Text

Supplemental Information No.1 for R.E. Ginna Nuclear Power Plant to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 and 10 CFR 50.69, - Risk-Informed Categorization and Treatment of Structu
	ML21287A006
Person / Time
Site:	Ginna
Issue date:	10/14/2021
From:	David Gudger; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML21287A006 (66)
	v • d • e

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 October 14, 2021 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244

Subject:

Supplemental Information No.1 for R.E. Ginna Nuclear Power Plant to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 and 10 CFR 50.69, Risk-Informed categorization and treatment of structures, systems and components for nuclear power reactors."

References:

1. Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, dated May 20, 2021 (ML21140A324)

2. Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Application to Adopt 10 CFR 50.69, Risk-Informed categorization and treatment of structures, systems and components for nuclear power reactors, dated May 20,2021 (ML21141A009)

3. Letter from V. Sreenivas (Senior Project Manager, U.S. Nuclear Regulatory Commission) R.E. Ginna Nuclear Power Plant - Audit Plan in Support of Review of License Amendment Request Regarding TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4B and 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2021-LLA-0091 and L-2021-LLA-0092) dated August 25, 2021 (ML21222A114)

By letters dated May 20, 2021 (References 1 and 2), Exelon Generation Company, LLC (Exelon) requested to change the R.E. Ginna Nuclear Power Plant (Ginna) Technical Specification (TS). The proposed amendments would modify TS requirements to permit the use of Risk Informed Completion Times in accordance with TSTF-505, Revision 2, Provide

U.S. Nuclear Regulatory Commission Supplemental Information TSTF-505, Risk-Informed Extended Completion Times 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Docket No. 50-244 October 14, 2021 Page 2 Risk-Informed Extended Completion Times - RITSTF Initiative 4b, (ADAMS Accession No. ML21140A324) and modify the licensing basis in accordance with the application to adopt 10 CFR 50.69, Risk-Informed categorization and treatment of structures, systems and components for nuclear power reactors.

By letter dated August 25, 2021 (Reference 3), the NRC notified Exelon of their intent to conduct a regulatory virtual audit the week of September 13, 2021 with Exelon staff and associated contractors in support of the License Amendment Requests (LARs) in References 1 and 2. The letter contain a regulatory virtual audit plan with attached audit questions.

This letter is a supplement to both References 1 and 2 LARs. Attachment 2 to this letter provides a response to several of the audit questions posed by the NRC staff during the regulatory virtual audit. to this letter provides the revised TS markups to address the requested supplemental information. The insert for TS page 5.5-13 is revised to be aligned with the NRC-approved TSTF-505, Revision 2. The information provided in Attachment 1 to this letter supersedes the information provided in Attachments 2 and 3 of Reference 1 for TS pages 3.3.1-4, 3.3.1-5, 3.3.2-1, 3.3.2-2, 3.7.5-2, and 3.8.1-2 and their associated Bases pages and inserts. All other information in Attachments 2 and 3 of Reference 1 remains unchanged.

Exelon has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the NRC in Reference 1 and 2. The supplemental information provided in this letter does not affect the bases for concluding that the proposed license amendments do not involve a significant hazards consideration. Furthermore, the supplemental information provided in this letter does not affect the bases for concluding that neither an environmental impact statements nor an environmental assessment needs to be prepared in connection with the proposed amendments. contains a commitment for 10 CFR 50.69, Risk-Informed categorization and treatment of structures, systems and components for nuclear power reactors. There are no commitments contained in this response for TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b.

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

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

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

U.S. Nuclear Regulatory Commission Supplemental Information TSTF-505, Risk-Informed Extended Completion Times 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Docket No. 50-244 October 14, 2021 Page 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 14th day of October 2021.

Respectfully, David T. Gudger Senior Manager - Licensing Exelon Generation Company, LLC Attachments:

1. Proposed Technical Specification and Technical Specification Bases Marked-Up Pages

2. Response to NRC Audit Questions

3. Proposed Renewed Facility Operating License Changes (Mark-ups) cc: USNRC Region I, Regional Administrator w/ attachments USNRC Senior Resident Inspector, Ginna "

USNRC Project Manager, Ginna "

A. L. Peterson, NYSERDA "

ATTACHMENT 1 Proposed Technical Specification and Technical Specification Bases Marked-Up Pages R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 Docket No. 50-244 TS Pages 3.3.1-4 3.3.1-5 3.3.2-1 3.3.2-2 3.7.5-2 3.8.1-2 TS Bases Pages for information only B 3.3.1-34 B 3.3.1-35 B 3.3.2-26 B 3.3.2-28 B 3.8.1-11

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.

INSERT RICT Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months NOTE 2 R.E. Ginna Nuclear Power Plant 3.3.1-4 Amendment 144

RTS Instrumentation B 3.3.1 K.1 Condition K applies to the following reactor trip Functions:

Pressurizer Pressure-Low;

Reactor Coolant Flow-Low (Two Loops);

RCP Breaker Position (Two Loops);

Undervoltage-Bus 11A and 11B; and

Underfrequency-Bus 11A and 11B.

With one channel inoperable, the inoperable channel must be restored to OPERABLE status or placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or within the RICT. Placing the channel in the tripped condition results in a partial trip condition requiring only one additional channel to initiate a reactor trip. The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowed to place the channel in the tripped condition is consistent with Reference 9 if the inoperable channel cannot be restored to OPERABLE status. Alternatively, a COMPLETION TIME can be determined in accordance with the Risk Informed Completion Time Program.

Allowance of this time interval takes into consideration the redundant capability provided by the remaining redundant OPERABLE channel(s),

and the low probability of occurrence of an event during this period that may require the protection afforded by the Functions associated with Condition K.

For the Reactor Coolant Flow-Low (Two Loops) Function, Condition K applies on a per loop basis. For the RCP Breaker Position (Two Loops)

Function, Condition K applies on a per RCP basis. This Function measures only the discrete position (open or closed) of the RCP breaker, using a position switch. Each RCP breaker has a position switch. Function (10b) requires both breakers to open to cause a reactor trip. However, each breaker position switch has two separate auxiliary contacts: one set of contacts feeds an A train logic relay and the other contact feeds the B train logic relay. Loss of function is dependent on what component fails. If one RCP breaker or one RCP breaker position switch is failed, the function will be lost, but if only one set of contacts or one downstream logic component (e.g., relay) fails, then the other set of contacts and logic train will still be able to provide the function. For Undervoltage-Bus 11A and 11B and underfrequency-Bus 11A and 11B, Condition K applies on a per bus basis. This allows one inoperable channel from each loop, RCP, or bus to be considered on a separate condition entry basis.

The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months while performing surveillance testing of the other channels. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time limit is consistent with Reference 9. The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is applied to each of the remaining OPERABLE channels.

R.E. Ginna Nuclear Power Plant B 3.3.1-34 Revision 61

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.

INSERT RICT 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 INSERT RICT Action A.1 and referenced OPERABLE status. NOTE 2 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.

INSERT RICT 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 B 3.3.1 L.1 If the Required Action and Completion Time of Condition K is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in MODE 1 < 8.5% RTP at which point the Function is no longer required. An alternative is not provided for increasing THERMAL POWER above the P-8 setpoint for the Reactor Coolant Flow-Low (Two Loops) and RCP Breaker Position (Two Loops) trip Functions since this places the plant in Condition M.

The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 1 < 8.5% RTP from full power conditions in an orderly manner and without challenging plant systems.

M.1 Condition M applies to the Reactor Coolant Flow-Low (Single Loop) reactor trip Function. Condition M applies on a per loop basis. With one channel per loop inoperable, the inoperable channel must be restored to OPERABLE status or placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or within the RICT. The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowed to restore the channel to OPERABLE status or place in trip is consistent with Reference 9.

Alternatively, a COMPLETION TIME can be determined in accordance with the Risk Informed Completion Time Program.

The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months while performing surveillance testing of the other channels. The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is applied to each of the two OPERABLE channels. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time limit is consistent with Reference 9.

N.1 Condition N applies to the RCP Breaker Position (Single Loop) trip Function. Condition N applies on a per loop basis. This Function measures only the discrete position (open or closed) of the RCP breaker, using a position switch. Each RCP breaker has a position switch. Function (10a) requires either breaker to open to cause a reactor trip. However, each breaker position switch has two separate auxiliary contacts: one set of contacts feeds an A train logic relay and the other contact feeds the B train logic relay. Loss of function is dependent on what component fails. If one RCP breaker or one RCP position switch is failed, the function will be lost, but if only one set of contacts or one logic component (e.g., relay) fails, then the other set of contacts and logic train will still be able to provide the function. There is one breaker position device per RCP breaker. With one channel per RCP inoperable, the inoperable channel must be restored to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or within the RICT. The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowed to restore the channel to OPERABLE status is consistent with Reference 9.

R.E. Ginna Nuclear Power Plant B 3.3.1-35 Revision 102

ESFAS Instrumentation 3.3.2 3.3 INSTRUMENTATION 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation LCO 3.3.2 The ESFAS instrumentation for each Function in Table 3.3.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.2-1.

ACTIONS

- NOTE -

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Enter the Condition Immediately with one channel or train referenced in Table 3.3.2-1 inoperable. for the channel or train.

B. As required by Required B.1 Restore channel to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months INSERT RICT Action A.1 and referenced OPERABLE status. NOTE 2 by Table 3.3.2-1.

C. Required Action and C.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B not met.

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

by Table 3.3.2-1.

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

by Table 3.3.2-1.

R.E. Ginna Nuclear Power Plant 3.3.2-1 Amendment 109

ESFAS Instrumentation B 3.3.2 B.1 Condition B applies to the AFW-Trip of Both MFW Pumps ESFAS Function. Each MFW pump breaker is equipped with a position sensing device. Both breakers need to trip to start both MDAFW pumps. The two-out-of-two logic requires both a MFW pump A breaker contact and a MFW pump B breaker contact to close. However, each breaker position switch has two separate auxiliary contacts: one set of contacts feeds one train of 2-out-of-2 logic and the other contact feeds the second train of 2-out-of-2 logic. Loss of function is dependent on what component fails. If one MFW pump breaker or one breaker position switch is failed, the function will be lost, but if only one set of contacts or one downstream logic component (e.g., relay) fails, then the other set of contacts and logic train will still be able to provide the function. If a channel is inoperable, 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or within the RICT is allowed to return it to OPERABLE status. These specified Completion Times of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is are reasonable considering the nature of this Function, the available redundancy, and the low probability of an event occurring during this interval.

C.1 If the Required Action and Completion Time of Condition B is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The allowed Completion time is reasonable, based on operating experience, to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems.

D.1 Condition D applies to the following ESFAS Functions:

Manual Initiation of SI;

Manual Initiation of Steam Line Isolation; and

AFW-Undervoltage-Bus 11A and 11B.

If a channel is inoperable, 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or within the RICT is allowed to restore it to OPERABLE status. These specified Completion Times of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is are reasonable considering that there are two automatic actuation trains and another manual initiation channel OPERABLE for each manual initiation Function, additional AFW actuation channels available besides the Undervoltage- Bus 11A and 11B AFW Initiation Function, and the low probability of an event occurring during this interval.

E.1 Condition E applies to the automatic actuation logic and actuation relays R.E. Ginna Nuclear Power Plant B 3.3.2-26 Revision 42

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 and 5b, 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.

INSERT RICT 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 INSERT RICT Action A.1 and referenced OPERABLE status. NOTE 2 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 INSERT RICT 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. INSERT RICT 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Place channel in trip.

R.E. Ginna Nuclear Power Plant 3.3.2-2 Amendment 144

ESFAS Instrumentation B 3.3.2 G.1 If the Required Actions and Completion Times of Conditions D, E, or F are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

H.1 Condition H applies to the following ESFAS functions:

Manual Initiation of CS; and

Manual Initiation of Containment Isolation.

If a channel is inoperable, 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is allowed to restore it to OPERABLE status. The specified Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or within the RICT is reasonable considering that there are two automatic actuation trains and another manual initiation channel OPERABLE for each Function (except for CS) and the low probability of an event occurring during this interval. Because manual actuation of CS requires actuation of 2/2 pushbuttons, RICT cannot be applied to that function.

I.1 Condition I applies to the automatic actuation logic and actuation relays for the following Functions:

SI;

CS; and

Containment Isolation.

Condition I addresses the train orientation of the protection system and the master and slave relays. If one train is inoperable, a Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is allowed to restore the train to OPERABLE status. This Completion Time is reasonable considering that there is another train OPERABLE, and the low probability of an event occurring during this interval. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is consistent with Reference 7.

Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.

R.E. Ginna Nuclear Power Plant B 3.3.2-28 Revision 42

AFW System 3.7.5 CONDITION REQUIRED ACTION COMPLETION TIME D. All AFW trains to one or D.1 more SGs inoperable.

- NOTE -

LCO 3.0.4.b is not applicable.

Restore one AFW train or 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months TDAFW flowpath to each INSERT RICT affected SG to OPERABLE status.

INSERT E. One SAFW train E.1 Restore SAFW train to 14 days RICT inoperable. OPERABLE status.

F. Both SAFW trains F.1 Restore one SAFW train to 7 days inoperable. OPERABLE status.

G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time for Condition A, B, AND C, D, E, or F not met.

G.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months H. Three AFW trains and H.1 both SAFW trains inoperable. - NOTE -

LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one MDAFW, TDAFW, or SAFW train is restored to OPERABLE status.

Initiate action to restore one Immediately MDAFW, TDAFW, or SAFW train to OPERABLE status.

