameters and conditions, and engineered safety features which can_reduce pressure are modeled). This includes utilizing a plant-specific initial containment temperature, and only taking credit for containment coatings which are qualified and outside of the break zone-of-influence.

Compliance The containment pressure calculation for the Millstone Unit 3 analysis was performed consistent with the NRC-approved methodology. Appropriate design parameters and conditions were modeled, as were the engineered safety features which can reduce the containment pressure. A plant-specific initial temperature associated with normal full-power operating conditions was modeled, and no coatings were credited on any of the containment structures.

Limitation and Condition Number 4 Summary The decay heat uncertainty multiplier will be sampled consistent with the NRC-approved methodology for the FSLOCA EM. The analysis simulations for the FSLOCA EM will not be executed for1onger than 10,000 seconds following reactor trip unless the decay heat model is appropriately justified. The sampled values of the decay heat unce1tainty multiplier for the cases which produced the Region I and Region II analysis results will be provided in the analysis submittal in units of sigma and absolute units.

Compliance Consistent with the NRC-approved methodology, the decay heat uncertainty multiplier was sampled consistent with the restdctions of the NRC-approved methodology for the FSLOCA EM. The analysis simulations were all executed for no longer than 10,000 seconds following reactor tdp. The sampled values of the decay heat unce1tainty multiplier for the cases which produced the Region II analysis results have been provided in units of sigma and approximate absolute units in Table 7.

Limitation and Condition Number 5 Summary The maximum assembly and rod length-average bumup must remain below the limits contained in the NRC-approved methodology for the FSLOCA EM.

Compliance The maximum analyzed assembly and rod length-average bumups for Millstone Unit 3 analysis are less than or equal to the litnits contained in the NRC-approved methodology for the FSLOCA EM.

Lintitation and Condition Number 6 Summary The fuel performance data for analyses with the FSLOCA EM should be based on the PADS code (at present), which includes the effect ofthetmal conductivity degradation. The nominal fuel pellet average

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 5 of 33 Westinghouse Non-Proprietary Class 3 temperatures and rod internal pressures should be the maximum values, and the generation of all the PADS fuel petformance data should adhere to the NRC-approved PADS methodology.

Compliance PADS fuel petformance data were utilized in the Millstone Unit 3 analysis with the FSLOCA EM. The analyzed fuel pellet average temperatures bound the maximum values calculated in accordance with Section 7.5.1 of Reference 2, and the analyzed rod internal pressures were calculated in accordance with Section 7.5.2 ofReference 2.

Limitation and Condition Number 7 Summary The YDRAG uncertainty parameter for Region I analyses should be treated as specified in the NRC-approved methodology for the FSLOCA EM.

Compliance A Region I uncertainty analysis was not performed in this application of the FSLOCA EM, so this Limitation and Condition is not applicable.

Limitation and Condition Number 8 Summary Ce1tain uncertainty contributors will be treated for Region I analyses as specified in the NRC-approved methodology for the FSLOCA EM.

Compliance A Region I uncertainty analysis was not performed in this application of the FSLOCA EM:, so this Limitation and Condition is not applicable.

Limitation and Condition Number 9 Summary For PWR designs which are not Westinghouse 3-loop P,VRs, a sensitivity study will be executed to confirm modeling selections for Region I analyses.

Compliance Information addressing Limitation and Condition 9 of the NRC 's SER for Reference 1 was docketed in Reference 12 in support of application of the FSLOCA EM to Westinghouse 4-loop plants.

Limitation and Condition Number 10 Summary For PWR designs which are not \Vestinghouse 3-loop P ~ , a sensitivity study will be executed to demonstrate that the applied break size boundary for Region I analyses serves the intended goal.

Additionally, the minimum sampled break area for the analysis of Region II should be 1 :ft:2.

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 6 of 33 Westinghouse Non-Proprietary Class 3 Compliance Information addressing Limitation and Condition 10 of the NRC's SER for Reference 1 was docketed in Reference 12 in support of application of the FSLOCA EM to Westinghouse 4-loop plants.

Limitation and Condition Number 11 Summary There are various aspects of this Limitation and Condition, which are summarized below:

1. Certain infotmation regarding the Region I and Region II analyses must be declared and documented prior to performing the uncertainty analysis, and will not be changed throughout the remainder of the analysis once they have been declared and documented.

2. If the analysis inputs are changed after they have been declared and documented, for the intended purpose of demonstrating compliance with the applicable acceptance criteria, then the changes and associated rationale for the changes will be provided in the analysis submittal. Additionally, the preliminary values for PCT, MLO, and CWO which caused the input changes will be provided. These preliminary values are not subject to Appendix B verification, and archival of the supporting information for these preliminary values is not required.

