AEP-NRC-2024-03, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

ML24073A234
Person / Time
Site:	Cook
Issue date:	03/06/2024
From:	Ferneau K Indiana Michigan Power Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
AEP-NRC-2024-03
Download: ML24073A234 (1)

Text

INDIANA MICHIGAN POWER An AEP Company BOUNDLESS ENERGY March 6, 2024 Docket Nos.:

50-315 50-316 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, D. C. 20555-0001 Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 indianamichiganpower.com AEP-NRC-2024-03 10 CFR 50.69 10 CFR 50.90 Donald C. Cook Nuclear Plant Units 1 and 2 Application to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors" In accordance with the provisions of Sections 50.90 and 50.69 of Title 10 of the Code of Federal Regulations (10 CFR), Indiana Michigan Power Company (I&M) is requesting an amendment to the renewed facility operating licenses (FOL) of Donald C. Cook Nuclear Plant (CNP) Units 1 and 2.

The proposed amendment would modify the CNP Units 1 and 2 licensing basis, by the addition of a License Condition, to allow for the implementation of the provisions of 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors."

The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety. to this letter provides an affirmation statement. Enclosure 2 to this letter provides the basis for the proposed change to the CNP Units 1 and 2 Operating Licenses. The categorization process being implemented through this change is consistent with Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, dated July 2005, which was endorsed by the Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1, dated May 2006. Attachment 1 of Enclosure 2 provides a list of categorization prerequisites. Use of the categorization process on a plant system will only occur after these prerequisites are met. Enclosure 3 contains a marked up copy of the applicable Unit 1 and Unit 2 FOL page.

Approval of the proposed amendment is requested commensurate with the NRC's normal review schedule. The license amendment will be implemented within 90 days of NRC approval.

U. S. Nuclear Regulatory Commission Page2 There are no new or revised regulatory commitments made in this submittal.

AEP-NRC-2024-03 In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated Michigan state officials.

Should you have any questions concerning this submittal, please contact Mr. Michael K. Scarpello, Regulatory Affairs Director at (269) 466-2649.

Sincerely, W} } 'e Kelly J. Ferneau Site Vice President Indiana Michigan Power Company JMT/sjh

Enclosures:

1. Affirmation

2. Evaluation of the Proposed Change

3. Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Renewed Facility Operating Licenses (FOL)

Pages Marked to Show Proposed Change 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, wlo enclosure to AEP-NRC-2024-03 AFFIRMATION I, Kelly J. Ferneau, being duly sworn, state that I am the 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 Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME

-w 7His Ge DAY or

[rc Notary Public ww comics6 ss,es/as) 203o

, 2024

.J to AEP-NRC-2024-03 Evaluation of the Proposed Change

Subject:

Application to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors."

1.

SUMMARY

DESCRIPTION

2.

DETAILED DESCRIPTION 2.1 Current Regulatory Requirements 2.2 Reason for the Proposed Change 2.3 Description of the Proposed Change

3.

TECHNICAL EVALUATION 3.1 Categorization Process Description (10 CFR 50.69(b)(2)(i))

3.1.1 Overall Categorization Process 3.1.2 Passive Categorization Process 3.2 Technical Adequacy Evaluation (10 CFR 50.69(b)(2)(ii))

3.2.1 Internal Events and Internal Flooding 3.2.2 Fire Hazards 3.2.3 Seismic Hazards 3.2.4 Other External Hazards 3.2.5 Low Power & Shutdown 3.2.6 Probabilistic Risk Assessment (PRA) Maintenance and Updates 3.2. 7 PRA Uncertainty Evaluations 3.2.8 Modeling of FLEX 3.3 PRA Review Process Results (10 CFR 50.69(b)(2)(iii))

3.4 Risk Evaluations (10 CFR 50.69(b)(2)(iv))

3.5 Feedback and Adjustment Process

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration Determination Analysis 4.3 Conclusions

5.

ENVIRONMENTAL CONSIDERATION

6.

REFERENCES ATTACHMENTS

1. List of Categorization Prerequisites

2. Description of PRA Models Used in Categorization

3. Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items

4. External Hazards Screening

5. Progressive Screening Approach for Addressing External Hazards

6. Disposition of Key Assumptions/Sources of Uncertainty to AEP-NRC-2024-03 Page2 1.0

SUMMARY

DESCRIPTION The proposed amendment modifies the licensing basis to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors."

The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance (LSS), alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance (HSS), requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

2.0 2.1 DETAILED DESCRIPTION Current Regulatory Requirements The Nuclear Regulatory Commission (NRG) has established a set of regulatory requirements for commercial nuclear reactors to ensure that a reactor facility does not impose an undue risk to the health and safety of the public, thereby providing reasonable assurance of adequate protection to public health and safety. The current body of NRG regulations and their implementation are largely based on a "deterministic" approach.

This deterministic approach establishes requirements for engineering margin and quality assurance in design, manufacture, and construction. In addition, it assumes that adverse conditions can exist (e.g., equipment failures and human errors) and establishes a specific set of design basis events (DBEs). The deterministic approach then requires that the facility include safety systems capable of preventing or mitigating the consequences of those DBEs to protect public health and safety. The Structures, Systems and Components (SSCs) necessary to defend against the DBEs are defined as "safety-related," and these SSCs are the subject of many regulatory requirements, herein referred to as "special treatments," designed to ensure that they are of high quality and high reliability, and have the capability to perform during postulated design basis conditions. Treatment includes, but is not limited to, quality assurance, testing, inspection, condition monitoring, assessment, evaluation, and resolution of deviations. The distinction between "treatment" and "special treatment" is the degree of NRG specification as to what must be implemented for particular SSCs or for particular conditions.

Typically, the regulations establish the scope of SSCs that receive special treatment using one of three different terms: "safety-related," "important to safety," or "basic component." The terms "safety related" and "basic component" are defined in the regulations, while "important to safety," used principally in the general design criteria (GDC) of Appendix A to 10 CFR Part 50, is not explicitly defined.

2.2 Reason for Proposed Change A probabilistic approach to regulation enhances and extends the traditional deterministic approach by allowing consideration of a broader set of potential challenges to safety, providing a logical means for prioritizing these challenges based on safety significance, and allowing consideration of a broader set of resources to defend against these challenges. In contrast to the deterministic approach, PRAs address credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for common cause failures. The probabilistic approach to regulation is an extension and enhancement of traditional regulation by considering risk in a comprehensive manner.

to AEP-NRC-2024-03 Page 3 To take advantage of the safety enhancements available through the use of PRA, in 2004 the NRC published a new regulation, 10 CFR 50.69. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of LSS, alternative treatment requirements can be implemented in accordance with the regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

The rule contains requirements on how a licensee categorizes SSCs using a risk-informed process, adjusts treatment requirements consistent with the relative significance of the SSC, and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four risk-informed safety class (RISC) categories. The determination of safety significance is performed by an integrated decision-making process, as described by Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline" [1 ]), which uses both risk insights and traditional engineering insights. The safety functions include the design basis functions, as well as functions credited for severe accidents (including external events). Special or alternative treatment for the SSCs is applied as necessary to maintain functionality and reliability and is a function of the SSC categorization results and associated bases.

Finally, periodic assessment activities are conducted to make adjustments to the categorization and/or treatment processes as needed so that SSCs continue to meet all applicable requirements.

The rule does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility. Instead, the rule enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as high safety significant, existing treatment requirements are maintained or enhanced. Conversely, for SSCs that do not significantly contribute to plant safety on an individual basis, the rule allows an alternative risk-informed approach to treatment that provides reasonable, though reduced, level of confidence that these SSCs will satisfy functional requirements.

Implementation of 10 CFR 50.69 will allow Indiana Michigan Power Company (l&M) to improve focus on equipment that has safety significance resulting in improved plant safety.

2.3 Description of The Proposed Change l&M proposes the addition of the following condition to the renewed facility operating licenses (FOL) of CNP Units 1 and 2 to document the NRC's approval of the use 10 CFR 50.69:

"The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC) RISC-1, RISC-2, RISC3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the Individual Plant Evaluation-External Events (IPEEE) Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; as specified in License Amendment No. [XXX] dated

[DATE].

to AEP-NRC-2024-03 Page4 Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach)."

3.0 TECHNICAL EVALUATION

10 CFR 50.69 specifies the information to be provided by a licensee requesting adoption of the regulation. This request conforms to the requirements of 10 CFR 50.69(b)(2), which states:

A licensee voluntarily choosing to implement this section shall submit an application for license amendment under 10 CFR 50.90 that contains the following information:

(i)

A description of the process for categorization of RISC-1, RISC-2, RISC-3 and RISC-4 SSCs.

(ii)

A description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown (including the plant-specific probabilistic risk assessment (PRA), margins-type approaches, or other systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs.

(iii)

Results of the PRA review process conducted to meet 10 CFR 50.69(c)(1)(i).

(iv)

A description of, and basis for acceptability of, the evaluations to be conducted to satisfy 10 CFR 50.69(c)(1 )(iv). The evaluations must include the effects of common cause interaction susceptibility, and the potential impacts from known degradation mechanisms for both active and passive functions, and address internally and externally initiated events and plant operating modes (e.g., full power and shutdown conditions).

Each of these submittal requirements are addressed in the following sections.

Though routine maintenance updates have been applied, the NRC has previously reviewed and accepted the technical adequacy of the CNP Units 1 and 2 PRA models identified in this application for TSTF 425 [44] and for 50.54f Fukushima Response [45].

3.1 CATEGORIZATION PROCESS DESCRIPTION(10 CFR 50.69(8)(2)(1))

3.1.1 Overall Categorization Process l&M will implement the risk categorization process in accordance with NEI 00-04, Revision 0, as endorsed by Regulatory Guide (RG) 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance" [2]. NEI 00-04 Section 1.5 states "Due to the varying levels of uncertainty and degrees of conservatism in the spectrum of risk contributors, the risk significance of SSCs is assessed separately from each of five risk perspectives and used to identify SSCs that are potentially safety-significant." A separate evaluation is appropriate to avoid reliance on a combined result that may mask the results of individual risk contributors.

The process to categorize each system will be consistent with the guidance in NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," [1] as endorsed by RG 1.201 [2]. RG 1.201 states that "the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to to AEP-NRC-2024-03 Page 5 providing reasonable confidence" and that "all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by 10 CFR 50.69(c)(l)(iv)." However, neither RG 1.201 nor NEI 00-04 prescribe a particular sequence or order for each of the elements to be completed. Therefore, the order in which each of the elements of the categorization process (listed below) is completed is flexible and as long as they are all completed, they may even be performed in parallel. Note that NEI 00-04 only requires Item 3 to be completed for components/functions categorized as LSS by all other elements. Similarly, NEI 00-04 only requires Item 4 to be completed for safety related active components/functions categorized as LSS by all other elements.

1. PRA-based evaluations (e.g., the internal events, internal flooding, fire, and seismic PRAs)

2. Non-PRA approaches (e.g., other external events screening and shutdown assessment)

3. Seven qualitative criteria in Section 9.2 of NEI 00-04

4. The defense-in-depth assessment

5. The passive categorization methodology Categorization of SSCs will be completed per the NEI 00-04 process, as endorsed by RG 1.201 [2],

which includes the determination of safety significance through the various elements identified above.

