Text

Unit 2 License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating
	ML24094A288
Person / Time
Site:	Cook
Issue date:	04/03/2024
From:	Ferneau K; Indiana Michigan Power Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	AEP-NRC-2024-02
	Download: ML24094A288 (1)
	v • d • e

Indiana Michigan Power INOIANA Cook Nuclear Plant MICHIGAN One Cook Place POWIR. Bridgman, Ml 49106 indianamichiganpower.com

An MP Company

BOUNDLESS ENERGY-

April 3, 2024 AEP-NRC-2024-02 10 CFR 50.90

Docket No: 50-315 50-316

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Donald C. Cook Nuclear Plant Unit 1 and Unit 2 License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the Licensee for Donald C.

Cook Nuclear Plant (CNP) Unit 1 and Unit 2, proposes to amend Facility Operating License DPR-58 and DPR-74 Technical Specification (TS) 3.8.1, "AC Sources-Operating." For both Unit 1 and Unit 2, TS 3.8.1 requires two qualified offsite circuits and separate and independent Diesel Generators (DGs) for each train to ensure availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an anticipated operational transient or a postulated design basis accident (OBA). If one of the offsite circuits is inoperable, TS 3.8.1 Condition A requires that the circuit be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant be in Mode 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

A modification has been developed to replace the 12AB (Train 8) Loop Feed Enclosure (LFE) and associated bus for the Train B reserve feed preferred power source coming into CNP with a cable bus design. This reserve feed power source feeds both Unit 1 and Unit 2 12AB Reserve Auxiliary Transformers. This modification is being performed to improve the long-term reliability of the 12AB LFE which has experienced failures from environmental conditions created by moisture from CNP's steam generator blowdown system effluent. Due to the scope of the modification, the completion of the modification will exceed the TS 3.8.1, Required Action A.3 Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Based on time estimates for the modification, l&M is requesting an additional 9 days for a one-time extension of the Completion Time to 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months ( 12 days).

Enclosure 1 to this letter provides an affirmation affidavit pertaining to the proposed amendment.

Enclosure 2 provides a detailed description and safety analysis to support the proposed amendment, including an evaluation of no significant hazards considerations pursuant to 10 CFR 50.92(c), and an environmental assessment. Enclosure 3 and 4 provide the TS page marked to show the proposed change for Unit 1 and Unit 2 respectively.

The modification is currently scheduled to start at 0400 hours0.00463 days 0.111 hours 6.613757e-4 weeks 1.522e-4 months on September 16, 2024. The start of this modification work requires entry into both Unit 1 and Unit 2 TS 3.8.1, Required Action A.3, which has a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time. l&M requests approval of the proposed amendment by U.S. Nuclear Regulatory Commission AEP-NRC-2024-02 Page2

September 9, 2024. l&M realizes that this requested approval is significantly shorter than the NRC's normal 12 month review and approval time frame. Since the degradation of the 12AB LFE was first identified in 2021, l&M has been working towards a long-term solution. This required an interim short-term fortification of the 12AB LFE, a study to determine possible design solutions and the development of the detailed design of the modification that was selected to address this issue. Based on plant-specific operating considerations, detailed below, the window for implementing the modification is mid-September to early October. It is highly desired to implement this modification at the earliest opportunity, and a delay in approval past September of 2024 is expected to delay implementation of the modification one full year to September 2025.

Per TS Limiting Condition for Operation (LCO) 3.8.1, the required qualified electrical sources from one Unit must include the qualified electrical sources from the other Unit that are required to support the Essential Service Water (ESW) System when the ESW unit cross-tie valves are open.

From TS LCO 3.8.1 Bases, When an ESW train is not isolated from the opposite Unit, the opposite Unit's AC sources are required to be OPERABLE and capable of supplying the appropriate Unit class 1 E Distribution subsystems. If an opposite Unit's qualified circuit is not supplying the appropriate Unit Class 1 E Distribution subsystem, then the required opposite Unit's preferred qualified circuit must be OPERABLE with the capability to fast transfer to the appropriate Class 1 E Distribution subsystem.

The preferred qualified circuit in question is TR101AB (TR201AB). Since the ESW system is shared between Units, and they supply the opposite train from Unit to Unit (Unit 1 East ESW supplies Unit 2 West ESW, and vice versa), the cross-tie valves between Units must be closed when the preferred offsite circuit is INOPERABLE.

When the ESW cross-tie valves between each Unit are closed, all 4 ESW pumps are required to be in operation to supply each Unit's respective East & West ESW headers. Each ESW Pump has a minimum required flow, per design, to ensure longevity of the pump (2000 GPM). To achieve this flow, the ESW cooling water flow through the Component Cooling Water Heat Exchanger (CCW) is adjusted. When lake water temperatures are low (below approximately 50-55° Fahrenheit (°F), it is very difficult to maintain required CCW temperatures (while maintaining minimum flow requirements) within required temperature limits. This, in turn, leads to adverse impacts with cooling loads such as Reactor Coolant Pump (RCP) seals and Reactor Coolant System (RCS) demineralizer boron affinity changes, which would impact reactivity management.

The period in which Lake Michigan temperatures are low, impacting minimum flow requirements, is typically between the months of October through May. This is when work to remove the preferred qualified circuit is not desired. In addition, during late spring and summer storm periods, as well as elevated Lake Michigan temperatures, it is also not desired to remove the qualified circuits from service based on risk and cooling ability of ESW-supplied systems. This period is between June and early September.

The desired time to remove the preferred qualified circuit from service is when there is not an operational challenge due to lake temperature and adverse summer weather conditions. This period of time is a short window in mid-to-late September, or early October.

U. S. Nuclear Regulatory Commission AEP-NRC-2024-02 Page 3

Copies of this letter and its attachments are being transmitted to the Michigan Public Service Commission and Michigan Department of Environment, Great Lakes, and Energy (EGLE), in accordance with the requirements of 10 CFR 50.91.

There are no new regulatory commitments in this letter. Should you have any questions, please contact Mr. Michael Scarpello, Regulatory Affairs Director, at (269) 466-2649.

Sincerely,

~)~

Kelly J. Ferneau Site Vice President

JMT/sjh

Enclosures:

1. Affirmation

2. License Amendment Request for One-Time Extension of Completion Time for an Inoperable AC Source - Operating

3. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked to Show Proposed Changes

4. Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked to Show Proposed Changes

c: EGLE - RMD/RPS J. B. Giessner - NRC Region Ill NRC Resident Inspector N. Quilico - MPSC R. M. Sistevaris - AEP Ft. Wayne, w/o enclosure S. P. Wall-NRC Washington D.C.

A. J. Williamson - AEP Ft. Wayne, w/o enclosure to AEP-NRC-2024-02 Page 1

AFFIRMATION

I, Kelly J. Ferneau, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (l&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf of l&M, and that the statements made and the matters set forth herein pertaining to l&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company

~!~

Kelly J. Ferneau Site Vice President

SWORN TO AND SUBSCRIBED BEFORE ME

THIS 3 DAYOF z\\>r,\\,2024

~ Public

My Commission Expires '=, / ~ 3 / ';to "5 0 Enclosure 2 to AEP-NRC-2024-02

LICENSE AMENDMENT REQUEST FOR ONE-TIME EXTENSION OF COMPLETION TIME FOR AN INOPERABLE AC SOURCE - OPERA TING

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specification Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change

3. TECHNICAL EVALUATION 3.1 Regulatory Compliance 3.2 Defense in Depth 3.2.1 Compensatory Actions 3.2.2 Safety Margins 3.2.3 Other Defense-in-depth Considerations 3.3 Evaluation of Risk Impact 3.3.1 PRA Quality 3.4 PRA Results 3.5 Conclusions 3.6 Identification of Risk-Significant Configurations

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES to AEP-NRC-2024-02 Page2

1.0

SUMMARY

DESCRIPTION

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2, proposes to amend Facility Operating Licenses DPR-58 and DPR-74. l&M proposes a one-time extension of the Technical Specification (TS} 3.8.1, Condition A.3, Completion Time from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months for an inoperable AC Electrical Source. This one-time extension is to support implementation of a modification to improve the long-term reliability of the Train B feed of the off-site preferred power source.

2.0 DETAILED DESCRIPTION

The proposed change would add the following Footnote to the 72-hour Completion Time for TS 3.8.1, Condition A.3:

"For Train B only, the Completion Time that Train B can be inoperable as specified by Required Action A.3 may be extended beyond the "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months " up to "288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months ," to support modification of the Train B Reserve Feed. Upon completion of the modification and restoration this footnote is no longer applicable. Compensatory measures described within CNP letter AEP-NRC-2024-02, dated April 3, 2024 will remain in effect during the extended period. The one-time extension shall expire upon completion of the modification and restoration of operability for Train B."

2.1 System Design and Operation

As shown on Figure 1 (a red star denotes the portion of reserve feed that is being modified) and Figure 2 at the end of this Section, the onsite alternating current (AC) electric power distribution system for each unit contains four, 4160V ( 4.16 kV) non-safety-related buses designated 1 A, 1 B, 1 C, and 1 D for Unit 1 and 2A, 28, 2C, and 2D for Unit 2. These buses are referred to as the "RCP" buses because they power the reactor coolant pumps. Each of the non-safety-related RCP buses feed a downstream safety-related 4.16 kV bus. These safety-related buses are designated T11 A, T11 B, T11 C, and T11 D for Unit 1 and T21 A, T21 B, T21 C, and T21 D for Unit 2.

These buses are referred to as the "T" buses. With the main generator on-line, the RCP buses are normally fed from the Unit Auxiliary Transformers (UATs ), which receive power from the main generator.

Upon a trip of the main generator, the station auxiliaries, which include the RCP buses and "T" buses, are automatically fast transferred to the preferred offsite power source, which is the reserve auxiliary transformers (RATs) TR101AB and TR101CD for Unit 1 and TR201AB and TR201CD for Unit 2. This assures continued power to equipment when the main generator is off-line. There is no loss of power to the Engineered Safety Features (ESF) loads during a fast transfer. The RATs supply the reserve auxiliary power for both units.