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

Compliance This Limitation and Condition was met for the Millstone Unit 3 analysis as follows:

1. The information specified in the NRC-approved methodology for the FSLOCA EM was declared and documented prior to analysis execution, and was not changed after it was declared and documented.

2. The analysis inputs were not changed once they were declared and documented.

3. The plant operating ranges which were sampled within the uncertainty analyses are provided for Millstone Unit 3 in Table 1.

Limitation and Condition Number 12 Summary The plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves must be adequately accounted for in analysis with the FSLOCA EM.

Compliance This Limitation and Condition applies to Region I (small break) transients. Region I calculations were not pe1fo1med for Millstone Unit 3; as such, the first stage main steam safety valve setpoint was used as a representative basis for the main steam safety valve setpoint calculation for the Region II analysis.

Furthe1more, Region II transients do not result in secondary side pressurization such that the MSSV setpoint pressures would be reached.

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 7 of 33 Westinghouse Non-Proprietary Class 3 Limitation and Condition Number 13 Summary In plant-specific models for analysis with the FSLOCA EM, specific modeling considerations for the upper head spray nozzles should be followed as required by the NRC-approved methodology.

Compliance These specific modeling requirements for the upper head spray nozzles were adhered to for the Millstone Unit 3 Region II analysis.

Limitation and Condition Number 14 Summary For analyses with the FSLOCA EM to demonstrate compliance against the current 10 CFR 50.46 oxidation criterion, the transient time-at-temperature will be converted to an equivalent cladding reacted (ECR) using either the Baker-Just or the Cathcati-Pawel correlation. In either case, the pre-transient corrosion will be summed with the LOCA transient oxidation. If the Cathcart-Pawel correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 13 percent limit. If the Baker-Just correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 17 percent limit.

Compliance For the Ivlillstone Unit 3 analysis, the Baker-Just correlation was used to convert the LOCA transient time-at-temperature to an ECR. The resulting LOCA transient ECR was then summed with the pre-existing corrosion for comparison against the 10 CFR 50.46 local oxidation acceptance critetion of 17 percent.

Limitation and Condition Number 15 Summmy The Region II analysis will be executed twice; once assuming loss-of-offsite power (LOOP) and once assuming offsite power available (OPA). The results from both analysis executions should be shown to be in compliance with the 10 CFR 50.46 acceptance c1ite1ia.

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

Compliance The Region II unce1tainty analysis for Jvfillstone Unit 3 was perfonned twice; once assuming a LOOP and once assuming OP A. The results from both analyses that were perfmmed are in compliance with the 10 CFR 50.46 acceptance ctiteria (see Section 4.0).

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

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 8 of 33 Westinghouse Non-Proprietary Class 3 3.0 REGION II ANALYSIS 3.1 Desc1iption of Representative Transient A large-break LOCA transient can be divided into phases in which specific phenomena are occuning. A convenient way to divide the transient is in terms of the various heatup and cool down phases that the fuel assemblies undergo. For each of these phases, specific phenomena and heat transfer regimes are impotiant, as discussed below.

Blow down - Critical Heat Flux (CHF) Phase In this phase, the break flow is subcooled, the discharge rate of coolant from the break is high, the core flow reverses, the fuel rods go through departure from nucleate boiling (DNB), and the cladding rapidly heats up and the reactor is shut down due to the core voiding.

The regions of the RCS with the highest initial temperatures (upper core, upper plenum, and hot legs) begin to flash during this period. This phase is tenninated when the water in the lower plenum and downcomer begins to flash. The mixture level swells and a saturated mixture is pushed into the core by the intact loop RCPs, still rotating in single-phase liquid. As the fluid in the cold leg reaches saturation conditions, the discharge flow rate at the break decreases significantly.

Blowdown - Upward Core Flow Phase Heat transfer is increased as the two-phase mixture is pushed into the core. The break discharge rate is reduced because the fluid becomes saturated at the break. This phase ends as the lower plenum mass is depleted, the fluid in the loops become two-phase, and the RCP head degrades.

Blowdown - Downward Core Flow Phase The break flow begins to dominate and pulls flow down through the core as the RCP head degrades due to increased voiding, while liquid and entrained liquid flows also provide core cooling. Heat transfer in this period may be enhanced by liquid flow from the upper head. Once the system has depressurized to less than the accumulator cover pressure, the accumulators begin to inject cold water into the cold legs.