The results of these elements are used as inputs to arrive at a preliminary component categorization (i.e., High Safety Significant (HSS) or Low Safety Significant (LSS)) that is presented to the Integrated Decision-Making Panel (IDP)). Note: the term "preliminary HSS or LSS" is synonymous with the NEI 00-04 term "candidate HSS or LSS." A component or function is preliminarily categorized as HSS if any element of the process results in a preliminary HSS determination in accordance with Table 3-1 below. The safety significance determination of each element, identified above, is independent of each other and therefore the sequence of the elements does not impact the resulting preliminary categorization of each component or function. Consistent with NEI 00-04, the categorization of a component or function will be "preliminary" until it has been confirmed by the IDP. Once the IDP confirms that the categorization process was followed appropriately, the final Risk Informed Safety Class (RISC) category can be assigned.

The IDP may direct and approve detailed categorization of components in accordance with NEI 00-04, Section 10.2. The IDP may always elect to change a preliminary LSS component or function to HSS, however the ability to change component categorization from preliminary HSS to LSS is limited.

This ability is only available to the IDP for select process steps as described in NEI 00-04 and endorsed by RG 1.201 [2].

Table 3-1 summarizes these IDP limitations in NEI 00-04. The steps of the process are performed at either the function level, component level, or both, which is also summarized in the Table 3-1. A component is assigned its final RISC category upon approval by the IDP.

i to AEP-NRC-2024-03 Table 3-1: Categorization Evaluation Summary Page 6 Categorization Step - NEI IDP Change Drives Element Evaluation Level Associated 00-04 Section HSS to LSS Functions Internal Events Base Case Not Allowed Yes

- Section 5.1 Risk Fire, Seismic and Other External Hazards Base Allowable No (PRA Case Component Modeled)

PRA Sensitivity Studies Allowable No Integral PRA Assessment -

Not Allowed Yes Section 5.6 Risk Fire, Seismic and Other Component Not Allowed No (Non-External Hazards -

modeled)

Shutdown - Section 5.5 Function/Component Not Allowed No Defense-in Core Damage - Section 6.1 Function/Component Not Allowed Yes Depth Containment-Section 6.2 Component Not Allowed Yes Qualitative Considerations -

Function Allowable1 N/A Criteria Section 9.2 Passive Passive -- Section 4 Segment/Component Not Allowed No Notes:

1 The assessments of the qualitative considerations are agreed upon by the /DP in accordance with Section 9.2. In some cases, a 50.69 categorization team may provide preliminary assessments of the seven considerations for the IDP's consideration, however the final assessments of the seven considerations are the direct responsibility of the /DP.

The seven considerations are addressed preliminarily by the 50.69 categorization team for at least the system functions that are not found to be HSS due to any other categorization step.

Each of the seven considerations requires a supporting justification for confirming (true response) or not confirming (false response) that consideration. If the 50.69 categorization team determines that one or more of the seven considerations cannot be confirmed, then that function is presented to the /DP as preliminary HSS. Conversely, if all the seven considerations are confirmed, then the function is presented to the /DP as preliminary LSS.

The System Categorization Document, including the justifications provided for the qualitative considerations, is reviewed by the /DP. The /DP is responsible for reviewing the preliminary assessment to the same level of detail as the 50.69 team. (i.e., all considerations for all functions are reviewed). The /DP may confirm the preliminary function risk and associated justification or may direct that it be changed based upon their expert knowledge. Because the Qualitative Criteria are the direct responsibility of the /DP, changes may be made from preliminary HSS to LSS or from preliminary LSS to HSS at the discretion of the /DP. If the /DP determines any of the seven considerations cannot be confirmed (false response) for a function, then the final categorization of that function is HSS The mapping of components to system functions is used in some categorization process steps to facilitate preliminary categorization of components. Specifically, functions with mapped components that are determined to be HSS by the PRA-based assessment (i.e., Internal events PRA or Integral PRA assessment) or defense-in-depth evaluation will be initially treated as HSS. However, NEI 00-

=

to AEP-NRC-2024-03 Page 7 04, Section 10.2, allows detailed categorization which can result in some components mapped to HSS functions being treated as LSS; and Section 4.0 discusses additional functions that may be identified (e.g., fill and drain) to group and consider potentially LSS components that may have been initially associated with an HSS function but which do not support the critical attributes of that HSS function. Note that certain steps of the categorization process are performed at a component level (e.g., Passive, Non-PRA-modeled hazards - see Table 3-1 above). These components from the component level assessments will remain HSS (IDP cannot override) regardless of the significance of the functions to which they are mapped. Therefore, if an HSS component is mapped to an LSS function, that component will remain HSS. If an LSS component is mapped to an HSS function, that component may be driven HSS, based on Table 3-1 above, or may remain LSS.

The following are clarifications to be applied to the NEI 00-04 categorization process:

The Integrated Decision-Making Panel (IDP) will be composed of a group of at least five experts who collectively have expertise in plant operation, design engineering, system engineering, safety analysis, and probabilistic risk assessment. At least three members of the IDP will have a minimum of five years of experience at the plant, and there will be at least one member of the IDP who has a minimum of three years of experience in the modeling and updating of the plant-specific PRA.

The IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address at a minimum the purpose of the categorization; present treatment requirements for SSCs including requirements for design basis events; PRA fundamentals; details of the plant specific PRA including the modeling, scope, and assumptions, the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and the defense-in-depth philosophy and requirements to maintain this philosophy.

The decision criteria for the IDP for categorizing SSCs as safety significant or low safety-significant pursuant to 10 CFR 50.69(f)(1) will be documented in l&M procedures.

Decisions of the IDP will be arrived at by consensus.

Differing opinions will be documented and resolved, if possible. However, a simple majority of the panel is sufficient for final decisions regarding High Safety Significant (HSS) and Low Safety Significant (LSS).

Passive characterization will be performed using the processes described in Section 3.1.2. of this Enclosure, consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

An unreliability factor of three (3) will be used for the sensitivity studies described in Section 8 of NEI 00-04 for LSS components. The factor of three (3) was chosen as it is representative of the typical error factor of basic events used in the PRA model.

NEI 00-04, Section 7, requires assigning the safety significance of functions to be preliminary HSS if it is supported by an SSC determined to be HSS from the PRA-based assessment in Section 5, but does not require this for SSCs determined to be HSS from non-PRA-based, deterministic assessments in Section 5. This requirement is further clarified in the Vogtle Safety Evaluation Report [4] which states "...if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth to AEP-NRC-2024-03 Page 8 assessment (Section 6), the associated system function(s) would be identified as HSS."

Once a system function is identified as HSS, then all the components that support that function are preliminary HSS. The IDP must intervene to assign any of these HSS Function components to Low Safety Significant (LSS).

With regard to the criteria that considers whether the active function is called out or relied upon in the plant Emergency/Abnormal Operating Procedures, l&M will not take credit for alternate means unless the alternate means are proceduralized and included in Licensed Operator training.

The risk analysis to be implemented for each modeled hazard is described below.

Note: previous versions of the following PRA models were submitted and accepted by the NRC for the listed applications. Any methodology upgrades have been peer reviewed consistent with RG 1.200 Revision 3.

Internal Event Risks: Internal events including internal flooding PRA CNP FPIE 2023-RO, April 2023. Revision 15MORW R1 was previously accepted by NRC for TSTF 425 [45].

Fire Risks: Fire PRA model CNP FPRA 2023-R0, February 2023. The 2014 Revisions of the FPRA was previously accepted by NRC for TSTF 425 [44].

Seismic Risks: Seismic PRA model CNP SPRA 2023-R0. Revision CNP SPRA 2019-R0 was previously accepted by NRC for 50.54f Fukushima Response [45].

Other External Risks (e.g., tornados, external floods): Hazards are dispositioned in of this enclosure. The other external hazards were determined to be insignificant contributors to plant risk.

Low Power and Shutdown Risks: Qualitative defense-in-depth (DID) shutdown model for shutdown Configuration Risk Management (CRM) based on the framework for DID provided in NUMARC 91-06, "Guidance for Industry Actions to Assess Shutdown Management" [3],

which provides guidance for assessing and enhancing safety during shutdown operations.

A change to the categorization process that is outside the bounds specified above (e.g.,

change from a seismic margins approach to a seismic PRA approach) will not be used without prior NRC approval. The SSC categorization process documentation will include the following elements:

1. Program procedures used in the categorization

2. System functions, identified and categorized with the associated bases

3. Mapping of components to supported function(s)

4. PRA model results, including sensitivity studies

5. Hazards analyses, as applicable

6. Passive categorization results and bases

7. Categorization results including all associated bases and RISC classifications

8. Component critical attributes for HSS SSCs

9. Results of periodic reviews and SSC performance evaluations

10. IDP meeting minutes and qualification/training records for the IDP members to AEP-NRC-2024-03 3.1.2 Passive Categorization Process Page 9 For the purposes of 10 CFR 50.69 categorization, passive components are those components that have a pressure retaining function. Passive components and the passive function of active components will be evaluated using the Arkansas Nuclear One (ANO)

Risk-Informed Repair/Replacement Activities (RI-RRA) methodology, consistent with the related Safety Evaluation (SE) issued by the Office of Nuclear Reactor Regulation [5].

The RI-RRA methodology is a risk-informed safety categorization and treatment program for repair/replacement activities for pressure retaining items and their associated supports. In this method, the component failure is assumed with a probability of 1.0 and only the consequence evaluation is performed. It additionally applies deterministic considerations (e.g., DID, safety margins) in determining safety significance. Consistent with the ANO-2 Safety Evaluation (ADAMS Accession No. ML090930246), pipe supports were not required to be in the scope of the evaluation, but may be included in the scope at the licensee's discretion. Component supports, if categorized, are assigned safety based upon one of the following approaches:

Supports should have the same safety significance as the highest passively ranked component within the bounds of the associated analytical pipe stress model.

Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

A combination of restraints or supports such that the LSS piping and associated SSCs attached to the HSS piping are included in scope up to a boundary point that encompasses at least two supports in each of three orthogonal directions [27, 28].

The use of this method was previously approved to be used for a 10 CFR 50.69 application by NRC in the final Safety Evaluation for Vogtle dated December 17, 2014 [4]. The RI-RRA method as approved for use at Vogtle for 10 CFR 50.69 does not have any plant specific aspects and is generic.

It relies on the conditional core damage and large early release probabilities associated with postulated ruptures. Safety significance is generally measured by the frequency and the consequence of the event. However, this RI-RRA process categorizes components solely based on consequence, which measures the safety significance of the passive component given that it ruptures. This approach is conservative compared to including the rupture frequency in the categorization as this approach will not allow the categorization of SSCs to be affected by any changes in frequency due to changes in treatment. The passive categorization process is intended to apply the same risk-informed process accepted by the NRC in the ANO2-R&R-004 for the passive categorization of Class 2, 3, and non-class components [5]. This is the same passive SSC scope the NRC has conditionally endorsed in ASME Code Cases N-660 and N-662 as published in Regulatory Guide 1.14 7, Revision 15. Both code cases employ a similar risk-informed safety categorization of SSCs in order to change the repair/ replacement requirements of the affected LSS components. All categorized ASME Code Class 1 SSCs with a pressure retaining function, as well as categorized supports, will be assigned HSS for passive categorization which will result in HSS for its risk-informed safety categorization and cannot be changed by the IDP. Therefore, this methodology and scope for passive categorization is acceptable and appropriate for use at CNP for 10 CFR 50.69 SSC categorization.

3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii))

The following sections demonstrate that the quality and level of detail of the processes used in categorization of SSCs are adequate. The PRA models described below have been independently peer reviewed and there are no PRA upgrades that have not been peer reviewed.

to AEP-NRC-2024-03 3.2.1 Internal Events and Internal Flooding Page 10 The CNP categorization process for the internal events and internal flooding hazards will use a technically acceptable and independently peer reviewed plant-specific PRA model. The CNP risk management process ensures that the PRA model used in this application reflects the as-built as-operated plant for CNP, Units 1 and 2. Attachment 2 of this enclosure identifies the applicable internal events (including internal flooding) PRA models.

