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

INOIANA MICHIGAN POWIR.

An MP Company BOUNDLESS ENERGY-Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 indianamichiganpower.com April 3, 2024 AEP-NRC-2024-02 Docket No:

50-315 50-316 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 10 CFR 50.90 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). to this letter provides an affirmation affidavit pertaining to the proposed amendment. 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 Page2 AEP-NRC-2024-02 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 Page 3 AEP-NRC-2024-02 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,

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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

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Public My Commission Expires '=, / ~ 3 / ';to "5 0 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:

2.1 "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."

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 Tm llrwll2 8Arct, 1 T11A 4KV LEGEND 785KV-345KV-69KV-34.5KV-28KV-12KV-4KV-600V -

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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.

AND A.2 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 Declare required feature(s) with no offsite power available inoperable when its redundant required feature(s) is inoperable.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of no offsite power to one train concurrent with 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 G. Required Action and G.1 Be in MODE 3.

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

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.

G.2 Be in MODE 5.

2.3 Reason for the Proposed Change Page 8 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours 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 CONDITION A. One required offsite circuit inoperable.

A.1 AND A.2 AND A.3 REQUIRED ACTION

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

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

Perform SR 3.8.1.1 for required OPERABLE offsite circuit.

Declare required feature(s) with no offsite power available inoperable when its redundant required feature(s) is inoperable.

Restore required offsite circuit to OPERABLE status.

Page 9 COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of no offsite power to one train concurrent with inoperability of redundant required feature(s) 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months <a>

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 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.

Page 17

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 Full Power Internal Events F&O 2-19 (2015 Full Scope) 4-4 (2015 Full Scope) 6-19 (2015 Full Scope)

Status Open PR Open SRs LE-D5 LE-C1-C5 LE-C9-C13 LE-E2 MET: CC-I SY-A2:

MET SY-A4:

CC-I SY-C2:

MET DA-C13 F&O Description The Large Early Release Frequency (LERF) analysis uses NUREG/CR-6595 type evaluations for some portions of the LERF evaluation based on conservative assessments. These portions of the LERF analysis do not meet Capability Category II of the standard. One example of these conservatisms is assuming Stem Generator Tube Rupture (SGTR) is 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 interviews with knowledgeable plant personnel were conducted to confirm that the systems analysis adequately reflected the as-built, as-operated plant and that plant-specific data was appropriately collected where required. However, a record of such interviews was not provided as part of the notebook documentation.

Plant walkdowns are discussed in a generic walkdown document created in June 1991. There is no record of recent system walkdowns conducted with knowledgeable plant personnel.

Even if the system configuration has not changed during that time, there should be a confirmatory walkdown to document that.

Based on a discussion with Cook Nuclear Plant (CNP) Probabilistic Risk Assessment (PRA) Engineer, the CNP model conservatively Page 1 Disposition Pressure and temperature induced SGTR events are modeled as progressing directly to LERF.

This ensures an over-estimation of the significance of this assumption.

There is not expected to be a significant deviation between what is modeled in the PRA and actual plant condition such that there would be a

substantive impact on numerical model results.

Some walkdowns and interviews have been performed and did not identify any necessary modeling changes, the same outcome is expected for those systems that still need walkdowns and interviews performed.

Recent outage durations have been long due to work related to replacement to Enclosure 2 of AEP-NRC-2024-02 F&O 2-4 (2017 Hydrogen FSPR)

Status Open SRs LE-D3:

CC-I F&O Description models the opposite unit's outage unavailability by assuming 45 days outages with train unavailability equal to an equivalent portion of the outage (e.g., a 2 train system would assume one train unavailability is 22.5 days).

All containment failures caused by hydrogen combustion are assumed to contribute to LERF in report 01 VO 15-RPT-0 1.

There is no discussion in the report that provides a basis for this approach to scenario assignment based on containment failure location considerations. PRA-NB-LER, Revision 0 also does not relate the assignment of LERF scenarios to containment failure location.

The latter document identifies the most likely containment failure location from the containment capacity report (Stevenson report) and includes a historical discussion that provides an argument against assigning scenarios that involve that 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.

Page 2 Disposition of baffle bolts in the reactor

vessel, and therefore a

value informed by recent Operating Experience would result in an 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.

The current implementation of the modeling is conservative and will result in an overestimation in the risk significance of sequences that could potentially be screened out based on containment failure location. However, an improvement of this modeling would not result in a significant improvement in the overall realism of the model results.

to Enclosure 2 of AEP-NRC-2024-02 Fire PRA F&O CFA2-01 (2010 Full Scope)

FQ-D1-02 (2022 Focused Scope)

PP 01 (2022 Focused Scope)

Status Partially Open Open Open SRs CF-A2 Met FQ-D1 PRM-82 PP-83 IGN-A7 F&O Description Observation: Parametric uncertainties of applied hot short probabilities have not been incorporated into the model.