During this pedod, due to steam upflow in the downcomer, a portion of the injected ECCS water is bypassed around the downcomer and sent out through the break. As the system pressure continues to decrease, the break flow and consequently the downward core flow are reduced. The system pressure approaches the containment pressure at the end of this last pe1iod of the blowdown phase.

Duling this phase, the core begins to heat up as the system approaches containment pressure, and the phase ends when the reactor vessel begins to refill with ECCS water.

Refill Phase The core continues to heat up as the lower plenum refills with ECCS water. This phase is charactedzed by a rapid increase in fuel cladding temperature at all elevations due to the lack of liquid and steam flow in the core region. The water completely refills the lower plenum and the refill phase ends. As ECCS water enters the core, the fuel rods in the lower core region begin to quench and liquid entrainment begins, resulting in increased fuel rod heat transfer.

Reflood Phase During the early reflood phase, the accumulators begin to empty and nitrogen is discharged into the RCS.

The nitrogen surge forces water into the core, which is then evaporated, causing system re-pressudzation

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 9 of 33 Westinghouse Non-Proprietary Class 3 and a temporary reduction of pumped ECCS flow; this re-pressurization is illustrated by the increase in RCS pressure. During this time, core cooling may increase due to vapor generation and liquid entrainment, but conversely the early reflood pressure spike results in loss of mass out through the broken cold leg.

The pumped ECCS water aids in the filling of the downcomer tlu*oughout the reflood period. As the quench front progresses further into the core; the PCT elevation moves increasingly higher in the fuel assembly.

As the transient progresses, continued injection of pumped ECCS water refloods the core, effectively removes the reactor vessel metal mass stored energy and core decay heat, and leads to an increase in the reactor vessel fluid mass. Eventually the core invent01y increases enough that liquid entrainment is able to quench all the fuel assemblies in the core.

3.2 Analysis Results The Millstone Unit 3 Region II analysis was performed in accordance with the NRC-approved methodology in Reference 1, with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The analysis was pe1formed assuming both LOOP and OPA, and the results of both of the LOOP and OPA analyses are compared to the 10 CFR 50.46 acceptance criteria. The most limiting ECCS single failure of one ECCS train is assumed in the analysis as identified in Table 1. 111e results of the Millstone Unit 3 Region II LOOP and OPA uncertainty analyses are summarized in Table 5. The sampled decay heat uncertainty multipliers for the Region II analysis cases are provided in Table 7.

Table 6 contains a sequence of events for the transient ( OPA offsite power assumption) that produced the more limiting analysis PCT result relative to the offsite power assumption. Figures 1 through 15 illustrate the key system thennal hydraulic parameters for this transient.

The containment pressure is calculated for each LOCA transient in the analysis using the COCO code (References 10 and 11). The COCO containment code is integrated into the WCOBRA/TRAC-1F2 thermal-hydraulic code. The transient-specific mass and energy releases calculated by the thermal-hydraulic code at the end of each timestep are transferred to COCO. COCO then calculates the containment pressure based on the containment model (the inputs are summarized in Tables 2 and 3) and the mass and energy releases, and transfers the pressure back to the the1mal-hydraulic code as a boundary condition at the break, consistent with the methodology in Reference 1. The containment model for COCO calculates a conse1vatively low containment pressure, including the effects of all the installed pressure reducing systems and processes such as assuming that all trains of containment spray are operable. The containment backpressure for the transient that produced the analysis PCT result is provided in Figure 9.

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 10 of 33 Westinghouse Non-Proprietary Class 3 4.0 COMPLIANCE WITH 10 CFR 50.46 It must be demonstrated that there is a high level of probability that the following criteria in 10 CFR 50.46 are met:

(b)(l) The analysis PCT corresponds to a bounding estimate of the 95tt* percentile PCT at the 95-percent confidence level. Since the resulting PCT is less than 2,200°F, the analysis with the FSLOCA EM confums that 10 CFR 50.46 acceptance criterion (b)(l), i.e., "Peak Cladding Temperature does not exceed 2,200°F," is demonstrated.

The results are shown in Table 5 for Millstone Unit 3.

(b)(2) The analysis MLO corresponds to a bounding estimate of the 95 th percentile MLO at the 95-percent confidence level. Since the resulting MLO is less than 17 percent when converting the time-at-temperature to an equivalent cladding reacted using the Baker-Just correlation and adding the pre-transient co1rnsion, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(2), i.e., "Maximum Local Oxidation of the cladding does not exceed 17 percent,"

is demonstrated.

The results are shown in Table 5 for Millstone Unit 3.