R.E. Ginna Nuclear Power Plant 3.7.5-2 Amendment 88

AC Sources - MODES 1, 2, 3, and 4 3.8.1 CONDITION REQUIRED ACTION COMPLETION TIME B.2 Declare required feature(s) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from supported by the inoperable discovery of DG inoperable when its Condition B required redundant concurrent with feature(s) is inoperable. inoperability of redundant required feature(s)

AND B.3.1 Determine OPERABLE DG 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is not inoperable due to common cause failure.

OR B.3.2 Perform SR 3.8.1.2 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE DG.

AND INSERT B.4 Restore DG to OPERABLE 7 days RICT status.

C. Offsite power to one or

- NOTE -

more 480 V safeguards Enter applicable Conditions bus(es) inoperable.

and Required Actions of LCO 3.8.9, "Distribution AND Systems - MODES 1, 2, 3, and 4," when Condition C is One DG inoperable. entered with no AC power source to one distribution train.

C.1 Restore required offsite 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months INSERT RICT circuit to OPERABLE NOTE 2 status.

OR INSERT C.2 Restore DG to OPERABLE 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months RICT status.

R.E. Ginna Nuclear Power Plant 3.8.1-2 Amendment 109

AC Sources - MODES 1, 2, 3, and 4 B 3.8.1 C.1 and C.2 With offsite power to one or more 480 V safeguards bus(es) and one DG inoperable, redundancy is lost in both the offsite and onsite AC electrical power systems. Since power system redundancy is provided by these two diverse sources of power, the AC power sources are only degraded and no loss of safety function has occurred since at least one DG and potentially one offsite AC power source are available. Also the capability to backfeed through the main transformer using a flexible connection that can be tied to the plant auxiliary transformer to supply required loads is available. However, the plant is vulnerable to a single failure which could result in the loss of multiple safety functions. Therefore, a Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or within the RICT is provided to either restore the offsite power circuit or the DG to OPERABLE status.

This Completion Time is less than that for an inoperable offsite power source or an inoperable DG due to the single failure vulnerability of this configuration.

As discussed in LCO 3.0.6, the AC electrical power distribution subsystem ACTIONS would not be entered even if all AC sources to either train were inoperable, resulting in de-energization. Therefore, the Required Actions of this Condition are modified by a Note which states that the Required Actions of LCO 3.8.9, "Distribution Systems - MODES 1, 2, 3, and 4" must also be immediately entered with no AC power source to one distribution train. This allows Condition C to provide requirements for the loss of an offsite power circuit and one DG, without regard to whether a train is de-energized. LCO 3.8.9 provides the appropriate restrictions for a de-energized train.

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

E.1 If both DGs are inoperable, a loss of safety function would exist if offsite power were unavailable; therefore, LCO 3.0.3 must be entered.

R.E. Ginna Nuclear Power Plant B 3.8.1-11 Revision 74

ATTACHMENT 2 Response to NRC Audit Questions R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 Docket No. 50-244

Attachment 2 Response to NRC Audit Questions Total 47 pages APLA QUESTION 01 - INTERNAL EVENTS AND INTERNAL FLOODING PEER REVIEW [10 CFR 50.69 AND TSTF-505 APPLICABLE]

Regulatory Guide 1.200, Revision 2 (ADAMS Accession No. ML090410014) and Revision 3 (ADAMS Accession No. ML19308B636) provide guidance for addressing PRA acceptability. RG 1.200, Revision 2, describes a peer review process using the ASME/ANS PRA standard ASME/ANS-RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," as one acceptable approach for determining the technical acceptability of the PRA.

For 10 CFR 50.69, LAR Section 3.3, states that the Ginna FPIE PRA model was peer reviewed in June 2009 using the NEI 05-04 process, the PRA Standard (ASME/ANS RA-Sc-2007) and Regulatory Guide 1.200, Revision 1. For TSTF-505, LAR Enclosure 2, states that the Ginna Internal Events PRA model was peer reviewed in June 2009 using the NEI 05-04 process, the PRA Standard (ASME/ANS RA-Sc-2007) and Regulatory Guide 1.200, Revision 1. The Ginna National Fire Protection Association (NFPA) 805 program NRC Safety Evaluation (SE) dated November 23, 2015 (ADAMS Accession No. ML1527A101) states that the June 2009 internal events PRA peer review was performed against the ASME RA-Sb-2005 version of the PRA Standard. Peer review Report LRT-RAM-II-049 indicates that the June 2009 internal events PRA peer review was performed relative to the ASME RA-Sb-2005 version of the PRA Standard.

The Ginna NFPA 805 SE states that the licensee performed a gap assessment between the ASME RA-Sb-2005 as clarified by RG 1.200, Revision 1 and ASME/ANS RA-Sa-2009 as clarified by RG 1.200, Revision 2, and that the licensee did not identify any significant issues for the FPRA from this gap assessment. However, the Ginna NFPA 805 Licensing Amendment Request (LAR) (ADAMS Accession No ML13093A066) states concerning the gaps between the two versions of the PRA Standard and RG 1.200 that the most significant changes [in requirements] occurred in the Internal Flooding portion and it acknowledged differences in internal event and fire PRA requirements between the two versions of the PRA Standard that did not impact the NFPA 805 application.

In light of these observations:

a) Clarify which version of PRA Standard was used in the June 2009 internal events PRA (including flooding) peer review.

1

Response

The Ginna Internal events (including flooding) PRA model was peer reviewed against the ASME RA-Sb-2005 standard as described in Peer review Report LTR-RAM-II-049. This is a correction from the previously noted ASME RA-Sc-2007.

b) Justify that the differences in the Supporting Requirements in the version of the PRA standard used for the June 2009 internal events peer review as clarified by RG 1.200, Revision 1 and the 2009 version of the PRA Standard endorsed by RG 1.200, Revision 2 have an inconsequential impact on the RICT calculations. Include a description of the results of a gap assessment that was performed to evaluate the impact of the gap on the RICT program.

Response

A gap assessment was performed in 2011 to identify which SRs wording and intent has changed in the 2009 version of the standard compared with the 2005 standard (against which the Ginna PRA Peer review was performed). A review of this gap assessment was performed in September 2021 for impact to the RICT and 50.69 programs and it was determined to have an inconsequential impact on these program calculations.

Many of the SR deltas between versions of the standard were addressed with the F&O closures since they were closed to ASME/ANS-RA-Sa-2009. The remaining 9 tasks that remain open to meet the ASME/ANS-RA-Sa-2009 are listed below along with justification of inconsequential impact on the RICT and 50.69 calculations. The remaining open F&O SC-A2-01 applies to task SC.1, SC.2, SC.3, and SC.4 below.

Gap Task Task Description RICT/50.69 Discussion IE.2009.1 Revise the IE Notebook to TSTF 505 and 50.69 LARs addresses Key reflect 2009 Standard text assumptions and sources of Uncertainty to the changes [F&O not programs. Per ASME ANS RA-Sa-2009, "An applicable] assumption is labeled "key" when it may influence (i.e., have the potential to change) the decision being made. Therefore, a key assumption is identified in the context of an application".

Similarly, for Model uncertainty, "A source of model uncertainty is labeled "key" when it could impact the PRA results that are being used in a decision, and consequently, may influence the decision being made. Therefore, a key source of model uncertainty is in the context of an 2

application.". Since the Key assumptions and sources of uncertainty have been addressed in the context of this application, this open item IE.2009.1 will not impact the PRA results that are being used in the decisions of the application.

AS.2009.1 Revise the AS Notebook to Same as response to IE.2009.1 reflect 2009 Standard text changes [F&O not applicable]

SC.1 Review PCTRAN Using core exit temperature or recommended peak calculations to determine node temperature consistent with Capability impact of change in Category II will likely reduce the CDF and LERF "success" definition [F&O estimates. Therefore, potential applications in SC-A2-01] FPIE analysis may be conservative for RICT and 50.69.

SC.2 Set up and run PCTRAN Same as response to SC.1 calculations and review results [F&O SC-A2-01]

SC.3 Update success criteria and Same as response to SC.1 revise PRA as needed and re-quantify the PRA model

[F&O SC-A2-01]

SC.4 Update documentation Same as response to SC.1

[F&O SC-A2-01]

HR.1 Update Human Reliability This task is Documentation only and has no impact Analysis documentation to RICT/50.69 quantifications. F&O items

[F&Os HR-G3-01 and HR- associated with this task are closed.

I1-01]

QU.2 Re-quantify and document Exelon Model of record and Application Specific

[F&O not applicable] Model ongoing process addresses this item. This item will be addressed for RICT and 50.69 by creating a Applications Specific model (ASM) that requires quantification and documentation.

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APLA QUESTION 03 - CREDIT FOR FLEX EQUIPMENT AND ACTIONS [10 CFR 50.69 AND TSTF-505]

The NRC memorandum dated May 30, 2017, Assessment of the Nuclear Energy Institute 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, Guidance for Risk-Informed Changes to Plants Licensing Basis (ADAMS Accession No. ML17031A269), provides the NRCs staff assessment of challenges to incorporating FLEX equipment and strategies into a probabilistic risk assessment (PRA) model in support of risk-informed decision-making in accordance with the guidance of RG 1.200, Revision 2 (ADAMS Accession No. ML090410014).

With regards to equipment failure probability, in the May 30, 2017 memo, the NRC staff concludes (Conclusion 8):

The uncertainty associated with failure rates of portable equipment should be considered in the PRA models consistent with the ASME/ANS PRA Standard as endorsed by RG 1.200. Risk-informed applications should address whether and how these uncertainties are evaluated.

With regards to human reliability analysis (HRA), NEI 16-06 Section 7.5 recognizes that the current HRA methods do not translate directly to human actions required for implementing mitigating strategies. Sections 7.5.4 and 7.5.5 of NEI 16-06 describe such actions to which the current HRA methods cannot be directly applied, such as: debris removal, transportation of portable equipment, installation of equipment at a staging location, routing of cables and hoses; and those complex actions that require many steps over an extended period, multiple personnel and locations, evolving command and control, and extended time delays. In the May 30, 2017 memo, the NRC staff concludes (Conclusion 11):

Until gaps in the human reliability analysis methodologies are addressed by improved industry guidance, [Human Error Probabilities] HEPs associated with actions for which the existing approaches are not explicitly applicable, such as actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06, along with assumptions and assessments, should be submitted to NRC for review.

With regard to uncertainty, Section 2.3.4 of NEI 06-09, Revision 0-A, states that PRA modeling uncertainties shall be considered in application of the PRA base model results to the RICT program and that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties which could potentially impact the results of a RICT calculation. NEI 06-09, Revision 0-A, also states that the insights from the sensitivity studies should be used to develop appropriate RMAs, including highlighting risk 4

significant operator actions, confirming availability and operability of important standby equipment, and assessing the presence of severe or unusual environmental conditions.

Uncertainty exists in PRA modeling of FLEX strategies, related to the equipment failure probabilities for FLEX equipment used in the model, the corresponding operator actions, and pre-initiator failure probabilities. Therefore, FLEX modeling assumptions can be key assumptions and sources of uncertainty for RICTs proposed in this application.

Neither the 10 CFR 50.69 or TSTF-505 LARs state or indicate that FLEX strategies are credited in the PRA models that will be used to support the RICT program or SSC categorization. TSTF-505 LAR Enclosure 4, Section 4 does refer to the availability of FLEX fuel trucks and trailers and portable fuel pumps. Portal report G1-MISC-026 Assessment of Key Assumptions and Sources of Uncertainty for R. E. Ginna PRA indicates that credit for FLEX modeling could be a source of generic modeling uncertainty. However, the report states that no PRA credit is modelled for the initiation of the Emergency Response Organization (ERO) and that any use of FLEX equipment would be driven by emergency and abnormal plant procedures. Accordingly, FLEX strategies do not appear to be credited in the PRA models with two noted exceptions.

Report G1-MISC-026 states that [i]n the level 2 models some credit is given for scrubbing of eluent from a steam generator tube rupture. The report also states that the PRA could also implicitly credit ERO response in the Human Reliability Analysis (HRA) for failure of operators in the long term. If these exceptions are the only instances in which FLEX strategies are credited in the PRA models, then address parts (a) and (b) below. If other credit is taken for FLEX, then address parts (c) through (g) below:

a) Describe the FLEX strategy that is used to credit some scrubbing of effluent from a steam generator tube rupture (SGTR).

i. Describe the equipment used to model this credit and explain whether this includes portable or temporary equipment.

Response

SG scrubbing as credited in the PRA is not a FLEX strategy. Steam Generator scrubbing actions are part of Emergency Operating Procedures.

ii. If the response to part (i) above indicates that the equipment used to credit some scrubbing of effluent from a SGTR is portable or temporary equipment, then address this equipment using the requests in part (e) and (g) below.

5

Response

N/A iii. Describe the operator actions needed to initiate operation of the equipment identified in part (i) above.

Response

N/A iv. If the response to part (ii) above is that the equipment used to credit some scrubbing of effluent from a SGTR is portable or temporary equipment, then address the HRA performed to model actions identified in part (iii) above using the requests in parts (f) and (g) below for these actions.

Response

N/A b) Pertaining to the implicit credit is taken for FLEX in modeling operator failure of long-term actions.

i. Describe how implicit credit is taken for FLEX in modeling operator failure of long-term actions. Include explanation of the term implicit credit and how it is different from taking explicit credit.

