ML22234A169

R. E. Ginna Nuclear Power Plant, Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR)

ML22234A169
Person / Time
Site:	Ginna
Issue date:	08/22/2022
From:	David Gudger Constellation Energy Generation
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML22234A168	List: ML22234A169
References
Download: ML22234A169 (97)

Text

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 200 Exelon Way Kennett Square, PA 19348 www.constellation.com 10 CFR 50.90 August 22, 2022 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244

Subject:

Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR)

References:

1. Letter from Thomas G. Mogren (CENG, a joint venture of Constellation Energy & EDF) to U.S. NRC, Thermal Conductivity Degradation Impact on R.E. Ginna Large Break Loss of Coolant Accident Analysis with ASTRUM.

Response to Request for Additional Information, dated June 19, 2013 In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) is submitting a request for an amendment to the Technical Specifications (TS) for Renewed Facility Operating License No. DPR-18 for R.E. Ginna Nuclear Power Plant (Ginna).

The amendment request proposes to revise TS 5.6.5, Core Operating Limits Report (COLR). The proposed change revises TS 5.6.5 to replace the current NRC approved Loss-of-Coolant Accident (LOCA) methodologies with a single, newer NRC approved LOCA methodology, the FULL SPECTRUMTM LOCA Evaluation Model (FSLOCATM EM),

that is contained in WCAP-16996-P-A, Rev. 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), and was used for LOCA reanalysis for Ginna. Use of WCAP-16996-P-A for 2-Loop plants is currently under NRC review with the issuance of Westinghouse Letter LTR-NRC-21-22, Extension of FULL SPECTRUMTM LOCA (FSLOCATM) Evaluation Methodology to 2-loop Westinghouse Pressurized Water Reactors (PWRs) with Information to Satisfy Limitations and Conditions Specific to 2-loop Plant Types (Proprietary/Non-Proprietary), dated September 16, 2021.

contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachment 3, this document is decontrolled.

License Amendment Request Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR)

Docket No. 50-244 August 22, 2022 Page 2 The attached request is subdivided as follows:

• provides a description and evaluation of the proposed changes.

• provides the Affidavit of Withholding, Proprietary Information Notice, and Copyright Notice for FSLOCATM LAR Input Document, APPLICATION OF WESTINGHOUSE FULL SPECTRUMTM LOCA EVALUATION MODEL TO THE R.E. GINNA NUCLEAR POWER PLANT.

• provides the Proprietary Class 2 Version of FSLOCATM LAR Input Document, APPLICATION OF WESTINGHOUSE FULL SPECTRUMTM LOCA EVALUATION MODEL TO THE R.E. GINNA NUCLEAR POWER PLANT.

• provides the Non-Proprietary Class 3 Version of the FSLOCATM LAR Input Document, APPLICATION OF WESTINGHOUSE FULL SPECTRUMTM LOCA EVALUATION MODEL TO THE R.E. GINNA NUCLEAR POWER PLANT.

• provides the markup of the affected TS pages for Ginna.

• provides the TS Bases pages marked to show the proposed changes for information only for Ginna.

The proposed change has been reviewed by Ginna Plant Operations Review Committees in accordance with the requirements of the CEG Quality Assurance Program.

CEG requests approval of the proposed license amendment request within one year of this submittal date; i.e., by August 15, 2023. Once approved, the amendment shall be implemented according to the following table. It will be implemented during the defueled window in the specified Refueling Outage (Calendar Year provided for reference):

Unit Date Ginna Refueling Outage G1R45 (October 2024)

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (a)(1), the analysis about the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is being provided to the Commission.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), CEG is notifying the State of New York of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

Should you have any questions concerning this submittal, please contact Jessie Hodge at (610) 765-5532.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 15th day of August 2022.

License Amendment Request Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR)

Docket No. 50-244 August 22, 2022 Page 3 Respectfully, David T. Gudger Sr Manager - Licensing Constellation Energy Generation, LLC Attachments:

1. Evaluation of Proposed Changes

2. Affidavit of Withholding, Proprietary Information Notice, and Copyright Notice for FSLOCATM LAR Input Document, APPLICATION OF WESTINGHOUSE FULL SPECTRUMTM LOCA EVALUATION MODEL TO THE R.E. GINNA NUCLEAR POWER PLANT

3. Proprietary Class 2 Version of FSLOCATM LAR Input Document, APPLICATION OF WESTINGHOUSE FULL SPECTRUMTM LOCA EVALUATION MODEL TO THE R.E. GINNA NUCLEAR POWER PLANT

4. Non-Proprietary Class 3 Version of the FSLOCATM LAR Input Document, APPLICATION OF WESTINGHOUSE FULL SPECTRUMTM LOCA EVALUATION MODEL TO THE R.E. GINNA NUCLEAR POWER PLANT

5. Proposed Technical Specification Changes (Mark-Ups)

6. Proposed Technical Specification Bases Changes (Mark-Ups) cc:

Regional Administrator, NRC Region I NRC Resident Inspector NRC Project Manager A. L. Peterson (NYSERDA)

ATTACHMENT 1 Evaluation of Proposed Changes R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 Docket No. 50-244

Subject:

License Amendment Request to revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR) 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY ANALYSIS

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

License Amendment Request Application to Revise Technical Specifications 5.6.5, Page 1 of 11 Core Operating Limits Report (COLR)

Docket No. 50-244 Description and Assessment

1.0 DESCRIPTION

In accordance with 10 CFR 50.90, Application for amendment of license, construction permit or early site permit, Constellation Energy Generation, LLC (CEG) requests amendments to Renewed Facility Operating License No. DPR-18 for R.E. Ginna Nuclear Power Plant (Ginna).

This amendment request proposes to revise Technical Specifications (TS) 5.6.5, Core Operating Limits Report (COLR). The proposed change revises TS 5.6.5 to replace the current NRC approved Loss-of-Coolant Accident (LOCA) methodologies with a single, newer NRC approved LOCA methodology, the FULL SPECTRUMTM LOCA Evaluation Model (FSLOCATM EM), that is contained in WCAP-16996-P-A, Rev. 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology),

(Reference 1) and was used for LOCA reanalysis for Ginna. Use of WCAP-16996-P-A for 2-Loop plants is currently under NRC review with the issuance of Westinghouse Letter LTR-NRC-21-22, Extension of FULL SPECTRUMTM LOCA (FSLOCATM) Evaluation Methodology to 2-loop Westinghouse Pressurized Water Reactors (PWRs) with Information to Satisfy Limitations and Conditions Specific to 2-loop Plant Types (Proprietary/Non-Proprietary), dated September 16, 2021.

The proposed change to the Emergency Core Cooling System (ECCS) LOCA EM for Ginna does not involve any changes to fuel type, peaking factors, fuel structural analysis, or boric acid precipitation methodology.

This license amendment request (LAR) completes CEGs action to perform an ECCS LOCA reanalysis in accordance with 10 CFR 50.46(a)(3)(ii) as described in CENG Letter Thermal Conductivity Degradation Impact on R.E. Ginna Large Break Loss of Coolant Accident Analysis with ASTRUM. Response to Request for Additional Information dated June 19, 2013 and documented as ML13175A357 (Reference 11).

2.0 DETAILED DESCRIPTION The proposed changes to TS 5.6.5 reflect the NRC approved LOCA methodology that was used for the LOCA reanalysis for Ginna. Attachment 5 to this amendment request provide the markup pages of the existing TS to show the proposed change.

2.1 Proposed Changes The proposed TS changes are described below.

TS 5.6.5.b.2 and TS 5.6.5.b.6 through 5.6.5.b.9 currently states:

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

6. WCAP-10054-P-A and WCAP-10081-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985. (Methodology for LCO 3.2.1.)

License Amendment Request Application to Revise Technical Specifications 5.6.5, Page 2 of 11 Core Operating Limits Report (COLR)

Docket No. 50-244 Description and Assessment

7. WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and COSI Condensation Model," July 1997. (Methodology for LCO 3.2.1)

8. WCAP-11145-P-A, "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study With the NOTRUMP Code," October 1986. (Methodology for LCO 3.2.1)

9. WCAP-10079-P-A, "NOTRUMP - A Nodal Transient Small Break and General Network Code," August 1985. (Methodology for LCO 3.2.1)

Revised TS 5.6.5.b.1 through 5.6.5.b.12 will state:

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

6. Not Used.

7. Not Used

8. Not Used

9. Not Used.

3.0 TECHNICAL EVALUATION

3.1 Compliance with the Limitations and Conditions in the Revised NRC Final Safety Evaluation (FSE) for Westinghouse WCAP-16996-P-A, Rev. 1 Attachments 3 and 4 to this amendment request show Ginna analyses are in compliance with the Limitations and Conditions with the exception of Number 2. Number 2 states:

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.

R. E. Ginna is a 2-loop PWR with Upper Plenum Injection (UPI), so it is not within the NRC-approved methodology. The purpose of LTR-NRC-21-22, Extension of FULL SPECTRUMTM LOCA (FSLOCATM) Evaluation Methodology to 2-loop Westinghouse Pressurized Water Reactors (PWRs) with Information to Satisfy Limitations and Conditions Specific to 2-loop Plant Types (Proprietary/Non-Proprietary) dated September 2021 and docketed by the NRC as ML13175A357 (Reference 11) is to extend the applicability of the FSLOCA EM to 2-loop Westinghouse-designed PWRs with UPI. Additional discussion on this Limitation and Condition can be found in Attachments 3 and 4.

License Amendment Request Application to Revise Technical Specifications 5.6.5, Page 3 of 11 Core Operating Limits Report (COLR)

Docket No. 50-244 Description and Assessment 3.2 Changes and Corrections to the FSLOCATM EM in Westinghouse WCAP-16996-P-A, Revision 1 Westinghouse has issued letter LTR-NRC-18-30 (Reference 2) and LTR-NRC-19-6 (Reference

3) to the NRC which document several changes and corrections that have been made to the FSLOCATM EM in Westinghouse WCAP-16996-P-A, Revision 1, after the NRC issued the revised Final Safety Evaluation. Those changes and corrections that are applicable to the FSLOCATM methods amount to minimal impacts and will be handled in accordance with 10 CFR 50.46 requirements. Furthermore, any future changes or corrections to the methods or the specific Ginna Analysis of Record being implemented here will also be in accordance with 10 CFR 50.46 requirements.

3.3 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 not exceeded:

Peak Cladding Temperature (10 CFR 50.46(b)(1)) - The analysis Peak Cladding Temperature (PCT), corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level. Since the resulting PCT is less than 2,200 °F, the analyses with the FSLOCATM EM confirm that 10 CFR 50.46 acceptance criterion (b)(1), i.e., that PCT not exceed 2,200 °F, is satisfied.

The results are shown in Table 7 of Attachments 3 and 4 for Ginna.

Maximum Cladding Oxidation (10 CFR 50.46(b)(2)) - The analysis Maximum Local Oxidation (MLO) corresponds to a bounding estimate of the 95th 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 corrosion, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(2), i.e.,

that the MLO of the cladding not exceed 17 percent of the total cladding thickness before oxidation, is satisfied.

The results are shown in Table 7 of Attachments 3 and 4 for Ginna.

Maximum Hydrogen Generation (10 CFR 50.46(b)(3)) - The analysis Core-Wide Oxidation (CWO) corresponds to a bounding estimate of the 95th percentile 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 criterion (b)(3), i.e., that CWO not exceed 1 percent of the total hypothetical amount, is satisfied.

The results are shown in Table 7 of Attachments 3 and 4 for Ginna.

Coolable Geometry (10 CFR 50.46(b)(4)) - This criterion 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 criteria (b)(1), (b)(2), and (b)(3), and by assuring that fuel assembly grid deformation due to combined LOCA and seismic loads is

License Amendment Request Application to Revise Technical Specifications 5.6.5, Page 4 of 11 Core Operating Limits Report (COLR)

Docket No. 50-244 Description and Assessment specifically addressed. Table 7 of Attachments 3 and 4 show Criteria (b)(1), (b)(2), and (b)(3) have been met for Ginna.