(b)(3) The analysis CWO corresponds to a bounding estimate of the 95thpercentile CWO at the 95-percent confidence level. Since the resulting CWO is less than 1 percent, the analysis confirms that 10 CFR 50.46 acceptance critetion (b)(3), i.e., "Core-Wide Oxidation does not exceed 1 percent," is demonstrated.

The results are shown in Table 5 -for Jv1illstone Unit 3.

(b)(4) 10 CFR 50.46 acceptance critetion (b)(4) requires that the calculated changes in core geometry are such that the core remains in a coolable geometry.

This criterion is met by demonstrating compliance with cdteria (b)(l), (b)(2), and (b)(3), and by assuring that fuel assembly gdd defotmation due to combined LOCA and seismic loads is specifically addressed. Critetia (b)(l), (b)(2), and (b)(3) have been met for Millstone Unit 3 as shown in Table 5.

It is discussed in Section 32.1 of the NRC-approved FSLOCA EM (Reference 1) that the effects ofLOCA and seismic loads on the core geometry do not need to be considered unless fuel assembly gdd deformation extends to inboard assemblies beyond the core periphety (i.e.,

deformation in a fuel assembly with no sides adjacent to the core baffle plates). Inboard gdd defo1mation due to combined LOCA and seismic loads is not calculated to occur for Millstone Unit 3.

(b)(5) 10 CFR 50.46 acceptance cdterion (b)(5) requires that long-te1m core cooling be provided following the successful initial operation of the ECCS.

Long-term cooling is dependent on the demonstration of the continued delivery of cooling water to the core. The actions that are currently in place to maintain long-term cooling are not impacted by the application of the NRC-approved FSLOCA EM (Reference 1).

Based on the analysis results for Region II presented in Table 5 for Millstone Unit 3, it is concluded that Millstone Unit 3 complies with the cdtetia in 10 CFR 50.46.

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 11 of 33 Westinghouse Non-Proprietary Class 3

5.0 REFERENCES

1. "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMLOCA Methodology)," WCAP-16996-P-A, Revision 1, November 2016.

2. "Westinghouse Performance Analysis and Design Model (PADS)," WCAP-17642-P-A, Revision 1, November 2017.

3. "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume 39, Number 3, Janumy 1974.

4. "Information Notice 98-29: Predicted fucrease in Fuel Rod Cladding Oxidation," USNRC, August 1998.

5. "U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Rep01iing for 2017," LTR-NRC-18-30, July 2018.

6. "Emergency Core Cooling Systems: Revisions to Acceptance Criteria," Federal Register, V53, N180, pp. 35996-36005, September 1988.

7. "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment OfUncerfainty Method (ASTRUM)," WCAP-16009-P-A, January 2005.

8. "Best Estimate Calculations of Emergency Core Cooling System Performance," Regulatory Guide 1.157, USNRC, May 1989.

9. "Transient and Accident Analysis Methods," Regulatory Guide 1.203, USNRC, December 2005.

10. "Westinghouse Emergency Core Cooling System Evaluation Model - Summmy," WCAP-8339, June 1974.

11. "Containment Pressure Analysis Code (COCO)," WCAP-8327, June 1974.

12. '"Inf01mation to Satisfy the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology Plant Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs)' (Proprietmy/Non-Proprietary)," LTR-NRC-18-50, July 2018.

13. "U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2018," LTR-NRC-19-6, February 2019.

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 12 of 33 Westinghouse Non-Proprietary Class 3 Table 1. Plant Operating Range Analyzed and Key Parameters for Millstone Unit 3 Parameter As-Analyzed Value or Range 1.0 Core Parameters a) Core power :S 3723 MWt +/- 0% Uncertainty b) Fuel type 17x17RFA-2, OptimizedZIRLOTh' Cladding with Intermediate Flow :Mixers (IFMs), Integral Fuel Burnable Absorber (IFBA) c) Maximum total core peaking factor (FQ), 2.60 including uncertainties d) Maximum hot channel enthalpy rise 1.65 peaking factor (F MI), including uncertainties e) Axial flux difference (AFD) band at 100% -12%to +9%

power f) Maximum transient operation fraction  :<;0,5 2.0 Reactor Coolant System Parameters a) Thermal design flow (TDF) 90,800 gpm/loop b) Vessel average temperature (TAvo) 576.5°F :S TAvo:S 594.5°F c) Pressurizer pressure (PRcs) 2203.2 psia :S PRcs :S 2296.8 psia d) Reactor coolant pump (RCP) model and Model 93Al, 7000 hp power 3.0 Containment Parameters a) Containment modeling Calculated for each transient using transient-specific mass and ene1gy releases and the infonnation in Tables 2 and3 4.0 Steam Generator (SG) and Secondary Side Parameters a) Steam generator tube plugging level  :<;10%

b) 11fain feedwater temperature Nominal (418.95°F) c) Auxiliary feedwater temperature Nominal (120°F) d) Auxiliary feedwater flow rate 61.625 gpm/SG