Response

Operator actions credited in the Ginna PRA are scoped in from Emergency operating procedures and abnormal operating procedures. ERO actions are not included within this scope. The term implicit credit for ERO response is referring to the ERF review [cognitive recovery factor] that could be applied to a credited operator action using the CBDTM methodology in the HRA calculator.

As applied to the Ginna PRA, this recovery should be considered explicit credit, as the ERF factor can impact the HRA probability. If there is enough time for this ERF recovery to be applied to FLEX (or any HRA) actions, it would be applied using the CBDTM methodology. Currently, the Ginna PRA does not apply this ERF factor to any FLEX actions.

6

ii. Justify that this modeling treatment has an inconsequential impact on the RICT calculations and the categorization of SSCs for 10 CFR 50.69. Alternatively, explain how the results of real time risk model will be used to identify RMAs prior to the implementation of the RICT program, consistent with the guidance in Section 2.3.4 of NEI 06-09, Revision 0-A.

Response

Since this ERF factor would impact an HRAs probability (if applied), it would be reflected in the RICT calculations and the categorization of SSCs for 10 CFR 50.69.

To complete the NRC staffs review of the FLEX strategies modeled in the PRA, the NRC staff requests the following information for the internal events PRA (includes internal floods) and fire PRA, as appropriate:

c) Discuss whether Ginna has credited FLEX equipment or mitigating actions into the Ginna internal events, including internal flooding, or fire PRA models.

If not incorporated or their inclusion is not expected to impact the PRA results used in the RICT program and 10 CFR 50.69, no additional response is requested, and remainder of this question is not applicable.

Response

As part of our NFPA 805 compliance strategy (NFPA805 Safety Evaluation [2015 ML15271A101; 2017 ML15271A101]), installed and portable equipment is credited.

Some of this equipment is also credited as part of the FLEX strategy. Equipment credited for NFPA 805 is also used to mitigate non-fire events which was used as part of the internal events risk offset described in our NFPA805 submittal. To ensure this equipment can be used for all hazards, use of this equipment is directly referenced through the emergency response procedures. Further, the actions to align the equipment required to comply with NFPA 805 are considered time critical actions. All of the risk significant equipment is part of the NFPA 805 monitoring program to ensure the target failure rates and unavailabilities are maintained. This includes: KDG08 (1000 kW DG),

KDG09 (1000 kW DG), and PCH02 (Alternate RCS Injection Pump).

d) [Not applicable to this supplemental LAR]

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e) Regarding the credited equipment:

i. Discuss whether the credited equipment (regardless of whether it is portable or permanently-installed) are like other plant equipment (i.e. SSCs with sufficient plant specific or generic industry data).

If all credited FLEX equipment is similar to other plant equipment credited in the PRA (i.e., SSCs with sufficient plant-specific or generic industry data), responses to items ii and iii below are not necessary.

ii. Discuss the data and failure probabilities used to support the modeling and provide the rationale for using the chosen data. Discuss whether the uncertainties associated with the parameter values are in accordance with the ASME/ANS PRA Standard as endorsed by RG 1.200, Revision 2.

iii. Perform, justify, and provide results of specific sensitivity studies that assess impact on RICT or SSC categorization for 10 CFR 50.69 due to FLEX equipment data and failure probabilities. Part of the response include the following:

1. For 10 CFR 50.69 and TSTF-505, justify values selected for the sensitivity studies, including justification of why the chosen values constitute bounding realistic estimates.

2. For TSTF-505, provide numerical results on specific selected RICTs and discussion of the results.

3. For TSTF-505, describe how the results of the sensitivity studies will be used to identify RMAs prior to the implementation of the RICT program, consistent with the guidance in Section 2.3.4 of NEI 06-09, Revision 0-A.

Response

The portable equipment credited in the full power internal events (FPIE) and fire PRA (FPRA) models will use double the failure rate of similar installed plant equipment. The risk achievement worths (RAW) of the portable equipment ranges from a RAW of 1.01 to 1.11. Risk achievement worths are based on setting the equipment to failure. Even if the equipment is assumed to further double again this (i.e. less than a 10% absolute failure likelihood) as a sensitivity (i.e. one tenth of the hundred percent failure rate of a RAW calculation), this 8

would cause the risk contribution from the portable equipment to be less than 1%.

Further, when the portable equipment is in use due to a severe fire or global failure effect, the safety related equipment is generally failed due to that same global effect (e.g. severe fire on Bus 16 which affects Bus 14 on the level above).

Therefore, increases in portable equipment failure rates would not increase the importance of safety related equipment significantly only increase the overall likelihood of severe fires going to core damage and large early release. Given the lower overall safety significance of the portable equipment coupled with when the equipment is credited, no additional sensitivity evaluations are required.

Given the limited RAW impacts of portable equipment coupled with the already required human action (all portable equipment requires operator alignment) sensitivity study required for 50.69 completely eliminates the requirement for additional sensitivity evaluations in 50.69.

To address concerns with the portable equipment failure rates for risk informed completion time, changes in the allowed outage time for the six technical specifications risk informed completion time cases are analyzed to determine the impact on portable equipment failure rates being double again (i.e. four times the base values). This includes both the independent failure rates as well as the common cause failure rates.

For example in the FPRA model used for in the RICT submittal, the portable equipment credited with the associated mission failure likelihood are:

Total Failure Likelihood over the Mission Base Sensitivity EIN Description (i.e. 2x) 4x KBD01A 100kW DG 4th Option for Long Term Battery 6.51E-02 1.20E-01 Charging PBD04 Alternate RCS Injection Diesel Driven Pump 3rd 6.51E-02 1.20E-01 Back-up for RCP Seal LOCAs PBD02A Flex Fuel Pump - Long Term Fuel Supplies to 6.51E-02 1.20E-01 KDG08, KDG09 and PBD04 PBD02B Flex Fuel Pump - Long Term Fuel Supplies to 6.51E-02 1.20E-01 KDG08, KDG09 and PBD04 TBD01A Flex Fuel Pump Trailer - Long Term Fuel 6.51E-02 1.20E-01 Supplies to KDG08, KDG09 and PBD04 9

Total Failure Likelihood over the Mission Base Sensitivity EIN Description (i.e. 2x) 4x TBD01B Flex Fuel Pump Trailer - Long Term Fuel 6.51E-02 1.20E-01 Supplies to KDG08, KDG09 and PBD04 CBD03A SAFW Vent Fan - Similar to Household Plug-in 3.41E-03 4.92E-03 Fans - Stored in Required Location CBD03B SAFW Vent Fan - Similar to Household Plug-in 3.41E-03 4.92E-03 Fans - Stored in Required Location CBD03C SAFW Vent Fan - Similar to Household Plug-in 3.41E-03 4.92E-03 Fans - Stored in Required Location The resulting impact to technical specifications expected to have a significant impact are:

2x 4x Tech CDF LERF RICT CDF LERF RICT Spec Case Days Days Days Days Days Days Change TD AFW Pump 3-7-5-C Inop 109.3 90.2 30 100.5 90.0 30 N/A SAFW Train 3-7-5-E PSF1A Inop 117.0 89.3 30 116.2 89.3 30 N/A SW Pumps A 3-7-8-B & C Inop 68.6 76.8 30 67.8 76.8 30 N/A Off-site Power 3-8-1-A 7T Inop 58.5 37.5 30 58.0 37.5 30 N/A 3-8-1-B DG A Inop 95.3 89.8 30 92.6 89.7 30 N/A Off-Site Power 7T Inop 3-8-1-C and DG A Inop 16.2 10.5 10.5 16.0 10.5 10.5 0.3%

As expected, the mission failure likelihood of portable equipment doesnt have a significant impact on the RICT and 50.69 programs.

10

f) Regarding human reliability analysis (HRA), address the following:

i. Discuss whether any credited operator actions related to FLEX equipment contain actions described in Sections 7.5.4 and Sections 7.5.5 of NEI 16-06.

If any credited operator actions related to FLEX equipment contain actions described in Sections 7.5.4 and Sections 7.5.5 of NEI 16-06, answer either item ii or iii below:

ii. Perform, justify and provide results of LCO specific sensitivity studies that assess impact from the FLEX independent and dependent HEPs associated with deploying and staging FLEX portable equipment on the RICTs proposed in this application and the 10 CFR 50.69 program. Response should include the following:

1. Justify independent and joint HEP values selected for the sensitivity studies provided to support 10 CFR 50.69 and TSTF-505, including justification of why the chosen values constitute bounding realistic estimates.

2. Provide numerical results on specific selected RICTs and discussion of the results.

3. Discuss composite sensitivity studies of the RICT results and SSC categorization for the operator action HEPs and the equipment reliability uncertainty sensitivity study provided in response to part (4.c.ii) above.

4. Describe how the source of uncertainty due to the uncertainty in FLEX operator actions HEPs will be addressed in the RICT and 10 CFR 50.69 programs. For TSTF-505, describe specific RMAs being proposed, and how these RMAs are expected to reduce the risk associated with this source of uncertainty.

iii. Alternatively, to item ii) above, provide information associated with the following items listed in supporting requirements (SR) HR-G3 and HR-G7 of the ASME/ANS RA-Sa-2009 PRA Standard to support detailed NRC review:

11

1. the level and frequency of training that the operators and/or non-operators receive for deployment of the FLEX equipment (performance shaping factor (a)),

2. performance shaping factor (f), regarding estimates of time available and time required to execute the response,

3. performance shaping factor (g) regarding complexity of detection, diagnosis and decision making and executing the required response,

4. Performance shaping factor (h) regarding consideration of environmental conditions, and

5. Human action dependencies as listed in SR HR-G7 of the ASME/ANS RA-Sa-2009 PRA Standard.

Response

The human error failure (HEF) likelihoods are developed using CBDTM/THERP as in the Ginna NFPA 805 submittal (2015 ML15271A101; 2017 ML15271A101). All of the outside the control room actions required for our design basis compliance are per the rules of our operator action program considered time critical actions (TCA) . Further, these actions apply to all hazards as the procedures only focus on what equipment is lost not what hazard caused the loss.

The time critical actions related to the primary portable equipment are:

Aligning 100kW DG KBD01A to support the chargers during abandonment The time sensitive actions (TSA) related to primary portable equipment described in APLA 03 D are:

Align portable alt RCS injection Changes to these actions are also controlled through NFPA 805 plant change evaluation.

Periodic validations are controlled through our TSA and TCA programs.

Under the RICT program, the risk significant HEFs in a RICT entry are briefed as part of the RMAs. Under the 50.69 program, changes in the HEFs are examined to determine 12

the effects on categorization.

It should be noted alignment of some portable equipment is of such low importance that these actions do not appear in the TSA program. These include alignment of portable ventilation fans (similar to household 120V plug in fans) and refueling of equipment.

Refueling of KDG08 and KDG09 is a long-term action (if required) beyond 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Ventilation fans are only required if normal room cooling is lost, it is hot outside, and the KDG08 fan does not provide adequate cooling.

TCAs are considered part of our design basis. This same time validation is used in the HRA calculator development to ensure agreement between the NFPA 805 performance basis approach and our design program. Operations is required to monitor procedure changes or drill results related to these actions and ensure the times are feasible as part of design compliance (for cause evaluations). Failure to meet a design requirement can result in lost qualifications for operators until remediated. These are periodically assessed to ensure these remain accurate (routine assessment). This is a standard program in the Nuclear Fleet. NRC and Industry references within our procedures are:

NRC Information Notice 97-78, Crediting Of Operator Actions In Place OF Automatic Actions and Modification of Operator Actions, Including Response Times

ANSI/ANS-58.8-1984, Time Response Design Criteria for Nuclear Safety Related Operator Actions

IP 7.11111.11, NRC Inspection Procedure for Licensed Operator Requalification Program

IP 7.11111.21, NRC Inspection Procedure for Component Design Basis Inspections

PWROG-16030-NP, Revision 1, Time Critical Action/Time Sensitive Action Program Standard

WCAP 14996 ERG Operator Response Time Assessment Program Final Report

NEI 12-06 Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, Appendix E Validation Guidance The travel path issues, approaches, controls, and procedures for using the NFPA 805 related equipment are the same as the other actions. As such, the same approach for modeling the actions is appropriate in the Fire PRA Model and Full Power Internal Events Model. External damage from scenarios such as seismic events would require new approaches to estimate removal and or compensation for such damage. Even these could be used in the HRA calculator using stress effects and time delays. As our seismic 13

penalty factor is based on a plant level fragility, this bounds HRA effects.

g) The ASME/ANS RA-Sa-2009 PRA standard defines PRA upgrade as the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences. Section 1-5 of Part 1 of ASME/ANS RA-Sa-2009 PRA Standard states that upgrades of a PRA shall receive a peer review in accordance with the requirements specified in the peer review section of each respective part of this Standard.

Provide an evaluation of the model changes associated with incorporating FLEX mitigating strategies, which demonstrates that none of the following criteria is satisfied:

(1) use of new methodology, (2) change in scope that impacts the significant accident sequences or the significant accident progression sequences, and (3) change in capability that impacts the significant accident sequences or the significant accident progression sequences.

Response

No new approaches related to new uses of the NFPA 805 equipment have been implemented since our NFPA 805 submittal. The same approaches and methodologies that were used to model the portable equipment for the NFPA 805 submittal are also used for the FLEX mitigating strategies.