Section 32.1 of the NRC approved FSLOCATM EM (WCAP-16996-P-A, Revision 1) documents that the effects of LOCA and seismic loads on the core geometry do not need to be considered unless fuel assembly grid deformation extends beyond the core periphery (i.e., deformation in a fuel assembly with no sides adjacent to the core baffle plates). Inboard grid deformation due to the combined LOCA and seismic loads was calculated to not occur for Ginna. The FSLOCATM EM analyses did not invalidate the existing seismic / LOCA analysis.

Long-Term Cooling (10 CFR 50.46(b)(5)) - This criterion requires that long-term 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 FSLOCATM EM (WCAP-16996-P-A, Revision 1).

In summary, based on the analysis results for the small-break LOCA (SBLOCA, Region I) and large-break LOCA (LBLOCA, Region II) presented in Table 7 of Attachments 3 and 4 for Ginna and the discussion above relative to the criteria in 10 CFR 50.46(b)(4) and (b)(5), it is concluded that Ginna would continue to comply with the criteria in 10 CFR 50.46 upon approval of this LAR.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria Section 182.a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TS as part of the license. The Commissions regulatory requirements related to the content of the TS are contained in 10 CFR 50.36, Technical specifications. The TS requirements in 10 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs);

(3) surveillance requirements; (4) design features; and (5) administrative controls.

However, the rule does not specify the particular requirements to be included in a plants TS.

Under 10 CFR 50.36(c)(2)(ii), a limiting condition for operation must be included in TS for any item meeting one or more of the following four criteria:

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor, coolant pressure boundary.

Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that

License Amendment Request Application to Revise Technical Specifications 5.6.5, Page 5 of 11 Core Operating Limits Report (COLR)

Docket No. 50-244 Description and Assessment either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

The proposed change does not impact the TS safety limits, limiting safety system settings, and limiting control settings; LCOs; surveillance requirements; or design features.

The proposed replacement for NRC approved LOCA methodology will be included in the Administrative Controls section of the Ginna 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 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 FULL SPECTRUMTM LOCA (FSLOCATM) methodology described in WCAP-16996-P-A, Revision 1 for the performance of full spectrum LOCA 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 Westinghouse designed four loop Pressurized Water Reactors.

WCAP-16996-P-A, Revision 1 is applicable to Ginna, and the plant specific application of the FSLOCATM methodology to the LOCA analyses have been performed in accordance with the conditions and limitations of the topical report and the associated NRC safety evaluation. The plant specific analyses demonstrate that the requirements of 10 CFR 50.46 will continue to be met, thus ensuring continued safe plant operation.

NRC Generic Letter (GL) 88-16, Removal of Cycle-Specific, Parameter Limits from Technical Specifications, dated October 4, 1988 (Reference 4), provides that it is acceptable for licensees to control reactor physics 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.

Consistent with the guidance in NRC GL 88-16, Ginna TS 5.6.5, Core Operating Limits Report (COLR), requires the following:

License Amendment Request Application to Revise Technical Specifications 5.6.5, Page 6 of 11 Core Operating Limits Report (COLR)

Docket No. 50-244 Description and Assessment An NRC approved methodology is to be used to determine the core operating limits listed in TS 5.6.5.a; The specific NRC approved methodologies used to determine the core operating limits are to be listed in TS 5.6.5.b; and The COLR, including any midcycle revisions or supplements, is to be provided upon issuance for each reload cycle to the NRC in accordance with TS 5.6.5.d.

The COLR is defined in Section 1.1 of the TS and the reporting requirements in TS 5.6.5 require that a COLR be submitted to the NRC each operating cycle, or each time the COLR is revised.

The GL also required that the TS include a list of references of the NRC approved methodologies that are used to determine the cycle specific core operating limits. TS 5.6.5.b identifies the NRC approved analytical methodologies that are used to determine the core operating limits for Ginna. 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 proposed change to replace the previous NRC approved LOCA methodologies with the NRC approved LOCA methodology in WCAP-16996-P-A, Revision 1, which was used for the Ginna reanalysis, satisfies NRC GL 88-16.

4.2 Precedent The NRC has approved the following similar license amendment request to revise Core Operating Limits Report for Full Spectrum Loss-of Coolant Accident Methodology:

Letter from B.K. Singal (NRC) to J.M. Welsch (PG&E), Diablo Canyon Nuclear Power 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), dated January 9, 2020 (ML19316A109). (Reference 5)

Letter from J. S. Wiebe (NRC) to B. C. Hanson (Constellation), BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 219, 219, 223, AND 223 REGARDING REVISION OF TECHNICAL SPECIFICATIONS 5.6.5, CORE OPERATING LIMITS REPORT (COLR) (EPID L-2020-LLA-0038), dated December 28, 2020 (ML20317A001) (Reference 9)

The following are other similar amendment requests that has been reviewed by the NRC:

Letter from K. Green (NRC) to J. Barstow (Tennessee Valley Authority), WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 143 AND 50 REGARDING IMPLEMENTATION OF FULL SPECTRUM' LOSS-OF-COOLANT ACCIDENT ANALYSIS (LOCA) AND NEW LOCA-SPECIFIC TRITIUM PRODUCING BURNABLE ABSORBER ROD STRESS ANALYSIS METHODOLOGY (EPID L-2020-LLA-0005), dated February 26, 2021 (ML21034A169). (Reference 6)

Letter from G.E. Miller (NRC) to D. Stoddard (Dominion Energy Virginia), NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 ISSUANCE OF AMENDMENT NOS. 286 AND 269 TO

License Amendment Request Application to Revise Technical Specifications 5.6.5, Page 7 of 11 Core Operating Limits Report (COLR)

Docket No. 50-244 Description and Assessment REVISE TECHNICAL SPECIFICATIONS TO ALLOW USAGE OF A FULL SPECTRUM LOSS-OF-COOLANT-ACCIDENT (LOCA) METHODOLOGY (EPID L-2019-LLA-0236), dated October 29, 2020 (ML20302A179). (Reference 7)

Letter from V. Thomas (NRC) to D. Stoddard (Dominion Energy Virginia), SURRY POWER STATION, UNIT 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), dated October 28, 2020 (ML20301A452). (Reference 8) 4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, Application for amendment of license, construction permit or early site permit, CEG requests amendments to Renewed Facility Operating License Nos.

DPR-18 for R.E. Ginna Nuclear Power Plant (Ginna). This amendment request proposes to revise Technical Specifications (TS) 5.6.5, Core Operating Limits Report (COLR). The proposed change revises TS 5.6.5 to replace the current NRC approved Loss-of-Coolant Accident (LOCA) methodologies with a single, newer NRC approved LOCA methodology, the FULL SPECTRUMTM LOCA Evaluation Model (FSLOCATM EM), that is contained in WCAP-16996-P-A, Rev. 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), and was used for LOCA reanalysis for Ginna. The NRC approved WCAP-16996-P-A, Revision 1 in a safety evaluation dated October 19, 2016.

According to 10 CFR 50.92, Issuance of amendment, paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of any 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.

CEG has evaluated the proposed change, using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

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

Response: No.

The proposed change revises TS 5.6.5.b to replace the current NRC approved Loss-of-Coolant Accident (LOCA) methodologies listed in TS 5.6.5.b with another NRC approved methodology contained in WCAP-16996-P-A, Rev. 1, Realistic LOCA Evaluation

License Amendment Request Application to Revise Technical Specifications 5.6.5, Page 8 of 11 Core Operating Limits Report (COLR)

Docket No. 50-244 Description and Assessment Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology).

The proposed changes to the TS 5.6.5.b core operating limits methodologies, consists of replacing three current LOCA methodologies with a newer, single NRC approved methodology (the FSLOCATM EM). The NRC review of the FSLOCATM EM concluded that the analytical methods are acceptable as a replacement for the current LOCA analytical methods listed in TS 5.6.5.b.

The proposed change does not affect the design or function of any plant structures, systems, and components (SSCs). Thus, the proposed change does not affect plant operation, design features, or the capability of any SSC to perform its safety function. In addition, the proposed change does not affect any previously evaluated accidents in the Updated Final Safety Analysis Report (UFSAR), or any SSCs, operating procedures, and administrative controls that have the function of preventing or mitigating any accident previously evaluated in the UFSAR. Thus, the proposed use of the FSLOCATM EM will continue to assure that the plant operates in the same safe manner as before and will not involve an increase in the probability of an accident.

The analyses results determined by use of the proposed new methodology will not increase the reactor power level or the core fission product inventory and will not change any transport assumptions or the shutdown margin requirements of the Ginna TS. As such, Ginna will continue to operate within the power distribution limits and shutdown margins required by the TS and within the assumptions of the safety analyses described in the UFSAR. As such, the proposed changes do not involve a significant increase in the consequences of an accident.

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 accident from any accident previously evaluated?

Response: No.

The Proposed change revises TS 5.6.5.b to replace the current NRC approved Loss-of-Coolant Accident (LOCA) methodologies listed in TS 5.6.5.b with a single, newer NRC approved methodology contained in WCAP-16996-P-A, Rev. 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology). The NRC review of the FSLOCATM EM concluded that the analytical methods are acceptable as a replacement for the current LOCA analytical methods listed in TS 5.6.5.b.

The proposed change provides revised analytical methods and does not change any system functions or maintenance activities. The change does not involve physical alteration of the plant; that is, no new of different type of equipment will be installed. The change does not impact the ability of any SSC to perform its safety function consistent with the assumptions of the safety analyses and continues to assure the plant is operated within safe limits. As

License Amendment Request Application to Revise Technical Specifications 5.6.5, Page 9 of 11 Core Operating Limits Report (COLR)

Docket No. 50-244 Description and Assessment such, the proposed change does not create new failure modes or mechanisms that are not identifiable during testing, and no new accident precursors are generated.

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

Response: No.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed change does not physically alter safety-related systems, nor does it affect the way in which safety-related systems perform their functions. The setpoints at which protective actions are initiated are not altered by the proposed change. Therefore, sufficient equipment remains available to actuate upon demand for the purpose of mitigating an analyzed event. The NRC has reviewed and approved the new methodology for the intended use in lieu of the current methodologies; thus, the margin of safety is not reduced due to this change.

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

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

5.0 ENVIRONMENTAL CONSIDERATION

CEG has evaluated the proposed amendment and has determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents 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 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.

License Amendment Request Application to Revise Technical Specifications 5.6.5, Page 10 of 11 Core Operating Limits Report (COLR)

Docket No. 50-244 Description and Assessment

6.0 REFERENCES

1. 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. LTR-NRC-18-30, 10 CFR 50.46 Annual Notification and Reporting for 2017, July 2018 (ML19288A174).

3. LTR-NRC-19-6, 10 CFR 50.46 Annual Notification and Reporting for 2018, February 7, 2019 (ML19042A379).

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

5. Letter from B.K. Singal (NRC) to J.M. Welsch (PG&E), Diablo Canyon Nuclear Power 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), dated January 9, 2020 (ML19316A109).

6. Letter from K. Green (NRC) to J. Barstow (Tennessee Valley Authority), WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 143 AND 50 REGARDING IMPLEMENTATION OF FULL SPECTRUM' LOSS-OF-COOLANT ACCIDENT ANALYSIS (LOCA) AND NEW LOCA-SPECIFIC TRITIUM PRODUCING BURNABLE ABSORBER ROD STRESS ANALYSIS METHODOLOGY (EPID L-2020-LLA-0005), dated February 26, 2021 (ML21034A169).

7. Letter from G.E. Miller (NRC) to D. Stoddard (Dominion Energy Virginia), NORTH ANNA POWER STATION, UNIT 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), dated October 29, 2020 (ML20302A179).

8. Letter from V. Thomas (NRC) to D. Stoddard (Dominion Energy Virginia), SURRY POWER STATION, UNIT 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), dated October 28, 2020 (ML20301A452).