3.2.2 Fire Hazards The CNP categorization process for fire hazards will use a technically acceptable and independently peer reviewed plant-specific Fire PRA model. The internal Fire PRA model was developed consistent with NUREG/CR 6850 [15], R.G. 1.200 [7] and the 2009 ASME/ANS PRA Standard [10]. The CNP risk management process ensures that the PRA model used in this application reflects the as-built as-operated plant for CNP, Units 1 and 2. Attachment 2 of this enclosure identifies the applicable internal Fire PRA model.

3.2.3 Seismic Hazards The CNP categorization process for seismic hazards will use a technically acceptable and independently peer reviewed plant-specific seismic PRA model. The CNP risk management process ensures that the PRA model used in this application reflects the as-built as-operated plant for CNP, Units 1 and 2. Industry standard methods were utilized in the development of the seismic hazards for the seismic PRA. Updates to the seismic hazard curves will be reflected in the PRA used for the categorization in accordance with the PRA model maintenance process. Attachment 2 of this enclosure identifies the applicable seismic PRA model.

3.2.4 Other External Hazards All external hazards, except for seismic, were screened for applicability to CNP, Units 1 and 2, per a plant-specific evaluation using the external hazard screening significance process identified in American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard RA-Sa-2009 [1 0]. Attachment 4 of this enclosure provides a summary of the other external hazards screening results. Attachment 5 of this enclosure provides a summary of the progressive screening approach for external hazards.

3.2.5 Low Power & Shutdown Consistent with NEI 00-04 [1], the CNP categorization process will use the shutdown safety management plan described in NU MARC 91-06 [3] for evaluation of safety significance related to low power and shutdown conditions. The overall process for addressing shutdown risk is illustrated in Figure 5-7 of NEI 00-04 [1].

NUMARC 91-06 specifies that a defense-in-depth approach should be used with respect to each defined shutdown key safety function. The key safety functions defined in NUMARC 91-06 are evaluated for categorization of SSCs.

SSCs that meet either of the two criteria (i.e., considered part of a "primary shutdown safety system" or a failure would initiate an event during shutdown conditions) described in Section 5.5 NEI 00-04 will be considered preliminary HSS.

to AEP-NRC-2024-03 3.2.6 PRA Maintenance and Updates Page 11 The CNP risk management process ensures that the applicable PRA models used in this application continue to reflect the as-built and as-operated plant for CNP. The CNP process delineates the responsibilities and guidelines for updating and maintaining the PRA models current with the design and operation of the station and includes criteria for both regularly scheduled and interim PRA model updates. The CNP process includes provisions for monitoring potential areas affecting the PRA models (e.g., due to changes in the plant, equipment performance, errors or limitations identified in the model, and industry operational experience) for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files. The CNP process directs the assessment of the impact of these changes on the plant PRA model in a timely manner but no longer than once every two refueling outages. If there is a significant impact on the PRA model, the SSC categorization will be re-evaluated and presented to the IDP.

In addition, CNP will implement procedural guidance for a process that addresses the requirements in NEI 00-04 [1 ], Section 11, "Program Documentation and Change Control." The process will review the results of periodic and interim updates of the plant PRA that may affect the results of the categorization process. If the results are affected, adjustments will be made as necessary to the categorization or treatment processes to maintain the validity of the processes. In addition, any PRA model change that is determined to meet the criteria of an upgrade in accordance with industry guidance [14] will be peer reviewed prior to implementing those changes in the PRA model used for categorization.

3.2. 7 PRA Uncertainty Evaluations Uncertainty evaluations associated with any applicable baseline PRA model(s) used in this application were evaluated during the assessment of PRA technical adequacy and confirmed through the self-assessment and peer review processes as discussed in Section 3.3 of this enclosure.

Uncertainty evaluations associated with the risk categorization process are addressed using the processes discussed in Section 8 of NEI 00-04 and in the prescribed sensitivity studies discussed Section 5.

In the overall risk sensitivity studies, CNP will utilize a factor of 3 to increase the unavailability or unreliability of LSS components consistent with that approved for Vogtle [4]. Consistent with the NEI 00-04 guidance, CNP will perform both an initial sensitivity study and a cumulative sensitivity study.

The initial sensitivity study applies to the system that is being categorized. In the cumulative sensitivity study, the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in all identified PRA models for all systems that have been categorized are increased by a factor of 3. This sensitivity study together with the periodic review process assures that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed, through CNP's Corrective Action Program, before reaching the rate assumed in the sensitivity study.

The CNP internal events PRA model and Fire PRA model and documentation were reviewed for plant-specific and generic modeling assumptions and related sources of uncertainty. The process to evaluate uncertainties is defined in NUREG 1855 Rev. 1 [8] and Electric Power Research Institute (EPRI) Technical Reports TR-1016737 [9] and TR-1026511 [46] Each PRA model includes an evaluation of the potential sources of uncertainty for the base case models using the approach that is consistent with the ASME/ANS RA-Sa-2009 requirements for identification and characterization of uncertainties and assumptions.

to AEP-NRC-2024-03 Page 12 Each PRA element notebook was also reviewed for assumptions and sources of uncertainties. The characterization of assumptions and sources of uncertainties are based on whether the assumption and/or source of uncertainty is key to the 50.69 application in accordance with RG 1.200 Revision 2.

Key CNP PRA model specific assumptions and sources of uncertainty for this application were identified and dispositioned in Attachment 6 of this enclosure. The conclusion of this review is that no additional sensitivity analyses are required to address CNP PRA model specific assumptions or sources of uncertainty.

3.2.8 Modeling of FLEX The NRC has been issuing a "generic" Request for Additional Information (RAI) regarding crediting of FLEX equipment in PRA models. The Limerick RAI (ADAMS Accession No. ML19192A031) is summarized below:

The NRG 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 ", 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] 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 Rf CT 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 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 to AEP-NRC-2024-03 Page 13 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 ASMEIANS RA-Sa-2009 PRA Standard were evaluated.

2.

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

3.

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

c.

The ASMEIANS RA-Sa-2009 PRA Standard defines PRA upgrade as the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences. Section 1-5 of Part 1 of ASMEIANS RA-Sa-2009 PRA Standard states that upgrades of a PRA shall receive a peer review in accordance with the requirements specified in the peer review section of each respective part of this Standard. Provide an evaluation of the model changes associated with incorporating mitigating strategies, which demonstrates that none of the following criteria is satisfied: (1) use of new methodology, (2) change in scope that impacts the significant accident sequences or the significant accident progression sequences, and (3) change in capability that impacts the significant accident sequences or the significant accident progression sequences.

d.

Section 2. 3.4 of NE/ 06-09, Revision 0-A, states that PRA modeling uncertainties shall be considered in application of the PRA base model results to the RIGT 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. NEI 06-09, Revision 0-A, further states that sensitivity studies should be performed on the base model prior to initial implementation of the RIGT program on uncertainties which could potentially impact the results of a /UGT calculation. NRG staff notes that the impact of model uncertainty could vary based on the proposed RIGTs. NE/ 06-09, Revision 0-A, also states that the insights from the sensitivity studies should be used to develop appropriate compensatory RMAs including highlighting risk significant operator actions, confining

==

to AEP-NRC-2024-03 Page 14 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 RICT proposed in this application for which FLEX equipment and./or 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.

Describe 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 RICT where the FLEX equipment and/or operator actions are identified as key assumptions and sources of uncertainty.

Section 3.2.8.2 (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) and Fire (FPRA) PRA Models. No credit is taken for FLEX strategies within the CNP SPRA model.

The 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 SGs with either the Turbine-Driven Auxiliary Feedwater Pump (TDAFP) 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 to AEP-NRC-2024-03 Page 15 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 or until procedural electric power recovery actions have been successfully implemented [17].

Specifics into how each strategy is adapted into the PRA models are available in the FLEX System Notebook [18].

A focused Scope peer review was conducted at CNP to review the implementation of FLEX into the PRA model [19]. 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 ASNE/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 [19]. This memo identified several areas of improvement to bring the FLEX HRA into alignment with the memo requirements. FLEX Human Event Probabilities updated as a result of this review were included in the 2023 Internal Events and Fire PRA models of record. A peer review was performed on the incorporation of FLEX modeling into DC Cook's PRAs. HRA was included as an element of this review to confirm that the HRA modeling was consistent with the PRA standard.

3.3 PRA Review Process Results (10 CFR 50.69(B)(2)(lii))

The PRA models described in Section 3.2 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 [7], consistent with NRC RIS 2007-06 [20].

Facts and Observations (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) [11 ], as accepted by NRC in the letter dated May 3, 2017 [12]. 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 05-04 [23] process, the PRA Standard [1 O] and Regulatory Guide 1.200, Revision 2 (7]. This Peer Review (PWROG-15076-P [21])

was a full-scope peer review of the technical elements of the internal events and internal flooding, at-power PRA.

to AEP-NRC-2024-03 Page 16 The CNP FPIE PRA model underwent a focused-scope peer review in September 2017 [28] and subsequent closure review [29] in March 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 [27] and November 2020 [19] focusing on the treatment of pre-initiator HRA [28] and the implementation of FLEX, respectively.

The CNP FPIE PRA model underwent an F&O Closure review in November 2021 using the 05-04

[11] process, the PRA Standard [10], Regulatory Guide 1.200, Revision 2 [7] and Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, "Close-out of Facts and Observations" (F&Os) [11 ]). This Appendix X Closure Review [32] 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 3 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 [24] was performed in July 2010 using the NEI 07-12 [33]

the ASME/ANS PRA Standard, ASME/ANS RA-Sa-2009 [10], and Regulatory Guide 1.200, Revision 2 [7]. 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. The scope also included a review of the CNP PRA Configuration Control Program in accordance with Section 1.5 of the ASME/ANS Combined PRA Standard [10].

The CNP FPRA underwent additional focused-scope peer reviews in November 2015 [36], July 2017

[37], 2022 [38], 2023 [39], involving Level 2 PRA [LERF] (00403140002-1515), 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 [40]. 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 3 of this enclosure.

Seismic PRA (SPRA) model The seismic PRA model was reviewed in November 2018 [41]. This peer review was conducted against the technical elements in PRA Standard Code Case for Part 5 [43]. 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 [41]:

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 [7]. 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:

to AEP-NRC-2024-03 Page 17 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. In 2017, the JCNRM released the Code Case as an acceptable alternative to Part 5 of Addendum B. The NRC released a letter in March 2018 indicating the following:

The NRG staff has determined that the alternative approach described in the Code Case is consistent with Part 5 of the ASMEIANS PRA standard which the staff has reviewed and endorsed in Regulatory Guide 1.200.

The NRC 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 NRC on March 7, 2018.

The findings from the Seismic PRA peer review have been addressed in the Seismic PRA model. In August 2020 (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 3 of this enclosure.

This demonstrates that the PRA models are of sufficient quality and level of detail to support the categorization process and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC as required 10 CFR 50.69(1)(1)(1).

3.4 Risk Evaluations (10 CFR 50.69(b)(2)(iv))

The CNP 10 CFR 50.69 categorization process will implement the guidance in NEI 00-04. The overall risk evaluation process described in the NEI guidance addresses both known degradation mechanisms and common cause interactions and meets the requirements of 10 CFR 50.69(b)(2)(iv).