Finding & Observation (F&O) Closure Notes:

The CF and Uncertainty (UNC) notebooks were reviewed and confirmed that numerical parametric uncertainties are documented.

However, several inconsistencies were identified between the documentation and the values used in the model. Given that the majority of CF uncertainties have been correctly applied CF-A2 is now considered Met.

Some of the Internal Events LE SRs were classified as CC-I due to conservative modeling.

Therefore, the Fire LERF should also be limited to CC-I as appropriate for applications. Revise 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 separation was credited as a partition element.

This statement was the basis for the CC-I assessment in the original peer review. There is a disconnect between Section 4.3 and Section 3.1 and 3.2 that needs to be rectified.

Additionally, Section 3.1 discussed the subdivision of the yard into sub-compartments based on spatial separation, but these sub-compartments do not become separate listed fire zones and are not separated in PRA-NB-FIRE-IGN. No explanation is given for this in R1900-0041-0001. IF the intent is to separate them for fire modeling only (as suggested by Table 6-1 of 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. 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.

Page 3 Disposition The impact of this Finding is limited to small portion of the uncertainty

analysis, and thus does not impact the overall technical quality of the Fire PRA.

See disposition for FPIE F&O 2-19.

Review of the PRA implementation of fire modeling concluded that the issue described is a documentation disconnect, and therefore its resolution will not impact numerical model results or risk insights.

to Enclosure 2 of AEP-NRC-2024-02 Page4 Seismic PRA F&O 1-1 (2018 Full Scope) 20-3 (2018 Full Scope)

Status PR DO SRs SHA-I1:

MET SHA-I2:

MET SHA-J2:

MET F&O Description 1 ) -Only a single method was considered to evaluate the liquefaction triggering potential, liquefaction susceptibility, and post liquefaction volumetric strains. However, in F&O 20-7, Item 2, more than one method was requested to conduct the liquefaction hazard evaluation as "the choice of any single method does not address the epistemic uncertainty in the field (which is the underlying motivation of a recent National Academy study and report)".

2) - Lateral spreading hazard at the site does not address the evaluation of this potential hazard.

Lateral spreading can occur in slope gradients as flat as 0.5 percent (%) (without a free face)

(See DC COOK-PR-09).

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.

Disposition Seismic PRA (SPRA) results are not expected to be impacted. l&M (2014) initially performed a liquefaction triggering (using Youd et al., 2001) and settlement (using Tokimatsu and

Seed, 1987) analyses using the RLE and obtained comparable results. l&M considers that Figure 6-7 shows liquefaction at some boreholes for 1 E-6 motions, but shows no lateral continuity of the liquefiable boreholes.

Based on this information, l&M has concluded that the site can be screened out for site-wide lateral spreading.

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

Prediction Equations (GMPEs) in report DC COOK-PR-02, Section to Enclosure 2 of AEP-NRC-2024-02 Page5 F&O Status 2-1 (2019 FSPR) 0 SRs SFR-B3:

MET F&O Description 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.

Disposition 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 1.0*RLE. Figures 1 through 8 in will decrease the Attachment E present the shear stiffness and increase 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 F&O 22-2 (2018 Full Scope)

F&O 22-5 (2018 Full Scope) 28-2 (2018 Full Scope Status SRs SFR-E3:

DO CCII Status SRs DO SFR-E2:

CCII PR SFR-B3:

MET F&O Description Perform a sensitivity study to address items determined to be risk significant based on F-V importance greater-than or equal-to 0.005.

F&O Descriptions Perform a sensitivity study to address items determined to be risk significant based on F-V importance.

SPRA team has used the ASCE 4-16, Section 3.7.2 dynamic coupling criteria for single-point attachment to show that the current CB modeling approach and response are realistic. While the modeling approach use probably does not have an effect on overall response of the structure but that conclusion has not been demonstrated adequately.

Page6 Disposition SPRA results are not expected to be impacted.

l&M position is that additional studies for risk items not considered by risk significant as (defined in the SPRA quantification notebook) will not change risk insights.

Dispositions SPRA results are not expected to be impacted.

l&M position is that sensitivity studies documented in the SPRA quantification notebook envelope any small fragility changes that may be discovered by the additional sensitivity recommend here and will not change risk insights.

SPRA results are not expected to be impacted.