9. Letter from J. S. Wiebe (NRC) to B. C. Hanson (Constellation), BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 219, 219, 223, AND 223 REGARDING REVISION OF TECHNICAL SPECIFICATIONS 5.6.5, CORE OPERATING LIMITS REPORT (COLR) (EPID L-2020-LLA-0038), dated December 28, 2020 (ML20317A001)

10. LTR-NRC-21-22, Extension of FULL SPECTRUM' LOCA (FSLOCA TM) Evaluation Methodology to 2-loop Westinghouse Pressurized Water Reactors (PWRs) with Information

License Amendment Request Application to Revise Technical Specifications 5.6.5, Page 11 of 11 Core Operating Limits Report (COLR)

Docket No. 50-244 Description and Assessment to Satisfy Limitations and Conditions Specific to 2-Ioop Plant Types (Proprietary/Non-Proprietary), dated September 16, 2021 (ML21265A203).

11. Letter from T. Mogren (CENG) to NRC, Thermal Conductivity Degradation Impact on R.E.

Ginna Large Break Loss of Coolant Accident Analysis with ASTRUM. Response to Request for Additional Information dated June 19, 2013 (ML13175A357).

ATTACHMENT 2 License Amendment Request R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 Docket No. 50-244 Affidavit of Withholding, Proprietary Information Notice, and Copyright Notice for FSLOCATM LAR Input Document, APPLICATION OF WESTINGHOUSE FULL SPECTRUMTM LOCA EVALUATION MODEL TO THE R.E. GINNA NUCLEAR POWER PLANT

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-22-013 Page 1 of 3 Commonwealth of Pennsylvania:

County of Butler:

(1)

I, Camille Zozula, Manager, Regulatory Compliance and Corporate Licensing, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2)

I am requesting the proprietary portions of RGE-22-4 P-Attachment be withheld from public disclosure under 10 CFR 2.390.

(3)

I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4)

Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i)

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii)

The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.

(iii)

Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-22-013 Page 2 of 3 (5)

Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a)

The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b)

It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c)

Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d)

It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e)

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f)

It contains patentable ideas, for which patent protection may be desirable.

(6)

The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit.

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-22-013 Page 3 of 3 I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 3/24/2022 Signed electronically by Camille Zozula

ATTACHMENT 3 License Amendment Request R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 Docket No. 50-244 Proprietary Class 2 Version of FSLOCATM LAR Input Document, APPLICATION OF WESTINGHOUSE FULL SPECTRUMTM LOCA EVALUATION MODEL TO THE R.E. GINNA NUCLEAR POWER PLANT

ATTACHMENT 4 License Amendment Request R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 Docket No. 50-244 Non-Proprietary Class 3 Version of the FSLOCATM LAR Input Document, APPLICATION OF WESTINGHOUSE FULL SPECTRUMTM LOCA EVALUATION MODEL TO THE R.E.

GINNA NUCLEAR POWER PLANT

APPLICATION OF WESTINGHOUSE FULL SPECTRUM LOCA EVALUATION MODEL TO THE R. E. GINNA NUCLEAR POWER PLANT

1.0 INTRODUCTION

An analysis with the FULL SPECTRUM'loss-of-coolant accident (FSLOCA') evaluation model (EM) has been completed for the R. E. Ginna Nuclear Power Plant. This license amendment request (LAR) for R. E. Ginna requests approval to apply the Westinghouse FSLOCA EM.

The FSLOCA EM (Reference 1) was developed to address the full spectrum of loss-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 Large Break LOCAs (LBLOCAs).

The FSLOCA EM explicitly considers the effects of fuel pellet thermal conductivity degradation (TCD) and other burnup-related effects by calibrating to fuel rod performance data input generated by the PAD5 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 WCOBRA/TRAC-TF2 code used in the FSLOCA EM explicitly accounts for pellet TCD.

Three of the Title 10 of the Code of Federal Regulations (CFR) 50.46 criteria (peak cladding temperature (PCT), maximum local oxidation (MLO), and core-wide oxidation (CWO)) are considered directly in the FSLOCA EM. A high probability statement is developed for the PCT, MLO, and CWO that is needed to demonstrate compliance with 10 CFR 50.46 acceptance criteria (b)(1), (b)(2), and (b)(3) (Reference 3) via statistical methods. The MLO is defined as the sum of pre-transient corrosion and transient oxidation consistent with the 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)(1) and (b)(2), 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 Regulatory Commission (NRC) for Westinghouse 3-loop and 4-loop plants with cold leg Emergency Core Cooling System (ECCS) injection (Reference 1). Since R. E. Ginna is a Westinghouse designed 2-loop plant with upper plenum injection (UPI), the approved method (Reference 1) is not applicable. The purpose of Reference 13 is to extend the applicability of the approved method to 2-loop Westinghouse-designed PWRs with UPI. Information required to address Limitations and Conditions 2, 9, and 10 of the NRCs Safety Evaluation Report (SER) for Reference 1 was docketed in Reference 13 in support of application of the FSLOCA EM to Westinghouse 2-loop plants. Reference 13 also identifies three additional methodology limitations for 2-loop PWRs equipped with UPI.

FULL SPECTRUM and FSLOCA are trademarks of Westinghouse 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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This report summarizes the application of the Westinghouse FSLOCA EM to R. E. Ginna. The application of the FSLOCA EM to R. E. Ginna is consistent with the NRC-approved methodology (Reference 1) as amended by the methodology extension in Reference 13, with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The application of the FSLOCA EM to R. E.

Ginna is consistent with the conditions and limitations as identified in the NRCs SER for Reference 1 and in Reference 13, and is also applicable for the R. E. Ginna plant design and operating conditions.

Both Constellation and its analysis vendor (Westinghouse) have interface processes which identify plant configuration changes potentially impacting safety analyses. These interface 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 R. E. Ginna analysis with the FSLOCA EM are provided in Tables 1 through 6.

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. Westinghouses previously approved best-estimate LBLOCA EM is discussed in Reference 8. The EM is referred to as the Automated Statistical Treatment of Uncertainty Method (ASTRUM), and was developed following Regulatory Guide (RG) 1.157 (Reference 7).

When the FSLOCA EM was being developed, the NRC issued RG 1.203 (Reference 9) which expands on the principles of RG 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 of RG 1.157 which are more detailed than RG 1.203; therefore, both RGs were used for the development of the FSLOCA EM.

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

This 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 1D module based on TRAC-P.

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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-equilibrium between phases

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

4. Ability to represent important 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-TF2 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 uncertainties 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.

The applicability of the WCOBRA/TRAC-TF2 code to 2-loop PWRs with UPI has been assessed in Reference 13, and it was concluded that the WCOBRA/TRAC-TF2 code is acceptable.

2.3 Compliance with FSLOCA EM Limitations and Conditions The NRCs SER for Reference 1 contains 15 limitations and conditions on the NRC-approved FSLOCA EM. The applicability of the 15 limitations and conditions with respect to the extension of the FSLOCA EM to 2-loop PWRs equipped with UPI is contained in Section 4.1 of Reference 13. 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-term cooling.

Compliance The analysis for R. E. Ginna with the FSLOCA EM is only being used to demonstrate compliance with 10 CFR 50.46 (b)(1) through (b)(4).

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

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Compliance R. E. Ginna is a 2-loop PWR with UPI, so it is not within the NRC-approved methodology. The purpose of Reference 13 is to extend the applicability of the FSLOCA EM to 2-loop Westinghouse-designed PWRs with UPI. [

]a,c no other deviations were taken from the NRC-approved methodology documented in Reference 1, 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 12).

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 R. E. Ginna 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 [

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

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Compliance Consistent with the NRC-approved methodology, the decay heat uncertainty multiplier was [

]a,c for the R. E. Ginna analysis. The analysis simulations were all executed for no longer than 10,000 seconds following reactor trip. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results have been provided in units of sigma and approximate absolute units in Table

10.

Limitation and Condition Number 5 Summary The maximum assembly and rod length-average burnup is limited to [

]a,c respectively.

Compliance The maximum analyzed assembly and rod length-average burnup were less than or equal to [

]a,c respectively, for R. E. Ginna.

Limitation and Condition Number 6 Summary The fuel performance data for analyses with the FSLOCA EM should be based on the PAD5 code (at present), which includes the effect of thermal conductivity degradation. The nominal fuel pellet average temperatures and rod internal pressures should be the maximum values, and the generation of all the PAD5 fuel performance data should adhere to the NRC-approved PAD5 methodology.

Compliance PAD5 fuel performance data were utilized in the R. E. Ginna 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 of Reference 2.

Limitation and Condition Number 7 Summary The YDRAG uncertainty parameter should be [

]a,c Compliance Consistent with the NRC-approved methodology, the YDRAG uncertainty parameter was [

]a,c for the R. E. Ginna Region I analysis.

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Limitation and Condition Number 8 Summary The [

]a,c Compliance Consistent with the NRC-approved methodology, the [

]a,c for the R. E. Ginna Region I analysis.

Limitation and Condition Number 9 Summary For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to confirm that the [

]a,c for the plant design being analyzed. This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.

Compliance R. E. Ginna is a Westinghouse-designed 2-loop PWR with UPI. The requested sensitivity study was performed for a 2-loop Westinghouse-designed PWR with UPI and is discussed in Reference 13.

Limitation and Condition Number 10 Summary For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to: 1) demonstrate that no unexplained behavior occurs in the predicted safety criteria across the region boundary, and 2) ensure that the [

]a,c must cover the equivalent 2 to 4-inch break range using RCS-volume scaling relative to the demonstration plant. This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.

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

Compliance R. E. Ginna is a Westinghouse-designed 2-loop PWR with UPI. The requested sensitivity study was performed for a 2-loop Westinghouse-designed PWR with UPI and is discussed in Reference 13.

The minimum sampled break area for the R. E. Ginna Region II analysis was 1 ft2.

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

1. The [

]a,c the Region I and Region II analysis seeds, and the analysis inputs will be declared and documented prior to performing the Region I and Region II uncertainty analyses. The [

]a,c and the Region I and Region II analysis seeds 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 R. E. Ginna analysis as follows:

1. The [

]a,c the Region I and Region II analysis seeds, and the analysis inputs were declared and documented prior to performing the Region I and Region II uncertainty analyses. The [

]a,c and the Region I and Region II analysis seeds were not changed once they were declared and documented.

2. After the uncertainty analysis was documented, input discrepancies were discovered, resulting in post-analysis evaluations for Region I and a re-execution of the Region II analysis. The identified analysis inputs were updated to set them equal to the intended plant operating ranges or values, but they were not changed for the intended purpose of demonstrating compliance with the applicable acceptance criteria. See Notes 1 and 2 of Table 1 for additional information.

3. The plant operating ranges which were sampled within the uncertainty analyses are provided for R. E. Ginna 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 A bounding plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves (MSSVs) was modeled in the R. E. Ginna analysis.

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Limitation and Condition Number 13 Summary In plant-specific models for analysis with the FSLOCA EM: 1) the [

]a,c and 2) the

[

]a,c Compliance The [

]a,c in the analysis for R. E. Ginna. The [

]a,c in the 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 Cathcart-Pawel correlation. In either case, the pre-transient corrosion will be summed with the LOCA transient oxidation. If the Cathcart-Pawel correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 13 percent limit. If the Baker-Just correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 17 percent limit.

Compliance For the R. E. Ginna 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 criterion of 17%.

Limitation and Condition Number 15 Summary The Region II analysis will be executed twice; once assuming LOOP and once assuming OPA. The results from both analysis executions should be shown to be in compliance with the 10 CFR 50.46 acceptance criteria.

The [

]a,c Compliance The Region II uncertainty analysis for R. E. Ginna was performed twice; once assuming a LOOP and once assuming OPA. The results from both analyses that were performed are in compliance with the 10 CFR 50.46 acceptance criteria (see Section 5.0).

The [

]a,c Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-9

2.4 Compliance with Additional Methodology Limitations for 2-Loop PWRs Equipped with UPI Reference 13 contains three additional limitations on the extension of the FSLOCA EM methodology to two-loop PWRs with UPI. A summary of each additional methodology limitation and how it was met is provided below.