Sensitivity studies described in NEI 00-04 Section 8 [1] will be used to confirm that the categorization process results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF). The failure rates for equipment and initiating event frequencies used in the PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, and human errors). Subsequent performance monitoring and PRA updates required by the rule will continue to capture this data and provide timely insights into the need to account for any important new degradation mechanisms.

3.5 Feedback and_AdjustmentProcess If significant changes to the plant risk profile are identified, or if it is identified that a RISC-3 or RISC-4 SSC can (or actually did) prevent a safety significant function from being satisfied, a timely evaluation and review will be performed prior to the normally scheduled periodic review. Otherwise, the assessment of potential equipment performance changes and new technical information will be performed during the normally scheduled periodic review cycle.

Scheduled periodic reviews will be completed at least once every two refueling cycles and will evaluate new insights resulting from available risk information changes (i.e., PRA model or other analysis used in the categorization), design changes, operational changes, and SSC performance. If it is determined that these changes have affected the risk information or other elements of the categorization process such that the categorization results are more than minimally affected, then the risk information and the categorization process will be updated. This review will include:

to AEP-NRC-2024-03 Page 18 A review of plant modifications since the last review that could impact the SSC categorization A review of plant specific operating experience that could impact the SSC categorization, A review of the impact of the updated risk information on the categorization process results A review of the importance measures used for screening in the categorization process.

An update of the risk sensitivity study performed for the categorization In addition to the normally scheduled periodic reviews, if a PRA model or other risk information is upgraded, a review of the SSC categorization will be performed.

4

4.1 REGULATORY EVALUATION

Applicable Regulatory Requirements/Criteria The following NRC requirements and guidance documents are applicable to the proposed Change:

The regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors."

NRC Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1,

May 2006 [2].

Regulatory Guide 1.17 4, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, April 2015.

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

[7].

The proposed change is consistent with the applicable regulations and regulatory guidance.

4.2 No Significant Hazards Consideration Analysis Indiana Michigan Power Company (l&M) proposes to modify the licensing basis to allow for the voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR),

Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

to AEP-NRC-2024-03 Page 19 l&M has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment,"

as discussed below:

1.

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of Structures, Systems and Components (SSCs) subject to Nuclear Regulatory Commission (NRC) special treatment requirements and to implement alternative treatments per the regulations.

The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC.

Under the proposed change, no additional plant equipment will be installed. 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 will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

to AEP-NRC-2024-03 Page 20 Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, I&M concludes that the proposed change presents 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.

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

5 ENVIRONMENTAL CONSIDERATION A review 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 a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22( c )(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1.

NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, Nuclear Energy Institute, July 2005.

2.

NRC Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1, May 2006.

3.

NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management,"

December 1991.

4.

NRC letter to Southern Nuclear Operating Company, Inc., "Vogtle Electric Generating Plant Units 1 and 2-Issuance of Amendments Re: Use of 10 CFR 50.69 (TAC Nos. ME9472 and ME94473)," dated December 17, 2014 (ADAMS Accession No. ML14237A034).

5.

NRC letter to Entergy Operations, Inc., "Arkansas Nuclear One, Unit 2-Approval of Request for Alternative AN02-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems (TAC No. MD5250)," dated April 22, 2009 (ADAMS Accession No. ML090930246).

6.

Generic Letter 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities -10 CFR 50.54(f), Supplement 4," US NRC, June 1991.

7.

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

8.

NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision-making, Revision 1, March 2017.

to AEP-NRC-2024-03 Page 21

9.

EPRI TR-1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, December 2008.

10.

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.

11.

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).

12.

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).

13.

NRC Letter to Mr. Oliver Martinez, "U.S. Nuclear Regulatory Commission (NRC) Comments on

'Addenda to a Current ANS: ASME RA-SB - 20XX, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment For Nuclear Power Plant Applications," dated July 6, 2011 (ADAMS Accession No. ML111720076).

14.

PWROG-20037-NP, PRA Upgrade/Maintenance and Newly Developed Method Examples PA-RMSC-1647, Maioli, Revision 0-B, March 2022.

15.

NUREG/CR 6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, EPRI 1011989, September 2005.

16.

1-(2)-OHP-4023-ECA-0-0, Loss of All AC Power, Revision 46, November 2022.

17.

PRA-NB-SY-FLEX, FLEX System Notebook, Revision 1, April 2023.

18.

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

19.

NRC Memo Dated 5/6/2022, UPDATED ASSESSMENT OF INDUSTRY GUIDANCE FOR CREDITING MITIGATING STRATEGIES IN PROBABILISTIC RISK ASSESSMENTS.

20.

NRC RIS 2007-06, NRC REGULATORY ISSUE

SUMMARY

2007-06 REGULATORY GUIDE 1.200 IMPLEMENTATION, March 2007.

21.

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

22.

NEI 05-0, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, Revision 2, November 2008.

23.

PRA-NB-UNC, Internal Events Uncertainty Notebook, Revision 3, 10/5/2023

24.

PRA-FIRE-UNC, Fire PRA Uncertainty Notebook, Revision 1, March 2023.

25.

PRA-NB-SPRA-QU, Seismic PRA Quantification Notebook, Revision 1, 10/5/2023

26.

PRA-EXT-HAZ-SCRN, External Hazards Assessment for D.C. Cook Nuclear Plant, Revision 0, March 2022.

27.

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

to AEP-NRC-2024-03 Page 22 October 2016.

28.

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.

29.

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

30.

01 V015-RPT-01, Revision 2, "Donald C. Cook Containment Failure Assessment for Loss of Hydrogen lgniters", March 2018.

31.

1V042-RPT-01, Revision 0, "D.C. Cook PRA Finding Level Fact and Observation Independent Assessment", March 2022.

32.

LTR-RAM-I-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 Probabilistic Risk Assessment," July 2010.

33.

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

34.

GT 2010-13232, "Actions to Support Transition to NFPA 805," Assignment 31, "Conduct Focused Peer Review FPRA LERF Values," Initiated March 2014.

35.

AR 2015-13810, "Quality of Resolution of PRA NFPA 805 License Condition," Initiated October 2015.

36.

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

37.

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

38.

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.

39.

P3823-0001-001, Revision 0 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.

40.

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

41.

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

42.

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

43.

PRA Standard Code Case for Part 5 ASME/ANS RA-S CASE 1, Case for ASME/ANS RA-Sb-to AEP-NRC-2024-03 Page 23 2013 x Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications.

44.

NRC Letter to Mr. Joel P. Gebbie, DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 -

ISSUANCE OF AMENDMENTS RE: ADOPTION OF TSTF-425-A, REVISION 3, "RELOCATE SURVEILLANCE FREQUENCIES TO LICENSEE CONTROL-RISK INFORMED TECHNICAL SPECIFICATION TASK FORCE (RITSTF) INITIATIVE 5B" (CAC NOS. MF7114 AND MF7115)," dated March 31, 2017 (ADAMS Accession No. ML17045A150).

45.

NRC Letter to Mr. Joel P. Gebbie, "DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2-STAFF REVIEW OF SEISMIC PROBABILISTIC RISK ASSESSMENT ASSOCIATED WITH REEVALUATED SEISMIC HAZARD IMPLEMENTATION OF THE NEAR-TERM TASK FORCE RECOMMENDATION 2.1: SEISMIC (EPID NO. L-2019-JLD-017)" dated September 23, 2020 (ADAMS Accession No. ML20232A894).

46.

EPRI TR-1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, December 2012.

of Enclosure 2: List of Categorization Prerequisites l&M will establish procedure(s) prior to the use of the categorization process on a plant system.

The procedure(s) will contain the elements/steps listed below.

Integrated Decision-Making Panel (IDP) member qualification requirements.

Qualitative assessment of system functions. System functions are qualitatively categorized as preliminary High Safety Significant (HSS) or Low Safety Significant (LSS) based on the seven criteria in Section 9 of NEI 00-04 (see Section 3.1 of this enclosure for this license amendment request). Any component supporting an HSS function is categorized as preliminary HSS. Components supporting, an LSS function are categorized as preliminary LSS.

Component safety significance assessment. Safety significance of active components is assessed through a combination of Probabilistic Risk Assessment (PRA) and non-PRA methods, covering all hazards. Safety significance of passive components is assessed using a methodology for passive components.

Assessment of defense-in-depth (DID) and safety margin. Safety-related components that are categorized as preliminary LSS are evaluated for their role in providing DID and safety margin and, if appropriate, upgraded to HSS.

Review by the IDP. The categorization results are presented to the IDP for review and approval. The IDP reviews the categorization results and makes the final determination on the safety significance of system functions and components.

Risk sensitivity study. For PRA-modeled components, an overall bounding risk sensitivity study, which conservatively assumes increased component failure rates for LSS components, is used to confirm that the population of preliminary LSS components with increased failure rates results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF) and meets the acceptance guidelines of Regulatory Guide 1.17 4.

Periodic reviews are performed to ensure continued categorization validity and acceptable performance for those Structures, Systems, and Components that have been categorized.

Documentation requirements per Section 3.1.1 of this enclosure.

of Enclosure 2: Description of Probabilistic Risk Assessment (PRA)

Models Used in Categorization Unit Model Baseline CDF Baseline LERF Comments Full Power Internal Events (FPIE) PRA Model 1 2 Model 2.41£-5/2.40E-5 1.56E-6/1.51E-6 2023 FPIE CNP FPIE-2023 Model of Record (MOR)

Fire (FPRA) PRA Model 1 2 Model 3.64£-5/4.16E-5 3.04£-6/2.44E-6 2023 Fire PRA CNP FPRA 2023-R0 Model of Record (MOR)

Seismic (SPRA) PRA Model 1 2 Model 2.10E-5/ 2.10E-5 5.31E-6/5.72E-6 2023 Seismic CNP SPRA 2023-R0 PRA Model of Record (MOR) of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS Full Power Internal Events F&O Status SRs F&O Description Disposition The Large Early Release Frequency (LERF) analysis Pressure and temperature induced uses NUREG/CR-6595 type evaluations for some SGTR events are modeled as LE-D5 portions of the LERF evaluation based on conservative progressing directly to LERF. This 2-19 LE-C1-C5 assessments. These portions of the LERF analysis do ensures an over-estimation of the (2015 LE-C9-C13 not meet Capability Category II of the standard. One significance of this assumption.

Full Open LE-E2 example of these conservatisms is assuming Steam Scope)

Generator Tube Rupture (SGTR) is a containment bypass event without considering success of MET: CC-I secondary side isolation. However, these conservatisms generally do not impact the ability to perform PSA applications using LERF.

Most of the notebooks indicate that interviews with There is not expected to be a knowledgeable plant personnel were conducted to significant deviation between what is confirm that the systems analysis adequately reflected modeled in the PRA and actual the as-built, as-operated plant and that plant-specific plant condition such that there 4-4 data was appropriately collected where required.

would be a substantive impact on (2015 SY-A2: MET However, a record of such interviews was not provided numerical model results. Some Full PR SY-A4: CC-I as part of the notebook documentation.

walkdowns and interviews have Scope)

SY-C2: MET Plant walkdowns are discussed in a generic walkdown been performed and did not identify document created in June 1991. There is no record of any necessary modeling changes, recent system walkdowns conducted with the same outcome is expected for knowledgeable plant personnel. Even if the system those systems that still need configuration has not changed during that time, there walkdowns and interviews should be a confirmatory walkdown to document that.

performed.

