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R. E. Ginna Nuclear Power Plant - Issuance of Amendment No. 144 Implementation of WCAP-14333 and WCAP-15376, TSTF-411-A, and TSTF-418-A to Revise Reactor Trip and Engineered Safety Feature Actuation System Instrumentation
	ML20353A126
Person / Time
Site:	Ginna
Issue date:	03/11/2021
From:	V Sreenivas; Plant Licensing Branch 1
To:	Rhoades D; Exelon Generation Co, Exelon Nuclear
	Sreenivas V, NRR/DORL/LPL1, 415-2596
References
	EPID L-2020-LLA-0055
	Download: ML20353A126 (38)
	v • d • e

March 11, 2021 Mr. David P. Rhoades Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

R. E. GINNA NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT NO. 144 RE: IMPLEMENTATION OF WCAP-14333 AND WCAP-15376; TSTF-411-A, REVISION 1; AND TSTF-418-A, REVISION 2, TO REVISE REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (EPID L-2020-LLA-0055)

Dear Mr. Rhoades:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 144 to Renewed Facility Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant in response to your application dated March 25, 2020, as supplemented by letter dated September 4, 2020.

The amendment revises Technical Specification (TS) 3.3.1, Reactor Trip System (RTS)

Instrumentation, and TS 3.3.2, Engineered Safety Feature Actuation System (ESFAS)

Instrumentation. These changes are based on Westinghouse topical reports WCAP-14333-P-A, Revision 1, Probabilistic Risk Analysis of the RPS [Reactor Protection System] and ESFAS Test Times and Completion Times, and WCAP-15376-P-A, Revision 1, Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times, and are consistent with NRC-approved Technical Specifications Task Force (TSTF)

Travelers TSTF-411-A, Revision 1, Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-15376-P), and TSTF-418-A, Revision 2, RPS and ESFAS Test Times and Completion Times (WCAP-14333).

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

Sincerely,

/RA/

V. Sreenivas, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244

Enclosures:

1. Amendment No. 144 to Renewed License No. DPR-18

2. Safety Evaluation cc: Listserv

EXELON GENERATION COMPANY, LLC DOCKET NO. 50-244 R. E. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 144 Renewed License No. DPR-18

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

A. The application for amendment by Exelon Generation Company, LLC (the licensee), dated March 25, 2020, as supplemented by letter dated September 4, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-18 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 144, are hereby incorporated in the renewed license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by James G. James G. Danna Date: 2021.03.11 Danna 13:00:48 -05'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 11, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 144 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 Replace the following page of Renewed Facility Operating License No. DPR-18 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 3 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 3.3.1-2 3.3.1-2 3.3.1-4 3.3.1-4 3.3.1-5 3.3.1-5 3.3.1-6 3.3.1-6 3.3.1-7 3.3.1-7 3.3.2-2 3.3.2-2 3.3.2-3 3.3.2-3 3.3.2-4 3.3.2-4

(b) Exelon Generation pursuant to the Act and 10 CFR Part 70, to possess and use four (4) mixed oxide fuel assemblies in accordance with the RG&Es application dated December 14, 1979 (transmitted by letter dated December 20, 1979), as supplemented February 20, 1980, and March 5, 1980; (3) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required.

(4) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

(1) Maximum Power Level Exelon Generation is authorized to operate the facility at steady-state power levels up to a maximum of 1775 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 144, are hereby incorporated in the renewed license.

Exelon Generation shall operate the facility in accordance with the Technical Specifications.

(3) Fire Protection Exelon Generation shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensees amendment request dated March 28, 2013, supplemented by letters dated December 17, 2013; January 29, 2014; February 28, 2014; September 5, 2014; September 24, 2014; December 4, 2014; March 18, 2015; June 11, 2015; August 7, 2015; and as approved in the safety evaluation report dated November 23, 2015.

Except where NRC approval for changes or deviations is required R. E. Ginna Nuclear Power Plant Amendment 144

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D.1 Action A.1 and referenced by Table 3.3.1-1. - NOTE -

1. For Functions 2a, 2b, 5, 6, 7b, 8, and 13, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

2. The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing of other channels.

Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months E. As required by Required E.1 Reduce THERMAL POWER 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Action A.1 and referenced to < 5E-11 amps.

by Table 3.3.1-1.

OR E.2

- NOTE -

Required Action E.2 is not applicable when:

a. Two channels are inoperable, or

b. THERMAL POWER is

< 5E-11 amps.

Increase THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER to 8% RTP.

F. As required by Required F.1 Open RTBs and RTBBs Immediately upon Action A.1 and referenced upon discovery of two discovery of two by Table 3.3.1-1. inoperable channels. inoperable channels AND R.E. Ginna Nuclear Power Plant 3.3.1-2 Amendment 144

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME H.3 Restore channel to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months OPERABLE status.

I. Required Action and I.1 Initiate action to fully insert Immediately associated Completion all rods.

Time of Condition H not met. AND I.2 Place the Control Rod Drive 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months System in a condition incapable of rod withdrawal.

J. As required by Required J.1 Action A.1 and referenced - NOTE -

by Table 3.3.1-1. Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SDM.

Suspend operations Immediately involving positive reactivity additions.

AND J.2 Perform SR 3.1.1.1. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter K. As required by Required K.1 Action A.1 and referenced - NOTE -

by Table 3.3.1-1. 1. For Functions 7a and 9b, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

2. The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing of other channels.

Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months R.E. Ginna Nuclear Power Plant 3.3.1-4 Amendment 144

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME L. Required Action and L.1 Reduce THERMAL POWER 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion to < 8.5% RTP.

Time of Condition K not met.

M. As required by Required M.1 Action A.1 and referenced by Table 3.3.1-1. - NOTE -

1. For Function 9a, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

2. The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing of other channels.

Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months N. As required by Required N.1 Restore channel to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action A.1 and referenced OPERABLE status.

by Table 3.3.1-1.

O. Required Action and O.1 Reduce THERMAL POWER 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion to < 30% RTP.

Time of Condition M or N not met.

P. As required by Required P.1 Action A.1 and referenced by Table 3.3.1-1. - NOTE -

The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels.

Place channel in trip. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Q. Required Action and Q.1 Reduce THERMAL POWER 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Associated Completion to < 50% RTP.

Time of Condition P not met. AND R.E. Ginna Nuclear Power Plant 3.3.1-5 Amendment 144

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME Q.2.1 Verify Steam Dump System 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months is OPERABLE.

OR Q.2.2 Reduce THERMAL POWER 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months to < 8% RTP.

R. As required by Required R.1 Action A.1 and referenced by Table 3.3.1-1. - NOTE -

One train may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing provided the other train is OPERABLE.

Restore train to OPERABLE 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months status.

S. As required by Required S.1 Action A.1 and referenced by Table 3.3.1-1. -------------------

-NOTE-For Functions 16c, 16d, and 16e, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

Verify interlock is in required 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months state for existing plant conditions.

OR S.2 Declare associated RTS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Function channel(s) inoperable.

R.E. Ginna Nuclear Power Plant 3.3.1-6 Amendment 144

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME T. As required by Required T.1 Action A.1 and referenced by Table 3.3.1-1. - NOTE -

1. One train may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing, provided the other train is OPERABLE.

2. One RTB may be bypassed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for maintenance on undervoltage or shunt trip mechanisms, provided the other train is OPERABLE.

Restore train to OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months status.

U. As required by Required U.1 Restore at least one trip 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Action A.1 and referenced mechanism to OPERABLE discovery of two by Table 3.3.1-1. status upon discovery of two inoperable trip RTBs with inoperable trip mechanisms mechanisms.

AND U.2 Restore trip mechanism to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months OPERABLE status.

V. Required Action and V.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition R, S, T, or U not met.

W. As required by Required W.1 Restore at least one trip 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Action A.1 and referenced mechanism to OPERABLE discovery of two by Table 3.3.1-1. status upon discovery of two inoperable trip RTBs with inoperable trip mechanisms mechanisms.