Additional Limitation Number 1 Summary Westinghouse will limit the application of the FSLOCA EM to 2-loop PWRs with [

]a,c Compliance A nominal core power level of 1811 MWt with 0% uncertainty was modeled in the R. E. Ginna analysis, which is less than the maximum level.

Additional Limitation Number 2 Summary Westinghouse will limit the application of the FSLOCA EM to 2-loop PWRs with [

]a,c Compliance A nominal peripheral assembly average linear heat rate of 2.66 kW/ft was modeled in the R. E. Ginna analysis, which is less than the maximum level.

Additional Limitation Number 3 Summary Westinghouse will limit the application of the FSLOCA EM to 2-loop PWRs where [

]a,c Compliance The product of the nominal hot assembly relative power and the average core linear heat rate modeled in the R. E. Ginna analysis is 16.5 kW/ft, which is less than the maximum level.

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3.0 REGION I ANALYSIS 3.1 Description of Representative Transient The small break LOCA transient can be divided into time periods in which specific phenomena are occurring, as discussed below.

Blowdown The rapid depressurization of the RCS coincides with subcooled liquid flow through the break. Following the reactor trip on the low pressurizer pressure setpoint, the pressurizer drains, and safety injection is initiated on the low pressurizer pressure SI setpoint. After reaching this setpoint and applying the safety injection delays, high pressure safety injection flow begins.

Forced Circulation Continued pump operation after the break drives forced circulation of fluid which tends to homogenize the fluid condition and voids in the RCS. This quasi-equilibrium phase persists while the RCS pressure remains above the secondary side pressure. The system drains from the top down, while significant mass is continually lost through the break. Throughout this period, the core remains covered by a homogenized, two-phase mixture and the fuel cladding temperatures remain at the saturation temperature level.

If the RCPs had tripped during or shortly after blowdown, phase separation and natural circulation would occur. The vapor generated in the core would be trapped in the upper regions by the liquid remaining in the crossover leg loop seal. As the system drained, the liquid levels in the downhill side of the pump suction (crossover leg) would become depressed all the way to the bottom elevations of the piping, allowing the steam trapped during the natural circulation phase to vent to the break (i.e., a process called loop seal clearance). However, because offsite power is available, RCP trip is modeled 5 minutes after reactor trip on the low pressurizer pressure setpoint. Therefore, RCP trip does not occur early enough for the process of loop-seal clearance to occur in the Region I transient simulations.

Boil-Off Once the RCPs trip, the system stratifies under gravity force, and liquid held up in the uphill side of the steam generators, hot legs, and top elevations of the vessel drains down to the lower elevations of the vessel. The RCS depressurizes at a rate controlled by the critical flow, which continues to be a primarily high quality mixture of water and steam. The RCS pressure remains high enough such that safety injection flow cannot make up for the primary system fluid inventory lost through the break, leading to core uncovery and a fuel rod cladding temperature heatup.

Core Recovery The RCS pressure continues to decrease, and once it reaches that of the accumulator gas pressure, the introduction of additional ECCS water from the accumulators replenishes the reactor vessel inventory and recovers the core mixture level. The transient is considered over as the break flow is compensated by the injected flow.

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3.2 Analysis Results The R. E. Ginna Region I analysis was performed in accordance with the NRC-approved methodology in Reference 1 and the methodology extension in Reference 13, with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The transient that produced the analysis PCT result is a cold leg break with a break diameter of 4.3-inches, and the limiting offsite power configuration is OPA. The most limiting ECCS single failure of one ECCS train (leaving 1 LHSI and 2 HHSI pumps available) is assumed in the analysis as identified in Table 1. Control rod drop is modeled for breaks less than 1 square foot assuming a 1.8 second signal delay time and a 1.8 second rod drop time. A post-analysis evaluation was performed to assess a low pressurizer pressure reactor trip signal processing time of 2.0 seconds. See Note 2 of Table 1 for additional information. RCP trip is modeled to occur 5 minutes after reactor trip on the low pressurizer pressure setpoint due to operator action for OPA transients. When the low pressurizer pressure SI setpoint is reached, there is a delay to account for emergency diesel generator start-up, filling headers, etc., after which safety injection is initiated into the reactor coolant system.

The results of the R. E. Ginna Region I uncertainty analysis are summarized in Table 7. The sampled decay heat uncertainty multipliers for the Region I analysis cases are provided in Table 10.

Table 8 contains a sequence of events for the transient that produced the Region I analysis PCT result.

Figures 1 through 13 illustrate the calculated key transient response parameters for this transient.

4.0 REGION II ANALYSIS 4.1 Description of Representative Transient A large-break LOCA transient can be divided into phases in which specific phenomena are occurring. A convenient way to divide the transient is in terms of the various heatup and cooldown phases that the fuel assemblies undergo. For each of these phases, specific phenomena and heat transfer regimes are important, as discussed below.

Blowdown - 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 terminated 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 Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-12

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 period, 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 period of the blowdown phase.

During 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 characterized 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. The ECCS water consists of high head safety injection (HHSI) pumped into the cold leg as well as low head safety injection (LHSI) pumped into the vessel at the hot leg nozzle centerline elevation.

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-pressurization 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 throughout the reflood period. The liquid flows down through the low power region and then across the core into the average assemblies near the bottom of the core. 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 inventory increases enough that liquid entrainment is able to quench all the fuel assemblies in the core.

Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-13

4.2 Analysis Results The R. E. Ginna Region II analysis was performed in accordance with the NRC-approved methodology in Reference 1 and the methodology extension in Reference 13, with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The analysis was performed 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 (leaving 1 LHSI and 2 HHSI pumps available) is assumed in the analysis as identified in Table 1. The results of the R. E. Ginna Region II LOOP and OPA uncertainty analyses are summarized in Table 7. The sampled decay heat uncertainty multipliers for the Region II analysis cases are provided in Table 10.

Table 9 contains a sequence of events for the transient that produced the more limiting analysis PCT result relative to the offsite power assumption. Figures 14 through 27 illustrate the key response 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-TF2 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, 3, and 4) and the mass and energy releases, and transfers the pressure back to the thermal-hydraulic code as a boundary condition at the break, consistent with the methodology in Reference 1. The containment model for COCO calculates a conservatively low containment pressure, including the effects of all the installed pressure reducing systems and processes such as assuming all trains of containment spray are operable and assuming fan cooler operation. The containment backpressure for the transient that produced the analysis PCT result is provided in Figure 21.

Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-14

5.0 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)(1) The analysis PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level. Since the resulting PCT is less than 2,200°F, the analysis with the FSLOCA EM confirms that 10 CFR 50.46 acceptance criterion (b)(1), i.e., Peak Cladding Temperature does not exceed 2,200°F, is demonstrated.

The results are shown in Table 7 for R. E. Ginna.

(b)(2) The analysis MLO corresponds to a bounding estimate of the 95th 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 corrosion, 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 7 for R. E. Ginna.

(b)(3) The analysis CWO corresponds to a bounding estimate of the 95th percentile 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 criterion (b)(3), i.e., Core-Wide Oxidation does not exceed 1 percent, is demonstrated.

The results are shown in Table 7 for R. E. Ginna.

(b)(4) 10 CFR 50.46 acceptance criterion (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 criteria (b)(1) and (b)(2), and by assuring that fuel assembly grid deformation due to combined LOCA and seismic loads is specifically addressed. Criteria (b)(1) and (b)(2) have been met for R. E. Ginna as shown in Table 7.

It is discussed in Section 32.1 of the NRC-approved FSLOCA EM (Reference 1) that the effects of LOCA and seismic loads on the core geometry do not need to be considered unless fuel assembly grid deformation extends beyond the core periphery (i.e., deformation in a fuel assembly with no sides adjacent to the core baffle plates). Inboard grid deformation due to combined LOCA and seismic loads is not calculated to occur for R. E. Ginna.

(b)(5) 10 CFR 50.46 acceptance criterion (b)(5) requires that long-term 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 I and Region II presented in Table 7 for R. E. Ginna, it is concluded that R. E. Ginna complies with the criteria in 10 CFR 50.46.

Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-15

6.0 REFERENCES

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

2. Westinghouse Performance Analysis and Design Model (PAD5), 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, January 1974.

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

5. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting 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. Best Estimate Calculations of Emergency Core Cooling System Performance, Regulatory Guide 1.157, USNRC, May 1989.

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

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

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

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

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

13. Extension of FULL SPECTRUMTM LOCA (FSLOCATM) Evaluation Methodology to 2-loop Westinghouse Pressurized Water Reactors (PWRs) with Information to Satisfy Limitations and Conditions Specific to 2-loop Plant Types (Proprietary/Non-Proprietary), LTR-NRC-21-22, September 2021.

Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-16

Table 1. Plant Operating Range Analyzed and Key Parameters for R. E. Ginna Parameter As-Analyzed Value or Range 1.0 Core Parameters a) Core power 1811 MWt +/- 0% Uncertainty b) Fuel type 14x14, 422 Vantage +, Optimized ZIRLOTM Cladding, Integral Fuel Burnable Absorbers (IFBA) or Non-IFBA c) Maximum total core peaking factor (FQ),

including uncertainties 2.65 d) Maximum hot channel enthalpy rise peaking factor (FH), including uncertainties 1.75 e) Axial flux difference (AFD) band at 100%

power

-14% to +8%

f)

Maximum transient operation fraction 0.4 2.0 Reactor Coolant System Parameters a) Thermal design flow (TDF) 85,100 gpm/loop b) Vessel average temperature (TAVG) 560.6°F TAVG 580.0°F c) Pressurizer pressure (PRCS) 2190 psia PRCS 2310 psia d) Reactor coolant pump (RCP) model and power Model 93, No Weir, 6000 hp 3.0 Containment Parameters a) Containment modeling Region I: Constant pressure equal to initial containment pressure Region II: Calculated for each transient using transient-specific mass and energy releases and the information in Tables 2, 3, and 4 4.0 Steam Generator (SG) and Secondary Side Parameters a) Steam generator tube plugging level 10%

b) Main steam safety valve (MSSV) nominal set pressures, uncertainty and accumulation Table 6 c) Main feedwater temperature Nominal (412.5°F) d) Auxiliary feedwater temperature Nominal (104°F) e) Auxiliary feedwater flow rate 85 gpm/SG Optimized ZIRLO is a trademark of Westinghouse 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.

Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-17

Table 1. Plant Operating Range Analyzed and Key Parameters for R. E. Ginna Parameter As-Analyzed Value or Range 5.0 Safety Injection (SI) Parameters a) Single failure configuration ECCS: Loss of one train of pumped ECCS (1 LHSI and 2 HHSI available)

Region II containment pressure: All containment spray trains are available b) Safety injection temperature (TSI) 50°F TSI 104°F c) Low pressurizer pressure safety injection safety analysis limit 1715 psia d) Initiation delay time from low pressurizer pressure SI setpoint to full SI flow Low Head SI:

19 seconds (OPA) or 30 seconds (LOOP)

High Head SI:

21 seconds (OPA) or 32 seconds (LOOP) e) Safety injection flow Table 5 6.0 Accumulator Parameters a) Accumulator temperature (TACC) 60°F TACC 125°F b) Accumulator water volume (VACC) 1090 ft3 VACC 1140 ft3 (See Note 1) c) Accumulator pressure (PACC) 714.7 psia PACC 804.7 psia d) Accumulator boron concentration 2100 ppm 7.0 Reactor Protection System Parameters a) Low pressurizer pressure reactor trip signal processing time 2.0 seconds (See Note 2) b) Low pressurizer pressure reactor trip setpoint 1806 psia Notes:

1. The Region I and Region II uncertainty analyses were originally executed with an accumulator water volume range of 1110 ft3 to 1140 ft3, but the plant intended range was 1090 ft3 to 1140 ft3. A full Region II uncertainty analysis re-execution was performed to incorporate the intended range. The Region II results presented herein are from the re-executed Region II uncertainty analysis with the intended accumulator water volume range of 1090 ft3 to 1140 ft3. The difference in the accumulator water volume range was qualitatively evaluated for the Region I uncertainty analysis as a post-analysis evaluation. The evaluation concluded that the updated accumulator water volume range of 1090 ft3 to 1140 ft3 has a negligible impact and is acceptable for the Ginna Region I uncertainty analysis.