The preferred offsite power source for both units can be arranged so that Main Switchyard transformer No. 4 or transformer No. 9 supplies TR101CD and TR201CD and Main Switchyard transformer No. 5, or transformer No. 9 supplies TR101AB and TR201AB. Under certain plant conditions, it is possible for transformer No. 4, transformer No. 5, or transformer No. 9 to feed the entire plant auxiliary load. to AEP-NRC-2024-02 Page 3

The other qualified circuit required to be operable by TS Limiting Condition for Operation (LCO) 3.8.1.a is the alternate offsite circuit. The alternate qualified offsite circuit consists of a 69/4.16 kV transformer (TR 12EP-1 ), the cabling and switches to a 4.16 kV bus, designated as the EP Bus, which supplies breakers 1 EP and 2EP, and the cabling, switches, and breakers to the T buses. This alternate offsite source is 30° out of phase with the plant 4kV buses. The 4kV safety related bus source breaker from this source can only be manually closed if the other source breakers from the EDG and 4kV non-safety related bus are open. Connection of the T buses to transformer 12EP1 requires manual switch operations in the control room. The alternate offsite power source has the necessary capacity to operate one train of the engineered safeguard equipment in one unit while supplying one train of the safe shutdown power in the other. The T buses can also be powered from the emergency diesel generators (EDGs). TS LCO 3.8.1.a requires two qualified circuits between the offsite transmission network and the onsite Class 1 E AC Electrical Power Distribution System.

An additional independent onsite standby AC power source consisting of two supplemental diesel generators (SDGs) is provided to automatically supply power to the EP bus, which is normally supplied by the alternate qualified offsite circuit and can be manually aligned to directly supply the T buses. While the SDGs are available to supply power to the EP bus it does not have the load capacity to supply both units safeguard loads at the same time.

The modification that is being implemented affects only the 34.5kV 12AB (Train B) LFE. The 34.5kV Loop Feed Enclosure is part of the Offsite Power Supply (OFPW) system that provides power for the operation of plant auxiliary equipment and instrumentation when the Units 1 and 2 main generators are off-line. The OFPW system consists of two subsystems: 1) The Preferred Offsite Power Supply System, and 2) The Alternate Offsite Power Supply System. The LFE 12AB bus replacement belongs to the Preferred OFPW subsystem.

During normal power operation of the plant, the OFPW system (both Preferred and Alternate subsystems) is in standby mode, ready and capable to accept both units' auxiliary loads when required. The 34.5 kV circuit breakers 12AB, 12CD, BF, and BG, which connect the preferred offsite source to the plant auxiliary buses have controls from Control Rooms of both units. The non-segregated bus that has experienced failures, and is being replaced, is part of the 34.5kV service to the Train B RATs through LFE 12AB. LFE 12CD has not experienced any failures due to environmental conditions as the non-segregated bus of LFE 12CD is located inside a Missile Shielding Structure and it is not subjected to CNP steam generator blowdown effluent. to AEP-NRC-2024-02 Page4

Figure 1 - CNP Power Arrangement from Offsite Sources to the RCP Buses

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'IR118MC "T" "T" TRlll!I.C mi ~ Vlllll- - TR21CUC 'T to AEP-NRC-2024-02 Page 5

Figure 2 - CNP Power Arrangement from RCP Buses to the "T" Buses

12EP1..L. From

-,- Supplemen~

4 kV _c Diesel I I EP Bus Generators

~EP ~EP ----------------~

F,om FCP From FCP Ftom FCP From FCP I I Ftom FCP F,om FCP From FCP From RCP Bus 1A Bus 1B Bus 1C Bus 10 Bus2O

c + I 1~D... 1 c l-J 1,,c::~ I J,, I -, I I 4kV I'~ F\\ I I ( 012

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11 84 0,3 0s 1'210 T11AI I ~ ;1T11r ~ T11C 0; 0; T110 T21A ~T21 'T21C

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~AA ~1 4 600V ~1 (~ ' 480V (11 (1 V 600V~; (~ ~

,:,.,,~;;,,J; ~~ 11B [ii§] 110 11PHC 21PHA ~ 21B ~ 210 21PHC to AEP-NRC-2024-02 Page 6

2.2 Current Technical Specification requirements

CNP Unit 1 and Unit 2 TS 3.8.1, "AC Sources - Operating," requires two qualified offsite circuits and separate and independent Diesel Generators (DGs) for each train to ensure availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an anticipated operational transient or a postulated design basis accident (DBA) in Modes 1, 2, 3, and 4. When one offsite circuit is inoperable, Condition A is entered with a Required Action to restore the train to operable status with a Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If Condition A and its associated Completion Time are not met, then entry into Condition G is required. Condition G requires the action to place the unit (applicable to Unit 1 and Unit 2) in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . l&M is requesting a one-time-use extension of the TS 3.8.1, Condition A.3, Completion Time from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months (12 days) that would allow continued plant operation for only the additional time needed to modify the bus structure and restore the inoperable Train B reserve feed to Operable status. to AEP-NRC-2024-02 Page 7

CONDITION REQUIRED ACTION COMPLETION TIME

A. One required offsite A. 1 ---------------NOTE--------------

circuit inoperable. Not applicable if a required Unit 2 offsite circuit is inoperable.

Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months required OPERABLE offsite circuit. AND

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter

AND

A.2 Declare required feature(s) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from with no offsite power discovery of no offsite available inoperable when power to one train its redundant required concurrent with feature(s) is inoperable. inoperability of redundant required feature(s)

A.3 Restore required offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months circuit to OPERABLE status.

17 days from discovery of failure to meet LCO 3.8.1.a orb to AEP-NRC-2024-02 Page 8

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 of Condition A, C, AND D, E, or F not met.

G.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR

Required Action and Associated Completion Time of Required Action B.2, B.3, B.4.1, B.4.2, or B.5 not met.

2.3 Reason for the Proposed Change

The existing 34.5kV 12AB LFE at CNP has experienced failures contributing to the loss of, and potential loss of, the Train B Reserve Feed, consequently causing the station to enter LCO 3.8.1. The cause analyses of these incidents revealed that the failures were caused by multiple factors at the non-segregated bus duct of the LFE. The issues that resulted in the failures of the non-segregated bus included inadequate design of the bus duct and associated components resulting in moisture and environmental contamination intrusion.

A study was performed and it was identified that the best course of action was the complete replacement of existing LFE structure with a cable bus. This new design resolves the deficiencies of the current 12AB LFE that have led to fault induced failures and does not introduce any new system level failures to the Offsite Power Supply system.

However, implementation of this modification is expected to take approximately 8 days with approximately 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of contingencies. Therefore, a requested one time allowance of 12 days, or 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months , is being requested.

2.4 Description of the Proposed Change

The proposed license amendment would revise CONDITION A of TS LCO 3.8.1, by adding a new footnote (a) to REQUIRED ACTION A.3 COMPLETION TIME and a new footnote denoted by the (a) as follows: to AEP-NRC-2024-02 Page 9

CONDITION REQUIRED ACTION COMPLETION TIME

A. One required offsite A.1 ------ ---------NOTE------ --------

circuit inoperable. Not applicable if a required Unit 2 offsite circuit is inoperable. ----------

Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months required OPERABLE offsite circuit. AND

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter

AND

A.2 Declare required feature(s) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from with no offsite power discovery of no offsite available inoperable when power to one train its redundant required concurrent with feature(s) is inoperable. inoperability of redundant required feature(s)

AND

A.3 Restore required offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months <a>

circuit to OPERABLE status. AND

17 days from discovery of failure to meet LCO 3.8.1.a orb

(a) For Train B only, the Completion Time that Train B can be inoperable as specified by Required Action A.3 may be extended beyond the "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months " up to "288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months ," to support modification of the Train B Reserve Feed 12AB Loop Feed Enclosure. Upon completion of the modification and restoration this footnote is no longer applicable. Compensatory measures described within CNP letter AEP-NRC-2024-02, dated April 3, 2024 will remain in effect during the extended period. The one-time extension shall expire upon completion of the modification and restoration of operability for Train B. to AEP-NRC-2024-02 Page 10

3.0 TECHNICAL EVALUATION

The proposed license amendment would permit a one-time change to TS 3.8.1 Condition A to allow a one-time extension of the TS 3.8.1 Required Action A.3 Completion Time to allow for a modification to the Train B 12AB LFE. The one-time change would allow CNP to modify the 12 AB LFE while at power.

In proposing a one-time extension to TS 3.8.1 Required Action A.3 Completion Time, l&M applied Regulatory Guide (RG) 1.177, "Plant-Specific, Risk-Informed Decision-making: Technical Specifications." RG 1.177 describes acceptable methods for assessing the nature and impact of proposed TS changes, including one-time extensions, by considering engineering issues and applying risk insights. Each of the RG 1.177 principles is addressed below.

3.1 Regulatory Compliance

No exceptions or exemptions from applicable codes and standards relevant to safe plant operation are proposed by this amendment request. A Regulatory Evaluation is included in Section 4 of this Enclosure.

3.2 Defense in Depth

During the proposed one-time extension defense in depth measures will be applied to account for unknown and unforeseen failure mechanisms or other phenomena and thereby ensure safety function is maintained. Appropriate compensatory actions have been established to the extent practical and will be implemented prior to the one-time extended inoperability of Train B reserve feed in order to maintain defense in depth. By creating multiple independent and redundant layers of defense, compliance with applicable plant specific design criteria and general design criteria are maintained.

3.2.1 Compensatory Actions

Compensatory Measures

CNP will reduce plant risk exposure through a combination of risk management actions (RMAs) that prevent planned high-risk configurations and other non-quantifiable risk-reducing actions.