AND R.E. Ginna Nuclear Power Plant 3.3.1-7 Amendment 144

ESFAS Instrumentation 3.3.2 CONDITION REQUIRED ACTION COMPLETION TIME F. As required by Required F.1 Action A.1 and referenced by Table 3.3.2-1. - NOTE -

1. For Functions 4c, 5b, and 6c, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

2. The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing of the other channels.

Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months G.1 Be in MODE 3.

G. Required Action and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AND Time of Condition D, E, or F not met. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months G.2 Be in MODE 4.

H. As required by Required H.1 Restore channel to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Action A.1 and referenced OPERABLE status.

by Table 3.3.2-1.

I. As required by Required I.1 Restore train to OPERABLE 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action A.1 and referenced status.

by Table 3.3.2-1.

J. As required by Required J.1 Action A.1 and referenced by Table 3.3.2-1. - NOTE -

1. For Functions 1c, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

2. The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing of the other channels.

Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months R.E. Ginna Nuclear Power Plant 3.3.2-2 Amendment 144

ESFAS Instrumentation 3.3.2 CONDITION REQUIRED ACTION COMPLETION TIME K. Required Action andK.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated CompletionK.2 Time of Condition H,K.3 I, or AND J not met.

Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months L. As required by Required L.1 Action A.1 and referenced - NOTE -

by Table 3.3.2-1. 1. For Functions 1d and 1e, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

2. The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing of the other channels.

Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months M. Required Action and M.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition L not AND met.

M.2 Reduce pressurizer 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months pressure to < 2000 psig.

N. As required by Required N.1 Declare associated Auxiliary Immediately Action A.1 and referenced Feedwater pump inoperable by Table 3.3.2-1. and enter applicable condition(s) of LCO 3.7.5, "Auxiliary Feedwater (AFW)

System."

SURVEILLANCE REQUIREMENTS

- NOTE -

Refer to Table 3.3.2-1 to determine which SRs apply for each ESFAS Function.

SURVEILLANCE FREQUENCY SR 3.3.2.1 Perform CHANNEL CHECK In accordance with the surveillance Frequency Control Program R.E. Ginna Nuclear Power Plant 3.3.2-3 Amendment 144

ESFAS Instrumentation 3.3.2 SURVEILLANCE FREQUENCY SR 3.3.2.2 --------------------------------------

- NOTE-The ESFAS input relays are excluded from this surveillance for Functions 1c, 1d, 1e, 4c, 5b, and 6c.

Perform COT. In accordance with the Surveillance Frequency Control Program SR 3.3.2.3

- NOTE -

Verification of relay setpoints not required.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.2.4 - NOTE -

Verification of relay setpoints not required.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.2.5 Perform CHANNEL CALIBRATION In accordance with the Surveillance Frequency Control Program SR 3.3.2.6 Verify the Pressurizer Pressure-Low and Steam In accordance with Line Pressure-Low Functions are not bypassed the Surveillance when pressurizer pressure > 2000 psig. Frequency Control Program SR 3.3.2.7 Perform ACTUATION LOGIC TEST. In accordance with the Surveillance Frequency Control Program R.E. Ginna Nuclear Power Plant 3.3.2-4 Amendment 144

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 144 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 EXELON GENERATION COMPANY, LLC R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244

1.0 INTRODUCTION

By letter dated March 25, 2020 (Reference 1), as supplemented by letter dated September 4, 2020 (Reference 2), Exelon Generation Company, LLC (the licensee) submitted a risk-informed license amendment request (LAR) to revise R. E. Ginna Nuclear Power Plant (Ginna) Technical Specification (TS) 3.3.1, Reactor Trip System (RTS) Instrumentation, and TS 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation. These changes are based on Westinghouse topical reports (TRs) WCAP-14333-P-A, Revision 1, Probabilistic Risk Analysis of the RPS [Reactor Protection System] and ESFAS Test Times and Completion Times (Reference 3), and WCAP-15376-P-A, Revision 1, Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times (Reference 4).

The supplement dated September 4, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on May 5, 2020 (85 FR 26730).

The proposed amendment would revise TS 3.3.1 and TS 3.3.2 to allow certain functions in the RTS and ESFAS instrumentation to be in bypass for increased periods of time during surveillance testing and have extended limiting condition for operation (LCO) completion times (CTs). An additional change is proposed to allow bypassing of the Steam Generator (SG)

Level - Low Low function during ESFAS testing. The requested changes are consistent with NRC-approved Technical Specifications Task Force (TSTF) Travelers TSTF-411-A, Revision 1, Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-15376-P) (Reference 5), and TSTF-418-A, Revision 2, RPS and ESFAS Test Times and Completion Times (WCAP-14333) (Reference 6), or are supported by plant-specific analysis for those changes which are plant-specific, and therefore, not evaluated in these WCAPs.

Enclosure 2

2.0 REGULATORY EVALUATION

2.1 Regulatory Criteria and Guidance The NRC staffs evaluation is based on the following guidance and regulations.

Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical specifications, paragraph (a)(1), states, Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. As required by 10 CFR 50.36(c)(2)(i), the technical specifications will include limiting conditions of operation, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

Pursuant to 10 CFR 50.36(c)(2)(i), when a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met. The regulation at 10 CFR 50.36(c)(3) requires technical specifications to include items in the category of surveillance requirements, which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

Appendix A, General Design Criteria for Nuclear Power Plants (GDC), to 10 CFR Part 50, GDC 13, Instrumentation and control, requires that instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions, as appropriate, to assure adequate safety, including those variables and systems affecting the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.

GDC 20, Protection system functions, requires that the protection system(s) shall be designed (1) to initiate automatically the operation of appropriate systems, including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences, and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

GDC 21, Protection system reliability and testability, requires that the system be designed for high functional reliability and inservice testability, with redundancy and independence sufficient to preclude loss of the protection function from a single failure and preservation of minimum redundancy, despite removal from service of any component or channel.

GDC 22 through GDC 25 and GDC 29 require various design attributes for the protection system(s), including independence, safe failure modes, separation from control systems, requirements for reactivity control malfunctions, and protection against anticipated operational occurrences.

NUREG-1431, Revision 4.0, Volume 1, Standard Technical Specifications, Westinghouse Plants (Reference 7), contains the improved Standard Technical Specifications for Westinghouse plants.

Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis (Reference 8), describes a risk-informed approach with associated acceptance guidelines for

licensees to assess the nature and impact of proposed permanent licensing basis changes by considering engineering issues and applying risk insights. In implementing risk-informed decision-making, the NRC expects licensing basis changes to meet the acceptance guidelines and key principles of risk-informed regulations specified in RG 1.174.

RG 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications (Reference 9), describes an acceptable risk-informed approach and additional acceptance guidelines geared toward the assessment of proposed permanent TS completion time (CT) changes. RG 1.177 identifies a three-tiered approach for the licensees evaluation of the risk associated with a proposed TS CT change, as discussed below:

Tier 1 assesses the risk impact of the proposed change in accordance with acceptance guidelines consistent with the Commissions Safety Goal Policy Statement, as documented in RG 1.174 and RG 1.177. The first tier assesses the impact on operational plant risk based on the change in core damage frequency (CDF) and change in large early release frequency (LERF). It also evaluates plant risk while equipment covered by the proposed CT is out of service, as represented by incremental conditional core damage probability (ICCDP) and incremental conditional large early release probability (ICLERP). Tier 1 also addresses probabilistic risk assessment (PRA) quality, including the technical adequacy of the licensees plant-specific PRA for the subject application.