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APLA QUESTION 05 - TOTAL RISK AND ACCOUNTING FOR THE SOKC [10 CFR 50.69 AND TSTF-505]

RG 1.174 provides the risk acceptance guidance for total core damage frequency (CDF) (1E-04 per year) and LERF (1E-05 per year). NRC staff notes based on RG 1.174 and Section 6.4 of NUREG-1855, Revision 1, for a Capability Category II risk evaluation, the mean values of the risk metrics (total and incremental values) need to be compared against the risk acceptance guidelines. The mean values referred to are the means of the probability distributions that result from the propagation of the uncertainties on the PRA input parameters and model uncertainties explicitly reelected in the PRA models. In general, the point estimate CDF and LERF obtained by quantification of the cutset probabilities using mean values for each basic event probability does not produce a true mean of the CDF and LERF. Under certain circumstances, a formal propagation of uncertainty may not be required if it can be demonstrated that the state of knowledge (SOKC) is unimportant (i.e., the risk results are well below the acceptance guidelines).

For 10 CFR 50.69, LAR Attachment 2 presents the total internal events and internal fire baseline CDF and LERF. The Ginna TSTF-505 LAR presents more total risk information than just the total internal events and fire CDF and LERF. TSTF-505 LAR Enclosure 5, Section 2 states that the total CDF and LERF values presented in Table E5-1 for Ginna are point estimate values which are likely lower than the mean CDF and LERF values. The total CDF and LERF values presented in Table E5-1 include the seismic hazard contribution based on the seismic penalty values that will be used in the RICT calculations, but do not include the high wind hazard contribution. LAR Enclosure 4, Section 4 presents a high winds CDF penalty of 1E-05 per year and LERF penalty of 2E-06 per year for use in the RICT calculations for all plant configurations except those associated with LCO 3.7.5.F, 3.6.2.C, and 3.6.3.E. (For these exceptions, the CDF penalty could be as high as 7E-05 per year and the LERF penalty could be as high as 5E-06 per year.) NRC staff notes given the high wind risk contribution and the potential risk increase due to a possible PRA model update in response to information requests (e.g., requests concerning update of the fire PRA to incorporate internal event F&O resolutions) that the total risk could be higher than shown in the LARs. Therefore, the total CDF could potentially approach the RG 1.174, Revision 3 guidelines of 1E-05 per year when the total mean LERF is used accounting for the SOKC, the high winds risk contribution is included, and potential risk increases associated with model updates performed in response to NRC requests are considered. Therefore, address the following:

a) Demonstrate, that after the total mean internal events and fire CDF and LERF values are calculated to account for the SOKC, the high winds risk contribution is included, and potential risk increases associated with model updates performed in response to NRC 15

requests are considered, the total risk for Ginna is in conformance with RG 1.174 risk acceptance guidelines (i.e., CDF < 1E-04 and LERF < 1E-05 per year). Include identification of the fire PRA parameters that are assumed to correlated in the parametric uncertainty analysis of fire events.

Response

State of knowledge correlation analysis were performed for previous Fire and Internal Events models. This analysis used UNCERT software to perform Monte Carlo calculations to propagate individual basic event uncertainty parameters using random sampling. Epistemic uncertainty parameters are assigned for basic event types including component type failures modes, initiating events, fire ignition frequencies, human action events, and circuit failure probabilities. State of knowledge correlation is used for system and component failure type codes (e.g. all emergency diesel generators fail to start or all salt water pumps fail to run).

The Fire PRA uncertainty analysis shows a negligible - <1% increase- difference between the point estimate and mean for CDF and LERF. This 1% risk increase for Fire CDF and LERF are shown in the table below.

The Internal Events uncertainty analysis shows a larger difference between the point estimate and mean for CDF and LERF. CDF increase is <8% and LERF increase is

<4%. This 8% increase in internal events CDF and 4% increase in internal events LERF are shown in the table below. identifies that, like many other external hazards, high winds (missiles) hazard are screened for total risk at <1E-06 CDF. The conservative upper bound screening value of 1E-06 is shown in the table below for CDF. A 10% factor is assigned for a bounding LERF value for high winds.

Table APLA Adjusted Baseline Risk CDF LERF Source Contribution Contribution Adjusted Internal Events PRA Mean Value 8.10E-06 3.67E-07 Adjusted Fire PRA Mean Value 3.84E-05 5.45E-07 Seismic 3.40E-06 1.90E-06 Bounding High Winds 1.00E-06 1.00E-07 Total 5.09E-05 2.90E-06 16

These adjusted baseline risk values show that the total CDF and LERF values remain well below the RG 1.174 limits.

Note that high winds (missiles) do not screen for all configurations; therefore, a penalty factor is developed for tornado missile risk in the RICT. The RICT program requires a calculation of cumulative risk impact, consistent with guidance in NEI 06 A. The total average annual change in risk for extended completion times will be compared to the guidance of RG 1.174 as described in the guidance (1E-5 CDF and 1E-6 LERF). Corrective actions will be taken if these threshold values are exceeded. The cumulative risk calculation includes the high winds penalty. This process helps assure that the plant risk remains under the RG 1.174 limits.

b) Alternatively, propose a mechanism that ensures calculation of the mean internal events and fire CDFs and LERFs to account for the SOKC, the high winds risk contribution is included, and potential risk increases associated with model updates performed in response to NRC requests are considered prior to implementation of the RICT program.

The mechanism must also ensure confirmation that the updated total CDF and LERF values are still in conformance with the RG 1.174 risk acceptance guidance (i.e., CDF <

1E-04 and LERF < 1E-05 per year) prior to implementation of the RICT program or SSC categorization.

Response

Total CDF and LERF will be recalculated if model updates are performed in response to NRC requests for additional information and compared to the RG 1.174 limits. The updated risk totals will be presented in the RAI responses.

c) Discuss how the SOKC will be addressed for the RICT program and SSC categorization, and how this treatment is consistent with NUREG-1855, Revision 1 when the risk increase associated with SOKC is considered.

Response

SOKC is not identified as a significant source of uncertainty for RICT. Delta risk is dominated by the components out-of-service. There would not be a SOKC between the out-of-service component and the failure rate of like in-service components.

Furthermore, since the RICT program is a delta risk type application, where acceptability is based on the difference between a base model and the configuration-specific model, the potential increase in CDF and LERF using parametric mean values would be reflected in both the base PRA model results and the configuration-specific 17

PRA results.

The 50.69 program includes sensitivity studies to ensure that uncertainties in the PRA are not masking the importance of a SSC. One of these sensitivities is to increase common cause failures. While SOKC was not identified as a significant source of uncertainty for 50.69, potential SOKC uncertainty would be captured by the same steps used to address common cause failures, as the component groups are similar. For example, a sensitivity to increase common cause factors on service water pumps would be like a sensitivity on SOKC between failures of the identical service water pumps.

18

APLB QUESTION 03 AMENDED (FORMALLY APLB-01) - CURRENT FIRE PRA MODEL INCLUDES F&O RESOLUTIONS Regulatory Guide 1.200, Revision 2 (ADAMS Accession No. ML090410014) provides guidance for addressing PRA acceptability. RG 1.200, Revision 2, describes a peer review process using the ASME/ANS PRA standard ASME/ANS-RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," as one acceptable approach for determining the technical acceptability of the PRA. The primary results of peer review are the F&Os recorded by the peer review team and the subsequent resolution of these F&Os. The primary results of peer review are the F&Os recorded by the peer review team and the subsequent resolution of these F&Os. A process to close finding-level F&Os is documented in Appendix X to the Nuclear Energy Institute (NEI) guidance documents NEI 05-04, NEI 07-12, and NEI 12- 13, titled NEI 05-04/07-12/12-06 Appendix X:

Close-out of Facts and Observations (F&Os) (ADAMS Package Accession No. ML17086A431),

which was accepted by the NRC in a letter dated May 3, 2017 (ADAMS Accession No. ML17079A427).

LAR Enclosure 2, Section 3 states that Independent Assessments were performed in 2017 and 2020 to closeout internal events PRA F&Os after the model was updated to resolve F&Os from the 2009 full-scope peer review. LAR Enclosure 2, Section 4 states that the last full-scope peer review of the fire PRA was performed in June 2012 which is significantly before the internal events PRA F&O closure reviews in 2017 and 2020. The LAR does not indicate when the modeling updates to the internal events PRA to resolve F&Os occurred and whether applicable modeling updates were also performed for the fire PRA. Given that internal events PRA provides the modeling foundation for the fire PRA, it is not clear to NRC staff whether F&O resolutions made to the internal events PRA to close F&Os that could impact the fire PRA were incorporated into the fire PRA. Therefore, address the following:

a) Confirm that all internal events PRA modeling updates performed to resolve F&Os that could impact fire risk were incorporated into the fire PRA.

Response

All of the internal events PRA modeling updates that could impact the fire risk have been incorporated into the fire PRA (FPRA). This was originally documented in our NFPA 805 submittal. All of the non-fire only internal events (FPIE) finding closures associated with the 2017 F&O closure review either did not impact the FPRA or have been incorporated. Many of the FPIE findings did not impact the FPRA. This includes flooding findings, SBO recovery modeling, success criteria notebook documentation, etc. Fire induced SBOs are not 19

considered recoverable. In a similar fashion, the recent 2020 F&O closure review had no non-fire internal events findings that affected the fire risk evaluation. The 2020 non-fire only internal events finding closures were related to SGTR scrubbing and internal flooding. In the fire evaluation, post-core damage induced steam generator tube ruptures occur due to dry steam generators with no scrubbing opportunity.

b) If it cannot be confirmed in response to part (a) above that all internal events modeling updates performed to resolve F&Os that could impact fire risk were incorporated into the fire PRA, then propose a mechanism that ensures that all internal events modeling updates performed to resolve F&Os that could impact fire risk are incorporated into the fire PRA prior to implementation of the RICT program. Alternatively, justify that all the internal events modeling updates performed to resolve F&Os have an inconsequential impact on the RICT calculations.

Response

All of the internal events PRA modeling updates that could impact the fire risk have been incorporated into the fire PRA (FPRA).

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APLB QUESTION 03 - NFPA805 MODIFICATIONS AND IMPLEMENTATION ITEMS RG 1.200, Revision 3 and NEI 06-09, Revision 0-A state that the PRA models which support the risk-informed program must be maintained consistent with the as-built, as-operated plant. The Ginna NFPA 805 program NRC SE dated November 23, 2015 (ADAMS Accession No. ML1527A101) cites commitments made in Attachment S of the Ginna NFPA 805 LAR to perform plant modifications and complete implementation items (e.g., updated fire response procedures) before fully transitioning to the NFPA 805 program. These plant improvements were used to offset the risk increase associated with transitioning to the NFPA 805 program and show that the risk acceptance guidelines in RG 1.174, Revision 3 are met. It is not clear to NRC staff whether the promised NFPA 805 plant modifications and implementation items have been completed and whether credited but uncompleted improvements can impact the RICT program. In light of these observations, address the following:

a) Confirm that the fire PRA model used to support the RICT program and 10 CFR 50.69 reflects the as-built, as-operated plant (e.g., do not credit NFPA 805 plant modifications or implementation items that are not yet complete).

Response

The fire PRA (FPRA) model reflects the as-built/as-operated plant with all NFPA 805 modifications and procedure changes complete.

b) If in response to part (a), it cannot be confirmed that the fire PRA models used to support the RICT program and 10 CFR 50.69 reflect the as-built, as-operated plant, then justify that the modeling credit for NFPA 805 plant modifications and/or implementation items not yet completed but credited in the fire PRA do not have a consequential impact on the RICT calculations or 10 CFR 50.69.

Response

N/A per part (a) c) As an alternative to part (b) above, propose a mechanism that ensures that fire PRA models used to support the RICT calculations or 10 CFR 50.69 reflect the as-built, as-operated plant (e.g., do not credit NFPA 805 plant modifications or implementation items that are not complete) prior to implementation of the RICT program or 10 CFR 50.69.

Response

21 N/A per part (a)

Ginna APLC Supplemental Response APLC 02 - RISK CONTRIBUTION OF A SEISMIC EVENT [10 CFR 50.69]

In Title 10 of the Code of Federal Regulation (CFR) 50.69(b)(2)(ii) requires a description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown (including the plant-specific probabilistic risk assessment (PRA),

margins-type approaches, or other systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs.

Section 3.2.3, Seismic Hazards, of the enclosure to the LAR states that low seismic CDF and LERF estimates lead to reasonable confidence that seismic risk contributions would allow reducing an HSS to LSS Section 2.2.2 of the EPRI report identifies the contribution of seismic to total plant risk as a basis for the use of the proposed alternate seismic approach. Further, insights in the EPRI report are derived from the full spectrum of the seismic hazard (i.e., the entire hazard curve). The LAR does not provide information to support the claim that the plant-specific seismic risk is a small contribution to the total plant risk and thereby, the applicability of the proposed alternate seismic approach to the licensee. Based on Enclosure 5 of the Ginna TSTF-505 LAR, seismic CDF is about 7%

of total CDF and seismic LERF is about 68% of total LERF. It appears that SLERF is not a small contribution to the total plant risk.

Justify that the plant specific seismic LERF risk is low relative to the overall plant LERF risk such that the categorization results will not be significantly impacted to support the applicability of the proposed alternate seismic approach.

EXELON RESPONSE The following information is provided to support that seismic risk will not solely result in a high safety significance (HSS) determination based on integrated importance measures and therefore, will not challenge the use of the qualitative consideration of seismic risk in the proposed approach.