RMAs to prevent high risk configurations (due either to fire initiation or other significant plant events) and establish non-quantifiable actions to monitor for high risk (fire or other internal or external) events and provide readily usable alternate power sources are listed below.

These actions include a provision that if emergent plant conditions require actions to stabilize the unit(s),

and if any of those actions conflict with any of the RMAs below, then those actions should be taken without delay, and the RMA restored after the emergent condition has passed and the plant is stabilized.

Operation and Maintenance Restrictions

The following compensatory measures will be taken to reduce the risk during the period of extended TS 3.8.1 Required Action A.3 Completion Time:

The FLEX equipment and installation locations will be reviewed for use during a loss of offsite power event. The FLEX Deployment areas will be walked down to ensure readiness. The functional and load checks are performed on a quarterly basis for the FLEX Diesel and will be reviewed to be current to ensure the diesel will perform if required. to AEP-NRC-2024-02 Page 11

Operations will confirm that there are no planned operations on the grid that would present a challenge to the remaining offsite power system during the duration of the extended TS completion period.

The following equipment and systems will be guarded on both Unit 1 and Unit 2 in accordance with station procedures. Operators will be briefed on the increased importance of these systems.

Reactor Coolant Pump (RCP) Breakers (T11A9, T1181, T11C1, T11D12, T21A9, T2181, T21C1, T21D12)

Emergency Power (EP) Breakers (T11A12, T21A12, T1182, T2182, T11C2, T21C2, T11D1, and T21D1)

AB and CD Emergency Diesel Generators (EDGs)

Supplemental Diesel Generators

69 kV Switchyard

Turbine Driven Auxiliary Feed Water (AFW) Pump, Steam Supply Valves, and Motor Operated Discharge Valves

4kV Rooms

250 VDC Train A

Backup Nitrogen (N2) supply to Steam Generator (SG) PORV

OME-42 ( control air compressor)

The offsite power supply and switchyard will be guarded. This includes ensuring that switchyard access is restricted to only essential work and no elective maintenance within the switchyard is performed that would challenge offsite power availability. This will reduce the likelihood of a loss of offsite power occurring.

No intrusive surveillances or maintenance activities will be allowed that could potentially jeopardize plant operations, except for emergent issues. This reduces the likelihood of the unavailability of redundant trains during the period of extended TS completion. All required weekly surveillances will be performed prior to entering the period of extended TS 3.8.1 Required Action A.3 Completion Time or deferred until after restoration of reserve feed.

The following Fire Areas were identified as having increased fire risk significance during this configuration. For this reason, the following Fire Areas will have shiftly fire watches established, and will not have any hot work performed:

AA 14 Unit 1 CD EDG Room

AA23 Unit 2 CD EDG Room

AA 19 Unit 1 East Motor Driven AFW Pump Room

AA45B Unit 2 CD Train Switchgear Room

Isolate and drain Unit 11-FHC-220 and Unit 2 2-FHC-220 to remove them as a flooding source.

Verify and maintain availability of NFPA-805-credited suppression systems.

During the period of extended TS 3.8.1 Required Action A.3 Completion Time, l&M will take the following actions to reduce the probability and severity of initiating events:

A Will not perform any elective maintenance on components that are credited for accident mitigation in the CNP PRA models.

B. Will not perform any unnecessary switchyard work or work on Balance of Plant systems that may increase the probability that there is a Unit trip to AEP-NRC-2024-02 Page 12

C. The following procedures have increased importance due to the nature of this configuration.

Operations will review these procedures which will increase the likelihood that the associated actions are completed successfully:

a. Procedure 1(2)-OHP-4023-SUP-009 Restoration of 4kV Power from EP.

b. Procedure 1 (2)-OHP-4022-016-004, Loss of Component Cooling Water.

c. Procedure 1 (2)-OHP-4023-E-3, Steam Generator Tube Rupture.

d. Procedure 1 (2)-OHP-4023-ECA-0-0, Loss of all AC Power

e. Procedure 1(2)-OHP-4023-FR-H-1, Response to Loss of Secondary Heat Sink

f. In addition to these procedures, the significance of external events will be reduced by Operations reviewing the following actions

i. Transferring supply power for T21A orT21 B ii. Opening the ESW crosstie iii. Tripping CW on a flood signal

To ensure compensatory measures are correctly implemented for Operations during the period of extended TS 3.8.1 Required Action A.3 Completion Time:

Operations will follow written guidance for compensatory measures in 12-OHP-211 0-CCA-001, Compensatory Measures and Contingency Actions. Operations will carry the assigned compensatory measures in the turnover log and review and perform each shift.

The Unit Supervisor for each Unit will log in the Control Room Narrative Logs when the compensatory measures are performed and/or reviewed each shift.

3.2.2 Safety Margin

The proposed one-time amendment does not alter the design or capabilities of the emergency safeguards systems, will not result in plant operation in a configuration outside the design basis, and will not impact any assumptions or consequences specified in applicable safety analyses. Safety margins will be maintained in accordance with Cook safety analyses acceptance criteria, and no changes are proposed that affect any assumptions or inputs to applicable safety analyses. Sufficient equipment redundancy will exist due to the availability of emergency diesel generators during the proposed completion time extension to ensure power is available. The emergency diesel generators and auxiliaries will be reviewed for any issues to ensure their readiness for continuous service up to the full expected duration of the Train B reserve feed outage. As such, no safety margins are impacted by the proposed change.

3.2.3 Other Defense-in-Depth Considerations

A reasonable balance among the prevention of core damage and consequence mitigation will be preserved during the proposed completion time extension. No other systems, structures, and components (SSCs) will be affected by the proposed Completion Time extension, and no limits will be imposed on any SSC performing its specified function. Elevated risk awareness and the guarding of critical equipment will be executed (as shown in Section 3.2.1, Compensatory Actions, above) during the proposed Completion Time extension in accordance with existing plant procedures. However, these programmatic activities will be accompanied by pre-job and periodic (e.g., shift change) briefings, equipment walk-downs, progress updates, and increased operational and managerial scrutiny. As such, there will be no overreliance on programmatic activities as compensatory measures during the proposed Completion Time extension. The independence of the physical barriers to radiological releases will not be degraded as a result of the proposed Completion Time extension. The planned maintenance will not impact fuel cladding, Reactor Coolant System (RCS), or Containment integrity. No other SSC will be affected by the proposed Completion Time extension, and thereby no limits will be imposed on any SSC in performing its specified safety function. to AEP-NRC-2024-02 Page 13

Potentially risk significant plant configurations will not occur during the proposed one-time completion time extension due to online risk assessment tools and increased operational and managerial scrutiny of plant operations. During the planned maintenance, no risk significant plant equipment will be removed from service, and protective measures will be implemented to reduce the likelihood of challenges to risk significant equipment. As a result, the functional redundancy, independence, and diversity currently described in the CNP Updated Final Safety Analysis Report (UFSAR) will be maintained throughout the proposed Completion Time extension.

Defenses against potential common-cause failures (CCFs) will be maintained by limiting non-essential maintenance and operation of SSCs having mitigatory roles credited in accident analyses.

3.3 Evaluation of Risk Impact

3.3.1 PRA Quality

3.3.1.1 Peer Review History

The CNP PRA model is generally robust and suitable to support this amendment. All CNP PRA models have undergone Finding and Observation (F&O) closure reviews, and there are no PRA Upgrades that have not yet been Peer Reviewed. The ongoing PRA maintenance and update activities associated with the CNP PRA program ensure that the PRA models represent the as-built, as-operated plant. Therefore, the CNP PRA model has the technical adequacy required to support the amendment.

The CNP PRA models have been assessed against RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2 (Reference 1), consistent with NRC Regulatory Issue Summary 2007-06 (Reference 2).

Finding and Observation (F&O) closure reviews were conducted on the PRA models discussed in this section. Closed findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, "Close-out of Facts and Observations" (F&Os) (Reference 3), as accepted by the NRC in the letter dated May 3, 2017 (Reference 4). The results of this review have been documented and are available for NRC audit.

Full Power Internal Events (FPIE) and Internal Flooding (IF) PRA Model

The CNP FPIE PRA model was peer reviewed in July 2015 using the NEI 05-04 (Reference 5) process, the PRA Standard (Reference 6) and Regulatory Guide 1.200, Revision 2 (Reference 1 ). This Peer Review (PWROG-15076-P (Reference 7)) was a full-scope peer review of the technical elements of the Internal events and internal flooding, at-power PRA.

The CNP FPIE PRA model underwent a focused-scope peer review in September 2017 (Reference 8) and subsequent closure review (Reference 9) in May 2018 for what was determined to be a methodology upgrade for the Containment Hydrogen Analysis.

The CNP FPIE PRA model underwent focused-scope peer reviews in October 2016 (Reference 10) and November 2020 (Reference 11) focusing on the treatment of pre-initiator Human Reliability Analysis (HRA) and the implementation of FLEX, respectively.

The CNP FPIE PRA model underwent an F&O Closure review in November 2021 using the NEI 05-04 (Reference 5) process, the PRA Standard (Reference 6), Regulatory Guide 1.200, Revision 2 to AEP-NRC-2024-02 Page 14

(Reference 1) and Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, "Close-out of Facts and Observations" (F&Os) (Reference 3). This Appendix X Closure Review included a review of the open F&Os in the full power internal events and internal flooding PRA model.

Finding level F&Os for the FPIE PRA model are discussed in Attachment 1 of this enclosure. There are no remaining open F&Os involving the internal flooding PRA model.