Based on a discussion with Cook Nuclear Plant (CNP)

Recent outage durations have been Probabilistic Risk Assessment (PRA) Engineer, the long due to work related to CNP model conservatively models the opposite unit's replacement of baffle bolts in the 6-19 outage unavailability by assuming 45 days outages reactor vessel, and therefore a (2015 Open DA-C13 with train unavailability equal to an equivalent portion value informed by recent Operating Full of the outage (e.g., a 2 train system would assume one Experience would result in an Scope) train unavailability is 22.5 days).

overestimation of the risk associated with outage windows. The current estimate of 45-day outages bounds recent outage experience and is of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS F&O Status SRs F&O Description Disposition therefore either not expected to impact model results, or result in an overestimation.

All containment failures caused by hydrogen The current implementation of the combustion are assumed to contribute to LERF in modeling is conservative and will report 01V015-RPT-01. There is no discussion in the result in an overestimation in the report that provides a basis for this approach to risk significance of sequences that scenario assignment based on containment failure could potentially be screened out location considerations. PRA-NB-LER, Revision O also based on containment failure does not relate the assignment of LERF scenarios to location. However, an improvement containment failure location. The latter document of this modeling would not result in a 2-4 identifies the most likely containment failure location significant improvement in the (2017 from the containment capacity report (Stevenson overall realism of the model results.

Open LE-D3: CC-I report) and includes a historical discussion that Hydrogen provides an argument against assigning scenarios that FSPR) 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.

of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS Fire PRA F&O Status SRs F&O Description Disposition Observation: Parametric uncertainties of applied hot short The impact of probabilities have not been incorporated into the model.

this Finding is limited to small CFA2-Finding & Observation (F&O) Closure Notes:

portion of the 01 Partially CF-The CF and UNC notebooks were reviewed and confirmed that numerical parametric uncertainty (2010 A2 uncertainties are documented. However, several inconsistencies were identified analysis, and Full Open Met between the documentation and the values used in the model. Given that the majority thus does not Scope) of CF uncertainties have been correctly applied CF-A2 is now considered Met.

impact the overall technical quality of the Fire PRA.

FQ-D1-FQ-Some of the Internal Events LE SRs were classified as CC-I due to conservative See disposition 02 D1 modeling. Therefore, the Fire LERF should also be limited to CC-I as appropriate for for FPIE F&O 2-(2022 Open PRM-applications. Revise the internal events PRA to meet CC-II for relevant SRs and

19.

Focused B2 implement changes to FPRA.

Scope)

Section 4.3 of the PP report says no spatial separation was credited as a partition Review of the element. This statement was the basis for the CC-I assessment in the original peer PRA review. There is a disconnect between Section 4.3 and Section 3.1 and 3.2 that needs implementation to be rectified.

of fire modeling concluded that PP-B3-Additionally, Section 3.1 discussed the subdivision of the yard into sub-compartments the issue 01 PP-based on spatial separation, but these sub-compartments do not become separate described is a (2022 Open B3 listed fire zones and are not separated in PRA-NB-FIRE-IGN. No explanation is given documentation Focused IGN-for this in R1900-0041-0001. IF the intent is to separate them for fire modeling only ( as disconnect, and Scope)

A7 suggested by Table 6-1 of PRA-NB-FIRE-IGN), this should be stated with the PP therefore its analysis and reiterated in the PAU Table in Attachment 1 with a note for clarity.

resolution will Revise language in Section 4.3 regarding the use of spatial separation. Clarify not impact treatment of the yard sub-compartments by adding additional discussion to Section 3.1 numerical model with a clarifying note in Attachment 1. Alternatively, carry the sub-compartments results or risk forward as separate "fire zones" into Attachment 1 and the IGN consistent with the sub-insights.

divisions of the other fire compartments.

of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS Seismic PRA F&O Status SRs F&O Description Disposition

1) -Only a single method was considered to Seismic PRA (SPRA) results are not evaluate the liquefaction triggering potential, expected to be impacted. l&M (2014) liquefaction susceptibility, and post liquefaction initially performed a liquefaction volumetric strains. However, in F&O 20-7, Item 2, triggering (using Youd et al., 2001) more than one method was requested to conduct and settlement (using Tokimatsu and the liquefaction hazard evaluation as "the choice of Seed, 1987) analyses using the RLE any single method does not address the epistemic and obtained comparable results. l&M uncertainty in the field (which is the underlying considers that Figure 6-7 shows motivation of a recent National Academy study and liquefaction at some boreholes for 1E-report)".

6 motions, but shows no lateral continuity of the liquefiable boreholes.

1-1 2)- Lateral spreading hazard at the site does not Based on this information, l&M has (2018 PR SHA-I1: MET address the evaluation of this potential hazard.

concluded that the site can be Full SHA-I2: MET Lateral spreading can occur in slope gradients as screened out for site-wide lateral Scope) flat as 0.5 percent(%) (without a free face) spreading.

(See NA report). Additionally, Figures 6-7 and 6-9 shows that there is continuous layer of potentially liquefiable soils (in direction towards the lake) on borings B120, B124, B133, B142, and B141 between elevations of about 560 and 555 ft.

Therefore, the potential of lateral spreading and/or flow slides at the site should be evaluated.

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

1)- Include additional justification on why V/H ratios SPRA results are not expected to be should be used instead of vertical GMPEs in report impacted as this F&O has been 20-3 DC COOK-PR-02, Section 7.1 (e.g., inconsistency technically resolved.

(2018 DO SHA-J2: MET of controlling earthquakes between horizontal and Full vertical spectra if vertical GMPEs were used).

Scope)

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

of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS F&O Status SRs F&O Description Disposition While the cracking assessment for the Containment SPRA results are not expected to be Building (CB) and TB/SH has been resolved the impacted. The studies performed in cracking assessment for AB has not been fully 15C4313-RPT-003 "Summary of resolved. Several changes were made to the AB Building Response Analysis for the structural model in 15C4313-CAL-010, "Response Cook Nuclear Plant (CNP) Unit 1 &

Analysis of Auxiliary Building," Revision 2, in Unit 2 SPRA," Attachment E, show response to other SFR F&Os. The updated AB that while there may be some model was used in the cracking assessment with cracking, it is not widespread at the un-cracked section properties. The SPRA team RLE-level. Additionally, l&M performed cracking assessment at earthquake engineering judgement is that with the levels corresponding to 0.5*RLE and 1.0*RLE.

studies performed, cracking in the 2-1 Figures 1 through 8 in Attachment E present the structure will decrease the stiffness (2019 0

SFR-B3: MET shear stress contour plots on isometric views of the and increase the damping. These two FSPR)

AB model showing the exterior walls. The stress effects tend to affect the structural contour plots only suggest that the building is overly response in opposite ways. Finally, stressed in certain regions. For a complex structure many significant contributors have low such as the AB, this is not sufficient to conclude fragilities for which consideration of a that cracking will or will not occur in the building cracked model would be non-especially under dynamic loads. The SPRA conservative.

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.

Perform a sensitivity study to address items SPRA results are not expected to be determined to be risk significant based on F-V impacted. l&M position is that 22-2 importance greater-than or equal-to 0.005.

additional studies for risk items not (2018 DO SFR-E3: CCII considered by risk significant as Full

( defined in the SPRA quantification Scope) notebook) will not change risk insights.

of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS Perform a sensitivity study to address items SPRA results are not expected to be determined to be risk significant based on F-V impacted. l&M position is that 22-5 importance.

sensitivity studies documented in the (2018 SPRA quantification notebook Full DO SFR-E2: CCII envelope any small fragility changes Scope) that may be discovered by the additional sensitivity recommended here and will not change risk insights.

SPRA team has used the ASCE 4-16, SPRA results are not expected to be Section 3.7.2 dynamic coupling criteria for single-impacted. The l&M position is that the point attachment to show that the current CB simplified method used to modeling approach and response are realistic.

demonstrate that the CB modelling While the modeling approach use probably does simplifications have no impact on the 28-2 not have an effect on overall response of the response in 15C4313-RPT-003 (2018 PR SFR-B3: MET structure but that conclusion has not been Attachment B is sufficient to address Full demonstrated adequately.

the F&O. The close-out team Scope requested more detailed studies be performed to close the F&O, however the team stated that they believe the conclusion will most likely not change as a result.

of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS Appropriate damping was used for cracked and un-SPRA results are not expected to be cracked building sections in the building response impacted. The position of l&M is that sensitivity studies following the current industry and the conclusion provided in 15C4313-standard ASCE 4-16. The sensitivity practice RPT-003, Attachment E, is sufficient studies are documented in Attachments B and F of to justify the use of un-cracked 15C4313-RPT-003, respectively, for Containment damping for the AB model. See F&O 28-4 Building and Turbine Building/Screen House.

2-1 for further information.

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

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 SPRA results are not expected to be that due to the way that fragilities were developed, impacted. The sensitivity studies including the application of uncertainty with respect performed in 15C4313-RPT-003 to frequency was sufficient to allow no variation in between un-cracked and cracked structural properties. The variation in frequency is properties show that structural 28-11 intended to reflect uncertainty in the value of the variability has a minor impact on (2018 calculated frequency. The variation in structural response compared to the soil Full 0

SFR-B4: MET properties is intended to reflect uncertainty in those property variability. l&M will review the Scope) properties. Both effects must be considered when small number of impacted risk-developing fragilities.

significant components on a case-by-case basis, adjusting the FROI by an additional +/- 15% to ensure structural variability is captured in the fragility calculations.

of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS The gap in PSD as described in the F&O should be SPRA results are not expected to be addressed per latest fragility guidance document. If impacted. l&M position is that there it is confirmed that there is a gap in PSD at FRO! of are not significant gaps in energy near structure, then it is recommended to perform a frequencies that are important to risk-sensitivity study to assess the impact of the gap in significant fragilities. The PSDs as energy. The SPRA development team can perform presented were developed using a 28-13 this by comparing the PSD functions of the five-logarithmic frequency interpolation (2018 time histories that were generated by resolution of which tends to emphasize magnitude Full PR SFR-B4: MET F&O 28-09 to the PSD function of the artificial time variation at low frequencies. A review Scope) history, or the development team can integrate the of the non-interpolated PSDs and PSD function to show that a smooth curve is PSDs developed using a linear generated.

frequency interpolation supports the determination that the gaps identified in the F&O are not significant.

The documentation needs to be further SPRA results are not expected to be updated to provide a basis for not considering impacted, as this F&O has been SSSI effects. Subsequent to the closure review, technically resolved. Additional additional documentation was added to the quantitative justification added 28-19 calculations. However, the closure review team to Section 4.4 of 15C4313-(2018 does not consider this additional documentation to RPT-003 is adequate in showing that Full SFR-F2: MET be sufficient to address the concern originally SSSI effects do not control over RLE Scope)

DO identified.

demand for applicable components.

Also note that components associated in this documentation item are not risk significant.

of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS Resolved with Open Documentation In the SPRA SPRA results are not expected to be Model Quantification Notebook,Section 8.2.2, impacted as this F&O has been Revision 1, the cutset review included a statement technically resolved.

on non-significant cutsets - samples are covered 25-7 by the examination of G1 and G2 bins. G1 and G2 (2018 bins contain relatively fewer seismic-induced DO SPR-E3: CCII failures and the cutsets have features more like the Full internal events PRA. A recommendation is made to Scope) 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 No impact to SPRA results -

were met at CC-I only; therefore, this SR is met for The LERF modeling is built upon the CC-I only.

internal events LERF model and is essentially unchanged.

25-9 The SPRA LERF model includes (2018 seismic-specific aspects such as Full 0

SPR-E6: CCI unique containment failure Scope) probabilities. Whereas there are some internal events LE supporting SRs that meet both CC-I and CC-II, the majority of the SRs are met at CC-I. Therefore, this SR is considered to be met at CC-I only.

of Enclosure 2: EXTERNAL HAZARDS SCREENING A calculation PRA-EXT-HAZ-SCRN was developed for external Hazards screening. See Table 4-4 below.