.,. This record was final approved on 8/4/2020 1:27:07 PM. (This statement was added by the PRIME system upon its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 13 of 33 Westinghouse Non-Proprietary Class 3 Table 1. Plant Operating Range Analyzed and Key Parameters for Millstone Unit 3 Parameter As-Analyzed Value or Range 5.0 Safety Injection (SI) Parameters a) Single failure configuration ECCS: Loss of one train of pumped ECCS Region II containment pressure: All containment spray trains are available.

b) Safety injection temperature (Tsi) 40.0°F S Tsr S 75.0°F c) Low pressurizer pressure safety injection 1700 psia safety analysis limit d) Initiation delay time from low pressutizer S 45 seconds (OPA or LOOP) pressure SI setpoint to full SI flow e) Safety injection flow Minimum flows in Table 4 (Region II) 6.0 Accumulator Parameters a) Accumulator temperature (TAcc) 67°F S TAcc S 84°F b) Accumulator water volume (VAce) 884.S ft3 SVAcc S939.S ft' c) Accumulator pressure (PAce) 626 psia SPAce S 704 psia d) Accumulator boron concentration ~2600 ppm 7.0 Reactor Protection System Parameters a) Low pressutizer pressure reactor hip 1860 psia setpoint

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 14 of 33 Westinghouse Non-Proprietary Class 3 Table 2. Containment Data Used for Region II Calculation of Containment Pressure for Millstone Unit3 Parameter Value Maximum containment net free volume 2,350,000 ff Minimum initial containment temperature at full power operation g9op Refueling water storage tank (RWST) temperature for containment spray 40°F '.S TRWST '.S 75°F (TRwsr)

Minimum RWST temperature for broken loop spilling SI 40°F Minimum containment outside air I ground temperature -20°F (air) 55°F (ground)

Minimum initial containment pressure at nmmal full power operation 10.4 psia Minimum containment spray pump initiation delay from containment high 2 26.3 seconds (OPA) pressure signal time 2 37.3 seconds (LOOP)

Maximum containment spray flow rate from all pumps 6500 gpm Maximum number of containment fan coolers in operation dming LOCA 0 transient Maximum number of containment venting lines (including purge lines, 0 pressure relieflines or any others) which can be OPEN at onset of transient at full power operation Containment walls/ heat sink properties Table 3 SI spilling flows 274.83 lbm/sec

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 15 of 33 Westinghouse Non-Proprietary Class 3 Table 3. Containment Heat Sink Data Used for Region II Calculation of Containment Pressure for Millstone Unit 3 Wall Area (ft2) Thickness (ft) Material 1 866 0.0375, 2.16 Stainless Steel, Concrete 2 7674 0.0375, 1.50 Stainless Steel, Concrete 3 133277 1.36 Concrete 4 17926 2.11 Concrete 5 6563 3.00 Concrete 6 2007 1.75 Concrete 7 12269 2.0, 0.25, 10.00 Concrete, Carbon Steel, Concrete 8 24675 0.0428, 4.5 Carbon Steel, Concrete 9 38493 0.0428, 4.5 Carbon Steel, Concrete 10 34100 0.0462, 2.56 Carbon Steel, Concrete 11 1722 0.1075 Stainless Steel 12 552 0.0592 Carbon Steel 13 13230 0.02 Stainless Steel 14 2063 0.0548 Stainless Steel 15 8966 0.0231 Carbon Steel 16 1282 0.0825 Carbon Steel 17 514279 0.0182 Carbon Steel 18 182517 0,00925 Carbon Steel 19 11033 0.0304 Stainless Steel 20 37068 0.0651 Carbon Steel 21 21000 0.0119 Stainless Steel 22 866 2.16 Concrete 23 7674 1.50 Concrete

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 16 of 33 Westinghouse Non-Proprietary Class 3 Table 4. Safety Injection Flows (Total in Intact Loops) Used for Region II Calculation for Millstone Unit 3 Safety Injection High Charging System Residual Heat Removal Pressure/Head (SIR) /

(CHS) I High Head (RHR) / Low Head Pressure (psia) Intermediate Head Safety Injection Safety Injection (LHSI)

Safety Injection (IHSI)

(HHSI) Flow (gpm) Flow (gpm)