Fire PRA Model

The CNP Fire PRA (FPRA) peer review (Reference 12) was performed in July 2010 using the NEI 07-12 (Reference 13), the ASME/ANS PRA Standard, ASME/ANS RA-Sa-2009 (Reference 6), and Regulatory Guide 1.200, Revision 2 (Reference 1 ). The FPRA peer review was a full-scope review of the CNP at-power FPRA technical elements against the Part 4 technical elements of the ASME/ANS PRA Standard with the exception of the Qualitative Screening (QLS) element and the Quantitative Screening (QNS) Element as screening tasks were not performed in the FPRA. Not performing these screening elements ensures that none of the risk increase from this configuration is being missed in screened scenarios. The scope also included a review of the CNP's PRA Configuration Control Program in accordance with Section 1.5 of the ASME/ANS Combined PRA Standard (Reference 6).

The CNP FPRA underwent additional focused-scope peer reviews in November 2015 (Reference 14),

July 2017 (Reference 15), July 2022 (Reference 16), and February 2023 (Reference 17), involving Level 2 PRA [LERF] (D0403140002-1515), CAFTA Conversion/-FSS/-IGN items (PWROG-17027), -FSS items, and -FQ items (P3801-0001-01).

The findings from the Fire PRA peer review have been resolved in the Fire PRA model. An F&O Closure Review was conducted for CNP (Reference 18). The scope of the review included explicit review of previous fire peer review findings.

Finding level F&Os for the FPRA model are discussed in Attachment 1 of this enclosure.

Seismic PRA (SPRA) model

The seismic PRA model was peer reviewed in November 2018 (Reference 19). This peer review was conducted against the technical elements in PRA Standard Code Case for Part 5 (Reference 20). For supporting requirements in the Code Case that referred back to requirements in Part 2, Addendum B, of the PRA Standard (ANSE/ANS RA-Sb-2013) was utilized.

Per PWROG-18062-P (Reference 19):

This standard, ASME/ANS RA-Sb-2013 (Addendum B), was approved by ANSI in 2013, but has not been formally endorsed by the NRC through a revision to RG 1.200 (Reference 1 ). However, Part 5 (Requirements for Seismic Events At-Power PRA) of Addendum B of the PRA Standard is referenced in the Electric Power Research Institute (EPRI) report "Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic". NRC has endorsed this EPRI report as "one acceptable method for responding to the information requested in Enclosure 1 of the 50.54(f) letter" pertaining to Post-Fukushima Near Term Task Force (NTTF)

Recommendation 2.1 on seismic hazard re-evaluation. This effectively provides NRC endorsement of Part 5 of Addendum B of the PRA Standard, ASME/ANS RA-Sb-2013. In 2017, the Joint Committee on Nuclear Risk Management (JCNRM) released the Code Case as an acceptable alternative to Part 5 of Addendum B. By letter dated, March 7, 2018, the NRC stated the following: to AEP-NRC-2024-02 Page 15

The NRG staff has determined that the alternative approach described in the Code Case is consistent with Part 5 of the ASME/ANS PRA standard which the staff has reviewed and endorsed in Regulatory Guide 1. 200.

The NRG acceptance letter of the Code Case included limited clarifications.

Sections 1-6 and 5-3 of the ASME/ANS PRA Standard include explicit requirements for a peer review of SPRAs against the requirements of Part 5 in the PRA Standard using a written process. The industry has developed the PRA peer review process as defined in NEI 12-13 to perform the peer reviews for SPRAs and other external hazards PRAs. This was accepted with limited amendments by the NRG in the letter dated March 7, 2018.

The findings from the Seismic PRA peer review have been addressed in the Seismic PRA model. In August 2019 (AEPDCC-0058-REPT-001), an F&O Closure Review was conducted for CNP. Finding level F&Os for the SPRA model are discussed in Attachment 1 of this enclosure.

This demonstrates that the PRA models are of sufficient quality and level of detail to support this application.

3.3.1.2 Sources of Model Uncertainty

Based on evaluations supporting the 2023 PRA models of record, some key assumptions were identified as key model uncertainties. These assumptions are described below, but do not uniquely impact the model results generated for this license amendment request. Because of this, no sensitivity evaluations were developed beyond those already included in base PRA model uncertainty documentation.

Westinghouse Generation Ill Reactor Coolant Pump (RCP) Shutdown Seals:

The modeling of the Westinghouse Generation Ill RCP shutdown seals is a key source of model uncertainty for the CNP PRAs. If the new RCP seals do not actuate or fail to remain actuated, severe accident sequences become much more likely. Risk metrics such as GDF and LERF increase significantly if failure of the shutdown seals is assumed. The current PRA model utilizes the Pressurized Water Reactor Owners Group guidance for PRA modeling of the shutdown seals, supported by the Westinghouse Owners Group 2000 RCP seal failure model (Reference 21 ), both of which are industry consensus models. The 2015 peer review also found the modeling of the shutdown seals acceptable.

Credit for Backup Power to Distributed Ignition System (DIS):

1-GEN-DGDIS and 2-GEN-DGDIS are diesel generators designed to give backup power to the Distributed Ignition System (DIS) should its normal power supply fail. They were installed to reduce PRA-estimated Seismic LERF, but are credited in both the FPIE and SPRA models consistent with their implementation in CNP's Emergency Operating Procedures. In November 2023, it was identified that these components would be shed off DC power during deep DC load shed during an Extended Loss of AC Power (ELAP). Shedding these components from DC power prevents their function in these scenarios. (Reference 23)

This condition represents a source of model uncertainty for the FPIE and SPRA models. For the FPIE model, deep load shed is a specifically modeled action as part of FLEX response. While this makes it possible to quantify the impact of this condition in the FPIE model, this information is not generated as part of standard model quantification. For the Seismic PRA model, since FLEX is not credited, there is no way to determine if a scenario would include deep load shed actions. For both of these models, an upper bound of the impact from this condition can be estimated by failing 1-GEN-DGDIS and 2-GEN-to AEP-NRC-2024-02 Page 16

DGDIS within the model. This method is conservative as it will remove credit for the function of these components during non-load-shed scenarios where they would function normally. This sensitivity was performed for the most limiting Unit (Unit 1 ), and the results are presented below. The baseline used to calculate the deltas presented was not re-calculated to remove the DIS DG credit from the base model, and therefore a portion of the delta is attributable to removal of this credit and not the risk associated with this license amendment. This contributes to the bounding and conservative nature of this sensitivity. These results show that even with no credit for this backup power to the DIS, the risk impact from this amendment is still below the thresholds for a one-time change presented in RG 1.177.

Sensitivity Evaluation Results:

LERF Zero U1 Maintenance Sensitivity Case : Delta ICLERP (12 days) Case CLERP (Unadjusted)

FPIE LERF 1.39E-06 1.86E-06 4.70E-07 1.55E-08 Fire LERF 3.05E-06 (Unadjusted) 2.44E-06 (Unadjusted) 6.10E-07 2.01E-08 Seismic LERF 5.30E-06 1.12E-05 5.90E-06 1.94E-07 Total 2.29E-07

FLEX Treatment:

=

Background===

The NRC has been issuing a "generic" Request for Additional Information (RAI) regarding crediting of FLEX equipment in PRA models. The Limerick RAI is summarized below.

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 NRC's staff assessment of the challenges of incorporating diverse and flexible (FLEX) coping strategies and equipment into a PRA model in support of risk-informed decision-making in accordance with the guidance of RG 1.200, Revision 2 (ADAMS Accession No. ML090410014 ). Docketed information does not indicate if [PLANT NAME J has credited FLEX equipment or actions in the

[PRA MODEL]. As such, please address the following:

a. Discuss whether [UTILITY] has credited FLEX equipment or mitigating actions into the [PLANT NAME PRA MODEL]. If not incorporated or their inclusion is not expected to impact the PRA results used in the RICT program, no additional response is requested.

b. If FLEX equipment or operator actions have been credited in the PRA, address the following, separately for the internal events (including internal flooding), and other PRAs.

i. Summarize the supplemental equipment and compensatory actions, including FLEX strategies that have been quantitatively credited for each of the PRA models used to support this application. Include to AEP-NRC-2024-02 Page 17

discussion of whether the credited FLEX equipment is portable or permanently installed equipment.

ii. 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) and whether the credited operator actions are similar to other operator actions evaluated using approaches consistent with the endorsed ASMEIANS RA-Sa-2009 PRA Standard.

iii. If any credited FLEX equipment is dissimilar to other plant equipment credited in the PRA (i.e., SSCs with sufficient plant-specific or generic industry data), 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 ASMEIANS PRA Standard as endorsed by RG 1.200, Revision 2.

iv. If any operator actions related to FLEX equipment are evaluated using approaches that are not consistent with the endorsed ASMEIANS RA-Sa-2009 PRA Standard (e.g. using surrogates), discuss the methodology used to assess human error probabilities for these operator actions. The discussion should include:

1. A summary of how the impact of the plant-specific human error probabilities and associated scenario-specific performance shaping factors listed in (a)-(j) of supporting requirement HR-G3 of the ASME/ANS RA-Sa-2009 PRA Standard were evaluated.

2. Whether maintenance procedures for the portable equipment were reviewed for possible pre-initiator human failures that renders the equipment unavailable during an event, and if the probabilities of the pre-initiator human failure events were assessed as described in HLR-HR-D of the ASMEIANS RA-Sa-2009 PRA Standard.