Tier 2 identifies and evaluates any potential risk-significant plant equipment outage configurations that could result if equipment, in addition to that associated with the proposed license amendment, is taken out of service simultaneously, or if other risk-significant operational factors, such as concurrent system or equipment testing, are also involved. The purpose of this evaluation is to ensure that appropriate restrictions are in place such that risk-significant plant equipment outage configurations will not occur when equipment associated with the proposed CT is implemented.

Tier 3 addresses the licensees overall configuration risk management program (CRMP) to ensure that adequate programs and procedures are in place for identifying risk-significant plant configurations resulting from maintenance or other operational activities and that the licensee takes appropriate compensatory measures to avoid risk-significant configurations that may not have been considered during the Tier 2 evaluation. Compared with Tier 2, Tier 3 provides additional coverage to ensure that the licensee identifies risk-significant plant equipment outage configurations in a timely manner and appropriately evaluates the risk impact of out-of-service equipment before performing any maintenance activity over extended periods of plant operation. Tier 3 guidance can be satisfied by the Maintenance Rule (10 CFR 50.65(a)(4)), which requires a licensee to assess and manage the increase in risk that may result from activities such as surveillance testing and corrective and preventive maintenance, subject to the guidance provided in RG 1.177, Section 2.3.7.1, and the adequacy of the licensees program and PRA model for this application. The purpose of the CRMP is to ensure that the licensee will appropriately assess, from a risk perspective, equipment removed from service before or during the proposed extended CT.

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference 10), describes an acceptable approach for determining whether the technical adequacy of the PRA is sufficient to support using the PRA in regulatory decision-making. RG 1.200 endorses, with clarifications and

qualifications, the use of the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Standard, RA-Sa-2009 (Reference 11).

Section 50.65 of 10 CFR, known as the Maintenance Rule, requires licensees to monitor the performance or condition of systems, structures, and components (SSCs) against licensee-established goals in a manner sufficient to provide reasonable assurance that SSCs are capable of fulfilling their intended functions. The implementation and monitoring program guidance of Section 3 of RG 1.174 and Section 3 of RG 1.177 states that monitoring performance in compliance with the Maintenance Rule can be used when it is sufficient for the SSCs affected by the risk-informed application. In addition, the Maintenance Rule (10 CFR 50.65(a)(4)), as it relates to the proposed surveillance, bypass test times, and CTs, requires the assessment and management of the increase in risk that may result from the proposed maintenance activity.

2.2 Proposed TS Changes

The proposed changes would revise Ginna TS Tables 3.3.1-1 and 3.3.2-1 to allow various functions in the RTS and ESFAS instrumentation to be in bypass for increased periods of time during surveillance testing and have extended LCO CTs. Additionally, changes to LCO 3.3.2.F and Surveillance Requirement 3.3.2.2 would allow bypassing the Steam Generator (SG)

Level - Low Low function during ESFAS surveillance testing and allow the input relays to be excluded from channel operational testing. The proposed changes are as follows:

TS 3.3.1, Reactor Trip System (RTS) Instrumentation The proposed changes would revise the following functions in TS Table 3.3.1-1, consistent with the generic evaluations approved in either WCAP-14333 or WCAP-15376:

Function System Action Proposed TS Change 5 Overtemperature T D.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 7b Pressurizer Pressure - D.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months High and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 11 Undervoltage - Bus 11A K.1 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to and 11B 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 12 Underfrequency - Bus K.1 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 11A and 11B 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 17 Reactor Trip Breakers T.1 Increase bypass time from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and completion time from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

The following functions in TS Table 3.3.1-1 were not included in the generic evaluations approved in WCAP-14333. In order to apply the TS relaxations as proposed, changes are supported by a plant-specific evaluation.

Function System Action Proposed TS Change 2a Power Range Neutron D.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Flux - High 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 2b Power Range Neutron D.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Flux - Low 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 6 Overpower T D.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 7a Pressurizer Pressure - K.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Low 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 8 Pressurizer Water Level - D.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to High 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 9a Reactor Coolant Flow - M.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Low - Single Loop 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 9b Reactor Coolant Flow - K.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Low - Two Loops 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 10b Reactor Coolant Pump K.1 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to (RCP) Breaker Position - 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Two Loops 13 Steam Generator Water D.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Level - Low Low 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 16c Reactor Trip System S.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Interlocks - Power Range 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Neutron Flux, P-8 16d Reactor Trip System S.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Interlocks - Power Range 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Neutron Flux, P-9 16e Reactor Trip System S.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Interlocks - Power Range 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Neutron Flux, P-10

TS 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation:

The proposed changes would revise the following functions in TS Table 3.3.2-1, consistent with the generic evaluations approved in WCAP-14333:

Function System Action Proposed TS Change 1d Safety Injection - L.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Pressurizer Pressure - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Low to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 4c Steam Line Isolation - F.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Containment Pressure - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months High High to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 4d Steam Line Isolation - F.1 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to High Steam Flow 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Coincident With Safety Injection and Coincident With Tavg- Low 4e Steam Line Isolation - F.1 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to High High Steam Flow 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Coincident With Safety Injection 5b Feedwater Isolation - SG F.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Water Level - High 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 6c Auxiliary Feedwater - SG F.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Water Level - Low Low 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months The following functions in TS Table 3.3.2-1 were not included in the generic evaluations approved in WCAP-14333. In order to apply the TS relaxations as proposed, changes are supported by a plant-specific evaluation.

Function System Action Proposed TS Change 1c Safety Injection - J.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Containment Pressure - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months High to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1e Safety Injection - Steam L.1 Increase bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to Line Pressure - Low 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 2c Containment Spray, J.1 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Containment Pressure - 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months High High

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the Ginna incorporation of WCAP-15376 and WCAP-14333 using the three-tiered approach and the five key principles of risk-informed decision-making presented in RG 1.174, Revision 2, and RG 1.177, Revision 1, and the conditions and limitations for

WCAP-15376 and WCAP-14333 in safety evaluations (SEs) dated December 20, 2002, and April 29, 1998, respectively (Reference 12 and Reference 13).

3.1 Traditional Engineering Evaluation (Key Principles 1, 2, and 3)

The purpose of the deterministic assessment is to address the traditional engineering considerations for the proposed changes. The traditional engineering evaluation addresses Key Principles 1, 2, and 3 of the NRC staffs philosophy of risk-informed decision-making, which concerns compliance with current regulations, evaluation of defense in depth, and evaluation of safety margins.

The NRC staff reviewed the proposed LAR and determined that the application is in accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, since the proposed changes are fully described, and the rationale for changes is provided, as appropriate, in the LAR.

3.1.1 Evaluation of Changes to RTS and ESFAS The NRC staff reviewed the proposed changes against the approved TSTF-411, Revision 1, and TSTF-418, Revision 2. The NRC letter dated August 30, 2002 (Reference 14), approved TSTF-411, Revision 1, without changes and advised Westinghouse to include TSTF-411, Revision 1, with publication of approved WCAP-15376-P. Similarly, the NRC staff approved TSTF-418, Revision 2, pertaining to WCAP-14333, without changes by letter dated April 2, 2003 (Reference 15). Based on these approvals, TSTF-411, Revision 1, and TSTF-418, Revision 2, reflect the approved versions of WCAP-15376 and WCAP-14333.

The NRC staff reviewed the proposed changes and the associated action statements regarding RTS instrumentation pertaining to Functions 5, 7b, 11,12, and 17. The changes were compared with approved TSTF-411 and TSTF-418 and found acceptable. RTS Functions 2a, 2b, 6, 7a, 8, 9a, 9b, 10b, 13, 16c, 16d, and 16e were not approved generically as part of the approved TSTFs and were submitted as plant-specific changes to be evaluated by the NRC staff. These plant-specific changes are evaluated in Section 3.2 of this SE as deterministic assessments and found acceptable.

The NRC staff also reviewed the proposed changes and the associated action statements regarding ESFAS instrumentation pertaining to Functions 1d, 4c, 4d, 4e, 5b, and 6c. The changes were compared with approved TSTF-411 and TSTF-418 and found acceptable.