1. Ginna Seismic Hazard Meets EPRI 3002017583 Tier 1 Criteria: As discussed in Section 3.2.3 of the Ginna 50.69 LAR (NRC ADAMS Accession No. ML21141A009),

the Ginna seismic hazard curve meets the low hazard (Tier 1) criteria specified in EPRI 3002017583. As stated in the LAR, at these sites, the GMRS is either very low or within the range of the SSE such that unique seismic categorization insights are not expected. For additional perspective, the Ginna safe shutdown earthquake, SSE, design basis is significantly higher than the Ginna latest seismic hazard, i.e., the seismic ground motion response spectra, GMRS, developed in response to NRC request pursuant to 10 CFR 50.54(f) regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (NRC ADAMS Accession No. ML12053A340). A comparison of the Ginna SSE design basis against the Ginna latest GMRS is shown here in Figure APLC2-1 (excerpted from Figure A4-1 of the Ginna 50.69 LAR). As can be seen from Figure APLC2-1, the Ginna design basis is significantly higher than the latest GMRS, by factors of 3x to 6x higher in the 1 hz to 10 hz spectral response frequency range. The 1-10 hz spectral response 22

Ginna APLC Supplemental Response frequency range dominates seismic risk at nuclear power plants given the wide range of structures and key equipment with resonant frequencies in that range.

2. Limited Unique Seismic Insights: The NRC is correct that Section 3.2.3 of the Ginna 50.69 LAR includes statements that imply that low estimated seismic risk is key to the utility of the EPRI 3002017583 Tier 1 seismic alternative process. However, the EPRI seismic alternative report 3002017583 (Section 2.2.2.1) does not explicitly state that the relative contribution of seismic risk is (or needs to be) low compared with the overall plant risk to justify application of the Tier 1 seismic alternative process. In fact, based on the trial studies, EPRI 3002017583 concludes that for both low hazard sites and higher hazard sites that the potential of an SSC being identified as HSS uniquely due to seismic risk calculations is low likelihood:

The test cases described in Section 3 showed that even for plants with high seismic ground motions compared to their design basis, there would be very few if any SSCs designated HSS for seismic unique reasons. At the low seismic hazard sites in Tier 1, the likelihood of identifying a unique seismic condition that would cause an SSC to be designated HSS is very low.

3. Ginna Seismic Risk Low Relative to RG 1.174: The absolute values for seismic risk at Ginna (Seismic CDF (SCDF): 3.4E-06/yr, Seismic LERF (SLERF): 1.9E-06/yr) are low risk relative to NRC RG 1.174 thresholds (CDF: 1E-04/yr, LERF: 1E-05/yr).

4. Ginna FPIE PRA and FPRA More Refined: Comparing the CDF/LERF values from mature, peer reviewed PRA models (for internal events and fire) with an estimate performed for purposes of screening a seismic hazard and/or developing a RICT seismic penalty without consideration of the purpose of the estimate is an inappropriate comparison. The Ginna TSTF-505 SCDF and SLERF seismic penalties are comparatively simple estimations (based on an Individual Plant Evaluation of External Events, IPEEE, era seismic margin assessment) compared to the current very detailed and maintained Ginna FPIE PRA and Fire PRA models. The Ginna SCDF and SLERF seismic penalty calculations do not specifically acknowledge plant improvements since the IPEEE.

5. Conservatisms in Seismic Penalty SLERF Estimate: The estimated seismic LERF used in the Ginna RICT LAR is for the purpose to be used as an SLERF penalty value in RICT calculations and contains conservatisms in the calculation. The calculation of the Ginna RICT SLERF penalty value is performed by assuming a conservative seismic HCLPF (high confidence low probability of failure) to represent containment of radionuclide releases. The same conservative HCLPF used for the SCDF convolution calculation, 0.2g PGA HCLPF, was then also used in the SLERF convolution calculation.

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Ginna APLC Supplemental Response Use of a 0.2g HCLPF to conservatively represent containment of early releases results in a high seismic conditional large early release probability (SCLERP) of 0.56.

Although other TSTF-505 LARs have used this approach in the calculation of an SLERF penalty, it is a conservative approach. A more realistic SCLERP is likely (i.e.,

a Ginna SPRA does not exist to provide plant-specific confirmation) less than half the value used in the Ginna TSTF-505 LAR. Of the nine publicly available PWR NTTF 2.1 SPRA submittals, seven out of the nine exhibit SCLERPs of 0.27 and lower.

And, as noted in Point 4 above, the FPIE and FPRA models are refined whereas the SPRA risk is estimated, which creates an uneven comparison of CDFs and LERF values.

Although not explicitly addressed and accounted for in the NEI 00 04 construct for SSC categorization, a significant fraction of calculated SPRA LERF would not be directly applicable or useful for SSC categorization purposes. This fraction is comprised of seismic induced severe damage states (e.g., seismic induced Reactor Pressure Vessel (RPV) support failure, seismic induced containment failure) that are modeled in the SLERF calculation as leading directly to SLERF based on information from publicly available PWR NTTF 2.1 SPRA submittals. For the majority of the plant SSCs (i.e., other than SSCs such as primary system and safety structures that are already HSS), this portion of the calculated SLERF is not influenced by whether or not an SSC is categorized as HSS or LSS.

Other than the seismic induced direct to SLERF effects (which would not impact categorization), the severe accident progression phenomena and non-seismic failure related events would be quite similar between seismic and fire accident scenarios.

Fire LERF is low as most of the Ginna isolation valves fail safe. All spurious actuation likelihoods and durations from NUREG/CR-7150 Vol. 2 are credited for the containment isolation valves. This makes it very unlikely that an isolation valve would remain in an unsafe state due to fire effects. Further, the Ginna containment isolation procedures are extensive and provide directions for redundant isolation options. Prior to NFPA 805, these isolation procedures were not credited in all fire scenarios. Post-NFPA 805, these procedures have been updated to ensure these containment isolation options are credited in all fire scenarios. Significant seismic events have a very high likelihood of causing loss of auxiliary AC power and making instrument air unavailable which would cause containment isolation valves to fail (i.e., close) in a safe state. The same containment isolation procedures are applicable to seismic events even prior to NFPA 805.

6. Ginna Plant Improvements: A number of plant improvements (physical modifications as well as procedural enhancements) have been instituted at Ginna over the years and these are not explicitly captured in the previously discussed RICT LAR SCDF and SLERF estimates. Some of these plant improvements are discussed below:

24

Ginna APLC Supplemental Response

NFPA 805 mitigating strategies were developed to respond to all hazards including fire, seismic, high wind, station blackout, etc. The NFPA 805 strategies will reduce the risk contribution for SBO scenarios for beyond design basis scenarios (e.g.,

seismic events). Backup diesel generators and a new water storage tank have been added to allow the standby auxiliary feedwater (SAFW) pumps C and D to continue to supply water to the steam generators during station blackout scenarios.

Procedure changes allow the use of the NFPA 805 equipment for all hazards. This equipment includes, among others, the following equipment:

- KDG08: permanently installed, seismically-qualified diesel generator

- KDG09: permanently installed NFPA 805 diesel generator

- KDG01A: portable seismically-qualified 100kW diesel generator (as backup AC supply to a charging pump and to battery chargers)

- PBD01A: portable diesel driven alternate RCS injection pump (back-up to the installed non-portable alternate RCS injection pump)

- TCD05: 160,000 gallon seismically-qualified water storage tank for SAFW and alternate RCS injection

Alternative RCS Injection System: Dedicated pump located in S/B AFW Bldg. and powered by KDG08 or KDG09 (mentioned above). Suction is from Refuel Water Storage Tank (RWST) or spent fuel pool (SFP).

RCP Shutdown Seals: Ginna has a redundant shutdown seal that greatly minimizes seal leakage if the normal RCP seal fails.

Reactor Make-up Water Tank (RWM TK): Per insights from the IPEEE and an Extended Power Uprate commitment, the RWM TK was seismically retrofitted to reduce the likelihood of seismic-induced flooding of key equipment.

Hotwell Level Circuitry: Modification to condenser hotwell control circuitry such that loss of power to hotwell level transmitters will not drain the CST inventory to the hotwell.

25

Ginna APLC Supplemental Response FIGURE APLC2-1 GINNA SSE DESIGN BASIS VS GMRS 26

Ginna APLC Supplemental Response APLC 03 - OVERALL USE OF NEI 00-04 FIGURE 5-6 AND CONSIDERATIONS OF EXTREME WIND OR TORNADOES AND ICE COVER HAZARD [10 CFR 50.69]

NEI 00-04, Revision 0, Figure 5-6 provides guidance to be used to determine SSC safety.

The same document, states, in part, that if it can be shown that the component either did not participate in any screened scenarios or, even if credit for the component was removed, the screened scenario would not become unscreened, then it is considered a candidate for the LSS category.

LAR Section 3.2.4 states that [a]ll external hazards, except for seismic, were screened for applicability to Ginna per a plant-specific evaluation. LAR Attachment 4 lists all hazards as screened except internal events, internal flooding, internal fire, and seismic events for which there are PRA models or in the case of the seismic hazard an alternate approach is used. Except for the external flooding hazard entry in the Attachment 4 table of the LAR, the guidance in NEI 00-04, Figure 5-6 regarding SSCs that play a role in screening a hazard is not discussed in the LAR. Therefore, it appears to NRC staff that at the time an SSC is categorized it may not be evaluated using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard (except the external flooding hazard) because that evaluation has already been made.

The NRC staff notes that plant changes, operational experience, and identified errors or limitations in the PRA models could potentially impact the conclusion that an SSC is not needed to screen an external hazard. Therefore, address the following:

a) Clarify whether an SSC will be evaluated during categorization of the SSC using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard.

b) If SSCs will not be evaluated using the guidance in NEI 00-04, Figure 5-6 to ensure that the SSC are not credited in screening an external hazard at the time of categorization because that evaluation has already been made, then explain how plant changes, plant or industry operational experience, and identified errors or limitations that could change that decision are addressed.

With regards to the extreme wind or tornado hazard, the LAR appears to indicate that screening of this hazard was determined in part on the success of tornado missile barriers after upgrades to several of the barriers are made. Attachment 7 of the LAR indicates that several identified upgrades and modifications are needed to protect against 3-inch pipe missiles generated by tornadoes. The LAR refers to these identified upgrades and modifications as commitments, but it is not clear to the NRC staff what mechanism ensures that these commitments will be completed prior to implementation of the 10 CFR 50.69 program. It is also not clear whether SSCs, including those associated with the cited upgrades, will be evaluated during risk-informed categorization using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard. Therefore, address the following:

c) Clarify whether the tornado missile plant upgrades and modifications discussed in the LAR to protect against 3-inch pipe missiles generated by tornadoes are needed to support the screening of the extreme wind or tornado hazard. If they are needed 27

Ginna APLC Supplemental Response for screening, then propose a mechanism to ensure that the cited upgrades and modifications will be completed prior to implementation of the 10 CFR 50.69 program.

d) Confirm that SSCs will be evaluated using the guidance in NEI 00-04, Figure 5-6 to determine whether SSCs are credited in screening the extreme wind or tornado hazard during 10 CFR 50.69 risk-informed categorization.

With regards to the ice cover hazard, Attachment 4 of the LAR indicates the ice cover hazard (i.e., accumulation of frozen water on bodies of water such as lakes, rivers or on structures, systems, and components) is screened based on the criteria defined in Attachment 5 as C1 (Event damage potential is < events for which the plant was designed) and C4 (Event is included in the definition of another event). However, Section 5.0 of the SE for the Ginna Individual Plant Examination of External Events (IPEEE)

(ADAMS Accession No. ML003773799) states that: The licensee reported that in an earlier plant modification, the power for the heaters on the cooling water intake screens on Lake Ontario had been increased to protect against ice formation (slush). Therefore, it appears there is a potential for ice to form on the cooling water intake screens in the winter that could potentially fail the cooling water supply for such systems as the Ultimate Heat Sink particularly if the heaters (or power to the heaters) are unavailable. It appears that there might be SSCs (e.g., the heaters) credited in screening this hazard.

e) If any SSCs are credited in screening the ice cover hazard, then confirm that SSCs will be evaluated using the guidance in NEI 00-04, Figure 5-6 to ensure that the SSCs credited in screening the ice cover hazard at the time of10 CFR 50.69 risk-informed categorization are identified including heaters and power supplies.

EXELON RESPONSE Exelon Response APLC-03 a):

During the categorization of SSCs, consistent with the guidance in NEI 00-04, Figure 5-6 will be followed.

Exelon Response APLC-03 b):

See the response to APLC-03 a) above.

Exelon Response APLC-03 c):

The tornado missile plant upgrades and modifications discussed in Attachment 7 of the Ginna 50.69 LAR (ML21141A009) to protect against 3-inch pipe missiles generated by tornadoes are needed to support the screening of the extreme wind and tornado hazards.

Exelon proposes to submit the completion of these modifications as a license condition, as was done for the Ginna Risk-Informed Completion Time LAR (ML21140A324).

Exelon Response: APLC-03 d):

See the response to APLC-03 a) above.

Exelon Response APLC3-03 e):

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Ginna APLC Supplemental Response There are no SSCs credited in screening the ice cover hazard. Per UFSAR Section 10.6.2.1, as of September 2018, all 24 plant original heater racks were removed in order to mitigate the effects of frazil ice on plant operations; specifically, to prevent ice buildup on the unheated metal portions of the heater racks as these had been shown to restrict flow and lower Screenhouse level. Further, at conditions of full flow (354,600 gpm) the velocity at the intake screen racks is 0.8 ft/sec. Plant cooling requirements during accident conditions would only be 10,000 gpm with an inlet velocity of 0.02 ft/sec. Water enters the screen racks in a 360-degree circle, protecting against stoppage by a single large piece of material. The low velocity plus the submergence provides assurance that floating ice will not plug the intake.