3. If the procedures governing the initiation or entry into mitigating strategies are ambiguous, vague, or not explicit, a discussion detailing the technical bases for probability of failure to initiate mitigating strategies.

c. The ASMEIANS 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 ASMEIANS 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 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.

d. Section 2. 3.4 of NE/ 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. The NRG SE for NE/ 06-09, Revision 0, states that this consideration is consistent with Section 2.3.5 of RG

1. 177, Revision 1. NE/ 06-09, Revision 0-A, further states that sensitivity studies should be performed on the base model prior to initial implementation of the Rf CT program on uncertainties which could potentially impact the results of a I UC T calculation. NRG staff notes that the impact of model uncertainty could vary based on the proposed RICTs. NE/ 06-09, Revision to AEP-NRC-2024-02 Page 18

0-A, a/so states that the insights from the sensitivity studies should be used to develop appropriate compensatory RMAs including highlighting risk significant operator actions, confining availability and operability of important standby equipment, and assessing the presence of severe or unusual environmental conditions. Uncertainty exists in modeling FLEX equipment and actions related to assumptions regarding the 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. In light of these observations:

i. Describe the sensitivity studies that will be used to identify the RICTs proposed in this application for which FLEX equipment and.for operator actions are key assumptions and sources of uncertainty (e.g., use of generic industry data for non-safety related equipment). Explain and justify the approach (e.g., any multipliers for failure probabilities) used to perform the sensitivity studies.

ii. Described how the results of the sensitivity studies which identify FLEX equipment and/or operator actions as key assumptions and sources of uncertainty will be used to identify RMAs prior the implementation of the RICT program, consistent with the guidance in Section 2.3.4 of NE/ 06-09, Revision 0-A.

iii. Demonstrate the approaches described in items (i) and (ii) above using an example sensitivity study for the nominal configuration of a proposed R/CT where the FLEX equipment and/or operator actions are identified as key assumptions and sources of uncertainty.

The discussion section below provides responses, as applicable, to the above questions regarding modeling of FLEX equipment in the CNP PRA models. The responses are provided in a consolidated form instead of individual responses to each question.

Discussion

FLEX strategies are credited in the CNP internal events (IE), Fire (FPRA) and seismic PRA (S-PRA) models. Specifically, three strategies are modeled.

The CNP Flex Final Integrated Plan documents the initial implementation of FLEX. The FLEX strategies and procedures at CNP are largely broken down by the functions they support, which are identified as follows:

1. Reactor Core Cooling - This strategy provides reactor core cooling by feeding the Steam Generators with either the Turbine Driven Auxiliary Feed Pump or portable equipment.

2. Reactor Coolant System (RCS) Boration/lnventory Control - This strategy provides long-term RCS makeup and boration using portable equipment.

3. Spent Fuel Pool (SFP) Cooling - This strategy provides makeup and cooling to the SFP using portable equipment.

4. Containment - Analyses performed for the FLEX implementation show that no additional actions are necessary for containment heat removal during the assumed FLEX conditions. The hydrogen igniters are repowered as part of the Electric Power FLEX strategy during Phase 2.

5. Electric Power - This strategy provides electric power using deep load shed to preserve station battery power to last until portable generators are deployed.

Specifics into how each strategy is adapted into the PRA models are available in the CNP FLEX System Notebook (PRA-NB-SY-FLEX) to AEP-NRC-2024-02 Page 19

A focused Scope peer review was conducted at CNP to review the implementation of FLEX into the PRA model (PRA-NB-FSPR-FLEX). Modeling inclusion of FLEX has been performed in a manner that:

Is consistent with other modeling aspects used in the PRA model

Is commensurate with the supporting requirement of the ASME/ANS PRA Standard

Does not add any additional scope to the PRA

Does not and any new capability of the PRA

Does not significantly impact significant accident sequences or accident sequence progression

In addition, a gap assessment was performed in 2022 to review the CNP FLEX Human Reliability Analysis) (HRA) evaluation against the NRC Memo dated May 6, 2022, Updated Assessment Of Industry Guidance For Crediting Mitigating Strategies In Probabilistic Risk Assessments (Reference 22). This memo identified several areas of improvement to bring the FLEX HRA into alignment with the memo requirements. FLEX Human Event Probabilities that were updated, as a result of this review, were included in the 2023 PRA models of record.

Method of Risk Assessment

The PRA risk impact of operation with Unit 1 and Unit 2 Train B Reserve Feed unavailable can be quantitatively estimated using the CNP Full Power Internal Events (FPIE) model of record, the Fire PRA model (FPRA) model of record, and the CNP Seismic PRA (SPRA) model of record.

The risk assessment is quantified by analyzing the risk of the Unit 1 and Unit 2 Train B Reserve Feed unavailable and then subtracting the risk from a zero maintenance case (zero other equipment unavailability) core damage frequency (CDF) and large early release frequency (LERF) for the FPIE, Fire, and Seismic PRA models. The following assumptions are used for the plant configuration, in addition to the components undergoing maintenance:

The Unit 1 East - Unit 2 West and Unit 1 West-Unit 2 East ESW crosstie valves will be assumed to be closed for this risk analysis. They will be assumed open (normal configuration) in the base case.

No additional equipment will be unavailable for the duration of the Train B Reserve Feed outage.

No surveillance testing will occur for PRA credited equipment.

Compensatory actions, which are not quantitatively considered.

Fire watch tours and maintenance restrictions are not assumed to modify the likelihood of any fire or internal initiating events.

Since the plant electrical system normally receives power from the main generator, the units do not automatically trip off if offsite power is lost. For this reason, the likelihood of initiating events is not adjusted for the risk analysis. The loss of the Train A reserve auxiliary transformers is already accounted for in the FPIE model as a random failure. Fire and Seismic events are not considered to be any more likely due to the Train B Reserve Feed Outage. to AEP-NRC-2024-02 Page 20

3.4 PRA Results

The PRA models were re-quantified with the configuration information as discussed above. Incremental Conditional Core Damage Probability (ICCDP) and Incremental Conditional Large Early Probability (ICLERP) are calculated for the planned outage duration of 12 days (288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months ).

Table 1 - CDF PRA Results CDF Zero 3.8.1. LAR Reserve U1 Maintenance Feed Extension Delta CCDP ICCDP ( 12 days)

Case Case FPIE CDF 2.05E-05 2.59E-05 5.40E-06 1.78E-07 Fire-CDF 2.84E-05 3.24E-05 4.00E-06 1.32E-07 Seismic CDF 2.08E-05 2.13E-05 5.00E-07 1.64E-08 Total 9.90E -06 3.25E-07

~

U2 FPIE CDF 2.05E -05 2.58E-05 5.30E-06 1.74E-07 Fire-CDF 3.41 E-05 3.78E-05 3.?0E -06 1.22E-07 Seismic CDF 2.08E-05 2.12E-05 4.00E-07 1.32E-08 Total 9.40E-06 3.09E-07 to AEP-NRC-2024-02 Page 21

Table 2 - LERF PRA Results

LERF Zero 3.8.1. LAR Reserve Delta U1 Maintenance Feed Extension CLERP ICLERP (12 days)

Case Case FPIE LERF 1.39E-06 1.69E-06 3.00E-07 9.86E-09 Fire LERF 2.44E-06 3.05E-06 6.10E-07 2.01E-08 Seismic LERF 5.30E-06 5.33E-06 3.00E-08 9.86E-10 Total 9.40E-07 3.09E-08

U2 FPIE LERF 1.40E-06 1.60E-06 2.00E-07 6.58E-09 Fire LERF 2.08E-06 2.24E-06 1.60E-07 5.26E-09 Seismic LERF 5.71E-06 5.74E-06 3.00E-08 9.86E-10 Total 3.90E-07 1.28E-08

External Events - Other

Other external events, such as high winds and external flooding, would be expected to result in a loss of offsite power during the event. Similar to the discussion for seismic events, significant external events that would cause a reactor trip would be expected to also cause a loss of offsite power. For this reason, the external event risk due to the reserve feed outage time extension is considered to be negligible.

Results Summary

Total U1 CDF and LERF Results Case ICCDP ICLERP FPIE 1.78E-07 9.86E-09 Fire PRA 1.32E-07 2.01E-08 Seismic PRA 1.64E-08 9.86E-10 Total 3.25E-07 3.09E-08

Total U2 CDF and LERF Results Case ICCDP ICLERP FPIE 1.74E-07 6.58E-09 Fire PRA 1.22E-07 5.26E-09 Seismic PRA 1.32E-08 9.86E-10 Total 3.09E-07 1.28E-08 to AEP-NRC-2024-02 Page 22

Treatment of Common Cause Failures

The type of maintenance activity that has required the proposed AOT is planned maintenance, and not an emergent equipment failure. For this reason, it is not expected that the failure probability of any other plant components is impacted during the AOT. No adjustments were made to the currently modelled common cause failure probabilities.

3.5 Conclusions

The thresholds for low risk in R.G. 1.177, Revision 1, are ICCDP < 1 E-06 and ICLERP < 1 E-07; however, the thresholds of ICCDP < 1 E-05 and ICLERP < 1 E-06 are acceptable should appropriate compensatory measure be implemented to reduce the sources of risk.

l&M has evaluated the risk implications of the proposed amendment. The risk assessment was performed assuming a 10-day Completion Time. Therefore, the requested 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months (12-day)

Completion Time is bounded by the risk assessment and associated compensatory actions.

3.6 Identification of Risk-Significant Configurations

CNP plant risk associated with the proposed extended Train B Reserve Feed Completion Time is calculated using the CNP FPIE, Fire PRA, and Seismic PRA models (including internal flooding).

Associated actions to avoid or respond to these events on one or both units through function of onsite emergency backup power supplies, and inclusion of additional onsite emergency power, are discussed in Tier 3 information, below.

Ultimately for this extended Completion Time request, CNP provides assurance that any other risk significant plant equipment outage configurations will not occur during the extended Completion Time period by flatly ruling out elective maintenance on other PRA risk significant plant equipment and avoiding other activities that could challenge unit operation or cause fires in risk significant areas. Refer to actions discussed in Section 3.2.1. The Tier 3 actions mitigate additional plant risk due to events beyond those associated with Train B Reserve Feed unavailability represented in the ICCDP and ICLERP values furnished in the Tier 1 discussion above.

Impact On Internal Events (IE)

The internal events risk impact is included in the ICCDP and ICLERP metrics provided in Tier 1. The PRA model used in this assessment includes an Internal Events (IE) model and an internal flooding model. Based on a review of the results, risk management actions will focus on protecting the Train A equipment and the importance to realign power sources and trip the RCPs when necessary. Although steam generator tube rupture was identified as a contributor, the additional risk from this event is primarily due to the potential for loss of power, so an emphasis on protecting equipment is appropriate.