ESFAS Functions 1c, 1e, and 2c were not approved as part of the approved TSTFs and were submitted as plant-specific changes to be evaluated by the NRC staff. These plant-specific changes are evaluated in Section 3.2 of this SE as deterministic assessments and found acceptable.

Furthermore, the NRC staff determined that the proposed LAR introduced no physical design change to the RTS and ESFAS systems. As such, the intent of the original plant design function is maintained, and there is no adverse impact on the redundancy, independence, or diversity of the RTS and ESFAS systems.

To demonstrate consistency with the defense-in-depth design philosophy, the licensee has performed the following assessments in Section 4.1.1 of the LAR:

Preserve reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation Avoid over-reliance on programmatic activities as compensatory measures Preserve system redundancy, independence, and diversity Defense against and prevention of new common cause failures Maintain independence of physical barriers Defense against human error, and Maintain the intent of the plant design criteria.

The NRC staff reviewed the licensees statements in these assessments and finds that the discussion adequately addressed these items. The statements are acceptable to the staff since the proposed increases in the CTs and test bypass times do not unduly affect the design and operation of the RTS or ESFAS instrumentation systems. The RTS and ESFAS will remain capable of performing their required functions, and no new accidents or transient events are introduced with the proposed changes. Over-reliance on programmatic activities as compensatory measures is avoided since RTS and ESFAS systems will function in the same manner as before. The licensees method for calculating risk increase CFD, LERF, ICCDP, and ICLERP) is in accordance with RG 1.174 and RG 1.177 and was demonstrated to be very small and within the bound of RG 1.174 and RG 1.177, as documented in Table 5-14 through 5-19 of LAR Attachment 1. Therefore, no additional reliance on systems, procedures, or operator actions are needed to be put into place to compensate for any risk increase.

There is no physical design change to the RTS and ESFAS systems; the intent of the original plant design is maintained; and there is no adverse impact on the redundancy, independence, or diversity of the RTS and ESFAS systems. Similarly, since there are no system design changes, and the TS time extensions requested are not for a long or extended duration, the NRC staff determined that defenses against potential common cause failures are maintained, and no new common cause failure mechanisms are expected to arise. Increasing the CTs and bypass test times to the RTS and ESFAS systems also does not affect the independence of the fuel cladding, reactor coolant system, or containment, and therefore, there is no undue impact on the integrity of the physical barriers to limit leakage to the environment. The extension of TS CT and bypass test time introduced no new operator actions or additional operating, maintenance, or test procedures, and no new at-power tests or maintenance activities are expected to occur as a result of these changes; therefore, no new human errors are introduced.

Based on the above, the NRC staff determined that the defense-in-depth design philosophy is maintained and continues to be met.

Section 4.1.2 of the LAR addresses safety margin. To assess the safety margin, the licensee reviewed the adherence to codes and standards and the safety analysis acceptance criteria in the Final Safety Analysis Report to assure that the existing criteria are maintained. The discussion in the LAR for both these items has been reviewed by the NRC staff and found adequate. Since the design functions and operation of the RTS and ESFAS systems are not changed by the proposed increase of the CTs and bypass test times, redundancy and diversity will be maintained. The proposed changes also do not allow plant operation in a configuration outside the design basis, and all signals credited as primary or secondary and all operator actions credited in the accident analysis will remain the same. Therefore, the NRC staff determined that the proposed changes do not reduce the safety margin.

3.1.2 Traditional Engineering Evaluation (Key Principles 1, 2, and 3) Conclusion The licensee has submitted the LAR in accordance with 10 CFR 50.90 and clearly described the proposed TS changes and aligned those functions with the NRC-approved TSTF-411 and TSTF-418. For functions that are not generically evaluated in WCAP-14333 and WCAP-15376, the licensee has provided an acceptable plant-specific evaluation for these functions per the guidelines in WCAP-14333 and WCAP-15376 to ensure the defense-in-depth design principles and safety margins are not adversely impacted by the proposed changes. The assessments are discussed in Section 3.2 of this SE. Therefore, the NRC staff finds that Key Principles 1, 2, and 3 are adequately addressed, and current regulations and applicable requirements continue to be met. In addition, adequate defense in depth and sufficient safety margins are also maintained.

3.2 Risk Evaluation (Key Principle 4)

WCAP-14333 and WCAP-15376 are consistent with the NRC approach for using PRA in risk-informed decisions on plant-specific changes to the current licensing basis, as presented in RG 1.174 and RG 1.177. The risk evaluation considered the three-tiered approach, as presented by RG 1.177, for the extension to the RTS and ESFAS CTs. Tier 1, Probabilistic Risk Assessment Capability and Insights, assesses the impact of the proposed CT and bypass time change on CDF, LERF, ICCDP, and ICLERP. Tier 2, Avoidance of Risk-Significant Plant Configurations, considers potential risk-significant plant operating configurations. Tier 3, Risk-Informed Configuration Risk Management, is addressed when the TS CT change is implemented.

3.2.1 Tier 1: Risk Impact of the Proposed CT and Bypass Time Change In Tier 1, the NRC staff evaluates the impact of the proposed changes on plant operational risk based on the Ginna implementation of WCAP-14333 and WCAP-15376. The objective of the PRA review is to determine whether the Ginna PRA model used to implement WCAP-14333 and WCAP-15376 is of sufficient scope, detail, and technical acceptability for this application.

The NRC staff review involves the evaluation of the validity of the licensees PRA and its application to the proposed changes, and evaluation of the PRA results and insights based on the licensees proposed application. The licensee has provided analysis in LAR Attachment 1 to address the risk impact of the proposed allowed outage time (AOT) extensions for identified RTS and ESFAS instrumentation using the Ginna full power internal events (FPIE) PRA and fire PRA (FPRA) models. The analyses referenced by TSTF-411 and TSTF-418 can be applied to this license amendment for extended allowable outage times. The few remaining functions that are not covered in the TSTF documents are addressed in the risk analysis provided in the LAR.

3.2.1.1 PRA Capability Internal Events PRA In LAR Attachment 1, the licensee stated that the Ginna PRA model was peer reviewed in 2009 using Nuclear Energy Institute (NEI) 07-12, Revision 1; RG 1.200, Revision 1; and the ASME/ANS PRA Standard RA-Sc-2007. In response to request for additional information (RAI) 2, the licensee provided a summary of the changes to the PRA since the 2009 peer review. The licensee stated that changes were made to support the LAR for transition to National Fire Protection Association (NFPA) 805, Performance-Based Standard for Fire

Protection for Light Water Reactor Electric Generating Plants. The licensee additionally provided a summary of other key changes to the Ginna internal events PRA and provided justification on why they were considered PRA maintenance. Further, the licensee stated that facts and observation (F&O) closures were performed in 2017 and 2020 in accordance with the process documented in the NEI letter to the NRC entitled, Final Revision of Appendix X to NEI 05-04/07-12/12-13, Close-Out of Facts and Observations (F&Os), dated February 21, 2017 (Reference 16), as accepted by the NRC in letter dated May 3, 2017 (Reference 17).

The licensee provided in the LAR the finding-level F&Os identified by the 2009 peer review.

The NRC staff reviewed these F&Os and any additional information provided by the licensee in the response to the RAI associated with the Ginna application for TS change regarding risk-informed justification for the relocation of specific surveillance frequency requirements to a licensee-controlled program (Reference 18, Reference 19, and Reference 20).

F&O IE-C10-01 found that the PRA documentation provided no explanation of differences between plant-specific initiating events and generic initiating events. Disposition to F&O IE-C10-01 stated that this issue is a documentation only issue. The NRC staff noted that, in response to RAI 3 related to the Ginna application for TSTF-425, the licensee identified that there were differences with loss of bus initiating events that resulted in an entry in the updating requirement evaluation (UHrRE) database, and the NRC requested additional information on why this F&O would have no impact on the application. In response to RAI 3, the licensee clarified that the difference was the result of modeled operator recoveries. Since the licensees use of plant-specific value was justified, the NRC staff finds this TSTF-411 and TSTF-418 application acceptable.