2. The Region I uncertainty analysis was executed using a low pressurizer pressure reactor trip signal processing time of 1.8 seconds, but the plant intended input was 2.0 seconds. This difference was qualitatively evaluated as a post-analysis evaluation with respect to the Region I uncertainty analysis.

The evaluation concluded that the updated low pressurizer pressure reactor trip signal processing time of 2.0 seconds has a negligible impact and is acceptable for the Ginna Region I uncertainty analysis. The Region II uncertainty analysis does not credit control rod insertion, so this input is not used in the Region II uncertainty analysis. Therefore, the change to the low pressurizer pressure reactor trip signal processing time has no impact on the Region II transients.

Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-18

Table 2. Containment Data Used for Region II Calculation of Containment Pressure for R. E. Ginna Parameter Value Maximum containment net free volume 1,066,000 ft3 Minimum initial containment temperature at full power operation 60°F Refueling water storage tank (RWST) temperature for containment spray (TRWST) 50°F TRWST 104°F Minimum RWST temperature for broken loop spilling SI 50°F Minimum containment outside air / ground temperature

-20°F Minimum initial containment pressure at normal full power operation 14.5 psia Minimum containment spray pump initiation delay from containment high pressure signal time 9 seconds (OPA) or 14 seconds (LOOP)

Maximum containment spray flow rate from all pumps 3600 gpm Maximum number of containment fan coolers in operation during LOCA transient 4 Minimum fan cooler initiation delay time 0 seconds (conservative value for both OPA and LOOP)

Maximum heat removal rate per fan cooler as a function of containment temperature Table 3 Maximum number of containment venting lines (including purge lines, pressure relief lines or any others) which can be OPEN at onset of transient at full power operation 0

Containment walls / heat sink properties Table 4 SI spilling flows 465 gpm Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-19

Table 3. Fan Cooler Performance Data Used for Region II Calculation of Containment Pressure for R. E. Ginna Containment Temperature (°F)

Heat Removal Rate per Fan Cooler (BTU/sec) 30 0.0 220 32472 286 50500 Table 4. Containment Heat Sink Data Used for Region II Calculation of Containment Pressure for R. E. Ginna Wall Area (ft2)

Thickness (ft)

Material 1

36285.0 0.001583 0.104167 0.03125 0.5 3.0 Stainless Steel Insulation Carbon Steel Concrete Concrete 2

12370.0 0.03125 0.5 2.0 Carbon Steel Concrete Concrete 3

7230.0 0.5 1.5 0.02083 2.0 Concrete Concrete Carbon Steel Concrete 4

2480.0 0.5 1.5 0.02083 1.0 Concrete Concrete Carbon Steel Concrete 5

400.0 0.5 1.5 0.02083 1.0 Concrete Concrete Carbon Steel Concrete 6

6170.0 0.02083 0.5 2.0 Stainless Steel Concrete Concrete 7

1260.0 0.02083 0.5 1.5 Stainless Steel Concrete Concrete Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-20

Table 4. Containment Heat Sink Data Used for Region II Calculation of Containment Pressure for R. E. Ginna Wall Area (ft2)

Thickness (ft)

Material 8

6750.0 0.02083 0.5 2.0 Stainless Steel Concrete Concrete 9

20740.0 0.5 0.75 Concrete Concrete 10 10640.0 0.25 Concrete 11 13000.0 0.5 0.5 Concrete Concrete 12 4000.0 0.06167 Carbon Steel 13 1260.0 0.03917 Carbon Steel 14 8440.0 0.02167 Carbon Steel 15 2380.0 0.02542 Carbon Steel 16 470.0 0.04167 Carbon Steel 17 4810.0 0.0625 Carbon Steel 18 6780.0 0.0625 Carbon Steel 19 2060.0 0.5 1.5 Concrete Concrete 20 7000.0 0.01042 Carbon Steel Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-21

Table 5. Safety Injection Flow Used for Region I and II Calculation for R. E. Ginna Pressure (psia)

High Head Safety Injection (HHSI) Flow (gpm)

Low Head Safety Injection (LHSI) Flow (gpm) 14.7 300 1200 20 300 1176 60 300 980 80 300 866 100 300 735 120 300 570 140 300 220 140.01 300 0

214.7 300 0

314.7 300 0

414.7 300 0

514.7 300 0

614.7 289 0

714.7 273 0

814.7 253 0

914.7 229 0

1014.7 201 0

1114.7 167 0

1214.7 125 0

1314.7 62 0

1314.71 0

0 Table 6. Steam Generator Main Steam Safety Valve Parameters for R. E. Ginna Stage Set Pressure (psig)

Uncertainty (%)

Accumulation (%)

1 1085 1

3 2

1140 1.4 3

3 1140 1.4 3

4 1140 1.4 3

Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-22

Table 7. R. E. Ginna Analysis Results with the FSLOCA EM Outcome Region I Value Region II Value (OPA)

Region II Value (LOOP) 95/95 PCT 911°F 1957°F 1974°F 95/95 MLO 6.60%

7.33%

7.75%

95/95 CWO 0.00%

0.41%

0.45%

Table 8. R. E. Ginna Sequence of Events for the Region I Analysis PCT Case Event Time after Break (sec)

Start of Transient 0.0 Reactor Trip Signal 7.1 Safety Injection Signal 8.7 Safety Injection Begins 29.7 Top of Core Uncovered 348 Accumulator Injection Begins 400 PCT Occurs 417 Top of Core Recovered 428 Table 9. R. E. Ginna Sequence of Events for the Region II Analysis PCT Case Event Time after Break (sec)

Start of Transient 0.0 Safety Injection Signal 4.1 Fuel Rod Burst Occurs 4.5 Accumulator Injection Begins 9.5 End of Blowdown 15.0 PCT Occurs 32.1 Safety Injection Begins 34.2 Accumulator Empty 50.5 All Rods Quenched 156 Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-23

Table 10. R. E. Ginna Sampled Value of Decay Heat Uncertainty Multiplier, DECAY_HT, for Region I and Region II Analysis Cases Region Case DECAY_HT (units of )

DECAY_HT (absolute units)*

Region I PCT

+ 0.9369 4.83%

MLO

+ 1.6197 8.36%

Region II (OPA)

PCT

+ 0.1728 0.87%

MLO

+ 1.1904 6.08%

CWO

+ 1.1912 5.97%

Region II (LOOP)

PCT

+ 1.3521 6.96%

MLO

+ 1.1904 6.08%

CWO

+ 1.1477 5.73%

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

Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-24

Figure 1: R. E. Ginna Break Flow Void Fraction for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-25

Figure 2: R. E. Ginna Total Safety Injection Flow (not including Accumulator Injection Flow) and Total Break Flow for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-26

Figure 3: R. E. Ginna RCS Pressure for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-27

Figure 4: R. E. Ginna Hot Assembly Two-Phase Mixture Level (Relative to Bottom of Active Fuel) for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-28

Figure 5: R. E. Ginna Peak Cladding Temperature for all Rods for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-29

Figure 6: R. E. Ginna Vapor Mass Flow Rate through the Crossover Legs for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-30

Figure 7: R. E. Ginna Core Collapsed Liquid Levels (Relative to Bottom of Active Fuel) for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-31

Figure 8: R. E. Ginna Accumulator Injection Flow for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-32

Figure 9: R. E. Ginna Vessel Fluid Mass for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-33

Figure 10: R. E. Ginna Steam Generator Secondary Side Pressure for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-34

Figure 11: R. E. Ginna Normalized Core Power Shapes for the Region I Analysis PCT Case Note: The localized power decreases occur at grid elevations.

Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-35

Figure 12: R. E. Ginna Relative Core Power for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-36

Figure 13: R. E. Ginna Vapor Temperature and Void Fraction at Core Outlet (Hot Assembly Channel) for the Region I Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-37

Figure 14: R. E. Ginna Peak Cladding Temperature for all Rods for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-38

Figure 15: R. E. Ginna Peak Cladding Temperature Elevation (Relative to Bottom of Active Fuel) for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-39

Figure 16: R. E. Ginna Break Mass Flow Rate for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-40

Figure 17: R. E. Ginna Lower Plenum Collapsed Liquid Level (Relative to Inside Bottom of Vessel) for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-41

Figure 18: R. E. Ginna Vapor Mass Flow Rate at the Top Cell Face of the Core Average Channel not Under Guide Tubes for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-42

Figure 19: R. E. Ginna RCS Pressure for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-43

Figure 20: R. E. Ginna Accumulator Injection Flow per Loop for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-44

Figure 21: R. E. Ginna Containment Pressure for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-45

Figure 22: R. E. Ginna Vessel Fluid Mass for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-46

Figure 23: R. E. Ginna Collapsed Liquid Level for Each Core Channel (Relative to Bottom of Active Fuel) for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-47

Figure 24: R. E. Ginna Average Downcomer Collapsed Liquid Level (Relative to 31.195 inches Above the Inside Bottom of the Vessel) for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-48

Figure 25: R. E. Ginna Safety Injection Flow per Loop (not including Accumulator Injection Flow) for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-49

Figure 26: R. E. Ginna Normalized Core Power Shapes for the Region II Analysis PCT Case Note: The localized power decreases occur at grid elevations.

Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-50

Figure 27: R. E. Ginna Relative Core Power for the Region II Analysis PCT Case Westinghouse Non-Proprietary Class 3 RGE-22-4 NP-Attachment NP-51

ATTACHMENT 5 License Amendment Request R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 Docket No. 50-244 Proposed Technical Specification Changes (Mark-Ups)

I Reporting Requirements 5.6

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

(Methodology for 2.1, LCO 3.1.1, LCO 3.1.3, LCO 3.1.5, LCO 3.1.6, LCO 3.2.1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)

2.

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

3.

WCAP-10216-P-A, Rev. 1 A, "Relaxation of Constant Axial Offset Control I FQ Surveillance Technical Specification,"

February 1994.

(Methodology for LCO 3.2.1 and LCO 3.2.3.)

4.

WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995.

(Methodology for LCO 3.2.1.)

5.

WCAP 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.

(Methodology for LCO 3.4.1 when using ATOP.)

6.

WCAP-10054-P-AandWCAP-10081-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.

(Methodology for LCO 3.2.1.)

7.

WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and COSI Condensation Model," July 1997.

(Methodology for LCO 3.2.1)

8.

WCAP-11145-P-A, "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study With the NOTRUMP Code,"

October 1986.

(Methodology for LCO 3.2.1)

9.

WCAP-10079-P-A, "NOTRUMP - A Nodal Transient Small Break and General Network Code," August 1985.

(Methodology for LCO 3.2.1)

10.

WCAP-87 45-P-A, "Design Basis for the Thermal Overpower Delta T and Thermal Overtemperature Delta T Trip Functions," September 1986.

(Methodology for LCO 3.3.1.)

R.E. Ginna Nuclear Power Plant 5.6-3 Amendment 125 to LTR-LIS-21-166, Revision 0 Page 3 of 28

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

November 2016.

6. Not used.

7. Not used.

8. Not used.

9. Not used.

ATTACHMENT 6 License Amendment Request R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 Docket No. 50-244 Proposed Technical Specification Bases Changes (Mark-Ups)

FQ(Z)

B 3.2.1 B 3.2.1-2 Revision 42 R.E. Ginna Nuclear Power Plant Core monitoring and control under non-equilibrium conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and bank insertion, sequence, and overlap limits.

APPLICABLE This LCO precludes core power distributions that violate the following fuel SAFETY design criteria:

ANALYSES

a.

During a large break loss of coolant accident (LOCA), the peak cladding temperature (PCT) must not exceed 2200ºF (Ref. 1),

b.

During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition,

c.

During an ejected rod accident, the energy deposition to the fuel must not exceed 200 cal/gm (Ref. 2), and

d.

The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).

Limits on FQ(Z) ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid. Other criteria must also be met (e.g., maximum cladding oxidation, maximum hydrogen generation, coolable geometry, and long term cooling). However, the peak cladding temperature is typically most limiting.