Impact On Internal Flooding

Internal flooding as noted above is part of the updated PRA model used to determine the PRA metrics provided in Tier 1. Two internal flooding scenarios are shown to be significant for the internal flooding model.

The first is flooding from fire protection hose stations, 1-FHC-218 and 2-FHC-220 with the potential to disable DC power. To compensate for this risk the hoses station will be drained for the duration of the work. to AEP-NRC-2024-02 Page 23

The second internal flooding scenario of concern, mainly for LERF is a large circulating water flood in the turbine building that could ultimately submerge the AFW pumps. Operators will be briefed on the importance of recognition of a large flood and the importance to trip the CW pumps if necessary.

Impact On Fire Risk

As discussed above, the fire risk impact is included in the ICCDP and ICLERP metrics provided in Tier

1. Based on the review of the results, risk management actions will focus on maintaining the Train A equipment as available and establishing hourly fire watches in areas identified as risk significant. Similar to internal events operator action review will focus on realigning alternate power sources. The compensatory measures in Section 3.2.1 include actions to assure that fire detection and suppression systems for these areas are functional, that likelihood of fire initiation from work or operating equipment in the area is reduced/eliminated, and that flammable transient material is not present in high-risk areas.

Impact On Seismic Risk

As discussed above, the seismic risk impact is included in the ICCDP and ICLERP metrics provided in Tier 1. Based on the review of the results, equipment failures are generally not dominating the risk results. One insight (similar to FPIE LERF result) is that the operator action to mitigate a large CW seismically induced flood dominates the seismic results. Operators will be briefed on the importance of recognition of a large flood and the importance to trip the CW pumps if necessary.

Summary

For the RMAs presented in this license amendment request, l&M will avoid risk significant plant configurations such as performing elective maintenance or intrusive surveillances on the listed plant equipment, and minimizing activities that could initiate plant transients or challenge continued operation.

RG 1.177 indicates that actions modifying plant design or operating procedures, or to obtain additional backup equipment, should be considered in the Tier 1 evaluation. However, no plant modifications have been made to reduce the risks associated with these Tier 2 considerations. Additional Tier 3 actions, which are developed to reduce the risk from risk-significant configurations identified in the quantitative analysis, are also included.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

10 CFR 50.36, Technical Specifications

10 CFR 50.36(c) provides that TS will include LCOs which are "the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee will shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met." The proposed changes involve extending the Completion Time for TS 3.8.1 Condition A, Required Action A.3 from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months on a one-time basis. The LCO itself remains unchanged, as do the required remedial actions or shut down requirements in accordance with 10 CFR 50.36. In addition, 10 CFR Requires that a licensee's TS be derived from the analyses and evaluation included in the safety analysis report. The proposed changes do not affect CNP's compliance with the intent of 10 CFR 50.36. to AEP-NRC-2024-02 Page 24

1 O CFR 50.63, Loss of all alternating current

10 CFR 50.63 requires that light water-cooled nuclear power plants licensed to operate be able to withstand for a specified duration and recover from an 580. The proposed changes do not alter CNP's duration (coping time) nor affect its compliance with the intent of 10 CFR 50.63.

Plant Specific Design Criterion 39 - Emergency Power

As described in CNP's Updated Final Safety Analysis Report Section 1.4, the Plant Specific Design Criteria (PSDC) define the principal criteria and safety objectives for the CNP design. The following PSDC are relevant to the proposed amendment.

Plant Specific Design Criterion 39 - Emergency Power - An emergency power source shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the functioning of the engineered safety features and protection systems required to avoid undue risk to the health and safety of the public. This power source shall provide this capacity assuming a failure of a single active component.

The above plant specific design criteria were used in the design of CNP and the proposed amendment will not affect compliance with these criteria

4.2 Precedent

Comanche Peak Nuclear Power Plant, Units 1 and 2 - Issuance of Amendments Re: Revision to Technical Specification 3.8.1, "AC Sources-Operating," for Extension of the Completion Time for the Offsite Circuits on a One-Time Basis from 72 Hours to 14 Days, dated October 29, 2010 (ML102810130).

Surry Power Station, Units 1 and 2 - Issuance of Amendments Revising Technical Specifications Section 3.16, "Emergency Power System," for a Temporary 21-Day Allowed Outage Time, dated October 5, 2018 (ML18261A099).

Peach Bottom Atomic Power Station, Units 2 And 3-Issuance Of Amendment Nos. 328 And 331 Revising Technical Specification Section 3.81, "Ac Sources - Operating," For A One-Time Extension Of A Completion Time, dated October 29, 2019 (ML19266A622).

Seabrook Station, Unit No. 1, Issuance of Amendment No. 173 Re: Revise Technical Specification 3/4.8.1 to Allow Replacement of Reserve Auxiliary Transformer (Emergency Circumstances) (EPID L-2024-LLA-0024), dated March 8, 2024 (ML24067A262).

4.3 No Significant Hazards Consideration Determination

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2, proposes to amend Facility Operating Licenses DPR-58 and DPR-74. l&M proposes a one-time extension of the Technical Specification (TS) 3.8.1, "AC Sources -

Operating," Condition A, Required Action A.3, Completion Time from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months for an inoperable AC Electrical Source. This one-time extension is to support implementation of a modification to ensure the long-term reliability of the Train B feed of the off-site preferred power source.

l&M has evaluated whether or not a significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below: to AEP-NRC-2024-02 Page 25

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

Response: No

The proposed change is a one-time extension of the Technical Specification (TS) 3.8.1, Condition A, Required Action A.3, Completion Time from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months for an inoperable AC Electrical Source. The proposed change does not alter any plant equipment or operating practices in such a manner that the probability of an accident is increased. The proposed change will not alter assumptions relative to the mitigation of an accident or transient event. Therefore, the proposed completion time does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.

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

Response: No

The proposed change is a one-time extension of the Technical Specification (TS) 3.8.1, Condition A, Required Action A.3, Completion Time from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months for an inoperable AC Electrical Source. The proposed amendment does not introduce any new equipment, create any new failure modes for existing equipment, or create any new limiting single failures. The plant equipment considered when evaluating the existing completion time remains unchanged. The extended completion time will permit completion of repair activities without incurring transient risks associated with performing a shutdown with one train of reserve feed unavailable. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No

The proposed change is a one-time extension of the Technical Specification (TS) 3.8. 1, Condition A, Required Action A.3, Completion Time from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months for an inoperable AC Electrical Source. The proposed completion time has been evaluated on a risk-informed basis. The proposed configuration controls and compensatory measures provide reasonable assurance that no significant reduction to the margin of safety will occur. Therefore, the proposed change does not involve a significant reduction in margin of safety.

In summary, based upon the above evaluation, l&M has concluded that the proposed change involves no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions

The proposed changes do not modify any plant equipment that provides emergency power to the safety-related 4160v buses in the event of a Loss of Offsite Power (LOOP). This amendment request for a one-time extended Completion Time for TS 3.8.1, Required Action A.3, has been prepared to comply with risk considerations from RG 1.177, Revision 1. Evaluation of the proposed changes has determined that the reliability of AC electrical sources is not significantly affected by the proposed changes and that applicable regulations and requirements continue to be met. to AEP-NRC-2024-02 Page 26

In conclusion, CNP has determined that the proposed change does not require any exemptions or relief from regulatory requirements, other than the TS, and does not affect conformance with any regulatory requirements/criteria.

5.0 ENVIRONMENTAL CONSIDERATION

l&M has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. l&M has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared concerning the proposed amendment.

6.0 REFERENCES

1. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009.

2. NRC RIS 2007-06, NRC REGULATORY ISSUE

SUMMARY

2007-06 REGULATORY GUIDE 1.200 IMPLEMENTATION, March 2007.

3. NEI Letter to NRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," February 21, 2017 (ADAMS Accession No. ML17086A450).

4. NRC Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os)," May 3, 2017 (ADAMS Accession No. ML17079A427).

5. NEI 05-04, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, Revision 2, November 2008.

6. ASME/ANS RA-Sa-2009, Standard for Level I/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, dated February 2009.

7. PWROG-15076-P, Peer Review of the D. C. Cook Nuclear Plant Internal Events Probabilistic Risk Assessment, Revision 0, September 2015.

8. AEPDCC-0036-REPT-001 Revision 0, "Cook Nuclear Plant Evaluation of Detailed Hydrogen Analyses (01V015-RPT-01) Against the LERF Support Requirements of ASME PRA Standard (2013)", September 2017

9. AEPDCC-00051-REPT-001, Revision 0, "Cook Nuclear Plant Seismic PRA Hydrogen Findings Closure Review", August 2018"

10. 1 BTIV001-RPT-01, Revision 0, "DC Cook Focused Scope Peer Review - Pre-Initiator HRA,"

October 2016

11. PRA-NB-FSPR-FLEX, DC Cook Focused-Scope Peer Review on the Incorporation of FLEX, Revision 0, November 2020.

12. L TR-RAM-11-10-041, "Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements from Section 4 of ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the D.C. Fire to AEP-NRC-2024-02 Page 27

Probabilistic Risk Assessment," July 2010.

13. NEI 07-12, FIRE PROBABILISTIC RISK ASSESSMENT (FPRA) PEER REVIEW PROCESS GUIDELINES, Revision 1, June 2010.

14. D0403140002-1515, "D.C. Cook Focused Scope Peer Review for Fire PRA," November 2015

15. PWROG-17027-P, "Focused Scope Peer Review of the DC Cook Internal Fire Probabilistic Risk Assessment," July 2017

16. P3801-0001-01, Revision 0, "Focused Scope Peer Review of the D.C. Cook Nuclear Plant (CNP) Fire PRA Model Against the ASME PRA Standard Requirements," July 2022

17. P3823-0001-001, Revision O OR Revision 1, of P3801-0001-01, " Focused Scope Peer Review of the D.C. Cook (CNP) Fire PRA Model Against ASME PRA Standard Requirements," February 2023.