Internal flooding F&O IF-B2-01 (and similarly, F&Os IF-D6-01 and IFEV-A7) identified that the Ginna internal flooding PRA had a limited attempt to address human-induced flooding mechanisms. In response to RAI 4.a, the licensee explained that these F&Os were reviewed by the 2017 F&O closure. The licensee explained that the closure process concluded that the approach taken to consider the impacts of human-induced flooding was adequate and appropriate. For F&O IF-D6-01, the licensee stated that the closure process found that a thorough search was made for floods that could be induced by human action or other failure during maintenance, and only one source of flooding was identified for explicit modeling.

Internal flooding F&O IF-D5a-01 identified that the internal flooding PRA did not adequately address plant-specific characteristics that might affect the manner in which the frequencies of flooding are estimated. In response to RAI 4.b, the licensee stated that characteristics including water hammer and human-induced floods have been addressed in the internal flooding PRA model and documentation. The license also stated that pipe break frequencies were based on Electric Power Research Institute (EPRI) Report 302000079 (Reference 21), and aging affects were included in the Ginna model.

Internal flooding F&O IF-F3-01 identified the lack of an adequate characterization of the sources of uncertainty associated with the flood analysis. In response to RAI 4.c, the licensee provided a list of assumptions and sources of uncertainty and their disposition for the application. The licensee explained that the list is exhaustive and not limited to key uncertainties. The licensee stated that no new assumptions or sources of uncertainty were identified unique to this application.

Internal Fire PRA The licensee stated that a peer review of the fire PRA was conducted in June 2012. This peer review used NEI 07-12, Revision 1, to evaluate the model against the PRA Standard ASME/ANS RA-Sa-2009, along with the NRC clarifications provided in RG 1.200, Revision 2.

Further, the licensee stated that since the 2012 peer review, there were several updates to the Ginna fire PRA. An earlier NRC staff review of the technical adequacy of the fire PRA was performed during the staffs review of the licensees NFPA 805 LAR (Reference 22). As a result of the NRC staffs review of the NFPA 805 LAR, as supplemented, the staff concluded in issuance of Amendment No. 119, dated November 23, 2015 (Reference 23), that (1) the fire PRA model adequately represents the current, as-built, as-operated configuration, and is, therefore, capable of modeling the plant as needed; (2) the fire PRA model conforms sufficiently to the applicable industry PRA standards at an appropriate capability category, considering the acceptable disposition of the peer review and NRC staff review findings; and (3) the fire modeling used to support the development of the FPRA has been confirmed as appropriate and acceptable. The NRC staff identified no information that would invalidate the staffs NFPA-805 conclusion that the fire PRA is technically acceptable to support risk calculations. Therefore, the NRC staff concludes that the fire PRA is technically acceptable to support this application.

Based on the above, the staff finds the use of the fire PRA acceptable since the licensee has followed the appropriate PRA standard and NRC guidance in developing and maintaining the fire PRA.

PRA Maintenance and Update In the LAR, the licensee described the process for PRA maintenance and update and how the PRA model is an accurate reflection of the as-built and as-operated plant. The licensee stated that the overall Exelon Generation Company, LLC risk-management program defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, industry operating experience, etc.), and for controlling the model and associated computer files.

The licensee stated that Ginna maintains a database of URE to track all enhancements, corrections, and unincorporated plant changes. The licensee stated that a review was performed on 57 high or medium priority open UREs for FPIE and FPRA models, and no open items were identified to have more than a negligible impact on the TSTF-411 and TSTF-418 delta risk analysis or results. The NRC staff requested in RAI 1 that the licensee describe the types of open UREs of high priority and explain how it was concluded that they have a negligible impact on application. In response to RAI 1, the licensee stated that high priority UREs are defined as items that could significantly impact the application or challenge the criteria for an unscheduled PRA update. The licensee provided a list of the high priority UREs and their associated disposition. The NRC staff finds the licensees RAI response adequate since the high priority UREs are identified, described, and dispositioned or downgraded to illustrate that outstanding UREs have no impact on this application.

Seismic Events The licensee stated that seismic event risk evaluation for Ginna is based on the individual plant examination of external events (IPEEE). Since the IPEEE completion in 1997, new seismic hazard information has been developed. The licensee stated that the impact of the updated hazard was considered negligible since Ginna was screened out of the seismic hazard per the

NRC/EPRI seismic reevaluations. However, for conservatism, the licensee performed a standalone bounding calculation for the potential seismic risk of the ESFAS and RTS AOT extension as further reviewed in this SE.

Other External Event Hazards The licensee stated that other external event hazards were evaluated in the Ginna IPEEE. The licensee stated that high winds, external floods, and transportation accidents were reviewed against the standard review plan and met the standard review plan criteria.

In RAI 6.a, the NRC staff requested the licensee to further discuss, in the context of the current plant and its environs, the applicability of the IPEEE conclusions for the current LAR. In response to RAI 6.a, the licensee stated that there are no significant changes to the plant or environs; however, additional analyses have been performed since the IPEEE to update the assessments of various hazards. Regarding high wind events, the licensee stated that Ginna committed to include tornado and tornado missiles into its licensing basis and that all necessary modifications were completed. Further, the licensee stated that Ginna performed an extensive review of the current licensing basis and potential exposure of SSCs to tornado-generated missiles in response to NRC Regulatory Issue Summary 2015-06, Tornado Missile Protection, dated June 10, 2015 (Reference 24). The licensee stated that all vulnerabilities identified were resolved to ensure that the required SSCs remain protected.

In RAI 6.b, the NRC staff requested the licensee provide technical justification for why the risk from external flooding is negligible, given the recent external flooding reevaluation performed in response to the Fukushima Near-Term Task Force recommendations. The licensee stated that all flood-causing mechanisms were bounded by the current licensing basis except for local intense precipitation (LIP) and combined effects river flood, which produces a probable maximum flood (PMF). The licensee stated that a reevaluation of LIP found the Battery Room and Diesel Generator Rooms are below the peak water surface elevation; however, these rooms are provided with normally closed watertight doors to protect against flood water intrusion. Further, the licensee stated that the service water screen house is also impacted by LIP; however, the service water system is not credited for providing cooling water during an external flood event. The licensee stated that a PMF resulting from the combined effects river flood would inundate the site, and the current licensing basis requires temporary barriers to be installed prior to the arrival of flood waters. In a letter dated March 10, 2017 (Reference 25), the licensee concluded the site has an adequate site response and available physical margin to mitigate the effects from the PMF. Based on the above, the NRC staff finds the licensee has provided adequate justification in concluding external flood risk is negligible because physical flood protection features and procedures are available to mitigate the impacts of severe flooding.

3.2.1.2 PRA Results and Insights Satisfaction of the fourth key principle of risk-informed decision-making may be demonstrated with reasonable assurance by comparing risk metrics that reflect the proposed TS changes to the numerical risk acceptance guidelines in RG 1.174, Revision 3, and RG 1.177, Revision 1.

Furthermore, Condition and Limitation Nos. 2 and 3 provided in Section 5.0 of the NRC SE for WCAP-15376 (Reference 12) delineate information the NRC staff requested the licensee provide in the LAR to assess the applicability of the plant-specific TS change to the approved TR.