FQ(Z) limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the FQ(Z) limit assumed in safety analyses for other postulated accidents. Therefore, this LCO provides conservative limits for other postulated accidents FQ(Z) satisfies Criterion 2 of 10 CFR 50.36.

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B 3.2.1-4 Revision 78 R.E. Ginna Nuclear Power Plant FQ(Z)

B 3.2.1 The expression for FQ W(Z) is:

FQ W(Z)=FQ C(Z)W(Z) where W(Z) is a cycle dependent function that accounts for power distribution transients encountered during normal operation. W(Z) is included in the COLR. The FQ W(Z) is calculated at equilibrium conditions.

The FQ(Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200°F during either a large or small break LOCA.

This LCO requires operation within the bounds assumed in the safety analyses. Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation that it can stay within the LOCA FQ(Z) limits. If FQ C(Z) cannot be maintained within the LCO limits, reduction of the core power is required and if FQ W(Z) cannot be maintained within the LOCA limits, reduction of the AFD limits is required. Note that sufficient reduction of the AFD limits will also result in a reduction of the core power.

Violating the LCO limits for FQ(Z) produces unacceptable consequences if a design basis event occurs while FQ(Z) is outside its specified limits.

APPLICABILITY The FQ(Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses.

Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power.

ACTIONS A.1 Reducing THERMAL POWER by 1% RTP for each 1% by which FQ C(Z) exceeds its limit, maintains an acceptable absolute power density FQ C(Z) is FQ M(Z) multiplied by a factor accounting for manufacturing tolerances and measurement uncertainties. FQ M(Z) is the measured value of FQ(Z). The Completion Time of 15 minutes provides an acceptable time to reduce power in an orderly manner and without allowing the plant to remain in an unacceptable condition for an extended period of time. The maximum allowable power level initially determined to LTR-LIS-21-166, Revision 0 Page 5 of 28 ensure that the 10 CFR 50.46 acceptance criteria are met

B3.2.2-2 Revision 21 FN'H B 3.2.2 R.E. Ginna Nuclear Power Plant The COLR provides peaking factor limits that ensure that the design basis value for departure from nucleate boiling ratio (DNBR) is met for normal operation, operational transients, and any transient condition arising from events of moderate frequency. The DNB design basis precludes DNB and is met by limiting the minimum local DNB heat flux ratio. All DNB limited transient events are assumed to begin with an FN'H value that satisfies the LCO requirements.

Operation outside the LCO limits may produce unacceptable consequences if a DNB limiting event occurs. The DNB design basis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant.

APPLICABLE Limits on FN'H preclude core power distributions that exceed the SAFETY following fuel design limits:

ANALYSES

a.

During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hottest fuel rod in the core does not experience a DNB condition;

b.

During a large break loss of coolant accident (LOCA), peak cladding temperature (PCT) must not exceed 2200°F (Ref. 1);

c.

During an ejected rod accident, the energy deposition to the fuel will be below 200 cal/gm (Ref. 2); and

d.

The control rods must be capable of shutting down the reactor with a minimum required SOM with the highest worth control rod stuck fully withdrawn (Ref. 3).

For transients that may be DNB limited, the Reactor Coolant System flow and FN'H are the core parameters of most importance. The limits on FN'H ensure that the DNB design basis is met for normal operation, operational transients, and any transients arising from events of moderate frequency (i.e., Condition 1 events as described in Reference 4). The DNB design basis is met by limiting the minimum DNBR to the 95/95 DNB criterion.

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FN

¨H B 3.2.2 B 3.2.2-3 Revision 21 R.E. Ginna Nuclear Power Plant The allowable FN'H limit increases with decreasing power level. This functionality in FN'H is included in the analyses that provide the Reactor Core Safety Limits (SLs) of SL 2.1.1. Therefore, any DNB events in which the calculation of the core limits is modeled implicitly use this variable value of FN'H in the analyses. Likewise, all transients that may be DNB limited are assumed to begin with an initial FNAH as a function of power level defined by the COLR limit equation.

The LOCA safety analysis indirectly models FN'H as an input parameter.

The Nuclear Heat Flux Hot Channel Factor (FQ(Z)) and the axial peaking factors are inserted directly into the LOCA safety analyses that verify the acceptability of the resulting peak cladding temperature (Ref. 1).

The fuel is protected in part by Technical Specifications, which ensure that the initial conditions assumed in the safety and accident analyses remain valid. The following LCOs ensure this: LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," LCO 3.1.6, "Control Bank Insertion Limits," LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FN'H)," and LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))."

FNAH is measured periodically using the movable incore detector system.

Measurements are generally taken with the core at, or near, steady state conditions. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on AFD, QPTR, and Bank Insertion, Sequence and Overlap Limits.

FN'H satisfies Criterion 2 of the NRC Policy Statement.

LCO FN'H shall be maintained within the limits of the relationship provided in the COLR.

The FN'H limit identifies the coolant flow channel with the maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for DNB.

The limiting value of FN'H, described by the equation contained in the COLR, is the design radial peaking factor used in the plant safety analyses.

to LTR-LIS-21-166, Revision 0 Page 7 of 28 The Nuclear Enthalpy Rise Hot Channel Factor (FNH),

the Nuclear Heat Flux Hot Channel Factor (FQ(Z)),

and the axial peaking factors are supported by the LOCA safety analyses that verify compliance with the 10 CFR 50.46 acceptance criteria (Ref. 1).

FN

¨H B 3.2.2 B 3.2.2-4 Revision 21 R.E. Ginna Nuclear Power Plant A power multiplication factor in this equation includes an additional margin for higher radial peaking from reduced thermal feedback and greater control rod insertion at low power levels. The limiting value of FN'H is allowed to increase 0.3% for every 1% RTP reduction in THERMAL POWER APPLICABILITY The FN'H limits must be maintained in MODE 1 to prevent core power distributions from exceeding the fuel design limits for DNBR and PCT.

Applicability in other modes is not required because there is neither sufficient stored energy in the fuel nor sufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power. Specifically, the design bases events that are sensitive to FN'H in MODES 2, 3, 4, and 5 have significant margin to DNB, and therefore, there is no need to restrict FN'H in these modes.

ACTIONS Reducing THERMAL POWER by 2 1% for each 1% by which FN'H exceeds its limit maintains an acceptable DNBR margin. When the FN'H limit is exceeded, the DNBR limit is not likely violated in steady state operation, because events that could significantly perturb the FN'H value (e.g., static control rod misalignment) are considered in the safety analyses. However, the DNBR limit may be violated if a DNB limiting event occurs. Reducing THERMAL POWER increases the DNB margin.

The Completion Time of 15 minutes provides an acceptable time to reduce power in an orderly manner and without allowing the plant to remain in an unacceptable condition for an extended period of time.

A reduction of the Power Range Neutron Flux-High trip setpoints by 1%

for each 1% by which FN'H exceeds its specified limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions and ensures that continuing operation remains at an acceptable low power level with adequate DNBR margin.

This reduction shall be made as follows, given that the FN'H limit is exceeded by 3% and the Power Range Neutron Flux-High setpoint is 108%, the Power Range Neutron Flux-High setpoint must be reduced by at least 3% to 105%. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient, considering the small likelihood of a severe transient in this period, and the preceding prompt reduction in THERMAL POWER in accordance with required action A.1.

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QPTR B 3.2.4 B 3.2.4-1 Revision 42 R.E. Ginna Nuclear Power Plant B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

BASES BACKGROUND The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing, after refueling, and periodically during power operation.

Quadrant Power Tilt is a core tilt that is measured with the use of the excore power range flux detectors. A core tilt is defined as the ratio of maximum to average quadrant power. The QPTR is defined as the ratio of the highest average nuclear power in any quadrant to the average nuclear power in the four quadrants. Limiting the QPTR prevents radial xenon oscillations and will indicate any core asymmetries.

The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," and LCO 3.1.6, "Control Bank Insertion Limits," provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses.

APPLICABLE Limits on QPTR preclude core power distributions that violate the SAFETY following fuel design criteria:

ANALYSES

a.

During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition;

b.

During a large break loss of coolant accident (LOCA), the peak cladding temperature (PCT) must not exceed 2200°F (Ref. 1);

c.

During an ejected rod accident, the energy deposition to the fuel will be below 200 cal/gm (Ref. 2); and The control rods must be capable of shutting down the reactor with a minimum required SOM with the highest worth control rod stuck fully withdrawn (Ref. 3).

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B 3.3.2-4 Revision 42 ESFAS Instrumentation B 3.3.2 R.E. Ginna Nuclear Power Plant Generally, three or four channels of process control equipment are used for the signal processing of plant parameters measured by the field transmitters and sensors. If a parameter is used only for input to the protection circuits, three channels with a two-out-of-three logic are typically sufficient to provide the required reliability and redundancy. If one channel fails in a direction that would not result in a partial Function trip, the Function can still be accomplished with a two-out-of-two logic. If one channel fails in a direction that a partial Function trip occurs, a trip will not occur unless a second channel fails or trips in the remaining one-out-of-two logic.

If a parameter is used for input to the protection system and a control function, four channels with a two-out-of-four logic are typically sufficient to provide the required reliability and redundancy.

This ensures that the circuit is able to withstand both an input failure to the control system, which may then require the protection function actuation, and a single failure in the other channels providing the protection function actuation. Therefore, a single failure will neither cause nor prevent the protection function actuation. These requirements are described in IEEE-279-1971 (Ref. 5).

The actuation of ESF components is accomplished through master and slave relays. The protection system energizes the master relays appropriate for the condition of the plant. Each master relay then energizes one or more slave relays, which then cause actuation of the end devices.

APPLICABLE Each of the analyzed accidents can be detected by one or more ESFAS SAFETY Functions. One of the ESFAS Functions is the primary actuation signal ANALYSES, for that accident. An ESFAS Function may be the primary actuation LCO, AND signal for more than one type of accident. An ESFAS Function may also APPLICABILITY be a secondary, or backup, actuation signal for one or more other accidents. For example, SI-Pressurizer Pressure-Low is a primary actuation signal for small break loss of coolant accidents (LOCAs) and a backup actuation signal for steam line breaks (SLBs) outside containment. Functions such as manual initiation, not specifically credited in the accident safety analysis, are qualitatively credited in the safety analysis and the NRC staff approved licensing basis for the plant. These Functions may provide protection for conditions that do not require dynamic transient analysis to demonstrate Function performance. These Functions may also serve as anticipatory actions to Functions that were credited in the accident analysis (Ref. 4).

This LCO requires all instrumentation performing an ESFAS Function to be OPERABLE. A channel is considered OPERABLE when:

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B 3.3.2-6 Revision 42 ESFAS Instrumentation B 3.3.2 R.E. Ginna Nuclear Power Plant and the previous as-left trip setpoint does not exceed the COT Acceptance Criteria, the bistable is considered OPERABLE.

The Nominal Trip Setpoint is the value at which the bistable is set and is the expected value to be achieved during calibration. The Nominal Trip Setpoint value ensures the LSSS and the safety analysis limits are met for surveillance interval selected when a channel is adjusted based on stated channel uncertainties. Any bistable is considered to be properly adjusted when the as-left trip setpoint is within the tolerance band assumed in the uncertainty analysis. The bistable is still operable even if the as-left trip setpoint is non-conservative with respect to the LSSS provided that the as-left trip setpoint is within the established calibration tolerance band as specified in the Ginna Instrument Setpoint Methodology.

Trip setpoints consistent with the requirements of the LSSS ensure that SLs are not violated during DBAs (and that the consequences of DBAs will be acceptable, providing the unit is operated from within the LCOss at the onset of the DBA and the equipment functions as designed).

The required channels of ESFAS instrumentation provide plant protection in the event of any of the analyzed accidents. ESFAS protection functions provided in Table 3.3.2-1 are as follows:

1.

Safety Injection Safety Injection (SI) provides two primary functions:

1.

Primary side water addition to ensure maintenance or recovery of reactor vessel water level (coverage of the active fuel for heat removal, clad integrity, and for limiting peak clad temperature to < 2200ºF); and

2.