Flow (gpm) 14.7 281.58 402.23 2596.92 34.7 279.585 399 2061.88 54.7 277.59 395.77 1502.615 74.7 275.595 392.54 1011.75 94.7 273.6 389.31 623.2 114.7 271.605 386.08 119.415 114.71 271.605 386.08 0.0 134.7 269.61 382.28 154.7 267.615 378.48 174.7 265.62 374.585 194.7 263.625 370.785 214.7 261.63 366.985 314.7 250.8 346.75 414.7 239.97 325.47 514.7 228.855 303.24 614.7 217.645 280.06 714.7 206.055 254.41 814.7 194.18 227.145 914.7 181.925 192.85 1014.7 169.385 152.95 1114.7 156.56 105.83 1214.7 142.595 42.085 1214.71 142.595 0.0 1314.7 127.775 1414.7 112.29 1514.7 95.95

-This record was final approved on 8/4/2020 1 :27:07 PM. (This statement was added by the PRIME system upon its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 17 of 33 Westinghouse Non-Proprietary Class 3 Table 4. Safety Injection Flows (Total in Intact Loops) Used for Region II Calculation for Millstone Unit 3 Safety Injection High Charging System Residual Heat Removal Pressure/Head (SIH) /

(CHS) / High Head (RHR) I Low Head Pressure (psia) Intehuediate Head Safety Injection Safety Injection (LHSI)

Safety Injection (IHSI)

(HHSI) Flow (gpm) Flow (gpn1)

Flow(gpm) 1614.7 79.135 1714.7 61.56 1814.7 41.325 1914.7 15.485 2014.7 0.0 Table 5. Millstone Unit 3 Analysis Results with the FSLOCA EM Outcome Region II Value (OPA) Region II Value (LOOP) 95/95 PCT 1712°F 1661°F 95/95 M[,0 7.82% 7.81%

95/95 cwo 0.12% 0.08%

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 18 of 33 Westinghouse Non-Proprietary Class 3 Table 6. Millstone Unit 3 Sequence of Events for the Region II Analysis PCT Case (OPA)

Event Time after Break (sec)

Start of Transient 0.0 Safety Iajection Signal 6.6 Accumulator Injection Begins 16.0 End ofBlowdown 29.0 Fuel Rod Burst Occurs 30.8 Bottom of Core Recovery ~40.0 Safety Injection Begins 51.6 Accumulator Empty 62.5 PCT Occurs 121 All Rods Quenched 246 End of Analysis Time 600 Table 7. Millstone Unit 3 Sampled Value of Decay Heat Uncertainty Multiplier, DECAY_HT, for Region II Analysis Cases Region Case DECAY_HT (units of a) DECAY_HT (absolute units)*

PCT +0.8651<1 4.17%

Region II MLO +1.3603cr 7.02%

(OPA) cwo +0.4502cr 2.16%

PCT +0.6203a 3.04%

Region II Ivfl.O + 1.3452cr 6.94%

(LOOP) cwo +1.9041 cr 9.25%

• Approximate uncertainty in total decay heat power at 1 second after shutdown as defined by the ANSJ/ ANS-5.1-1979 decay heat standard for 235 U, 239Pu, and 238U assuming infinite operation.

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 19 of 33 Westinghouse N on-Ptoprietaty Class 3 PCT 6 0 0 Dummy Rod 1 PCT 7 0 0 Durnm~ Rod 2 PCT 1 0 0 Hot .o d PCT 2 0 0 Hot Assembly l'CT 3 0 0 Averoge Rod 1 PCT 4 0 0 Averoge Rod 2 PCT 5 0 0 Low Power Rod 1aoo-,--------------------------,

1600 r;:-

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~

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0 100 200 300 400 500 600 700 Time After Break (sec) 619153357 Figure 1: Millstone Unit 3 Peak Cladding TBnp irature for all Rods for the Region II Analysis PCT Case Assuming OPA

"' THs record 'Mis firal approved on 8/4120201 :27:07 PM. (This statement 'MIS added by the PRIME system lf)On Its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 20 of 33 Westinghouse Non-Proprietary Class 3 PCJ-LOC 0 0 PEAK CLAD TEMP LOC, 12-----------------------------,

10 8

-,1

::'.