18. P3823-001-02, Revision 0, "F&O Closure Review of the DC Cook Nuclear Plant (CNP) Fire PRA Against the ASME/ANS PRA Standard Requirements", June 2023

19. PWROG-18062-P, Revision 0, "Peer Review of the D.C. Cook Nuclear Plant, Units 1 & 2, Seismic Probabilistic Risk Assessment", January 2019

20. PRA Standard Code Case for Part 5 ASME/ANS RA-S CASE 1, Case for ASME/ANS RA-Sb-2013 x Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications

21. WOG 2000, Reactor Coolant Pump Seal Leakage Model for Westinghouse PWRs, WCAP-15603, Revision 1-A, June 2003

22. NRC Memo Dated 5/6/2022, UPDATED ASSESSMENT OF INDUSTRY GUIDANCE FOR CREDITING MITIGATING STRATEGIES IN PROBABILISTIC RISK ASSESSMENTS (ADAMS Accession No. ML22014A084 ).

23. AR 2023-8421, "Potential inability to run DIS diesels during deep load shed" to Enclosure 2 of AEP-NRC-2024-02

PRA Open F&O's to Enclosure 2 of AEP-NRC-2024-02 Page 1

Full Power Internal Events

F&O Status SRs F&O Description Disposition The Large Early Release Frequency Pressure and (LERF) analysis uses NUREG/CR-temperature induced 6595 type evaluations for some SGTR events are portions of the LERF evaluation modeled as based on conservative progressing directly to LE-D5 assessments. These portions of the LERF. This ensures LE-C1-C5 LERF analysis do not meet an over-estimation of 2-19 LE-C9-Capability Category II of the the significance of this (2015 Open C13 standard. One example of these assumption.

Full LE-E2 conservatisms is assuming Stem Scope) Generator Tube Rupture (SGTR) is MET: CC-I a containment bypass event without considering success of secondary side isolation. However, these conservatisms generally do not impact the ability to perform Probabilistic Safety Assessment (PSA) aoolications using LERF.

Most of the notebooks indicate that There is not expected interviews with knowledgeable plant to be a significant personnel were conducted to deviation between confirm that the systems analysis what is modeled in the adequately reflected PRA and actual plant the as-built, as-operated plant and condition such that that plant-specific data was there would be a SY-A2: appropriately collected where substantive impact on 4-4 MET required. However, a record of such numerical model (2015 SY-A4: interviews was not provided as part results. Some Full PR CC-I of the notebook documentation. walkdowns and Scope) SY-C2: Plant walkdowns are discussed in a interviews have been MET generic walkdown document created performed and did not in June 1991. There is no record of identify any necessary recent system walkdowns modeling changes, the conducted with knowledgeable plant same outcome is personnel. Even if the system expected for those configuration has not changed systems that still need during that time, there should be a walkdowns and confirmatory walkdown to document interviews performed.

that.

6-19 Based on a discussion with Cook Recent outage (2015 Open DA-C13 Nuclear Plant (CNP) Probabilistic durations have been Full Risk Assessment (PRA) Engineer, long due to work Scope) the CNP model conservatively related to replacement to Enclosure 2 of AEP-NRC-2024-02 Page 2

F&O Status SRs F&O Description Disposition models the opposite unit's outage of baffle bolts in the unavailability by assuming 45 days reactor vessel, and outages with train unavailability therefore a value equal to an equivalent portion of the informed by recent outage (e.g., a 2 train system would Operating Experience assume one train unavailability is would result in an 22.5 days). overestimation of the risk associated with outage windows. The current estimate of 45-day outages bounds recent outage experience and is therefore either not expected to impact model results, or result in an overestimation.

All containment failures caused by The current hydrogen combustion are assumed implementation of the to contribute to LERF in report modeling is 01 VO 15-RPT-0 1. There is no conservative and will discussion in the report that provides result in an a basis for this approach to scenario overestimation in the assignment based on containment risk significance of failure location considerations. PRA-sequences that could NB-LER, Revision 0 also does not potentially be screened relate the assignment of LERF out based on scenarios to containment failure containment failure location. The latter document location. However, an identifies the most likely containment improvement of this 2-4 failure location from the containment modeling would not (2017 Open LE-D3: capacity report (Stevenson report) result in a significant Hydrogen CC-I and includes a historical discussion improvement in the FSPR) that provides an argument against overall realism of the assigning scenarios that involve that model results.

failure location to LERF. This indicates that there is some uncertainty about the release size from this most likely failure location which would constitute an effect of failure location on event classification that is not discussed.

However, this SR is considered met at CC-I because the failure location was assessed in a conservative manner. to Enclosure 2 of AEP-NRC-2024-02 Page 3

Fire PRA F&O Status SRs F&O Description Disposition Observation: Parametric uncertainties of applied The impact of hot short probabilities have not been this Finding is incorporated into the model. limited to small portion of the CFA2-Finding & Observation (F&O) Closure Notes: uncertainty 01 Partially CF-The CF and Uncertainty (UNC) notebooks were analysis, and (2010 Open A2 reviewed and confirmed that numerical thus does not Full Met parametric uncertainties are documented. impact the Scope) However, several inconsistencies were identified overall between the documentation and the values used technical in the model. Given that the majority of CF quality of the uncertainties have been correctly applied CF-A2 Fire PRA.

is now considered Met.

FQ-D1-Some of the Internal Events LE SRs were See disposition 02 FQ-classified as CC-I due to conservative modeling. for FPIE F&O (2022 Open D1 Therefore, the Fire LERF should also be limited 2-19.

Focused PRM-to CC-I as appropriate for applications. Revise Scope) 82 the internal events PRA to meet CC-II for relevant SRs and implement chanQes to FPRA.

Section 4.3 of the PP report says no spatial Review of the separation was credited as a partition element. PRA This statement was the basis for the CC-I implementation assessment in the original peer review. There is of fire modeling a disconnect between Section 4.3 and Section concluded that 3.1 and 3.2 that needs to be rectified. the issue described is a Additionally, Section 3.1 discussed the documentation subdivision of the yard into sub-compartments disconnect, based on spatial separation, but these sub-and therefore PP compartments do not become separate listed fire its resolution 01 PP-zones and are not separated in PRA-NB-FIRE-will not impact (2022 Open 83 IGN. No explanation is given for this in R1900- numerical Focused IGN- 0041-0001. IF the intent is to separate them for model results Scope) A7 fire modeling only (as suggested by Table 6-1 of or risk insights.

PRA-NB-FIRE-IGN), this should be stated with the PP analysis and reiterated in the PAU Table in Attachment 1 with a note for clarity.

Revise language in Section 4.3 regarding the use of spatial separation. Clarify treatment of the yard sub-compartments by adding additional discussion to Section 3.1 with a clarifying note in Attachment 1. Alternatively, carry the sub-compartments forward as separate "fire zones" into Attachment 1 and the IGN consistent with the sub-divisions of the other fire compartments. to Enclosure 2 of AEP-NRC-2024-02 Page4

Seismic PRA

F&O Status SRs F&O Description Disposition 1 ) -Only a single method was Seismic PRA (SPRA) considered to evaluate the results are not expected liquefaction triggering potential, to be impacted. l&M liquefaction susceptibility, and (2014) initially performed post liquefaction volumetric a liquefaction triggering strains. However, in F&O 20-7, (using Youd et al., 2001)

Item 2, more than one method and settlement (using was requested to conduct the Tokimatsu and Seed, liquefaction hazard evaluation as 1987) analyses using the "the choice of any single method RLE and obtained does not address the epistemic comparable results. l&M uncertainty in the field (which is considers that Figure 6-7 the underlying motivation of a shows liquefaction at recent National Academy study some boreholes for 1 E-6 and report)". motions, but shows no lateral continuity of the 1-1 SHA-I1: 2) - Lateral spreading hazard at liquefiable boreholes.

(2018 MET the site does not address the Based on this Full PR SHA-I2: evaluation of this potential hazard. information, l&M has Scope) MET Lateral spreading can occur in concluded that the site slope gradients as flat as 0.5 can be screened out for percent (%) (without a free face) site-wide lateral (See DC COOK-PR-09). spreading.

Additionally, Figures 6-7 and 6-9 shows that there is continuous layer of potentially liquefiable soils (in direction towards the lake) on borings 8120, 8124, 8133, 8142, and B 141 between elevations of about 560 and 555 ft. Therefore, the potential of lateral spreading and/or flow slides at the site should be evaluated.

3) - Provide a full reference to all citations included in the report.

1 )- Include additional justification SPRA results are not 20-3 on why Vertical-to-Horizontal expected to be impacted (2018 DO SHA-J2: (V/H) ratios should be used as this F&O has been Full MET instead of vertical Ground Motion technically resolved.

Scope) Prediction Equations (GMPEs) in report DC COOK-PR-02, Section to Enclosure 2 of AEP-NRC-2024-02 Page5

F&O Status SRs F&O Description Disposition 7.1 (e.g., inconsistency of controlling earthquakes between horizontal and vertical spectra if vertical GMPEs were used).

2) - Perform a thorough editorial review of the reference citations and list of references.

While the cracking assessment for SPRA results are not the Containment Building (CB) expected to be impacted.

and TB/SH has been resolved, the The studies performed in cracking assessment for AB has 15C4313-RPT-003 not been fully resolved. Several "Summary of Building changes were made to the AB Response Analysis for structural model in 15C4313-CAL-the Cook Nuclear Plant 010, "Response Analysis of (CNP) Unit 1 & Unit 2 Auxiliary Building," Revision 2, in SPRA," Attachment E, response to other SFR F&Os. The show that while there updated AB model was used in the may be some cracking, it cracking assessment with un-is not widespread at the cracked section properties. The RLE-level. Additionally, SPRA team performed cracking l&M engineering assessment at earthquake levels judgement is that with corresponding to 0.5*RLE the studies performed, (Review Level Earthquake) and cracking in the structure 2-1 SFR-B3: 1.0*RLE. Figures 1 through 8 in will decrease the (2019 0 MET Attachment E present the shear stiffness and increase FSPR) stress contour plots on isometric the damping. These two views of the AB model showing the effects tend to affect the exterior walls. The stress contour structural response in plots only suggest that the building opposite ways. Finally, is overly stressed in certain many significant regions. For a complex structure contributors have low such as the AB, this is not fragilities for which sufficient to conclude that cracking consideration of a will or will not occur in the building cracked model would be especially under dynamic loads. non-conservative.