Table 4-4: Hazard Dispositions 50.69 Hazard Screening Disposition for 10 CFR 50.69 Criterion Aircraft Impacts PS2 Per Updated Final Safety Analysis Report (UFSAR) Section 2.1.4 [6.3], there are two airports within a 15-mile vicinity of the DC Cook Nuclear Plant (CNP): Southwest Michigan Regional Airport located approximately 12 miles North East (NE) of the plant on the NE edge of Benton Harbor and Andrews University Airport located approximately 10 miles East of the plant near Berrien Springs.

For airports beyond this 15-mile radius, the orientation of runways and normal flight patterns are not in the direction of the plant or the normal glidepath heights are not within the plant vicinity so that aircraft utilizing the facilities of these airports would not normally fly over the plant site.

Southwest Michigan Regional Airport, data from 2017 shows approximately 18,350 operations (take-offs or landings) [6.10]. Due to the north-easterly location of the airport and the orientation of the runways, normal glide paths would not approach the vicinity of the plant.

Andrews University Airport/Airpark has one runway. Due to the easterly location of the airport and the orientation of the runway, normal glide paths would not approach the vicinity of the plant. For 2016, which is the most current data for this airport, there were approximately 7,300 operations

[6.10].

The annual movements are below the critical number at which a probability analysis for aircraft accidents would be required according to Regulatory Guide 1. 70 [6.11]. Therefore, the probability of aircraft crashing into the site is considered to be remote, and airplane crashes need not be considered for design basis events.

Additionally, the Individual Plant Examination of External Events (IPEEE) [6.12] checked against the requirements of the US Nuclear Regulatory Commission (NRC) Standard Review Plan. The first requirement is met and was previously explained. The second requirement is the plant is at least five (5) statute miles from the edge of military training routes, including low level training routes. According to the IPEEE, this requirement is satisfied as the nearest military training route, VR 1640, is approximately 50 miles from the site. The third requirement is the plant is at least two (2) statute miles beyond the nearest edge of a federal airway, holding pattern, or approach pattern.

This requirement is not met as a low altitude flight path V526 is within the two (2)-mile limit and of Enclosure 2: EXTERNAL HAZARDS SCREENING 50.69 Hazard Screening Disposition for 1 0 CFR 50.69 Criterion required further analysis. The high-altitude flight path is J584 and is at least five (5) miles from the site at its closest approach and does not require further analysis. The IPEEE calculated the probability per year of an aircraft crashing into the plant for flight pattern V526 for 1990 (with data from 1988) and it was less than the limit of 1E-7 and thus precluded further analysis. The IPEEE concluded the contribution to plant risk is insignificant relative to other initiating events.

Based on this review, the aircraft impact hazard can be considered to be negligible.

Avalanche C3 The location of CNP precludes the possibility of an avalanche.

Based on this review, the avalanche impact hazard can be considered to be negligible.

Biological Events c5 The only biological event that may credibly affect CNP is zebra mussel blockage of circulating water system intakes. According to the IPEEE [612], effects of mussel buildup are continuously monitored, and the plant would have sufficient warning if conditions warranted shutdown.

Additionally, according to UFSAR Section 2.6.4 (6.31. biocides supplemented by mechanical cleaning and design changes including strainers, filters, screens, and chemical delivery systems, work to protect plant systems. A zebra mussel monitoring program utilizing side-stream and artificial substrate monitors, along with diver and heat exchanger inspections, is used to evaluate the effectiveness of chemical and physical control measures.

Based on this review, the Biological Event impact hazard can be considered to be negligible.

Coastal Erosion c5 Per UFSAR Section 2.3.3 [6.3], shoreline erosion is not evident at the site.

The long-time periods required to produce sufficient coastal erosion to endanger the plant would provide sufficient time for plant shutdown, and, therefore, no further analysis was performed.

Based on this review, the Coastal Erosion impact hazard can be considered to be negligible.

Drought c5 Drought is a slowly developing hazard allowing time for orderly plant reductions, including shutdowns. According to the IPEEE [6.12], the depth of the intake cribs at the Cook site (about 10 feet below the record low lake level) precludes further analysis.

Based on this review, the drought impact hazard can be considered to be negligible.

of Enclosure 2: EXTERNAL HAZARDS SCREENING External Flooding 1

The evaluation of the impact of the external flooding hazard at the site was updated as a result of the NRC's post-Fukushima 50.54(f) Request for Information. The station's FHRR was submitted to NRC for review on March 6, 2015 [6.6). The results indicate all flood-causing mechanisms, except LIP, are bounded by the Current Licensing Basis and do not pose a challenge to the plant.

Modifications to prevent critical plant equipment from being adversely impacted by flood water intrusion from a Beyond Design Basis LIP event were performed. Specific details are discussed in Section 4.2.1 of PRA-EXT-HAZ-SCRN.

Consistent with Figure 5-6 in NEI 00-04 [6.9], an evaluation was performed for the screening of the external flooding mechanism LIP to determine if there are any components that participate in screened scenarios and whose failure would result in an unscreened scenario. There are several components whose failure to be in their normal position (e.g., doors in the closed position) or function appropriately (e.g., roof scuppers shedding water to not allow water to pool) during a LIP event to limit the ingress of water to the Auxiliary and Turbine Buildings would result in an unscreened scenario. These components should be categorized as HSS in accordance with NRC approved guidance.

With credit taken for these components during a postulated LIP, external flooding mechanisms are screened as not impacting 10 CFR 50.69 categorization.

The components are identified in Table 4-3 of PRA-EXT-HAZ-SCRN.

Based on this analysis of the external flooding hazard for CNP, the hazard has negligible impact.

Extreme winds C1,PS4 Based on information in the UFSAR [6.3], the plant design for wind pressure and the low frequency and Tornadoes of design tornadoes, a demonstrably conservative estimate of Core Damage Frequency (CDF) associated with high wind hazard (other than wind generated missiles) is much less than 1E-6/yr.

Details are provided in Section 4.1 of PRA-EXT-HAZ-SCRN.

Section 1.4.7 of the UFSAR [6.3] documents that a limited number of SSCs located near openings/penetrations in Seismic Category I structures or located outside of such structures have been evaluated and do not require additional physical tornado missile protection features. These Safety System Components have been evaluated with respect to the overall risk resulting from tornado-generated missiles upon potential off-site dose consequences exceeding the guidelines of Regulatory Guide 1.183 and 10 CFR 50.67; the acceptance criterion is being less than 1.0E-06 per reactor-year. TORMIS determines the probability of tornado generated missiles striking targets, of Enclosure 2: EXTERNAL HAZARDS SCREENING which may include, but are not limited to, walls and roofs of buildings, penetrations of Seismic Category I structures, and exposed portions of systems/components. The probability is calculated by simulating many tornado strike events at the site. This results in a calculated probability per unit area of striking any target. Then, the exposed surface area of each component is factored in to determine the probability of striking each item. The TORMIS analysis for CNP is documented in SD-990930-004 Revision 12 [6.4].

The TORMIS analysis for CNP is updated for this assessment with a bounding analysis which incorporates the CNP FPIE PRA MOR [6.2] and updates the target (e.g., component) to basic event mapping.

The results of the bounding analysis, presented in Table 4-2 of PRA-EXT-HAZ-SCRN, indicate this hazard screens because the CDF is below 1.0E-06 per year (PS4). Based on this analysis of the high winds/ tornadoes hazard for CNP, the hazard has negligible impact.

Fog C4 Fog can increase the frequency of occurrence of accidents or events. Fog is implicitly included in data for transportation accidents and, if freezing fog, included in the weather-related loss of offsite power initiating event in the internal events PRA.

Based on this review, the Fog impact hazard can be considered to be negligible.

Forest Fire /

C1 External/Forest fires have the potential to cause a grid-related or switchyard loss of offsite power External Fire event or cause the control room to become uninhabitable. The IPEEE [6.12] screened out these two scenarios from the analysis due to low likelihood and small probabilities of adverse effects.

The loss of offsite power initiator is included in the internal events PRA. The control room habitability scenario is not considered a problem because the resultant smoke will not cause an equipment failure or a reactor trip and because the control room personnel would be notified almost immediately of a major fire by the security patrols. Additionally, control room ventilation would be placed in recirculation (isolation) mode.

Based on this review, the External/Forest Fire impact hazard can be considered to be negligible.

Frost Hail High Summer Temperature High Tide / High Lake Level Hurricane Ice Cover of Enclosure 2: EXTERNAL HAZARDS SCREENING C1 The principal effects of such events would be to cause a loss of offsite power and are addressed in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for CNP.

C1 The principal effects of such events would be to cause a loss of offsite power and are addressed in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for CNP C1 The principal effects of such events would result in elevated lake temperatures, which are monitored by station personnel. Should the ultimate heat sink (UHS) temperature exceed the DC Cook Technical Specification (TS) 3.7.9 temperature limit [6.13], an orderly shutdown would be initiated in accordance with appropriate TS required actions.

Additionally, if the containment temperature (due to inadequate cooling because of high UHS temperature) exceeds the DC Cook TS 3.6.5 limit (6.13], then an orderly shutdown would be initiated.

Based on this review, the High Summer Temperature impact hazard can be considered to be negligible.

C4 According to UFSAR Section 2.6.2.3 [6.31. the plant is flood protected from the maximum (monthly mean) high lake water level; however, a design basis seiche occurring when the lake is at its maximum recorded level will cause flooding in the Turbine Building Screen-house. Safety-related components located in the Turbine Building Screen-house have been evaluated for the condition and flood sensitive components (associated with ESW System) have been protected. These flood sensitive components had their terminations in their respective local terminal boxes reworked to bring them above a postulated seiche flood level, which were performed via EC-46977 [6.14], EC-46978 [6.15], EC-46979 [6.16], and EC-46980 [6.17]. Therefore, protection has been provided for safety-related equipment from flooding, waves, ice storms and other lake-related hazards.

See also "External Flooding."

C3 The location of CNP precludes the possibility of a hurricane.

Based on this review, the Hurricane impact hazard can be considered to be negligible.

C1, C4, C5 The principal effects of such events would be to cause a loss of offsite power and are addressed in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for CNP(C1).

The potential hazard of ice clogging the intake is prevented by the De-icing System. This system opens Unit 1 and 2 motor operated valves 1-WMO-16 and 2-WMO-26, respectively, which are in the 591' elevation of the Screen House and provides circulating water discharge from the of Enclosure 2: EXTERNAL HAZARDS SCREENING condensers to the de-icing tunnel to prevent ice buildup in the forebay [6.18] (C5). Therefore, these components (1-WMO-16 and 2-WMO-26) credited for mitigating ice clogging the intake and which are pertinent to this hazard screening will be treated as HSS if categorized, in accordance with the guidance provided in NEI 00-04 Figure 5-6 [6.9]. The IDP will be informed of the basis for ice clogging the intake (Ice Cover) screening during their reviews of categorization results.

See also "External Flooding" (C4).

Based on this review, the Ice Cover impact hazard can be considered to be small.

Industrial or C3 According to UFSAR Section 2.1.2 [6.3], there are no military installations, missile sites, or Military Facility industrial facilities located beyond the DC Cook Nuclear Plant Site boundaries at which an accident Accident might cause interaction with the plant so as to affect public health and safety.

Based on this review, the Industrial or Military Facility Accident impact hazard can be considered to be negligible.

Internal Fire N/A The CNP Internal Fire PRA includes evaluation of risk from internal fire events.