The licensee explained the PRA model changes performed for this analysis in LAR Section 5.6.4.2. The licensee stated that the testing and maintenance event probabilities associated with the instrumentation are adjusted to reflect the extended AOT. The licensee stated it added basic events to model unavailability of some relays and channels. The licensee explained also how common cause failure (CCF) was modeled. The licensee explained that the Ginna FPIE model includes a common cause event for failure of all RTS signals, which was derived from generic industry data sources with a value of 1.6E-7. The licensee further stated that since a CCF development for each individual signal modeled in RTS is not available, this event was increased to 1.6E-6 to account for the increased likelihood of CCF if one instrument channel is inoperable. The licensee stated this factor of 10 increase is conservatively used to bound the WCAP-14333-P-A factor of 2 increase chosen for the sensitivity study to address the uncertainty in the CCF analysis for the risk calculations. The licensee further stated in the LAR that Ginna did not model common cause for the ESFAS functions analyzed. In response to RAI 5.b, the licensee explained that while the WCAP analysis may be deemed sufficient to address various common cause combinations of ESFAS failures, it is not solely relied upon to justify this LAR. The licensee explained that the plant-specific calculations also modeled Train A/B ESFAS events to conservatively cover all failure modes, including common cause within and among different signals and channels. The NRC staff finds the licensees approach to common cause risk evaluation reasonable and acceptable for the application.

In LAR Attachment 1, Tables 5-14 through 5-19, the licensee provided the baseline CDF/LERF values and the ICCDP/ICLERP risk metrics for the proposed change due to unavailability of ESFAS/RTS instrumentation resulting from implementing WCAP-14333 and WCAP-15376. The licensee estimated an increase in CDF of 1.13E-07/year and an increase in LERF of 1.02E-09/year. The NRC staff finds that increase in CDF and LERF satisfies the RG 1.174 acceptance guidelines and is considered very small change in risk.

The licensee also presented ICCDP/ICLERP results in LAR Table 5-19. In response to RAI 7, the licensee clarified that some ICCDP/ICLERP values were reported as zero because the impacts of those specific failed conditions fall below the truncation levels. The ICCDP and ICLERP for each ESFAS/RTS instrumentation AOT is well below the RG 1.177 acceptance guidelines of 1E-06 ICCDP and 1E-07 ICLERP.

The licensee also performed a standalone bounding calculation for the potential seismic risk of the ESFAS and RTS AOT extension using inputs from the FPIE. The licensee identified two primary seismic events of interest for this assessment as seismically induced loss-of-offsite power and small loss-of-coolant accident. The licensee estimated the risk impact related to the change in signal unavailability to be negligible.

The NRC staff reviewed the above results, insights, and subsequent clarifications provided by the licensee and determined them to be acceptable. The staff also reviewed the change in CDF/LERF with all the cases combined due to the proposed extended CT and AOT and determined that they are well below the acceptance guideline of RG 1.174 for Region III.

Similarly, the change in ICCDP/ICLERP is also well below the RG 1.177 acceptance guideline for TS changes. Therefore, the staff concluded that the PRA results and insights are acceptable.

3.2.2 Tier 2: Identification and Evaluation of Potential Risk-significant Plant Equipment Outage Configurations RG 1.177, Revision 1, states that the licensee should provide reasonable assurance that risk-significant plant equipment outage configurations will not occur when specific plant equipment is out of service, consistent with the proposed TS change. Tier 2 evaluates the capability of the licensee to recognize and avoid risk-significant plant configurations that could result if equipment, in addition to that associated with the proposed change, is taken out of service simultaneously or if other risk-significant operational factors, such as concurrent system or equipment testing, are also involved.

The NRC staff reviewed the analysis provided by the licensee in LAR Attachment 1 and determined that since the licensee has adequately demonstrated in the Tier 1 evaluations that ICCDP and ICLERP are far below acceptance criteria, it is unlikely that the plant will enter a risk-significant configuration while the equipment covered in this LAR is out of service due to the proposed extended outage time. Should any specific failures dominate the results, existing administrative procedure and compensatory measures are in place to cover equipment affected by this LAR.

In LAR Attachment 1, Sections 5.5.1.2 and 5.5.2.2, the licensee provided Tier 2 restrictions for implementation of WCAP-14333 and WCAP-15376. Should equipment failure occur during the extended CT or AOT for the ESFAS and RTS, the licensee will use its Tier 3 CRMP to assess the emergent condition and direct activities (e.g., restore the inoperable logic train and exit the TS action or fully implement the Tier 2 restrictions). Ginna will establish administrative controls to implement the following restrictions during the mode of applicability for the specified equipment:

To preserve ATWS [anticipated transient without scram] mitigation capability, activities that degrade the ability of the AFW [auxiliary feedwater] system, reactor coolant system (RCS) pressure relief systems (pressurizer power operated relief valves (PORVS) and safety valves), ATWS mitigating systems actuation circuitry (AMSAC), or turbine trip should not be scheduled when a logic train is inoperable.

To preserve loss-of-coolant accident (LOCA) mitigation capability, one complete emergency core cooling system (ECCS) train that can be actuated automatically must be maintained when a logic train is inoperable.

To preserve reactor trip and safeguards actuation capability, activities that cause relays in the available train to be unavailable, and activities that cause analog channels to be unavailable should not be scheduled when a logic train is inoperable.

Activities in electrical systems (e.g., AC [alternating current] and DC [direct current] power) and cooling systems (e.g., service water and component cooling water) that support the systems or functions listed in the first three bullets should not be scheduled when a logic train is inoperable. That is, one complete train of a function that supports a complete train of a function noted above must be available.

Based on the above, the NRC staff concludes that the licensee has Tier 2 restrictions and administrative procedures in place to ensure a complete train of safety logic function is available or can be restored when either the ESFAS or RTS is in extended CT or AOT.

3.2.3 Tier 3: Configuration Risk Management Program Section 2.3 of RG 1.177 discusses Tier 3 of the three-tiered approach for evaluating risk associated with proposed changes to TS CT. Tier 3 is the establishment of a risk-informed plant configuration control program (i.e., a CRMP) to ensure that other potentially lower probability, but nonetheless risk-significant, configurations resulting from maintenance and other operational activities are identified and compensated for. Because the Maintenance Rule, as codified in 10 CFR 50.65(a)(4), requires licensees to assess and manage the potential increase in risk that may result from activities such as surveillance testing and corrective and preventive maintenance, a licensee may use its existing Maintenance Rule program to satisfy Tier 3.

The NRC staff reviewed the Ginna CRMP to ensure there is no significant risk increase while instrument maintenance is being performed. The licensee stated that Ginna uses the PARAGON PRA software tool, supplemented by operations and work management procedure, as the CRMP to implement 10 CFR 50.65(a)(4). The licensee stated that the use of PARAGON is procedurally controlled. Any plant configuration will be evaluated to assess and manage the risk. The licensee stated that Ginna procedures recognize there are limitations in PARAGON, and specifically, direct consideration of external events and site activities that can result in significant plant events. The licensee stated that high risk evolutions, site activities, and external events, such as lightning, high winds, and solar magnetic disturbances, are addressed qualitatively in the PARAGON model based on Ginnas redundancy to mitigate the affected transient.

In response to RAI 8.a, the licensee clarified that fire risk is controlled through procedures. The licensee stated that the list of risk-significant equipment is validated or updated when the fire model is updated or an issue is discovered. The licensee also explained that pre-established compensation measures are also implemented when risk-significant equipment is removed for maintenance to manage risk. In response to RAI 8.b, the licensee also stated that only pre-identified risk-significant RTS and ESFAS systems are modeled in the PARAGON online risk monitoring software. The licensee stated that modeled components include AMSAC, reactor trip breakers, power supplies that support RTS/ESFAS functions, master relays, and undervoltage relays. The licensee stated other low-level components not explicitly included in PARAGON are assumed to have a negligible impact on CRMP calculations based on their pre-determined low risk significance. The NRC staff finds the licensees methodology adequately ensures risk-significant equipment is properly tracked and evaluated in the Ginna CRMP model.

Based on the above discussion, the NRC staff finds that the licensee has adequately implemented the CRMP to ensure no significant risk would arise while RTS and ESFAS instruments are in maintenance for the proposed extended outage time.