APLC 04 - EVALUATION OF SEISMIC INDUCED LOSS OF OFFSITE POWER [TSTF-505]

Section 2.3.1, Item 7, of NEI 06-09, Revision 0-A (ADAMS Accession No. ML12286A322),

states that the impact of other external events risk shall be addressed in the [Risk Managed Technical Specifications] RMTS program, and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated

[Risk-Informed Completion Time] RICT. The NRC staffs safety evaluation for NEI 06-09 (ADAMS Accession No. ML071200238) states that [w]here [probabilistic risk assessment] PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.

Section 4 of Enclosure 4 to the LAR does not address the incremental risk associated with seismic-induced loss of offsite power (LOOP) that may occur following the design basis seismic event. The accident scenarios associate with seismically-induced (and therefore unrecoverable) LOOP frequency could already be addressed to some extent in the internal events PRA for unrecovered LOOP events, but this is not explained either.

Demonstrate that seismic-induced LOOP will have an inconsequential impact on the RICT calculations.

EXELON RESPONSE The Ginna TSTF-505 LAR Enclosure 4 seismic penalty calculation is intended to address the fraction of seismic events within (i.e., at or below) the design basis by conservatively including very low magnitude seismic events (as low as 0.0005g peak ground acceleration, PGA, i.e., 1/400th of the Ginna Safe Shutdown Earthquake, SSE of 0.20g PGA) in the SCDF and SLERF convolution calculations. These very low magnitudes are even a factor of 100 lower than the Ginna Operating Basis Earthquake, OBE of 0.08g PGA; the plant is reasonably expected to remain online for seismic events below the OBE.

In response to this question, more discussions and calculations are provided below regarding the inconsequential impact on RICT calculations from plant challenges associated with seismic-induced LOOP from earthquakes within the design basis.

The approach used in the discussion below is the same as used in past LARs that have explicitly discussed this topic, i.e., 1) estimate the annual frequency of seismic-induced 29

Ginna APLC Supplemental Response LOOP; 2) assume no offsite AC recovery within 24 hrs; and 3) compare the result with the internal events PRA frequency estimate for non-recovered LOOP. The methodology used for computing the seismically-induced LOOP frequency is to convolve the Ginna mean seismic hazard curve with an offsite power seismic fragility. Previous TSTF-505 applications have approached this discussion conservatively by performing the seismic-induced LOOP convolution calculation over the entire hazard curve (not just the portion of the hazard curve below the design basis). That same approach is used in this response.

Table APLC4-1 provides the Ginna mean PGA seismic hazard data (this is the same hazard curve data as used in Enclosure 4 of the Ginna TSTF-505 LAR) and the LOOP seismic-induced failure probability (increasing with increasing seismic magnitude) based on the seismic fragility of offsite power. The seismic-induced LOOP convolution calculation in Table APLC4-1 includes the entire seismic hazard curve from earthquakes magnitudes well below the Ginna operating basis earthquake to well beyond the Ginna safe shutdown earthquake.

The failure probabilities for seismic-induced LOOP are represented by failure of ceramic insulators in the offsite AC power distribution system, based on the following seismic fragility data from Table A-0-4 of the NRC RASP Handbook, Volume 2 (NRC ADAMS Accession No. #ML17349A301). This is a common offsite power seismic fragility used for Central and Eastern US SPRAs and seismic risk calculations:

Offsite Power Seismic Capacity (ceramic insulators):

Median Acceleration Capacity, Am = 0.30g PGA

Randomness uncertainty, R = 0.30

Modeling uncertainty, U = 0.45 Given the mean frequency and failure probability for each seismic hazard interval, it is straightforward to compute the estimated frequency of seismically induced loss of offsite power for the Ginna site by multiplying the hazard interval occurrence frequency and the offsite power fragility failure probability. The hazard interval frequency calculation approach and the fragility failure probability calculation approach are the same as that described in Section 3 of Enclosure 4 of the Ginna TSTF-505 LAR. As shown in Table APLC4-1, the total seismic-induced LOOP frequency across the entire seismic hazard curve is estimated at 1.3E 5/yr. Note that this overstates the within design basis challenge frequency but is conservative for this purpose.

The Ginna full-power internal events, FPIE, PRA models LOOP from plant-centered, switchyard-centered, grid-related, and weather-related events. Based on the Ginna FPIE PRA, the total 24-hr non-recovered LOOP frequency is 1.9E-3/yr, as shown in Table APLC4-2.

Assuming offsite AC recovery failure probability of 1.0 for 24 hrs for seismic-induced LOOP, the total (i.e., across the entire hazard curve) 24-hr non-recovered seismic-induced LOOP frequency is 0.7% of the total 24-hr non-recovered LOOP frequency already addressed in the FPIE PRA. The within design basis (i.e., up to the SSE) 24-hr non-30

Ginna APLC Supplemental Response recovered seismic-induced LOOP frequency is approximately 0.5% of the total 24-hr non-recovered LOOP frequency already addressed in the FPIE PRA.

As can be seen, the 24-hr non-recovered seismic-induced LOOP frequency is a very small percentage of the frequency of such challenges already captured in the FPIE PRA (which is explicitly used in RICT calculations) such that it will not significantly impact the RICT Program calculations, and it can be omitted from explicit analysis in RICT calculations.

31

Ginna APLC Supplemental Response TABLE APLC4-1 GINNA SEISMIC-INDUCED LOOP FREQUENCY ESTIMATE (ACROSS ENTIRE SEISMIC HAZARD CURVE) 32

Ginna APLC Supplemental Response TABLE APLC4-2 LOSS OF OFFSITE POWER (LOOP) NON-RECOVERY FREQUENCY Ginna FPIE PRA Ginna FPIE PRA LOOP Initiator Probability of Ginna FPIE PRA Contributor Non-Recovery of 24-hr Non-Frequency(1) Offsite AC by 24 Recovered LOOP LOOP Contributor (/yr) Hrs(2) Frequency (/yr)

Plant-Centered 2.98E-03 8.90E-03 2.4E-05 Switchyard-Centered 1.33E-02 1.26E-02 1.5E-04 Grid-Related 2.95E-02 2.11E-02 5.7E-04 Weather-Related 5.69E-03 2.22E-01 1.2E-03 Total: 1.9E-03 (1) Values per Table 4-1 of Ginna Initiating Events Notebook, G1-PRA-001, Rev. 5, March 2019.

(2) Values per Table 5-1 of Ginna Data Analysis Notebook, G1-PRA-010, Rev. 6, March 2019.

APLC 05 - HIGH WINDS PENALTY FACTORS [TSTF-505]

Section 2.3.1, Item 7, of NEI 06-09, Revision 0-A (ADAMS Accession No. ML12286A322),

states that the impact of other external events risk shall be addressed in the [Risk Managed Technical Specifications] RMTS program, and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated

[Risk-Informed Completion Time] RICT. The NRC staffs safety evaluation for NEI 06-09 (ADAMS Accession No. ML071200238) states that [w]here [probabilistic risk assessment] PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.

In Section 4 of Enclosure 4 to the LAR, the licensee provides its extreme winds analysis, and stated that the high wind hazards can be screened from consideration for the TSTF-505 application, except tornado missiles for certain maintenance configurations.

However, the licensee does not provide the basis for these high wind penalty factors. The LAR states that the CDF penalty is 1E-05 per year and the LERF penalty is 2E-06 per year for all plant configurations associated with LCO conditions encompassed in the RICT program with the exceptions of LCO conditions 3.7.5.F, 3.6.2.C, and 3.6.3.E. For these exceptions, other penalty factors are presented but no explanation of how these additional penalty factors were derived is provided in the LAR.

Enclosure 2 of the LAR provides a completion time of 7 days for the original LCO 3.7.5.F.

Table E1-2 of Enclosure 1 to the LAR shows that a completion time of RICT estimate is 1.4 days for LCO 3.7.5.F, which is much lower than its original completion time.

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Ginna APLC Supplemental Response In light of above observations, address the following:

a) Discuss the calculational basis for each of the extreme wind and tornado CDF and LERF penalty factors presented in the LAR for use in the RICT calculations.

b) Justify that each penalty factor proposed represents a reasonable bounding value per the guidance in NEI 06-09.

c) Discuss the completion times for LCO 3.7.5.F, in which the estimated RICT is lower than its original CT. Is the lower estimated RICT caused by the conservative penalty factor of delta CDF (7E-5 /yr) for LCO 3.7.5.F?

Answers to these questions will support the staffs review of the licensees RICT program implementation and technical basis for seismic penalty factors.

EXELON RESPONSE Exelon Response APLC-05 a):

A conservative risk model was developed to support the screening of the tornado missiles [1]. This risk model was also used to determine the CDF and LERF associated with RICT LCO maintenance configurations. The model was developed from the internal events PRA model and includes the following conservative assumptions:

1. A tornado event is assumed to result in a loss of offsite power (LOOP) with a probability of 1.0 and is considered to be unrecoverable. A recent EPRI report [5]

determines that the offsite power recovery curve (probability of non-recovery vs time) following wind events less than 165 mph is comparable to the typical internal events offsite power recovery curve.

2. Unprotected SSCs, such as the SW pumps, CR HVAC, and A train switchgear (Bus

14) are considered to fail with a probability of 1.0 during a tornado event. It is likely that some of this equipment will not fail for most tornados, except for high intensity tornados with low frequencies. These assumed failures result in limited mitigation paths and thus higher tornado missile risk estimates.

In addition to the conservative model elements, the development of the penalty factor includes the following conservatisms [2]:

1. The estimated CDF associated with tornado missiles is used for the penalty factor, as opposed to CDF. This is comparable to assuming that the zero maintenance CDF for tornado missiles is equal to 0. This results in a penalty factor approximately 35% higher than if the CDF were used. The LERF penalty factor is also based on an estimate of total LERF as opposed to LERF.

2. The highest CDF (with the exception of the case that results in a CCDP of 1.0) is used for all configurations, even though the CDF for most LCO configurations will be significantly less than the values chosen for the penalty factor.

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Ginna APLC Supplemental Response

3. The highest CDF and highest CLERP/CCDP ratio (with the exception of the two cases that result in CCDP and CLERP of 1.0) are used to calculate the penalty factor for LERF, for all configurations. This combination is higher than any actual configurations, since the configuration with the highest CLERP/CCDP ratio and highest CLERP are associated with lower CDF configurations.

Finally, the penalty factor calculation [2] considers two wind-speed intervals:

100 - 150 mph. It is assumed that wind speeds less than 100 mph are not capable of developing damaging tornado missiles [3]. Wind speeds less than 100 mph are considered part of the extreme weather LOOP in the internal events PRA. The frequency of tornados between 100 - 150 mph is 6.6E-5/yr, based on interpolation of the EF-scale tornado wind speeds in Table 6-1 of NUREG/CR-4461, Revision 2 [4].

Greater than 150 mph. 150 mph (3-second gust) is equivalent to the Ginna 132-mph (fastest-mile) design basis wind speed [1]. The frequency of tornados greater than 150 mph is 3E-6/yr, based on interpolation of the EF-scale tornado wind speeds in Table 6-1 of NUREG/CR-4461 [4]. A CCDP of 1.0 are assumed for all tornados greater than 150 mph. It is noted that although successful mitigation of tornado events can be achieved with equipment in structures protected against tornado missiles for wind speeds greater than 150 mph (e.g., B EDG and the SAFW Building and Annex), no credit is taken for mitigation for wind speeds greater than 150 mph.

The model was quantified for maintenance configurations associated with all LCOs in the RICT scope. CCDP and CLERP were determined with the model; CCDP values and CLERP/CCDP ratios were used in the penalty factor calculations, as shown in Tables APLC5-1 through APLC5-4 [2].

CDF Penalty Factor With the exception of LCO 3.7.5.F, the highest CCDP for any of the LCO maintenance configurations is 0.102 (associated with LCOs 3.3.1.R, 3.3.2.I, and 3.7.5.E). This CCDP is used to determine the CDF penalty factor in the Table APLC5-1. The total CDF of 9.8E-6/yr is rounded to 1E-5/yr for the CDF penalty factor.

Table APLC5-1 CDF Calculations for Penalty Factor (not LCO 3.7.5.F)

Tornado Speed (mph) Frequency (/yr) CCDP CDF (/yr) 100-150 6.6E-05 0.102 6.8E-06

>150 3.0E-06 1 3.0E-06 Total 9.8E-06 35

Ginna APLC Supplemental Response The CCDP for LCO 3.7.5.F is 1.0, since the SAFW pumps are relied upon to prevent core damage based on the conservative assumptions previously stated. The total CDF for this LCO is 6.9E-5/yr, which is rounded to 7E-5/yr for the CDF penalty factor (Table APLC5-2).

Table APLC5-2 CDF Calculations for LCO 3.7.5.F Penalty Factor Tornado Speed (mph) Frequency (/yr) CCDP CDF (/yr) 100-150 6.6E-05 1 6.6E-05

>150 3.0E-06 1 3.0E-06 Total 6.9E-05 LERF Penalty Factor As noted in the LAR submittal, two LERF penalty factors are provided. For LCOs other than 3.6.2.C and 3.6.3.E, the LERF penalty factor is calculated in Table APLC5-3. The values used in Table APLC5-3 are based on:

- The highest CCDP other than for LCO 3.7.5.F is 0.102

- The highest CLERP/CCDP ratio other than for LCOs 3.6.2.C and 3.6.3.E is 0.228

- For wind speeds greater than 150 mph, a CCDP of 1.0 is assumed (as previously discussed) and the highest CLERP/CCDP ratio (0.228) is used.