C

.Q 0

Q) w 4

2 100 200 300 400 500 600 700 Time After Break (sec) 8l91WS7 Figure 2: Millstone Unit3 Peak CladdingTimp O'ature Elevatio.n (Relative to Bottom ofActive Fuel) for the Region II Analysis PCT Case Assuming OPA

"' TH s record 'MlS firal approved on 8/4120201: 27: 07 PM. (This statement \>\EIS added by !he PRIME system lf)0n Its vali da!lon)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 21 of 33 Westinghouse Non-Proprietary Class 3 RMVI.I 16 6 0 Breok Flow 70000-r-------------------------,

60000 50000 If)

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~

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-10000-+-'~~-.-'-'--'--r-~~,_,_...,__~-'--r~~_._,.......,....~~-,-,.~'-'-t 0 100 200 .300 400 500 600 700 Time After Break (sec) 8191~57 Figure 3: Millstone Unit 3 Break Mass Flow Rate for the Region II Analysis PCT Case Assuming OPA

... This record was final approved on 8/4/2020 1:27:07 PM. (This statement was added by the PRIME sy.;tem upon its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 22 of 33 Westinghouse Non-Proprietary Class 3 LO-LEVEL 0 0 COLLAPSED LIQ, LEVEL 12-.---------------------------,

10 ,-----------'--"-*~---*_.---~----_,

,._ 8 Q) 3

"'O

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0 0

u 4 2

100 200 300 400 500 600 700 Time After Break (sec) s191ms1 Figure 4: Millstone Unit 3 Lower Plenum Collapsed Liquid Level (Relative to Inside Bottom of Vessel) for the Region II Analysis PCT Case Assuming OPA

"'This record was final approved on 8/4/2020 1 :27:07 PM. (This statement was added by the PRIME system upon its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 23 of 33 Westinghouse N on-Proprietaty Class 3 FGIA 13 1 0 Bo t t om Ce I I FGM 13 15 0 Top Cell 2000...------------------------,

. II ti \

/ I 1500 _, ,.

I\

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-1000-1-.i......i.....1..-J-_,...J'-L-.l......,_.,....._..._................,_.....................,,-.,......................,__._.........._.i....i 0 5 10 15 20 25 30 Time After Break (sec) 819l5l3S7 Figure 5: Millstone Unit 3 Vapor Mass Flow .~teat the Top and Bottom Cell Faces of the Core Avl!'age Challllelnot Undm* Guide Tubes for the Region IIAnaly5isPCT Case Assuming OPA

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 24 of 33 WeGlinghouse N on-Proprietaty Class 3 FLM 13 1 0 Bo I I om Ce I I FLM 13 15 0 Top Ce I I 20000..-----------------------,

15000 "w 10000 E

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o::

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r

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-lOOOO-l-'--'"-'--'-r--L...L..J...JL-.r-"-J...Jc....L-,-.._,_........_..-1_,__,_.,__,......_...,_,._,..-1 0 5 10 15 20 25 JO Time After Break {sec) 8l915J357 Figure 6: Milhtone.Unit3 Liquid )\,tass Flow Rate at the Top and Bottom Cell Faces of tile CQre Avirage aramtel not Under Guide Tubes for tile Region II Analysis PCT Case~suming OPA

"' THs record Ms firal approved on 8/4120201 :27:07 PM. (This statement Ms added by the PRIME system lf.)On Its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 25 of 33 WestirtghouseNon-Proprietary Class 3 PN 39 0 PRESSURE 100-r---.------------------------,

80

---0

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CL 40 20 0-1--L-L-'--'-,--'--'--L-L~..L......l--'--'--.-'--'-..L......l~-'-L-L-'-.,........i--'--'--........-'-..L......l--'-l 0 100 200 300 400 500 600 700 Time After Break (sec)

Figure 7: Millstone Unit 3 RCS Pressure for the Region II Analysis PCT Case Assuming OPA

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Serial No.20-333 Docket No. 50-423 Attachment 3, Page 26 of 33 Westinghouse Non-Proprietary Class 3 I.HH00243 74 4 0 Comp on en t 74 I.ITH00247 84 4 O Component 84 I.ITH00251 94 4 0 Component 94 2000~----------------------~

1500-r z

r---

11000 0

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-500+--'-__._........._,--..___._........_..___._.._...._....,...._..____._._-,-....__._--la---4 0 20 40 60 80 100 Time After Break (sec) 819lill57 Figure 8: Millstone Unit3 Accumulator Injection Flow pa* Loop for the Region II Analysis PCT Ca~ Assuming OPA

"' THs reoorcf \/\es flral approved on 8/4120201:27:07 PM. (This stalement1J1es added bYthe PRIME system 1.pon Its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 27 of 33 Westinghouse Non-Proprietary Class 3 PN 4 0 Containment Pressure 23,---------------------------,