3.2.4 Addressing WCAP-14333 and WCAP-15376 Conditions and Limitations 3.2.4.1 Limitations for WCAP-14333 WCAP-14333 SE Condition 1 Condition 1 of the SE for WCAP-14333 is for the licensee to confirm the applicability of WCAP-14333 analysis for the plant.

WCAP-14333 and WCAP-15376 provide a generic PRA model for the evaluation of the CT and bypass test times. The NRC staff found this generic model and the WCAP-14333 and WCAP-15376 evaluations to be acceptable on a generic basis in its SEs dated April 29, 1998, and December 20, 2002, respectively. Although the SEs accepted the use of a representative model as generally reasonable, the application of the representative model and the associated results to a specific plant introduce a degree of uncertainty because of modeling, design, and operational differences. Therefore, each licensee adopting WCAP-14333 needs to confirm that the TR analyses and results are applicable to its plant.

To determine that WCAP-14333 is applicable to Ginna, the licensee listed the important parameters and assumptions made in the generic analysis that are relevant to the AOT evaluation in LAR Attachment 1, Tables 5-4 through 5-6. The evaluation provided by the licensee compared the analysis assumptions in WCAP-14333 to plant-specific parameters, including surveillance and maintenance intervals, operator actions, transient and ATWS frequencies, actuation signals, safety functions, and certain component failure probabilities.

Based on the above discussion and the NRC staffs Tier 1 evaluation in this SE, the staff finds the evaluation provided by the licensee confirms that the generic evaluation assumptions used in WCAP-14333 are applicable to Ginna. Therefore, the staff finds that WCAP-14333 SE Condition 1 is satisfied.

WCAP-14333 SE Condition 2 Condition 2 of the SE for WCAP-14333 is for the licensee to address the Tier 2 and Tier 3 analyses, including the CRMP insights, which confirm that these insights are incorporated into the decision-making process before taking equipment out of service.

Based on the NRC staffs Tier 2 and Tier 3 evaluation in Sections 3.2.2 and 3.2.3 of this SE, respectively, the staff finds that WCAP-14333 SE Condition 2 is satisfied.

3.2.4.2 Limitations for WCAP-15376 The evaluation of the five NRC staff SE conditions and limitations of WCAP-15376 is discussed below.

WCAP-15376 SE Condition 1 Condition 1 of the SE for WCAP-15376 is that the licensee is expected to confirm the applicability of the TR to its plant and to perform a plant-specific assessment of containment failures and address any design or performance differences that may affect the proposed changes.

The licensee addressed WCAP-15376 Condition 1 in two parts. The first part confirms the applicability of the TR to Ginna, and the second part addresses the containment failure assessment. To demonstrate the applicability of the WCAP-15376 analysis on a plant-specific basis, the licensee performed a comparison between the key generic analysis parameters and assumptions and plant-specific parameters and design. LAR Attachment 1, Tables 5-7 through 5-10, provide a list of the key analysis parameters and assumptions, along with the input used in the generic analysis to show that the analysis is applicable. The information in these tables is related to plant-specific signals that are available to actuate RTS and ESFAS and test and maintenance information for the components of the RPS. Information is also provided on Ginnas current calculated CDF, LERF, and the contribution to CDF from ATWS events. The NRC staff reviewed the information provided in Tables 5-7 through 5-10 and determined that the licensee adequately demonstrated the applicability of the WCAP-15376 analysis to Ginna since the current baseline CDF and LERF values for Ginna meet RG 1.174 criteria for determining that small increases in CDF and LERF are acceptable, and important safety functions have been compared for applicability between WCAP-15376 and Ginna plant-specific parameters.

Based on the above discussion and the NRC staffs Tier 1 evaluation in this SE, the staff finds the evaluation provided by the licensee confirms that the generic evaluation assumptions used in the WCAPs are applicable to Ginna. The staff finds the first part of WCAP-15376 SE Condition 1 is satisfied.

The second part of Condition 1 requires plant-specific assessment of containment failures and addressing any design or performance differences that may be affected by the proposed changes. Containment failure modes typically considered in PRA include containment isolation failure; containment bypasses from interfacing systems loss-of-coolant accident, steam generator tube rupture, and steam generator tube creep rupture; and containment failure from steam explosion, hydrogen burns, direct containment heating, and containment steam overpressurization. WCAP-15376 was based on a large, dry containment and assumed that the only contributions to LERF would come from containment bypass events and core damage events with the containment not isolated. The LERF analysis completed to support this LAR was based on a large, dry containment with LERF contributions from containment isolation failure and containment bypasses from interfacing systems loss-of-coolant accident and steam generator tube rupture events, excluding steam generator tube creep rupture, since it is generally a small contributor to LERF. The NRC staff evaluated the accident scenarios analyzed by the licensee for Ginna that would likely cause containment failure based on containment bypass and containment isolation failure, and the staff finds these accident scenarios are typical of a large, dry containment. In addition, based on the staffs Tier 1 evaluation in Section 3.2.1 of this SE, the NRC staff finds the licensees PRA model to be technically adequate to analyze containment failures, and the licensees containment assessment conclusions are acceptable. Therefore, the NRC staff finds the licensees containment failure assessment to be consistent with WCAP-15376. The NRC staff finds that the second part of WCAP-15376 SE Condition 1 is also satisfied.

WCAP-15376 SE Condition 2 Condition 2 of the SE for WCAP-15376 is for the licensee to address the Tier 2 and Tier 3 analyses, including risk-significant configuration insights and confirm that these insights are incorporated into the plant-specific CRMP.

Based on the NRC staffs Tier 2 and Tier 3 evaluation in Sections 3.2.2 and 3.2.3 of this SE, respectively, the staff finds that WCAP-15376 SE Condition 2 is satisfied.

WCAP-15376 SE Condition 3 Condition 3 of the SE for WCAP-15376 is that the risk impact of concurrent testing of one logic cabinet and associated reactor trip breaker (RTB) needs to be evaluated on a plant-specific basis to ensure conformance with the WCAP-15376 evaluation, RG 1.174, and RG 1.177.

WCAP-15376 did not specifically evaluate or preclude concurrent testing of one logic cabinet and its associated RTB. Based on this, the NRC staff questioned the applicability of the TR to this particular maintenance configuration. In response to the NRC staffs RAls on WCAP-15376, the applicant provided the ICCDP for both the logic cabinet and associated RTB out of service for preventive maintenance for a total time of 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , which is comprised of a CT of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , plus 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reach Mode 3. The ICCDP for a duration of 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months in this configuration is 3.2E-07, which meets the RG 1.177 acceptance guideline of 5E-07. Since this ICCDP value is based on the logic cabinet and RTB being out of service for 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months at the same time, bypassing one logic cabinet and associated RTB for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for testing per this LAR will also meet the RG 1.177 ICCDP guideline. As such, the NRC staff finds the generic analysis presented in WCAP-15376 for concurrent testing of one logic cabinet and associated RTB is applicable to Ginna. The staff also finds that the risk metrics are expected to be within the acceptance guidelines in RG 1.177 for this configuration, and therefore, the staff finds that WCAP-15376 SE Condition 3 is satisfied.

WCAP-15376 SE Condition 4 Condition 4 of the SE for WCAP-15376 is that in order for the licensee to ensure consistency with the referenced plant, the model assumptions for human reliability in WCAP-15376 should be confirmed to be applicable to the plant-specific configuration.

In LAR Attachment 1, Section 5.5.2.4 and Table 5-11, the licensee lists the operator actions credited in the WCAP-15376 analysis. In Table 5-11, the licensee indicated that the human reliability associated with the relevant operator actions applicable to Ginna is based on a plant-specific assessment and confirmed that sufficient time is available for operator actions, and that the procedures for manual actions are in place for each of the operator action categories. The NRC staff reviewed the operator actions provided in Table 5-11 and finds the actions to be acceptable and consistent with WCAP-15376, and that the WCAP-15376 model assumptions are, therefore, applicable to Ginna. The staff finds that Condition 4 of the WCAP-15376 SE is satisfied because the licensee confirmed the operator actions that are credited have sufficient time, and procedures are in place for the operator actions.