The LERF calculated is 2.2E-6/yr; this is rounded to 2E-6/yr for the LERF penalty factor.

Table APLC5-3 LERF Calculations for Nominal Penalty Factor Tornado Frequency CLERP/ LERF Speed CCDP

(/yr) CCDP (/yr)

(mph) 1.02E-100-150 6.6E-05 2.28E-1 1.5E-06 1

>150 3.0E-06 1 2.28E-1 6.8E-07 Total 2.2E-06 For LCOs 3.6.2.C and 3.6.3.E, there is no impact on CDF, since they only affect the containment function. Therefore, the zero-maintenance tornado missile CCDP of 2.67E-2 is used. For wind speeds greater than 150 mph, a CCDP of 1.0 is used, as discussed previously. The CLERP/CCDP ratio for these LCOs is 1. Table APLC5-4 provides the LERF calculations for these LCOs. The result of 4.7E-6/yr is rounded to 5E-6/yr for the LERF penalty factor.

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Ginna APLC Supplemental Response Table APLC5-4 LERF Calculations for LCO 3.6.2.C and 3.6.3.E Penalty Factor Tornado Frequency CLERP/

CCDP LERF (/yr)

Speed (mph) (/yr) CCDP 100-150 6.6E-05 2.67E-2 1 1.8E-06

>150 3.0E-06 1 1 3.0E-06 Total 4.7E-06 Exelon Response APLC-05 b):

The conservatisms described in response to part a) above and the following considerations result in reasonably bounding penalty factor values.

- The nominal penalty factors are applied to all RICT calculations, while most LCOs have a much lower risk. With the exception of a few LCOs that are primarily for train B SSCs that are tornado missile protected (i.e., 3.3.1.R, 3.3.2.I, 3.7.5.E, 3.8.9.A, 3.8.9.C, 3.8.4.A, 3.8.1.B, and 3.8.1.C), there is no or very little increase in tornado missile CDF and LERF. However, as stated in the LAR submittal, the penalty factors are applied to all configurations.

- Sensitivity cases were run to determine their impact on the penalty factors, specifically the impacts of (1) higher than normal temperatures and (2) decreased city water reliability. It was determined that the penalty factors were still bounding for the sensitivity cases [2].

- As stated in the summary to Enclosure 4, Section 4 of the TSTF-505 LAR submittal, the higher penalty factor (7E-5/yr 1) will be used when certain tornado missile risk significant SSCs are unavailable.

Exelon Response APLC-05 c):

The conservative CDF penalty factor (7E-5/yr) for LCO 3.7.5.F has very little impact on the RICT for that LCO. The CDF for fire risk alone is greater than 2E-3/yr [6] and is the primary driver for the short RICT for LCO 3.7.5.F.

1 During a review of Enclosure 4 of the Ginna TSTF-505 LAR submittal (ML21140A324) in support of the audit question, a typographic error was noticed at the bottom of page E4-19. It states: Additionally, if any of the following conditions exist, the higher CDF of 1E-5/yr should be used, regardless of the LCO(s).

The CDF should be 7E-5/yr in this statement since it is the higher CDF.

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Ginna APLC Supplemental Response References for Responses to APLC-05

[1] G1-MISC-028, Ginna High Winds and Tornado Missile Risk Assessment, Revision 0, April 2021.

[2] G1-MISC-032, Tornado Missile Penalty Factor Calculations, Revision 0, April 2021.

[3] NEI 17-02, Tornado Missile Risk Evaluator (TMRE) Industry Guidance Document, Revision 1B, September 2018.

[4] NUREG/CR-4461, Tornado Climatology of the Contiguous United States, Revision 2, US Nuclear Regulatory Commission, February 2007.

[5] EPRI 3002018232, High Wind Loss of Offsite Power Durations and Recovery, October 2020.

[6] G1-LAR-007, RICT Estimates for TSTF-505 (RICT) Program LAR Submittal, Revision 0, April 2021.

APLC 06 - EXTERNAL FLOODING [TSTF-505]

Section 2.3.1, Item 7, of NEI 06-09, Revision 0-A, states that the impact of other external events risk shall be addressed in the RMTS program, and explains that one method to do this is by documenting prior to the RMTS program that external events that are not modeled in the PRA are not significant contributors to configuration risk. The SE for NEI 06-09 states that[o]ther external events are also treated quantitatively, unless it is demonstrated that these risk sources are insignificant contributors to configuration-specific risk.

LAR Enclosure 4, Section 5 concluded that Ginnas external flooding hazard is screened out because the maximum event (i.e., a LIP combined with a River Flood) that could impact the plant has an exceedance frequency of < 1E-06 per year. However, NRC staff notes that, according to Section 6.1 of the Focused Evaluation report (ADAMS Accession No. ML17069A004), the site protection against flooding events depends on the combination of permanent and temporary passive flood protection barriers to prevent ingress of flood waters in areas with key SSCs. The report refers to temporary portable flood barriers at the Auxiliary Building and Standby Auxiliary Feedwater Pump Building Annex and installed water-resistant doors at the Battery and Diesel Generator Rooms.

The LAR did not describe any RMAs to ensure that the flood protection features, which are integral to flood protection and important for screening of external flooding, continue to be available and functional during the proposed RICTs.

Identify and justify the mechanism that will be used to ensure that the temporary portable flood barriers will be in installed and the water-resistant doors will be closed during a flood event to prevent impact on risk significant equipment.

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Ginna APLC Supplemental Response EXELON RESPONSE The Focused Evaluation (FE) references the probable maximum flood (PMF) as calculated in the Flood Hazard Reevaluation Report (FHRR). The PMF definition was determined utilizing a set of deterministic assumptions, inputs, and methods, as defined in NUREG 0800 and NUREG/CR-7046. Therefore, the PMF value represents a theoretical maximum flood height that does not take frequency of the event into consideration and the plant has provided protection measures for the water surface elevations (WSEs) up to that value.

For this risk-informed LAR, it was necessary to perform an evaluation determining the frequency of various flood heights at Ginna. Using the same flood model from the FHRR, the rainfall associated with a 1E-6/yr frequency was routed through the Ginna flood model to determine the WSE associated with that frequency. It was determined the WSE of a flood with a frequency of 1E-6/yr or less is not high enough to top the sites southern banks and no water is postulated to reach any safety related SSCs. The site does not rely on any temporary barriers or manual actions to screen the flood hazard using PS4 (EXT-C1 Criterion C in ASME/ANS Ra-Sb-2009). Conservatively assuming a conditional core damage probability (CCDP) of 1 for all floods with a frequency equal to or less than 1E-6/yr, the core damage frequency (CDF) is conservatively estimated to be less than 1E-6/yr.

For the local intense precipitation (LIP) mechanism, there are nine doors that are required to be in their normally closed position for the flood seals to operate correctly.

When the station receives warning of more than 4 inches of rain in a 24-hour period, the site enacts ER-SC.2 High Water (Flood) Plan. Section 6 of that document has operators check to ensure the appropriate doors are closed. These doors are credited in screening the LIP mechanism from consideration in the TSTF-505 program. After enacting ER-SC.2 and ensuring the doors are closed, the event damage potential is less than the events for which the plant is designed and therefore can be screened using criteria C1.

APLC 07 - ICE COVER HAZARD [TSTF-505]

Section 2.3.1, Item 7, of NEI 06-09, Revision 0-A, states that the impact of other external events risk shall be addressed in the RMTS program, and explains that one method to do this is by documenting prior to the RMTS program that external events that are not modeled in theP RA are not significant contributors to configuration risk. The SE for NEI 06-09 states that[o]ther external events are also treated quantitatively, unless it is demonstrated that these risk sources are insignificant contributors to configuration-specific risk.

LAR Enclosure 4, Table E4-4 indicates the ice cover hazard (i.e., The accumulation of frozen water on bodies of water such as lakes and rivers or on structures, systems, and components) is screened based on the Table E4-5 criteria of C1 (Event damage potential 39

Ginna APLC Supplemental Response is < events for which the plant was designed) and C4 (Event is included in the definition of another event). Section 5.0 of the SE for the Ginna Individual Plant Examination of External Events (IPEEE) states that: The licensee reported that in an earlier plant modification, the power for the heaters on the cooling water intake screens on Lake Ontario had been increased to protect against ice formation (slush). Therefore, it appears there is a potential for ice to form on the cooling water intake screens in the winter that could potentially fail the cooling water supply for such systems as the Ultimate Heat Sink, particularly if the heaters (or power to the heaters) are unavailable.

It is not clear to the NRC staff how the criteria cited above are used to screen this hazard event for all plant configurations encompassed in the RICT program. Section 6 of LAR states for configurations allowed by the RICT program that hazards for which the ability to achieve safe shutdown may be impacted by one or more plant configurations, the impact of the hazard to particular SSCs is assessed and a basis for the screening decision applicable to configurations impacting those SSCs is provided. Accordingly, given that ice cover could impact the ability to achieve safe shutdown, especially for certain configurations, address the following:

a) Explain how the C1 and C4 screening are used to screen the ice cover hazard from consideration for impact on RICT calculations given that ice cover events appear to be anticipated and could be a contributor to a core damage accident particularly for certain plant configurations. Include discussion of the heaters used to keep the ice clear from the cooling water intake screens.

b) If the screening criteria cited in the LAR are not sufficient to screen the ice cover hazard from consideration for impact on the RICTs for all plant configurations encompassed in the RICT program, then justify screening the ice cover hazard using another basis.

c) If it cannot be justified that the Ice Cover hazard can be screened for impact on the RICT calculations, then explain how the RICT program will mitigate or prevent the impact of the ice cover hazard during a RICT application.

EXELON RESPONSE Exelon Response APLC-07 a):

Per UFSAR Section 1.2.12, station water use is primarily via the circulating water and service water systems. Lake Ontario is the source of circulating water, which is taken through eight 17.3 ft. wide by 10 ft. high ports of the submerged octagonal intake structure that lies about 3100 ft. offshore in about 33 ft. of water at mean lake level, 244.7 ft.

Per UFSAR Section 10.6.2.1, as of September 2018, all 24 plant original heater racks were removed in order to mitigate the effects of frazil ice on plant operations; specifically, to prevent ice buildup on the unheated metal portions of the heater racks as these had been shown to restrict flow and lower Screenhouse level. Further, at conditions of full flow (354,600 gpm) the velocity at the intake screen racks is 0.8 ft/sec.

Plant cooling requirements during accident conditions would only be 10,000 gpm with an 40

Ginna APLC Supplemental Response inlet velocity of 0.02 ft/sec. Water enters the screen racks in a 360-degree circle, protecting against stoppage by a single large piece of material. The low velocity plus the submergence provides assurance that floating ice will not plug the intake. Therefore, the event damage potential is less than events for which the plant was designed (i.e.,

Criterion C1).

In addition, the ice cover can accumulate on SSCs in addition to bodies of water. As stated in Enclosure 4 of the Ginna TSTF-505 LAR (ML21140A324), the principal effects of ice accumulation would be to cause a loss of offsite power event, which is addressed in the internal events PRA model for Ginna. Therefore, the ice cover hazard is included in the definition of the LOOP event (i.e., Criterion C4).

Exelon Response APLC-07 b):

See the response to APLC-07 a) above.

Exelon Response APLC-07 c):

The response in APLC-07 a) above that justifies that the Ice Cover hazard can be screened for impact on the RICT calculations.

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EICB Question 01:

7-1 Page 15 of 359 of the LAR (ML21140A324) includes INSERT RICT NOTE 1 however, this insert does not appear to be used in the LAR. Please explain.

RESPONSE

In earlier drafts of the Ginna RICT submittal, there were some TSs being considered for RICT application that had concerns over trip capabilities. These TSs were removed in later drafts and the NOTE 1 insert was erroneously left in.

EICB Question 02:

7-2 LAR Enclosure 1 Table E1-1 seems to depict several loss of TS required functions, please clarify whether these items are or are not a loss of function.

Tech Function Design Explanation Spec Success (TS) Criteria 3.3.1.K (10b) RCP breaker One open TS Table 3.3.1-1 identifies one position (Two Loops) breaker per channel per RCP. If this RCP channel is lost, then the associated TS function is lost.

Since Condition K applies to other functions, should the proposed TS change be in to Insert RICT Note 2?

3.3.1.N (10a) RCP breaker One open TS Table 3.3.1-1 identifies one position (Single Loop) breaker per channel per RCP. If this RCP channel is lost, then the associated TS function is lost.

Should the proposed TS change be in to Insert RICT Note 2?

3.3.1.P (14) (b) Turbine Two of two A design success criteria of Stop Valve Closure channels two of two seems like a loss of TS function if one is out.

Since Condition P applies to other functions, should the proposed TS change be in to Insert RICT Note 2?

3.3.1.U (18) Reactor Trip One trip Needs some discussion to Breaker Undervoltage mechanism understand.

and Shunt Trip per RTB Mechanisms 3.3.2.B (6f) Auxiliary Two of two Need explanation of logical 42

Feedwater-Trip of Channels per arrangement (e.g., voting) of Both Main Feedwater MFW Pump channels.

Pumps 3.3.2.H Containment Spray; (2a) Two of A design success criteria of (2a) Manual Initiation two two of two seems like a loss pushbuttons of TS function if one is out.