Boration to ensure recovery and maintenance of SDM (keff

< 1.0).

These functions are necessary to mitigate the effects of high energy line breaks (HELBs) both inside and outside of containment.

The SI signal is also used to initiate other Functions such as:

Containment Isolation; Containment Ventilation Isolation; Reactor Trip; Feedwater Isolation; and Start of motor driven auxiliary feedwater (AFW) pumps. to LTR-LIS-21-166, Revision 0 Page 11 of 28 compliance with the 10 CFR 50.46 acceptance criteria (Ref. 11)

ESFAS Instrumentation B 3.3.2 B 3.3.2-35 Revision 102 R.E. Ginna Nuclear Power Plant REFERENCES

1.

Atomic Industrial Forum (AIF) GDC 15, Issued for Comment July 10, 1967.

2.

UFSAR, Chapter 7.

3.

UFSAR, Chapter 6.

4.

UFSAR, Chapter 15.

5.

IEEE-279-1971.

6.

EP-3-S-0505, "Instrument Setpoint/Loop Accuracy Calculation Methodology".

7.

WCAP-10271-P-A, Supplement 2, Rev. 1, June 1990.

8.

"Power Range Nuclear Instrumentation System Bypass Test Instrumentation for R. E. Ginna," WCAP-18298-P, September 2017.

9.

WCAP-14333-P-A, Revision 1, October 1998.

10.

Ginna PRA Analysis for ESFAS/RTS AOT Extension, G1-LAR-005.

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11. 10 CFR 50.46.

Accumulators B 3.5.1 B 3.5.1-1 Revision 44 R.E. Ginna Nuclear Power Plant B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1 Accumulators BASES BACKGROUND The functions of the ECCS accumulators are to supply water to the reactor vessel during theblowdown phase of a large break loss of coolant accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for a small break LOCA.

The blowdown phase of a large break LOCA is the initial period of the transient during which the RCS departs from equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. The reactor coolant inventory is vacating the core during this phase through steam flashing and ejection out through the break. The blowdown phase of the transient ends when the RCS pressure falls to a value approaching that of the containment atmosphere.

In the refill phase of a LOCA, which immediately follows the blowdown phase, the core is essentially in adiabatic heatup. The balance of accumulator inventory is available to reflood the core and help fill voids in the lower plenum and reactor vessel downcomer so as to establish a recovery level at the bottom of the core.

The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. The level transmitters for the accumulators measure the level over a 14" span for the corresponding 0-100% level indicated on the main control board. The accumulators are passive components, since no operator or control actions are required in order for them to perform their function. Internal accumulator tank pressure is sufficient to discharge the accumulator contents to the RCS, if RCS pressure decreases below the accumulator pressure.

Each accumulator is piped into an RCS cold leg via an accumulator line and is isolated from the RCS by a motor operated isolation valve and two check valves in series (see Figure B 3.5.2-1a). The motor operated isolation valves (841 and 865) are maintained open with AC power removed under administrative control when pressurizer pressure is

> 1600 psig. This feature ensures that the valves meet the single failure criterion of manually-controlled electrically operated valves per Branch Technical Position (BTP) ICSB-18 (Ref. 1). This is also discussed in References 2 and 3. to LTR-LIS-21-166, Revision 0 Page 13 of 28 large break Initial accumulator inventory which is injected into the reactor vessel is lost out the break.

help fill the lower plenum and reactor vessel downcomer, so as to establish a liquid level at the bottom of the core and initiate reflood of the core with the pumped SI liquid.

Accumulators B 3.5.1 B 3.5.1-2 Revision 44 R.E. Ginna Nuclear Power Plant The accumulator size, water volume, and nitrogen cover pressure are selected so that one of the two accumulators is sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that one accumulator is adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.

APPLICABLE SAFETY ANALYSES The accumulators are assumed OPERABLE in both the large and small break LOCA analyses at full power (Ref. 4). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits.

In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a large break LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. The large break LOCA also considers a case with offsite power available. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.

The largest break area considered for a large break LOCA is a double ended guillotine break at the discharge of the reactor coolant pump.

During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure. As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for SI signal generation, the diesels starting, and the pumps being loaded and delivering full flow.

During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA.

The worst case small break LOCA analyses also assume a time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators and safety injection pumps both play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators continues to decrease until they are not required to LTR-LIS-21-166, Revision 0 Page 14 of 28 transient must be considered, and in the RCS cold leg (for loss of offsite power assumption) for a large break LOCA in the modeling is assumed to inject into the reactor coolant system intermediate into small break LOCA

B 3.5.1-3 Revision 44 R.E. Ginna Nuclear Power Plant Accumulators B 3.5.1 and the safety injection pumps become solely responsible for terminating the temperature increase.

This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 5) will be met following a LOCA:

a.

Maximum fuel element cladding temperature is 2200ºF;

b.

Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation;

c.

Maximum hydrogen generation from a zirconium water reaction is 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; and

d.

Core is maintained in a coolable geometry.

Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46.

For the small break LOCA analysis, a nominal contained accumulator water volume is used. The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged. For small breaks, an increase in water volume is a peak clad temperature penalty due to the reduced gas volume. A peak clad temperature penalty is an assumed increase in the calculated peak clad temperature due to a change in an input parameter.

For large breaks, an increase in water volume can be either a peak clad temperature penalty or benefit, depending on downcomer filling and subsequent spill through the break during the core reflooding portion of the transient. The large break analysis uses a range of accumulator water volumes consistent with the approved methodology. The large break LOCA analysis also considers the line water volume, however, this volume is not ranged.

The minimum boron concentration limit is used in the post LOCA sump boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the time frame in which boron precipitation is addressed post LOCA. The maximum boron concentration limit is based on the coldest expected temperature of the accumulator water volume and on chemical to LTR-LIS-21-166, Revision 0 Page 15 of 28 There is a high level of probability that the peak does not exceed There is a high level of probability that the maximum does not exceed There is a high level of probability that the maximum does not exceed large break LOCA and the recovery phase of a small break LOCA The small and large break LOCA analyses use of 1090 ft3 to 1140 ft3 (Ref. 11)

The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged.

Both small and large break LOCA analyses use a nominal accumulator line water volume from the accumulator to the check valve.

B 3.5.1-4 Revision 44 Accumulators B 3.5.1 R.E. Ginna Nuclear Power Plant effects resulting from operation of the ECCS and the Containment Spray (CS) System. The maximum value of 3050 ppm would not create the potential for boron precipitation in the accumulator assuming a containment temperature of 60ºF (Ref. 6). Analyses performed to address 10 CFR 50.49 (Ref. 7) assumed a chemical spray solution resulting from 2550-3050 ppm boron concentration in the accumulator and 2750-3050 ppm boron concentration in the RWST (Ref. 6). The chemical spray solution impacts sump pH and the resulting effect of chloride and caustic stress corrosion on mechanical systems and components. The sump pH also affects the rate of hydrogen generation within containment due to the interaction of CS and sump fluid with aluminum components.

The small break LOCA analysis is performed at the minimum nitrogen cover pressure, since sensitivity analyses have demonstrated that higher nitrogen cover pressure results in a computed peak clad temperature benefit. The large break LOCA analysis considers a range of accumulator nitrogen cover pressures consistent with the approved methodology. The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation at 800 psig, and ultimately preserves accumulator integrity.

The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Refs. 8 and 9).

The accumulators satisfy Criterion 3 of the NRC Policy Statement.

LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Two accumulators are required to ensure that 100% of the contents of one accumulator will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than one accumulator is injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 5) could be violated.

For an accumulator to be considered OPERABLE, the motor-operated isolation valve must be fully open (see Figure B 3.5.2-1a), power removed above 1600 psig, and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met. to LTR-LIS-21-166, Revision 0 Page 16 of 28 and analyses use of 700 psig to 790 psig (Ref. 11) large break LOCA and the recovery phase of a small break LOCA

B 3.5.1-5 Revision 44 Accumulators B 3.5.1 R.E. Ginna Nuclear Power Plant APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure > 1600 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist.

This LCO is only applicable at pressures > 1600 psig. At pressures 1600 psig, the rate of RCS blowdown is such that the ECCS pumps can provide adequate injection to ensure that peak clad temperature remains below the 10 CFR 50.46 (Ref. 5) limit of 2200ºF.

In MODE 3, with RCS pressure 1600 psig, and in MODES 4, 5, and 6, the accumulator motor operated isolation valves are closed to isolate the accumulators from the RCS. This allows RCS cooldown and depressurization without discharging the accumulators into the RCS or requiring depressurization of the accumulators.

ACTIONS A.1 If the boron concentration of one accumulator is not within limits, it must be returned to within the limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the ability to maintain subcriticality or minimum boron precipitation time may be reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on core subcriticality during reflood since the accumulator water volume is very small when compared to RCS and RWST inventory. Boiling of ECCS water in the core during reflood concentrates boron in the saturated liquid that remains in the core. In addition, current analysis techniques demonstrate that the accumulators are not expected to discharge following a large steam line break. Even if they do discharge, their impact is minor and not a design limiting event. Thus, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to return the boron concentration to within limits.

B.1 If one accumulator is inoperable for a reason other than boron concentration, the accumulator must be returned to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In this Condition, the required contents of one accumulator cannot be assumed to reach the core during a LOCA. Due to the severity of the consequences should a LOCA occur in these conditions, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time to open the valve, remove to LTR-LIS-21-166, Revision 0 Page 17 of 28 acceptance criteria are met

Accumulators B 3.5.1 B 3.5.1-8 Revision 44 R.E. Ginna Nuclear Power Plant REFERENCES

1.

Branch Technical Position (BTP) ICSB-18 "Application of the Single Failure Criterion to Manually-Controlled Electrically Operated Valves."

2.

Letter from D. M. Crutchfield, NRC, to J. E. Maier, RG&E,

Subject:

"SEP Topics VI-7.F, VII-3, VII-6, and VIII-2," dated June 24, 1981.

3.

Letter from R. A. Purple, NRC, to L. D. White, RG&E,

Subject:

"Issuance of Amendment 7 to Provisional Operating License No.

DPR-18," dated May 14, 1975.

4.

UFSAR, Section 6.3.

5.

10 CFR 50.46.

6.

UFSAR, Section 3.11.

7.

10 CFR 50.49.

8.

UFSAR, Section 6.2.

9.

UFSAR, Section 15.6.

10.

WCAP-15049-A, Rev. 1, April 1999 to LTR-LIS-21-166, Revision 0 Page 18 of 28

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

B 3.5.2-5 R.E. Ginna Nuclear Power Plant Revision 76, 100 ECCS - MODES 1, 2, and 3 B 3.5.2 The ECCS subsystems are actuated upon receipt of an SI signal. The actuation of safeguard loads is accomplished in a programmed time sequence. If offsite power is available, the safeguard loads start immediately in the programmed sequence. If offsite power is not available, the Engineered Safety Feature (ESF) buses shed normal operating loads and are connected to the emergency diesel generators (EDGs). Safeguard loads are then actuated in the programmed time sequence. The time delay associated with diesel starting, sequenced loading, and pump starting determines the time required before pumped flow is available to the core following a LOCA.

The active ECCS components, along with the passive accumulators, the RWST, and the Containment Sump, are covered in LCO 3.5.1, "Accumulators," and LCO 3.5.4, "Refueling Water Storage Tank (RWST),"

and LCO 3.6.7, "Containment Sump," and provide the cooling water necessary to meet AIF-GDC 44 (Ref. 8).

APPLICABLE SAFETY ANALYSES The LCO helps to ensure that the following acceptance criteria for the ECCS, established by 10 CFR 50.46 (Ref. 9), will be met following a LOCA:

a.

Maximum fuel element cladding temperature is 2200ºF; b.

Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation; c.

Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; d.

Core is maintained in a coolable geometry; and e.

Adequate long term core cooling capability is maintained.