26 24

.----... 22 0

"in

....3:

e 20 ti) ti) e a._

18 16 14 12-1-'-'--'--'-,,-,...~.._.-r-..,_,__,__._-,--,_._..,_"-T-'-~~-r---L-'-'--'-,-'--'-~'-t 0 100 200 300 400 500 600 700 Time Arter Break (sec) s191ms1 Figure 9.: Millstone Unit 3 Containment Pressure for the Region II Analysis PCT Case Assuming OPA

"'This record was final approved on 8/4/2020 1 :27:07 PM. [fhis statement was added by the PRIME system upon its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 28 of 33 Westinghouse Non-Proprietary Class 3 VFMASS 0 0 0 VESSEL WATER MASS 250000-.--------------------------,

200000 150000 E

_o

=-

100000 50000 0-1-._..__,__._,_,__.__.__.__,__,_,_,__._,_.___,_..._,L-,-___._,__._.__-,-L.-..L_,_.._,_,___.__._,_-I 0 100 200 300 400 500 600 700 Time After Break (sec) 8191m57 Figure 10: Millstone Unit 3 Vessel Fluid Mass for the Region II Analysis PCT Case Assuming OPA

"'This record was final approved on 814/2020 1:27:07 PM. (This statement was added by the PRIME system upon its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 29 of 33 Westinghouse N on-Proprietaty Class 3 LO-LEVEL 6 0 O Hot As1embly Channel LO-LEVEL 4 0 0 N~n-GT Avg Channel LO-LEVEL 5 0 0 GT Avg Channel LO-LEVEL 3 0 0 Low Power Channel 12----------------------------,

10 4-

~ 8 Q)

(l)

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"' Tlis record 'MlS fiml approved on 8/4120201 :27:07 PM. (This statement \/\es added by the PRIME system lf.)On Its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 30 of 33 WeGl.inghouse Non-Proprietaty Class 3 IATH00218 7 0 0 COLLAPSED LIQ. LEVEL 35...--------------------------,

30 25 Q)

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100 200 300. 400 500 600 700 Time After Break (sec) 8191Sll57 Figure 12: Millstmte Uuit3 Average Dowucoml!' Collapsed Liquid Level (Relative to l!ottom of UppB' Tie Plate) fortl1e Regimi JlAilalyg.sPCT QiseAssu.min.g OPA

"' THs reoord 'MIS final approved on 8/4120201:27:07 PM. (This statement Vies added by the PRIME system lf)0n Its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 31 of 33 Westinghouse N on-Proprietaty Class 3 MTH00235 78 7 0 Loop 2 SI UTH00237 88 7 0 Loop ,3 SI MTH00:?39 98 7 O Loop 4 $1 160-,-------------------------~

140 120 1/)

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Figure 13: l\/Iilbtmte Unit3 Safety Inj oction Flow pa* Loop (not including Accumulator Injection Flow) for the Reglmt II Analysis PCT CaseAsswning OPA

"' THs rerord \'\tls firal approved on 8(4120201:27:07 PM. (This statement \'\tlS added bYlhe PRIME system ll)On Its validation)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 32 of 33 Westinghouse N on-Proprietaty Class 3 HRPWR 0 0 0 Hot Rod HAPWR 0 0 0 Hot Assembly AVGPWR 0 0 O Average Rod 25~--------------------------~

2

\,,..

~ 1,5 0

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z:

o-_,__,__._~__,_...._,._~.,__-'-_.__ _.____.,__,_~...._.,__..,_~_.__.__,_--l 0 2 4 6 8 10 12 Elevation (fl) 819!53JS7 Figure 14: Millstone Unit 3 Normalized Core Power Shapes for the RegiOll II Analysis PCT Case Asi;uming OPA

"' THs record 'MlS flral approved on 8(4120201 :27:07 PM. (This statement 'MlS added by the PRIME system lt)0n Its valldlrtlon)

Serial No.20-333 Docket No. 50-423 Attachment 3, Page 33 of 33 Westinghouse Non-Proprietary Class 3 POWERf 0 0 0 RELATIVE CORE POWER 1.2-.------------------------,

0

.Ec:; 0-8 0

z 0

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0 0 0.6 Lt Q) 0 CL Q.4 Q) 0 u

0.2 Q-1-"-'-'--'-,.-,..................-.-...._,._........,_-r-"_,_.._.,-,-_,_..__._._-,-b..A._..._-'-f~.......,~

0 100 200 300 400 500 600 700 Time After Break (sec)

Bl91!il357 figure 15: Milbtone Unit 3 Rdativi:i Core PowB' for the Region II Allllly~sPCT Case Assuming OPA

"'Tlis record W:ls final approved on 8/4120201:27:07 PM. (This statement w;s added by the PRIME system 1,pon Its velldatlon)