Since Condition H applies to other functions, should the proposed TS change be in to Insert RICT Note 2?

The TS Bases state: The operator can initiate CS at any time from the control room by simultaneously depressing two CS actuation pushbuttonsthe inoperability of either pushbutton fails both trains of manual initiation.

3.3.5.A Containment A design success criteria of Radiation Signal from one of one seems like a loss either of 2 channels: of TS function if one is out.

Gaseous: one of one Channel Particulate: Should the proposed TS one of one channel change be in to Insert RICT Note 2?

As stated in the TS Bases for these radiation monitors, they are of different design (R-11 is particulate and R-12 is gaseous),

and R-11 is more sensitive to Reactor Coolant Leakage. These are important distinctions for normal and off normal operating conditions, such as credit for minimal RCS leakage detection.

However, for purposes of Post-LOCA Containment Ventilation Isolation of the mini-purge valves, their TS 3.3.5 function, the response of both detectors is timely and effective. Both detectors will isolate the mini-purge valves within the time 43

constraints assumed in the accident analysis. Therefore, there does exist redundancy, and there is no loss of function.

RESPONSE

1. TS 3.3.1.K, function 10(b), RCP Breaker position (two loops)

Response - Per Bases, Condition K applies on a per RCP basis. This Function measures only the discrete position (open or closed) of the RCP breaker, using a position switch. Each RCP breaker has a position switch. Function (10b) requires both breakers to open to cause a reactor trip. However, each breaker position switch has two separate auxiliary contacts: one set of contacts feeds an A train logic relay and the other contact feeds the B train logic relay. Loss of function is dependent on what component fails. If one RCP breaker or one RCP breaker position switch is failed, the function will be lost, but if only one set of contacts or one downstream logic component (e.g., relay) fails, then the other set of contacts and logic train will still be able to provide the function. The proposed TS change should be in to Insert RICT Note 2 for TS 3.3.1.K and the Technical Specifications Basis will be revised to clarify what a loss of function would be for RCP breaker position (two loops).

2. TS 3.3.1.N, function (10a), RCP Breaker position (one loop)

Response - Per Bases, Condition N applies on a per loop basis. This Function measures only the discrete position (open or closed) of the RCP breaker, using a position switch. Each RCP breaker has a position switch. Function (10a) requires either breaker to open to cause a reactor trip. However, each breaker position switch has two separate auxiliary contacts: one set of contacts feeds an A train logic relay and the other contact feeds the B train logic relay. Loss of function is dependent on what component fails. If one RCP breaker or one RCP position switch is failed, the function will be lost, but if only one set of contacts or one logic component (e.g., relay) fails, then the other set of contacts and logic train will still be able to provide the function. The proposed TS change should be in to Insert RICT Note 2 for TS 3.3.1.N and the Technical Specifications Basis will be revised to clarify what a loss of function would be for RCP breaker position (single loop).

3. TS 3.3.1.P, function (14b), Turbine Stop Valve Closure Response - Based on the discussion with the reviewer, this question will be withdrawn

4. 3.3.1.U, function (18), Reactor Trip Breaker UV and STA mechanisms Response - Per the Bases, one diverse trip feature can be inoperable on each RTB (either UV or STA). Reactor trip capability is maintained for each of the trip breakers. Therefore, this is not a loss of function.

5. TS 3.3.2.B, function (6f), AFW initiate on loss of main feed pumps Each MFW pump breaker is equipped with a position sensing device. Both breakers need to trip to start both MDAFW pumps. The two-out-of-two logic requires both a MFW pump A breaker contact and a MFW pump B breaker contact to close. However, each breaker position switch has two separate auxiliary contacts: one set of contacts feeds one train of 2-out-of-2 logic and the other contact feeds the second train of 2-out-of-2 logic. Loss of function is dependent on what component fails. If one MFW pump breaker or one breaker position switch is failed, the function will be lost, but if only one set of contacts or one downstream logic component (e.g., relay) fails, then the other set of contacts and logic train will still be able to provide the function. The proposed TS change should be in to Insert RICT Note 2 for TS 3.3.2.B and the Technical Specifications 44

Basis will be revised to clarify what a loss of function would be for AFW initiation on loss of main feed pumps.

6. TS 3.3.2.H, function (2a), Containment Spray Manual Initiation Response - Since loss of either pushbutton fails both trains of manual initiation, Exelon agrees that this proposed TS change note should be stated as Insert RICT note 2.

7. TS 3.3.5.A, Containment Radiation from either of 2 channels Response - As stated in the TS Bases for these radiation monitors, they are of different design (R-11 is particulate and R-12 is gaseous), and R-11 is more sensitive to Reactor Coolant Leakage. These are important distinctions for normal and off normal operating conditions, such as credit for minimal RCS leakage detection. However, for purposes of Post-LOCA Containment Ventilation Isolation of the mini-purge valves, their TS 3.3.5 function, the response of both detectors is timely and effective. Both detectors will isolate the mini-purge valves within the time constraints assumed in the accident analysis. Therefore, there does exist redundancy, and there is no loss of function.

EICB Question 03:

7-3 LAR Attachment 1, page 3 pf 6 states:

TS 3.3.5.A.1- One rad monitor inoperable. Per UFSAR Section 6.2.4.3, there is no loss of function if R-11 or R-12 become inoperable. These radiation monitors actuate Containment Ventilation Isolation (CVI), for the mini-purge valves. CVI serves as a backup to the Containment Isolation (CI) signal, and is not specifically credited in the accident analysis.

This is a TS loss of function. If the function is in the TS, it does not matter whether the function is credited in the accident analysis. Generally. if a function is not credited in the accident analysis, then the loss of the function is most likely a reduction in defense-in-depth.

RESPONSE

TS 3.3.5.A, Containment Radiation from either of 2 channels Response - As stated in the TS Bases for these radiation monitors, they are of different design (R-11 is particulate and R-12 is gaseous), and R-11 is more sensitive to Reactor Coolant Leakage. These are important distinctions for normal and off normal operating conditions, such as credit for minimal RCS leakage detection. However, for purposes of Post-LOCA Containment Ventilation Isolation of the mini-purge valves, their TS 3.3.5 function, the response of both detectors is timely and effective. Both detectors will isolate the mini-purge valves within the time constraints assumed in the accident analysis. Therefore, there does exist redundancy, and there is no loss of function.

EEEB RAIs :

EEEB Request for Additional Information (RAI) No. 1 According to TS Condition 3.8.1.A, Action A.2, the Completion Time is modified by INSERT RICT Note 2 (Not applicable if there is a loss of function); whereas the Completion Time for Condition 3.8.1.C, Action C.1 is modified by INSERT RICT.

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Please explain why Note 2 of RICT is not applicable to Condition 3.8.1.C, Action C.1, although the scenario Offsite power to one or more 480 V safeguards bus(es) inoperable is common in both the Conditions 3.8.1.A and 3.8.1.C.

Question 1 response TS 3.8.1.C. 1 is comparable to 3.8.1.A.1 and should also have INSERT RICT NOTE 2 specified. Exelon will make that change.

EEEB RAI No. 2 According to TS Condition 3.8.7.B, Action B.2, the Completion Time is modified by INSERT RICT Note 2 (Not applicable if there is a loss of function).

Please explain why Note 2 of RICT is applicable to the Condition 3.8.1.C, Action B.2, for the scenario Class 1E CVT for AC Instrument Bus B inoperable.

Question 2 response TS 3.8.7.B is a Ginnaspecific condition. The use of a CVT is below the level of detail specified in TSTF505. Note 2 was used because inoperability of the CVT could APPEAR to be a loss of function. So we used an explanatory note in Table 11 to indicate that the inoperability of a CVT for Instrument Bus B can be accommodated, in that the Ginna design includes a nonClass 1E CVT to supply power to Instrument Bus B, so there is no loss of function.

Question Technical Specification 3.7.5, Condition D, is entered when All AFW trains to one or more SGs inoperable. and has a required action to Restore one AFW train or TDAFW flowpath to each affected SG to OPERABLE status, with a completion time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . In GINNA LAR, in Table E1-3: TSTF-505 Rev 2 Table 1 Technical Specifications (TS) that Require Additional Justification, for T.S. 3.7.5.D the following justification was provided.

As long as one of the five AFW pumps is available to one SG, there is no loss of function except for the low probability event of a FLB on that SG.

The above suggest that when in LCO 3.7.5.d, there is a potential loss of function in which case the RICT would not be applicable. UFSAR section 10.5.1 states that the auxiliary feedwater systems consist of a preferred auxiliary feedwater system and a standby auxiliary feedwater system (SAFW), and UFSAR Section 10.5.2 states that the main function of the auxiliary feedwater system is to maintain the steam generator water inventory when the normal 46

feedwater system is not available. Since the justification suggest that both preferred and standby AFW trains could be available, please clarify the following:

1. What AFW trains are considered inoperable per LCO 3.7.5.d

2. If SAFW trains are available, and they are capable of independently supply water to both steam generators, specify the loss of function associated with the FWLB that is referred to in for T.S. 3.7.5.d. in LAR Table E1-3

3. Provide a more detailed justification in LAR Table E1-3.

Response

1. 3.7.5.D refers to the preferred (MDAFW and TDAFW) trains only Standby AFW is covered by 3.7.5.E and F. NOTE 2 will be removed.

2. If all AFW flow (MDAFW and TDAFW) is lost to one of the two SGs, additional sources of cooling via the Standby Auxiliary Feedwater System is available. This system is independent of the preferred AFW system, having its own cooling source (160,000 gallon DI tank) and a separate power source available (KDG08) as described in our NFPA 805 submittal (ML18114A025).

3. LAR Table E 1-3 is being revised to reflect this information. It will be revised to be consistent with 2 above.

EEEB RAI No. 3 In Table E1-1 In Scope TS/LCO Conditions to Corresponding PRA Functions, of LAR, differences in Design Success Criteria versus PRA Success Criteria are noted in the following electrical power system related TS Conditions: 3.8.7.A (One inverter inoperable);

3.8.7.B (Class 1E CVT for AC Instrument Bus B inoperable); 3.8.9.B (One AC instrument bus electrical power distribution train inoperable).

Please explain how the Design Success Criteria versus PRA Success Criteria will be applied during calculations of the RICT, for the TS conditions 3.8.7.A; 3.8.7.B; and 3.8.9.B.

Response

Design function is to provide necessary power to RPS and ESF so that fuel, RCS, and containment design limits are not exceeded.

PRA function is to prevent or mitigate core damage. Core damage is prevented by providing decay heat removal, inventory control, and reactivity control. The instrument busses each have different functional impacts. This is reflected in the PRA model. Thus, the success in the PRA cannot be completely described by how many Instrument busses are available. The success criteria to mitigate core damage in the PRA is dependent on which higher level systems are impacted upon loss of an individual instrument bus. This dependency is explicitly modeled in the PRA model with each unique configuration. For example, any equipment powered from Instrument BUS (IB) A will be lost (e.g. NIS Channels, MFW Controls, CCW Temp Alarms, Charging Indication, etc.).

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ATTACHMENT 3 Proposed Renewed Facility Operating License Changes (Mark-ups)t R. E. Ginna Nuclear Power Plant Docket No. 50-244 FOL Page 9

ATTACHMENT 3 INSERT 1 (18) Adoption of Allowance to Implement the Provisions of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components in Nuclear Power Reactors Exelon is approved to implement 10 CFR 50.69, allowing a risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors. The categorization process is consistent with NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Rev. 0, dated July 2005, as endorsed by the NRC in Regulatory Guide 1.201, Rev. 1, dated May 2006.

Exelon will complete the implementation items listed in Attachment 7 of Exelon Letter to the NRC dated May 20, 2021, prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa -2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to the implementation of 10 CFR 50.69.

ATTACHMENT 3 (15) At least half the members of the CENG Board of Directors must be U.S. citizens.

(16) The CENG Chief Executive Officer, Chief Nuclear Officer, and Chairman of the CENG Board of Directors must be U.S. citizens.

These individuals shall have the responsibility and exclusive authority to ensure and shall ensure that the business and activities of CENG with respect to the facilitys license are at all times conducted in a manner consistent with the public health and safety and common defense and security of the United States.

INSERT 1 (17)

(18)

D. The facility requires an exemption from certain requirements of 10 CFR 50.46(a)(1). This includes an exemption from 50.46(a)(1), that emergency core cooling system (ECCS) performance be calculated in accordance with an acceptable calculational model which conforms to the provisions in Appendix K (SER dated April 18, 1978). The exemption will expire upon receipt and approval of revised ECCS calculations. The aforementioned exemption is authorized by law and will not endanger life property or the common defense and security and is otherwise in the public interest. Therefore, the exemption is hereby granted pursuant to 10 CFR 50.12.

E. Exelon Generation shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27827 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "R. E. Ginna Nuclear Power Plant Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan," submitted by letter dated May 15, 2006.

Exelon Generation shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The licensee's CSP was approved by License Amendment No. 113 and modified by License Amendment No. 117. The licensee has obtained Commission authorization to use Section 161A preemption authority under 42 U.S.C. 2201a for weapons at its facility.

v t e Letter Sequence Supplement
EPID:L-2021-LLA-0091, TSTF-505, TSTF-3, TSTF-505 (Approved, Closed)
Initiation Request, Request, Request Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement Administration Questions Answers Results Approval, Approval, Approval Other: ML21222A114
MONTHYEAR2022-04-28 [Table View]

ML21287A006

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