Internal Flooding N/A The CNP Internal Events PRA includes evaluation of risk from internal flooding events.

Landslide C3 According to the IPEEE [6.12], the topography is such that a landslide is not possible.

Based on this review, the Landslide impact hazard can be considered to be negligible.

Lightning C1 Lightning strikes are not uncommon in nuclear plant experience and can result in losses of offsite power. Loss of offsite power events are incorporated into the CNP internal events model through the incorporation of generic data [6.19].

Based on this review, the Lightning impact hazard can be considered to be negligible.

Low lake or river C4 Included under drought.

water level Low winter C4,C5 According to the IPEEE [6.12], the likelihood of the lake freezing solid to a depth of the intake cribs temperature is insignificantly small. Also, there would be ample warning time for the plant to shutdown with respect to freezing of the heat sink. A procedure is used to increase the forebay temperature when signs of localized freezing are observed; this is covered under the "Ice Cover" hazard (C4).

The procedure used to increase the forebay temperature (and level) with Circulating Water Deicing system in service is 1-OHP-4021-057-002, "Placing In/Removing from Service Circulating Water Deicing System" [6.20]. Other procedures that provide guidance during severe weather, such as snow and ice storms (which occur during low winter temperature), are 12-OHP-4022-001-010, of Enclosure 2: EXTERNAL HAZARDS SCREENING "Severe Weather" [6.21], PMP-5055-001-001, "Winterization/Summerization" [6.22], and PMP-5055-SWM-001, "Severe Weather Guidelines" [6.23).

Based on this review, the Low Winter Temperature impact hazard can be considered to be negligible.

Meteorite/Satellite PS4 The frequency of a meteor or satellite strike is judged to be so low as to make the risk impact from strikes such events insignificant. This hazard also was reviewed as part of the IPEEE submittal [6.12) and screened based on low frequency of occurrence.

Based on this review, the Meteorite or Satellite impact hazard can be considered to be negligible.

Pipeline Accident C3 According to the IPEEE [6.12], DC Cook receives no hazardous materials via pipeline. Additionally, the "Evaluation of Offsite Sources of Toxic Gas" [6.24] was reviewed for accidents from pipelines carrying hazardous waste; none were identified.

Based on this review, the pipeline accident hazard can be considered to be negligible.

Release of PS1 The impact of releases of hazardous materials stored on-site was evaluated in the IPEEE submittal Chemicals from

[6.12]. The IPEEE documents analyses that were performed to calculate the maximum control On-site Storage room concentrations given an accident involving on-site hazardous materials, which included ammonia (ammonium hydroxide) 29% solution, chlorine (gas), hydrazine (35% solution), and sulfuric acid. The analyses considered the consequences of a rupture of the single largest container and the largest container of the most concentrated solution of these chemicals in their respective locations, their dispersion, and subsequent build-up in the control room ventilation system. This analysis indicates the hazardous materials stored within the site boundaries of CNP present no threat to plant safety; however, the analyses remain only applicable for ammonia (ammonium hydroxide) 29% solution, chlorine (gas), and sulfuric acid. Analysis CA-91-04, which was used in the IPEEE to address a hydrazine spill, was superseded by CA-03-01 [627] because it did not address the then most recent setup of hydrazine storage.

CA-03-01 [6.27] reanalyzed the storage locations of hydrazine, including two 55-gallon drums on the east side of the Turbine Building and in the Auxiliary Chemical Feed Gallery and a 200-gallon tote stored in the west central section of the Turbine Building on the 591' elevation. CA-03-01 identified unacceptable conditions and that a postulated hydrazine spill event would be a challenge to control room habitability. Regulatory Guide 1. 78 established screening criterion for evaluation of control room habitability during a postulated hazardous chemical release. An evaluation is not needed when storage of the toxic chemical is within 0.3 miles from the control room if the storage is less than 100 lbs. of the chemical or partial vapor pressure of such chemical is less than 10

River Diversion Sandstorm Seiche C3 C3 C4 of Enclosure 2: EXTERNAL HAZARDS SCREENING mmHg. A new non-toxic chemical, carbohydrazide, was to replace hydrazine for water treatment.

However, since hydrazine is needed to be available for post-trip steam generator chemistry, hydrazine was to be stored in smaller containers (i.e., 55-gallon drums and in locations on the east side of the Turbine Building elevation 591). In addition, the partial pressure of hydrazine at the storage temperature exceeded 10 mm Hg. This new storage did not meet the requirements of Regulatory Guide 1. 78, and therefore, an analysis of control room habitability was needed and documented in CA-03-01 [6.27].

CA-03-01 was applicable to the post conversion conditions that used carbohydrazide for water treatment and that stored hydrazine in 55-gallon drums on the east side of Turbine Building elevation 591'. The postulated spill assumed a non-mechanistic failure of the worst-case container, a 55-gallon drum. The new storage location for the 55-gallon drums yielded favorable results with respect to control room habitability due to no freestanding fans in the vicinity of the drums. In addition, because of floor drains, the postulated spill event would be terminated with no operator actions required. Plant modifications were not required [6.27].

Airborne hydrazine is conservatively released from the nearest point on the Turbine Building to the control room intakes, and no credit is taken for Turbine Building roof ventilation units. The resulting 8-hour time-weighted average control room concentration is 0.79 ppm. This result is below the OSHA acceptance criterion of 1 ppm. Therefore, it is concluded that the control room remains habitable following a postulated accidental release of liquid hydrazine [6.27].

Based on this review, the Release of Chemicals in Onsite Storage impact hazard can be considered to be negligible.

The location of DC Cook along Lake Michigan precludes the possibility of a river diversion.

Based on this review, the River Diversion impact hazard can be considered to be negligible.

According to the IPEEE [6.12]. a sandstorm hazard is not relevant for this region.

Based on this review, the Sandstorm hazard can be considered to be negligible.

According to UFSAR Section 2.6.2.3 [6.3], the plant is flood protected from the maximum (monthly mean) high lake water level; however, a design basis seiche occurring when the lake is at its maximum recorded level will cause flooding in the Turbine Building Screen-house. Safety-related components located in the Turbine Building Screen-house have been evaluated for the condition and flood sensitive components (associated with ESW System) have been protected. These flood sensitive components had their terminations in their respective local terminal boxes reworked to bring them above a postulated seiche flood level, which were performed via EC-46977 [6.14], EC-of Enclosure 2: EXTERNAL HAZARDS SCREENING 46978 [6.15]. EC-46979 [6.16), and EC-46980 [6.17). Therefore, protection has been provided for safety-related equipment from flooding, waves, ice storms, and other lake related hazards.

See also "External Flooding."

Seismic Activity N/A The DC Cook Seismic PRA includes evaluation of risk from seismic events.

Snow C4,C5 This hazard is slow to develop (C5) and can be identified via monitoring and managed via normal plant processes. Potential flooding impacts covered under external flooding (C4).

Based on this review, the Snow impact hazard can be considered to be negligible.

Soil Shrink-Swell C1 According to the IPEEE [6.12], the site-suitability evaluation and site development for the plant are designed to preclude the effects of this hazard.

Based on this review, the Soil Shrink-Swell impact hazard can be considered to be negligible.

Storm Surge C4 The location of DC Cook along Lake Michigan precludes the possibility of a sea level driven storm surge. Potential flooding impacts by water levels of Lake Michigan are covered under external flooding.

Based on this review, the Storm Surge impact hazard can be considered to be negligible. See also "External Flooding."

Toxic Gas C4,PS1 According to Reference 6.24,an analysis was performed for cases that involved 34 mobile truck sources, 16 rail car sources and 2 stationary sources at Reliable Disposal. The analyses included both burst and leak cases, except for the one stationary source that was the below ground diesel fuel tank. Scenarios were analyzed for the different cases/chemicals. Each chemical that burst or leaked was analyzed along with the amount available for release and the release location. The evaluation showed whether the release scenario resulted in the control room concentration remaining below the toxicity limit for the chemical or there being at least 2 minutes in the time to act calculation. All release scenarios do not challenge the control room habitability with the exception of three chemicals that require further scrutiny to assess control room habitability. These are the acrolein, bromine, and hydrogen fluoride mobile truck burst scenarios. For the acrolein burst, 22 of the 672 parametric cases run resulted in a time to act of less than 2 minutes, with the minimum time to act as 1.5 minutes. Each of these 22 cases occurred with a combination of wind speed and atmospheric stability class that are usually incongruous (i.e., high wind speed with relatively stable conditions) and only occur less than 0.25% of the time at the plant. For the bromine burst, 16 of the 672 parametric cases run resulted in a time to act of less than 2 minutes, with the minimum time to act as 1. 75 minutes. These 16 cases were for meteorological conditions that occur less than 0.01% of the time at CNP. Similarly, for the hydrogen fluoride burst, only 4 of the 672 cases resulted in a time to act of less than 2 minutes, with the minimum time to act as 1. 75 minutes.

Transportation Accidents Tsunami C1, C4 C3 of Enclosure 2: EXTERNAL HAZARDS SCREENING These 4 hydrogen fluoride cases were for meteorological conditions that occur less than 0.01 % of the time at the plant. Given the low number of cases with time to act less than 2 minutes and the infrequency with which the meteorological conditions used to generate those cases actually occur, it is reasonably surmised that neither acrolein, bromine, nor hydrogen fluoride mobile truck burst scenarios pose a challenge to control room habitability (PS1 ).

See also "Release of Chemicals from On-site Storage," "Pipeline Accident," and "Industrial or Military Facility Accident" (C4).

Based on this review, the Toxic Gas impact hazard can be considered to be negligible.

For shipping impact hazards, according to the IPEEE [6.12], due to the physical location of DC Cook buildings and structures, the only danger to the plant from ships/barges are from those that run-aground that collapse the circulating water intake cribs resulting in flow obstruction of all three circulating water system intake lines. This results in the shutdown of both units; however, essential service water system flow can be maintained to remove heat from the component cooling water system and other essential service water system loads by opening sluice gates (1-WMO-17, 2-WMO-27) between the discharge chambers and forebay, which allows water from the discharge chambers to enter the forebay to supply the essential service water pumps [6.18]. Therefore, no plant damage leading to core damage or radiological release is expected as a result of a shipping accident. Additionally, DC Cook receives no hazardous materials via ship or barge (C1).

Therefore, sluice gates (1-WMO-17, 2-WMO-27) between the discharge chambers and forebay credited for allowing water from the discharge chambers to enter the forebay to supply the essential service water pumps (in case there is flow obstruction of all three circulating water system intake lines) and which are pertinent to this hazard screening will be treated as HSS if categorized, in accordance with the guidance provided in NEI 00-04 Figure 5-6 [6.9]. The IDP will be informed of the basis for shipping impact (transportation accident) screening during their reviews of categorization results.

See also "Aircraft Impacts" (C4).

Based on this review, the Transportation Accident impact hazard can be considered to be negligible.

The location of DC Cook along Lake Michigan precludes the possibility of a tsunami.

Based on this review, the Tsunami impact hazard can be considered to be negligible.

of Enclosure 2: EXTERNAL HAZARDS SCREENING Turbine-generated C1 Per UFSAR Section 14.1.13.1 [6.3], the turbine-generators are safe and reliable and the chance of Missiles a failure during normal operation, which could endanger the reactor and associated Seismic Class I nuclear systems, is extremely small. Additionally, the chance of a turbine running away out of control to destruction is also extremely small. For the Unit 1 and Unit 2 Alstom low-pressure turbines, turbine missile probability analysis indicates the probability of the generation of a turbine missile (including turbine overspeed conditions) is below the NRC limit which would require missile analysis. Therefore, no additional missile analysis is required for the Unit 1 and Unit 2 Alstom low-pressure turbines.