The l&M position is that the simplified method used to demonstrate that the Containment Building (CB) modelling simplifications have no impact on the response in 15C4313-RPT-003 Attachment B

is 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 28-4 (2018 Full Scope) 28-11 (2018 Full Scope)

PR 0

SRs SFR-B3:

MET SFR-B4:

MET 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 Turbine Building/Screen House. further information.

Appropriate damping is also used for AB response analysis model 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 added an argument that due to the way that fragilities were developed, including the application of uncertainty with respect to frequency was sufficient to allow no variation in structural properties.

The variation in frequency is intended to reflect uncertainty in the value of the calculated frequency.

The variation in structural properties is intended to reflect uncertainty in those properties.

Both effects must be considered when developing fragilities.

SPRA results are not expected to be impacted.

The sensitivity studies performed in 15C4313-RPT-003 between un-cracked and cracked properties show that structural variability has a

minor impact on response compared to the soil property variability.

l&M will review the small number of impacted risk-significant components on a case-by-case basis, adjusting the FROI by an additional +/-

15% to ensure structural to Enclosure 2 of AEP-NRC-2024-02 F&O 28-13 (2018 Full Scope) 28-19 (2018 Full Scope)

Status PR DO SRs SFR-B4:

MET SFR-F2:

MET F&O Descriptions The gap in Power Spectral Density (PSD) as described in the F&O should be addressed per latest fragility guidance document. If it is confirmed that there is a gap in PSD at Frequency Range of Interest (FROI) of structure, then it is recommended to perform a sensitivity study to assess the impact of the gap in energy. The SPRA development team can perform this by comparing the PSD functions of the five-time histories that were generated by resolution of F&O 28-09 to the PSD function of the artificial time history, or the development team can integrate the PSD function to show that a smooth curve is generated.

The documentation needs to be further updated to provide a basis for not considering SSSI effects.

Subsequent to the closure review, additional documentation was added to the calculations.

However, the closure review team does not consider this additional documentation to be sufficient to address the concern originally identified.

Page 8 Dispositions variability is captured in the fragility calculations.

SPRA results are not expected to be impacted.

l&M position is that there are not significant gaps in energy near frequencies that are important to risk-significant fragilities. The PSDs as presented were developed using a

logarithmic frequency interpolation which tends to emphasize magnitude variation at low frequencies. A review of the non-interpolated PSDs and PSDs developed using a linear frequency interpolation supports the determination that the gaps identified in the F&O are not significant.

SPRA results are not expected to be impacted, as this F&O has been technically resolved.

Additional quantitative justification added to Section 4.4 of 15C4313-RPT-003 is adequate in showing that SSSI effects do not 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 F&O 25-7 (2018 Full Scope) 25-9 (2018 Full Scope)

Status DO 0

SRs SPR-E3:

CCII SPR-E6:

CCI F&O Descriptions Resolved with Open Documentation In the SPRA Model Quantification Notebook, Section 6.;2.2, Revision1, the cutset review included a statement on non-significant cutsets samples are covered by the examination of G1 and G2 bins.

G 1 and G2 bins contain relatively fewer seismic-induced failures 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 events LE SRs were met at CC-I only; therefore, this SR is met for CC-I only.

Page 9 Dispositions SPRA results are not expected to be impacted as this F&O has been technically resolved.

No impact to SPRA results -

The LERF modeling is built upon the internal events LERF model and is essentially unchanged.

The SPRA LERF model includes seismic-specific aspects such as unique containment failure 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.

to AEP-NRC-2024-02 Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked to Show Proposed Changes

ACTIONS AC Sources - Operating 3.8.1

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.

AND A.2 AND 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 Declare required feature(s) with no offsite power available inoperable when its redundant required feature(s) is inoperable.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of no offsite power to one train concurrent with 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 <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 to AEP-NRC-2024-02 Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked to Show Proposed Changes

ACTIONS AC Sources - Operating 3.8.1

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

LCO 3.0.4.b is not applicable to DGs.

CONDITION A. One required offsite circuit inoperable.

A.1 AND A.2 AND A.3 REQUIRED ACTION

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

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

Perform SR 3.8.1.1 for required OPERABLE offsite circuit.

Declare required feature(s) with no offsite power available inoperable when its redundant required feature(s) is inoperable.

Restore required offsite circuit to OPERABLE status.

COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of no offsite power to one train concurrent with inoperability of redundant required feature(s) 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (al 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

Text

Enclosures:

SUMMARY

SUMMARY

3.0 TECHNICAL EVALUATION

Background

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

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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

3.0 TECHNICAL EVALUATION

Background

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

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