The LCO also limits the potential for a post trip return to power following an SLB event and helps ensure that containment temperature limits are met post accident. to LTR-LIS-21-166, Revision 0 Page 19 of 28 There is a high level of probability that the maximum There is a high level of probability that the maximum There is a high level of probability that the peak does not exceed does not exceed does not exceed

B 3.5.2-6 Revision 76 ECCS - MODES 1, 2, and 3 B 3.5.2 R.E. Ginna Nuclear Power Plant Both ECCS subsystems are taken credit for in a large break LOCA event at full power (Refs. 6 and 10). This event establishes the requirement for runout flow for the ECCS pumps, as well as the maximum response time for their actuation. The SI pumps are credited in a small break LOCA event. This event establishes the flow and discharge head at the design point for the pumps. The SGTR and SLB events also credit the SI pumps. The OPERABILITY requirements for the ECCS are based on the following LOCA analysis assumptions:

a.

A large break LOCA event, with limiting offsite power assumptions and a single failure disabling one ECCS train (both EDG trains are assumed to operate for heat removal and spray systems in containment backpressure calculation); and

b.

A small break LOCA event, with a loss of offsite power and a single failure disabling one ECCS train.

During the blowdown stage of a LOCA, the RCS depressurizes as primary coolant is ejected through the break into the containment. The nuclear reaction is terminated either by moderator voiding during large breaks or control rod insertion for small breaks. Following depressurization, emergency cooling water is injected by the SI pumps into the cold legs, flows into the downcomer, fills the lower plenum, and refloods the core. The RHR pumps inject directly into the core barrel by upper plenum injection.

The effects on containment mass and energy releases are accounted for in appropriate analyses (Refs. 10 and 11). The LCO ensures that an ECCS train will deliver sufficient water to match boiloff rates quickly enough to minimize the consequences of the core being uncovered following a large LOCA. It also ensures that the SI pumps will deliver sufficient water and boron during a small LOCA to maintain core subcriticality. For smaller LOCAs, the SI pumps deliver sufficient fluid to maintain RCS inventory. For a small break LOCA, the steam generators continue to serve as the heat sink, providing part of the required core cooling.

The ECCS trains satisfy Criterion 3 of the NRC Policy Statement.

to LTR-LIS-21-166, Revision 0 Page 20 of 28 limiting offsite power assumptions

Containment Pressure B 3.6.4 B 3.6.4-2 Revision 78 R.E. Ginna Nuclear Power Plant For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. Therefore, for the reflood phase, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the containment pressure response in accordance with 10 CFR 50, Appendix K (Ref. 2). Service Water System (LCO 3.7.8) temperature plays an important role in both maximizing and minimizing containment pressure following a DBA response.

Containment pressure satisfies Criterion 2 of the NRC Policy Statement.

LCO Maintaining containment pressure at less than or equal to the LCO upper pressure limit ensures that, in the event of a DBA, the resultant peak containment accident pressure will remain below the containment design pressure. Maintaining containment pressure at greater than or equal to the LCO lower pressure limit ensures that the containment will not exceed the design negative differential pressure. However, the lower pressure limit specified for this LCO is set at a more limiting pressure to ensure continued cooling of the reactor coolant pump motors inside containment which are required to be OPERABLE for a large portion of MODES 1, 2, 3, and 4.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material into containment. Since maintaining containment pressure within limits is essential to ensure initial conditions assumed in the accident analyses are maintained, the LCO is applicable in MODES 1, 2, 3 and 4.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining containment pressure within the limits of the LCO is not required in MODE 5 or 6.

ACTIONS A.1 When containment pressure is not within the limits of the LCO, it must be restored to within these limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is greater than the ACTIONS of LCO 3.6.1, "Containment," which requires that containment be restored to LTR-LIS-21-166, Revision 0 Page 21 of 28 For these calculations WCAP-16996-P-A, Revision 1

Containment Pressure B 3.6.4 B 3.6.4-3 Revision 77 R.E. Ginna Nuclear Power Plant to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. However, due to the large containment free volume and limited size of the containment Mini-Purge System, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is allowed to restore containment pressure to within limits. This is justified by the low probability of a DBA during this time period.

B.1 and B.2 If containment pressure cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS SR 3.6.4.1 Verifying that containment pressure is within limits ensures that plant operation remains within the limits assumed in the containment analysis.

This verification should normally be performed using PI-944. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Calibration of PI-944 or other containment pressure monitoring devices should be performed in accordance with industry standards.

REFERENCES

1.

UFSAR, Section 6.2.1.2.

2.

10 CFR 50, Appendix K.

to LTR-LIS-21-166, Revision 0 Page 22 of 28 WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.

B 3.6.6-4 Revision 76 CS, CRFC and NaOH Systems B 3.6.6 R.E. Ginna Nuclear Power Plant APPLICABLE SAFETY ANALYSES The CS System and CRFC System limit the temperature and pressure that could be experienced following a DBA. The limiting DBAs considered are the LOCA and the SLB which are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. No two DBAs are assumed to occur simultaneously or consecutively. The postulated DBAs are analyzed with regard to containment ESF systems, assuming the worst case single active failure.

The operability requirements for the CS System and the CRFC System are based on the following LOCA long-term containment response assumptions:

a.

A LOCA mass and energy event with a loss of offsite power, and a single failure of an EDG, which causes the loss of one of two containment spray pumps and two of four fan coolers; and

b.

For the LOCA long-term containment response the containment spray is credited only during the injection phase of the transient and is terminated during the transition to sump recirculation.

The analysis and evaluation show that under the worst case scenario, the highest peak containment pressure is 59.7 psig and the peak containment temperature is greater than 350ºF (both experienced during an SLB). Both results meet the intent of the design basis. (See the Bases for LCO 3.6.4, "Containment Pressure," and LCO 3.6.5,"

Containment Temperature," for a detailed discussion.) The analyses and evaluations assume a plant specific power level of 1817MWt, one CS train and one containment cooling train operating, and initial (pre-accident) containment conditions of 125ºF and 1.0 psig. The analyses also assume a response time delayed initiation to provide conservative peak calculated containment pressure and temperature responses.

For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the containment pressure response in accordance with 10 CFR 50, Appendix K (Ref. 7).

The effect of an inadvertent CS actuation is not considered since there is no single failure, including the loss of offsite power, which results in a spurious CS actuation. to LTR-LIS-21-166, Revision 0 Page 23 of 28 WCAP-16996-P-A, Revision 1 cooling

CS, CRFC and NaOH Systems B 3.6.6 B 3.6.6-13 Revision 77 R.E. Ginna Nuclear Power Plant are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

REFERENCES

1.

Atomic Industry Forum (AIF) GDC 49, 52, 58, 59, 60, and 61, issued for comment July 10, 1967.

2.

Branch Technical Position MTEB 6-1, "pH For Emergency Coolant Water For PWRs."

3.

Letter from D. M. Crutchfield, NRC, to J. E. Maier, RG&E,

Subject:

"SEP Topic VI-7.B: ESF Automatic Switchover from Injection to Recirculation Mode, Automatic ECCS Realignment, Ginna," dated December 31, 1981.

4.

NUREG-0821.

5.

UFSAR, Section 6.3.

6.

UFSAR, Section 6.1.2.4.

7.

10 CFR 50, Appendix K.

8.

UFSAR, Section 6.2.1.2.

9.

UFSAR, Section 6.2.2.2.

10.

UFSAR, Section 6.5.

11.

UFSAR, Section 6.2.2.1.

12.

ASME Code for Operation and Maintenance of Nuclear Power Plants.

13.

Regulatory Guide 1.52, Revision 2.

14.

Design Analysis DA-NS-2001-087, Large Break LOCA Offsite and Control Room Doses.

to LTR-LIS-21-166, Revision 0 Page 24 of 28 WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.

AFW System B 3.7.5 B 3.7.5-3 Revision 75 R.E. Ginna Nuclear Power Plant

b.

Loss of MFW (with and without offsite power);

c.

Steam Line Break (SLB);

d.

Small break loss of coolant accident (LOCA);

e.

Steam generator tube rupture (SGTR); and AFW is also used to mitigate the effects of an ATWS event (which is a beyond design basis event) and external events (tornados and seismic events) all of which are not addressed by this LCO.

The AFW System design is such that any of the above DBAs can be mitigated using the preferred AFW System or SAFW System. For the FWLB and SLB, (items a and c), the worst case scenario is the loss of all three preferred AFW trains due to a HELB in the Intermediate or Turbine Building. For these events, the use of the SAFW System within 14.5 minutes is assumed by the accident analyses. Since a single failure must also be assumed in addition to the HELB, the capability of the SAFW System to supply flow to an intact SG could be compromised if the SAFW cross-tie or intact SG flowpath is not available. For HELBs within containment, use of either the SAFW System (within 14.5 minutes) or the AFW System (within 1 minute) to the intact SG is assumed.

For the SGTR events (item e), the accident analyses assume that one AFW train is available upon a SI signal or low-low SG level signal.

Additional inventory is being added to the ruptured SG as a result of the SGTR such that AFW flow is not a critical feature for this DBA.

The loss of MFW (item b) is a Condition 2 event (Ref. 3) which places limits on the response of the RCS from the transient (e.g., no challenge to the pressurizer power operated relief valves due to a water solid pressurizer is allowed). This analysis has been performed assuming no AFW flow is available until 1 minute with acceptable results. The most limiting small break LOCA (item d) analysis has also been performed assuming no AFW flow with no adverse impact on peak cladding temperature.

In addition to its accident mitigation function, the energy and mass addition capability of the AFW System is also considered with respect to HELBs within containment. For SLBs and FWLBs within containment, maximum pump flow from all three AFW pumps is assumed for 10 minutes until operations can isolate the flow by tripping the AFW pumps or by closing the respective pump discharge flow path(s). Therefore, the motor operated discharge isolation valves for the motor MDAFW pump trains (4007 and 4008) are designed to limit flow to 235 gpm to limit the energy and mass addition so that containment remains within design limits for items a and c. The TDAFW train is assumed to be at runout conditions (i.e., 630 gpm). to LTR-LIS-21-166, Revision 0 Page 25 of 28

B 3.7.5-6 Revision 75 R.E. Ginna Nuclear Power Plant AFW System B 3.7.5 C.1 With the TDAFW train inoperable, or both MDAFW trains inoperable, or one TDAFW train flow path and one MDAFW train inoperable to opposite SGs, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

If the inoperable MDAFW train supplies the same SG as the inoperable TDAFW flow path, Condition D must be entered.

The combination of failures which requires entry into this Condition all result in the loss of one train (or one flow path) of preferred AFW cooling to each SG such that redundancy is lost. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on redundant capabilities afforded by the SAFW System, time needed for repairs, and the low probability of a DBA occurring during this time period.

Condition C is modified by a Note which prohibits the application of LCO 3.0.4.b with a TDAFW train inoperable, or both MDAFW trains inoperable, or one TDAFW train flow path and one MDAFW train inoperable to opposite SGs. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with a TDAFW train inoperable, or both MDAFW trains inoperable, or one TDAFW train flow path and one MDAFW train inoperable to opposite SGs consequently the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in these circumstances.

D.1 With all AFW trains to one or both SGs inoperable, action must be taken to restore at least one train or TDAFW flow path to each affected SG to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The combination of failures which require entry into this Condition all result in the loss of preferred AFW cooling to at least one SG. If a SGTR were to occur in this condition, preferred AFW is potentially unavailable to the unaffected SG. If AFW is unavailable to both SGs, the accident analyses for small break LOCAs and loss of MFW would not be met.

The two MDAFW trains of the preferred AFW System are normally used for decay heat removal during low power operations since air operated bypass control valves are installed in each train to better control SG level (see Figure B 3.7.5-1). Since a feedwater transient is more likely during reduced power conditions, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is provided to restore at least one train of additional preferred AFW before requiring a controlled cooldown. This will also provide time to find a condensate source other than the SW System for the SAFW System if all three AFW trains are inoperable. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time is reasonable, based on redundant capabilities afforded by the SAFW System, time needed for repairs, and the low probability of a DBA occurring during this time period. to LTR-LIS-21-166, Revision 0 Page 26 of 28