The SPRA development team has not assessed or documented the cracking assessment for the AB interior walls in a way that resolves the concern identified in the initial F&O issued by the peer review team. to Enclosure 2 of AEP-NRC-2024-02 Page6

F&O Status SRs F&O Description Disposition Perform a sensitivity study to SPRA results are not address items determined to be expected to be impacted.

risk significant based on F-V l&M position is that 22-2 importance greater-than or equal-additional studies for risk (2018 SFR-E3: to 0.005. items not considered by Full DO CCII risk significant as Scope) (defined in the SPRA quantification notebook) will not change risk insights.

F&O Status SRs F&O Descriptions Dispositions Perform a sensitivity study to SPRA results are not address items determined to be expected to be impacted.

risk significant based on F-V l&M position is that importance. sensitivity studies documented in the 22-5 SPRA quantification (2018 DO SFR-E2: notebook envelope any Full CCII small fragility changes Scope) that may be discovered by the additional sensitivity recommend here and will not change risk insights.

SPRA team has used the ASCE 4-SPRA results are not 16, expected to be impacted.

Section 3.7.2 dynamic coupling The l&M position is that criteria for single-point attachment the simplified method to show that the current CB used to demonstrate that modeling approach and response the Containment are realistic. While the modeling Building (CB) modelling 28-2 approach use probably does not simplifications have no (2018 PR SFR-B3: have an effect on overall response impact on the response Full MET of the structure but that conclusion in 15C4313-RPT-003 Scope has not been demonstrated Attachment B is adequately. sufficient to address the F&O. The close-out team requested more detailed studies be performed to close the F&O, however the team stated that they believe the conclusion to Enclosure 2 of AEP-NRC-2024-02 Page 7

F&O Status SRs F&O Descriptions Dispositions will most likely not change as a result.

Appropriate damping was used for SPRA results are not cracked and un-cracked building expected to be impacted.

sections in the building response The position of l&M is sensitivity studies following the that the conclusion current industry and standard provided in 15C4313-ASCE 4-16. The sensitivity RPT-003, Attachment E, practice studies are documented is sufficient to justify the in Attachments B and F of use of un-cracked 15C4313-RPT-003, respectively, damping for the AB for Containment Building and model. See F&O 2-1 for 28-4 Turbine Building/Screen House. further information.

(2018 PR SFR-B3: Appropriate damping is also used Full MET for AB response analysis model Scope) documented in 15C4313-CAL-010 Revision 2.

However, the focused scope peer review F&O 2-1 would require to reassess the cracking assessment of AB and appropriate damping should be used if cracking is assessed to be of significance.

The SPRA development team SPRA results are not added an argument that due to the expected to be impacted.

way that fragilities were The sensitivity studies developed, including the performed in 15C4313-application of uncertainty with RPT-003 between un-respect to frequency was sufficient cracked and cracked to allow no variation in structural properties show that 28-11 properties. The variation in structural variability has (2018 SFR-B4: frequency is intended to reflect a minor impact on Full 0 MET uncertainty in the value of the response compared to Scope) calculated frequency. The the soil property variation in structural properties is variability. l&M will intended to reflect uncertainty in review the small number those properties. Both effects of impacted risk-must be considered when significant components developing fragilities. on a case-by-case basis, adjusting the FROI by an additional +/- 15% to ensure structural to Enclosure 2 of AEP-NRC-2024-02 Page 8

F&O Status SRs F&O Descriptions Dispositions variability is captured in the fragility calculations.

The gap in Power Spectral Density SPRA results are not (PSD) as described in the F&O expected to be impacted.

should be addressed per latest l&M position is that there fragility guidance document. If it is are not significant gaps confirmed that there is a gap in in energy near PSD at Frequency Range of frequencies that are Interest (FROI) of structure, then it important to risk-is recommended to perform a significant fragilities. The sensitivity study to assess the PSDs as presented were impact of the gap in energy. The developed using a 28-13 SPRA development team can logarithmic frequency (2018 PR SFR-B4: perform this by comparing the interpolation which tends Full MET PSD functions of the five-time to emphasize magnitude Scope) histories that were generated by variation at low resolution of F&O 28-09 to the frequencies. A review of PSD function of the artificial time the non-interpolated history, or the development team PSDs and PSDs can integrate the PSD function to developed using a linear show that a smooth curve is frequency interpolation generated. supports the determination that the gaps identified in the F&O are not significant.

The documentation needs to be SPRA results are not further updated to provide a basis expected to be impacted, for not considering SSSI effects. as this F&O has been Subsequent to the closure review, technically resolved.

additional documentation was Additional quantitative added to the calculations. justification added However, the closure review team to Section 4.4 of 28-19 does not consider this additional 15C4313-RPT-003 is (2018 SFR-F2: documentation to be sufficient to adequate in showing that Full DO MET address the concern originally SSSI effects do not Scope) identified. control over Review Level Earthquake (RLE) demand for applicable components. Also note that components associated in this documentation item are not risk significant. to Enclosure 2 of AEP-NRC-2024-02 Page 9

F&O Status SRs F&O Descriptions Dispositions Resolved with Open SPRA results are not Documentation In the SPRA expected to be impacted Model Quantification Notebook, as this F&O has been Section 6.;2.2, Revision1, the technically resolved.

cutset review included a statement on non-significant cutsets -

samples are covered by the 25-7 examination of G1 and G2 bins.

(2018 SPR-E3: G 1 and G2 bins contain relatively Full DO CCII fewer seismic-induced failures Scope) and the cutsets have features more like the internal events PRA.

A recommendation is made to expand the review to other ground motion bins so that model logic related specifically to the SPRA can be confirmed to be appropriate and as intended.

Some of the supporting internal No impact to SPRA events LE SRs were met at CC-I results -

only; therefore, this SR is met for The LERF modeling is CC-I only. built upon the internal events LERF model and is essentially unchanged.

The SPRA LERF model 25-9 includes seismic-specific (2018 SPR-E6: aspects such as unique Full 0 CCI containment failure Scope) probabilities. Whereas there are some internal events LE supporting SRs that meet both CC-I and CC-II, the majority of the SRs are met at CC-I.

Therefore, this SR is considered to be met at CC-I only.

Enclosure 3 to AEP-NRC-2024-02

Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked to Show Proposed Changes AC Sources - Operating 3.8.1

ACTIONS

----------NOTE----------------------------------------------------

LCO 3.0.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME

A One required offsite A.1 ---------------NOTE-------

circuit inoperable. Not applicable if a required Unit 2 offsite circuit is inoperable.

Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months required OPERABLE offsite circuit. AND

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND

A.2 Declare required feature(s) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from with no offsite power discovery of no offsite available inoperable when power to one train its redundant required concurrent with feature(s) is inoperable. inoperability of redundant required feature(s)

AND

A.3 Restore required offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months <a>

circuit to OPERABLE status. AND

17 days from discovery of failure to meet LCO 3.8.1.a orb

(a) For Train B only, the Completion Time that Train B can be inoperable as specified by Required Action A.3 may be extended beyond the "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months " up to "288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months ," to support modification of the Train B Reserve Feed 12AB Loop Feed Enclosure. Upon completion of the modification and restoration this footnote is no longer applicable. Compensatory measures described within CNP letter AEP-NRC-2024-02, dated April 3, 2024 will remain in effect during the extended period.

The one-time extension shall expire upon completion of the modification and restoration of operability for Train B.

Cook Nuclear Plant Unit 1 3.8.1-2 Amendment No. 237, 291 Enclosure 4 to AEP-NRC-2024-02

Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked to Show Proposed Changes AC Sources - Operating 3.8.1

ACTIONS

NOTE----------------------------------------------------

LCO 3.0.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME

A. One required offsite A.1 ---------------NOTE--------------

circuit inoperable. Not applicable if a required Unit 1 offsite circuit is inoperable.

Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months required OPERABLE offsite circuit. AND

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND

A.2 Declare required feature(s) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from with no offsite power discovery of no offsite available inoperable when power to one train its redundant required concurrent with feature(s) is inoperable. inoperability of redundant required feature(s)

AND

A.3 Restore required offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (al circuit to OPERABLE status. AND

17 days from discovery of failure to meet LCO 3.8.1.a orb

(a) For Train B only, the Completion Time that Train B can be inoperable as specified by Required Action A.3 may be extended beyond the "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months " up to "288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months ," to support modification of the Train B Reserve Feed 12AB Loop Feed Enclosure. Upon completion of the modification and restoration this footnote is no longer applicable. Compensatory measures described within CNP letter AEP-NRC-2024-02, dated April 3, 2024 will remain in effect during the extended period.

The one-time extension shall expire upon completion of the modification and restoration of operability for Train B.

Cook Nuclear Plant Unit 2 3.8.1-2 Amendment No.~. 273

v t e Letter Sequence Request
EPID:L-2024-LLA-0040, Unit 2 License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance Supplement, Supplement Administration Questions 30 May 2024 Answers 2 July 2024 Results Approval Other: ML24115A215
MONTHYEAR2024-05-30 [Table View]

AEP-NRC-2024-02, Unit 2 License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating

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5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

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AEP-NRC-2024-02, Unit 2 License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating

Text

Enclosures:

SUMMARY

SUMMARY

=

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

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