WCAP-15376 SE Condition 5 Condition 5 of the SE for WCAP-15376 is that for future digital upgrades with increased scope, integration, and architectural differences beyond that of Eagle 21, the NRC staff finds the generic applicability of WCAP-15376 to future digital systems is not clear and should be considered on a plant-specific basis.

As stated in its LAR, this condition does not apply to the current Ginna LAR. Therefore, the NRC staff determined that any future digital upgrades should be considered on a plant-specific basis.

3.2.5 Risk Evaluation (Key Principle 4) Conclusion The licensee has demonstrated the technical acceptability and scope of its PRA model, and the PRA model can support implementation of WCAP-14333 and WCAP-15376. The risk metrics satisfy the acceptance guidelines in RG 1.174 and RG 1.177. The licensee has appropriate restrictions in place to provide reasonable assurance that risk-significant configurations will be prevented and avoided. The licensee also has a CRMP that is administratively controlled through plant procedures and updated by qualified staff. For functions that are not generically evaluated in WCAP-14333 and WCAP-15376, the licensee has provided an acceptable plant-specific evaluation for these functions per the guidelines in WCAP-14333 and WCAP-15376. Ginna satisfies the RG 1.174 and RG 1.177 guidelines for the risk evaluation for the proposed changes, and therefore, the NRC staff finds that Key Principle 4 is met.

3.3 Implementation and Monitoring Program (Key Principle 5)

RG 1.174 and RG 1.177 establish the need for an implementation and monitoring program to ensure that extensions to TS CTs and bypass test times do not degrade operational safety over time and that no adverse effects occur from unanticipated degradation or increases in common cause mechanisms. The purpose of an implementation and monitoring program is to ensure that the impact of the proposed TS change continues to reflect the reliability and availability of SSCs impacted by the change. In addition, the application of the three-tiered approach in evaluating the extensions to CTs and bypass test times provides additional assurance that the changes will not significantly impact the key principle of defense in depth. The licensee monitors the reliability and availability of the RTS and ESFAS instrumentation under the Maintenance Rule (10 CFR 50.65), which requires a licensee to monitor the performance or condition of SSCs against licensee-established goals. Ginna satisfies the RG 1.174 and RG 1.177 guidelines for an implementation and monitoring program for the proposed change, and therefore, the NRC staff finds that Key Principle 5 is met.

3.4 Conclusion The NRC staff finds that the licensee has demonstrated the applicability of WCAP-14333 and WCAP-15376 to Ginna and has met the limitations and conditions as outlined in the NRC staffs SE reports. The staff finds the plant-specific analysis is acceptable for the signals not evaluated in the TRs. The Tier 1 conditions were found to be acceptable, and the estimates for CDF, LERF, ICCDP, and ICLERP were found to be within the acceptance guidelines of RG 1.174 and RG 1.177. The licensees Tier 2 analysis evaluated concurrent outage configurations and confirmed the applicability of the risk-significant configurations identified by the WCAP-14333 and WCAP-15376 SE report limitations and conditions and TR analysis to ensure control of these configurations. The licensees Tier 3 CRMP at Ginna was found to be consistent with the CRMP guidance in RG 1.177 and the Maintenance Rule (10 CFR 50.65(a)(4)) for the implementation of WCAP-14333 and WCAP-15376. The licensee monitors the reliability and availability of the RTS and ESFAS instrumentation under the Maintenance Rule (10 CFR 50.65(a)(1)). The NRC staff concludes that the TS revisions proposed by the licensee are consistent with the CTs and bypass test times approved for WCAP-14333 and WCAP-15376 or meet the staffs SE report conditions and limitations for WCAP-14333 and WCAP-15376. For the reasons outlined in the evaluation above, the NRC staff finds that there is reasonable assurance that the technical specifications as amended by the proposed revisions would be consistent with the requirements of 10 CFR 50.36(c)(2) and (c)(3), and therefore are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on December 18, 2020. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on May 5, 2020 (85 FR 26730).

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

7.0 REFERENCES

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15. U.S. Nuclear Regulatory Commission, letter to Nuclear Energy Institute regarding review of Technical Specification Change Traveler, TSTF-418, Rev. 2, RPS and ESFAS Test Times and Completion Times (WCAP-14333) proposed changes to NUREG-1431, Rev. 2, Standard Technical Specifications Westinghouse Plants, dated April 2, 2003 (ADAMS Accession No. ML030920633).

16. Anderson, V., Nuclear Energy Institute, letter to U.S. Nuclear Regulatory Commission entitled, Final Revision of Appendix X to NEI 05-04, 07-12, 12-16, Close-Out of Facts and Observations (F&Os), dated February 21, 2017 (ADAMS Package Accession No. ML17086A431).

17. Giitter, J. and Ross-Lee, M. J., U.S. Nuclear Regulatory Commission, letter to Nuclear Energy Institute, U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os), dated May 3, 2017 (ADAMS Accession No. ML17079A427).

18. Barstow, J., Exelon Generation Company, LLC, letter to U.S. Nuclear Regulatory Commission, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3), dated June 4, 2015 (ADAMS Accession No. ML15166A075).

19. Gudger, D., Exelon Generation Company, LLC, letter to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information - Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program, dated February 3, 2016 (ADAMS Accession No. ML16034A139).

20. Gudger, D., Exelon Generation Company, LLC, letter to U.S. Nuclear Regulatory Commission, Supplemental Response to Request for Additional Information -

Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program, dated March 29, 2016 (ADAMS Accession No. ML16089A425).

21. Electric Power Research Institute (EPRI) Report 302000079, Revision 3, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments.

22. Pacher, J., Constellation Energy, letter to U.S. Nuclear Regulatory Commission, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection of Light Water Electric Generating Plants (2001 Edition), dated March 28, 2013 (ADAMS Accession No. ML13093A064).

23. Render, D., U.S. Nuclear Regulatory Commission letter to Exelon Nuclear, R. E. Ginna Nuclear Power Plant - Issuance of Amendment Regarding Transition to a Risk Informed, Performance-Based Fire Protection Program in Accordance with Title 10 of the Code of Federal Regulations Section 50.48(c) (CAC No. MF1393), dated November 23, 2015 (ADAMS Accession No. ML15271A101).

24. U.S. Nuclear Regulatory Commission, NRC Regulatory Issue Summary 2015-06, Tornado Missile Protection, dated June 10, 2015 (ADAMS Accession No. ML15020A419).

25. Barstow, J., Exelon Generation Company, LLC, letter to U.S. Nuclear Regulatory Commission, Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Focused Evaluation Summary Submittal, dated March 10, 2017 (ADAMS Accession No. ML17069A004).

Principal Contributors: T. Dinh M. Biro T. Sweat G. Singh V. Sreenivas Date: March 11, 2021

ML20353A126 *by memorandum OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DRA/APLB/BC(A)* NRR/DEX/EICB/BC(A)

NAME VSreenivas LRonewicz CMoulton RStattel DATE 12/30/2020 12/30/2020 12/04/2020 12/23/2020 OFFICE NRR/DSS/STSB/BC OGC - NLO NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME VCusumano CCarson JDanna VSreenivas DATE 01/13/2021 01/14/2021 03/11/2021 03/11/2021

v t e Letter Sequence Approval
EPID:L-2020-LLA-0055, TSTF-418 (Approved, Closed)
Initiation Request Acceptance Administration Questions 5 August 2020 Answers 4 September 2020 Results Approval Other:
MONTHYEAR2020-08-05 [Table View]

ML20353A126

Text

SUBJECT:

Dear Mr. Rhoades:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

2.2 Proposed TS Changes

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

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