Based on this review, the Turbine-Generated Missiles' impact hazard can be considered to be negligible.

Volcanic Activity C3 Not applicable to the site because of location (no active or dormant volcanoes located near plant site).

Based on this review, the Volcanic Activity impact hazard can be considered to be negligible.

Waves C4 Waves associated with external flooding are covered under that hazard.

Based on this review, the Waves impact hazard can be considered to be negligible.

PRA-EXT-HAZ-SCRN [26] References 6.1. Nuclear Energy Institute, "NEI 50.69 LAR Template," retrieved from https://www.nei.org.

6.2. Doc. No. PRA-NB-QU, Revision 5, "Internal Events Quantification Notebook," April 6, 2018.

6.3. Indiana and Michigan Power, D.C. Cook Nuclear Plant, Updated Final Safety Analysis Report, Revision 30.

6.4. Doc. No. SD-990930-004, Revision 12, "Probability of Tornado Missile Strike on Targets at D.C. Cook Nuclear Plant," May 22, 2019.

6.5. Doc. No. PRA-TORMIS-QNT-001, Revision O, "Risk Quantification to Support TORMIS Analysis for Cook Tornado Assessment," August 2, 2013.

6.6. Letter from J. P. Gebbie, Indiana Michigan Power Company (l&M), to the NRC, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2, Response to March 12, 2012, Request for of Enclosure 2: EXTERNAL HAZARDS SCREENING Information, Enclosure 2, 'Recommendation 2.1: Flooding,' Required Response 2, Hazard Reevaluation Report," dated March 6, 2015, AEP-NRC-2015-14, ADAMS Accession No. ML15069A334.

6.7. Letter from J.P. Gebbie, Indiana Michigan Power Company (l&M), to the NRC, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2, Update to Interim Action Plan re. Flood Hazard Reevaluation," dated December 14, 2015, AEP-NRC-2015-116.

6.8. PMP-5091-FLD-001, Revision 9, "Flood Protection Program Implementation," 8/30/2021.

6.9. Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline,"

Revision 0, July 2005.

6.10. Federal Aviation Administration, "Airport Data & Contact Information, retrieved from https://www.faa.gov/airports/airport_ safety/airportdata_ 5010.

6.11. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," LWR Edition, Revision 3, November 1978.

6.12. American Electric Power Service Corporation, Donald C. Cook Nuclear Plant Units 1 and 2, "Individual Plant Examination of External Events Summary Report," April 1992.

6.13. Donald C. Cook Nuclear Plant, Unit 1/2 Technical Specifications, September 5, 2019.

6.14. EC-46977, "Flood Control Wiring Protection for Unit 1 A Train," Revision 0, 9/21/06.

6.15. EC-46978, "Flood Control Wiring Protection for Unit 1 B Train," Revision 0, 9/21/06.

6.16. EC-46979, "Flood Control Wiring Protection for Unit 2 A Train," Revision 0, 9/21/06.

6.17. EC-46980, "Flood Control Wiring Protection for Unit 2 B Train," Revision 0, 9/21/06.

6.18. OP-12-5119, "Flow Diagram Circulating Water, Priming System & Screen Wash Units No.

1 & 2, Revision 88, May 26, 2017.

6.19. Idaho National Laboratory, INL/EXT-18-45359, "Analysis of Loss of Offsite Power Events: 1987-2017," August 2018.

6.20. 1-OHP-4021-057-002, "Placing In/Removing from Service Circulating Water Deicing System," Revision 26, 1/20/2020.

6.21. 12-OHP-4022-001-010, "Severe Weather," Revision 21, 7/21/2017.

6.22. PMP-5055-001-001, "Winterization/Summerization,", Revision 32, 9/25/2019.

6.23. PMP-5055-SWM-001, "Severe Weather Guidelines," Revision 10, 7/27/2017.

6.24. Calculation MD-12-MSC-003-N, "Evaluation of Offsite Sources of Toxic Gas," Revision 5, Red Wolf Associates, December 10, 2019.

6.25. NUREG/CR-4461, "Tornado Climatology of the Contiguous United States, Revision 2, February 2007.

of Enclosure 2: EXTERNAL HAZARDS SCREENING 6.26. EPRI 3002003107, "High Wind Risk Assessment Guidelines," June 2015.

6.27. CA-03-01, "Control Room Habitability Following Postulated Release of Hydrazine,"

Revision 3, 6/26/2017.

6.28. Calculation No. SD-991215-004, Revision 1, "Seismic Analysis of Tornado Missile Barrier for EDG Combustion Air Intake," 5/20/2000.

of Enclosure 2: PROGRESSIVE SCREENING APPROACH FOR ADDRESSING EXTERNAL HAZARDS Event Analysis Criterion Source Comments Initial Preliminary c1.

Event damage potential is < events for NUREG/CR-2300 and Screening which plant is designed.

ASME/ANS Standard RA-Sa-2009 C2.

Event has lower mean frequency and NUREG/CR-2300 and no worse consequences than other ASME/ANS Standard RA-Sa-2009 events analyzed.

C3.

Event cannot occur close enough to NUREG/CR-2300 and the plant to affect it.

ASME/ANS Standard RA-Sa-2009 C4.

Event is included in the definition of NUREG/CR-2300 and Not used to screen.

another event.

ASME/ANS Standard RA-Sa-2009 Used only to include within another event.

C5.

Event develops slowly, allowing ASME/ANS Standard RA-Sa-2009 adequate time to eliminate or mitigate the threat.

Progressive PS1. Design basis hazard cannot cause a ASME/ANS Standard RA-Sa-2009 Screening core damage accident.

PS2. Design basis for the event meets the NUREG-1407 and criteria in the NRC 1975 Standard ASME/ANS Standard RA-Sa-2009 Review Plan (SRP).

PS3. Design basis event mean frequency NUREG-1407 as is < 1E-5/y and the mean conditional modified in core damage probability is < 0.1.

ASME/ANS Standard RA-Sa-2009 PS4. Bounding mean CDF is < 1E-6/y.

NUREG-1407 and ASME/ANS Standard RA-Sa-2009 Detailed PRA Screening not successful. Probabilistic Risk NUREG-1407 and Assessment (PRA) needs to meet ASME/ANS Standard RA-Sa-2009 requirements in the ASME/ANS PRA Standard.

of Enclosure 2: DISPOSITION OF KEY ASSUMPTIONS/SOURCES OF UNCERTAINTY The Cook Nuclear Plant Probabilistic Risk Assessment (PRA) models and documentation were reviewed for plant-specific modeling assumptions and related sources of uncertainty. The FPIE (PRA-NB-UNC [23]) and Fire PRA (PRA-FIRE-UNC [24]) uncertainty notebooks as well as the seismic PRA quantification notebook (PRA-NB-SPRA-QU [25]) document sources of PRA modeling uncertainty.

They identify assumptions and determine if those assumptions are related to sources of model uncertainty and characterize that uncertainty, as necessary. The identified uncertainties in these references were reviewed for this application. Each PRA model includes an evaluation of the potential sources of uncertainty for the base case models using the approach that is consistent with the ASME/ANS RA-Sa-2009 [10] requirements for identification and characterization of uncertainties and assumptions. This evaluation identifies those sources of uncertainty that are important to the PRA results and may be important to PRA applications which meets the intent of steps C-1 and E-1 of NUREG-1855, Revision 1 [8].

The results of the base PRA evaluations were reviewed to determine which potential uncertainties could impact the 10 CFR 50.69 categorization process results. This evaluation meets the intent of the screening portion of steps C-2 and E-2 of NUREG-1855, Revision 1 [8].

Based on the evaluations of the PRA models of record, one key assumption was identified as key model uncertainty for the PRA models:

Westinghouse Generation Ill Reactor Coolant Pump (RCP) Shutdown Seals. The Modeling of the Westinghouse Generation 111 RCP Shutdown seals is the first key model uncertainty for the CNP PRA.

If the new RCP seals do not actuate or fail to remain actuated, severe accident sequences become much more likely. Risk metrics such as Core Damage Frequency and LERF increase significantly if the failure of the shutdown seals is assured. 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 [42], both of which are industry consensus models. The 2015 peer review [22] also found the modeling of the shutdown seals acceptable.

to AEP-NRC-2024-03 Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Renewed Facility Operating Licenses (FOL) Pages Marked to Show Proposed Change (19) Operation with Vacuum Fill:

The licensee is authorized to operate the facility using Reactor Coolant System (RCS) vacuum fill operation in accordance with TS 3.4.3, "RCS Pressure and Temperature (PIT) Limits," with corresponding revisions to Figure 3.4.3-1, "Reactor Coolant System Pressure versus Temperature Limits - Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 EFPY)," and Figure 3.4.3-2, "Reactor Coolant System Pressure versus Temperature Limits - Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY)," as approved in License Amendment No. 323 to Renewed Facility Operating License No. DPR-

58. This includes an approved extension to -14.7 pounds per square inch gage to bound the RCS conditions required to support vacuum fill operation. The licensee shall submit an analysis of the PIT curves in Figures 3.4.3-1 and 3.4.3-2 within one year of the date of issuance of License Amendment No. 323, which demonstrates consideration of all ferritic reactor vessel materials as defined in Appendix G to 10 CFR Part 50, including non-beltline ferritic reactor vessel materials."

(20) The licensee shall implement the items listed in Enclosure 2, Table 1, of l&M letter AEP-NRC-2016-69, dated September 9, 2016, prior to Surveillance Frequency Control Program implementation.

(21) 10 CFR 50.69 Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors:

The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC) RISC-1, RISC-2, RISC3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the Individual Plant Evaluation-External Events (IPEEE) Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; as specified in License Amendment No. [XXX] dated [DATE].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

D.

Physical Protection The Indiana Michigan Power Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, Renewed License No. DPR-58 Amendment No. 315, 319, 322, 323, 325, 333, 334 (II)

The first performance of the periodic measurement of CRE pressure, TS 5.5.16.d, shall be within 24 months, plus the 182 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within 182 days if not performed previously.

(gg) Operation with Vacuum Fill:

The licensee is authorized to operate the facility using Reactor Coolant System (RCS) vacuum fill operation in accordance with TS 3.4.3, "RCS Pressure and Temperature (PIT) Limits," with corresponding revisions to Figure 3.4.3-1, "Reactor Coolant System Pressure versus Temperature Limits - Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 EFPY)," and Figure 3.4.3-2, "Reactor Coolant System Pressure versus Temperature Limits - Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY)," as approved in License Amendment No. 306 to Renewed Facility Operating License No. DPR-74. This includes an approved extension to -14.7 pounds per square inch gage to bound the RCS conditions required to support vacuum fill operation. The licensee shall submit an analysis of the PIT curves in Figures 3.4.3-1 and 3.4.3-2 within one year of the date of issuance of License Amendment No. 306, which demonstrates consideration of all ferritic reactor vessel materials as defined in Appendix G to 10 CFR Part 50, including non-beltline ferritic reactor vessel materials."

(hh)

The licensee shall implement the items listed in Enclosure 2, Table 1, of l&M letter AEP-NRC-2016-69, dated September 9, 2016, prior to Surveillance Frequency Control Program implementation.

(ii) 10 CFR 50.69 Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors:

The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC) RISC-1, RISC-2, RISC3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the Individual Plant Evaluation-External Events (IPEEE) Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; as specified in License Amendment No. [XXX] dated [DA TE].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

Renewed License No. DPR-74 Amendment No. 299, 303, 305, 306, 308, 315, 316