NMP1L3484, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Reatment of Structures, Systems and Components for Nuclear Power Reactors

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

ML22349A521
Person / Time
Site:	Nine Mile Point
Issue date:	12/15/2022
From:	David Gudger Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NMP1L3484
Download: ML22349A521 (1)

Text

200 Constellation Way Kennett Square, PA 19348 www.exeloncorp.com

10 CFR 50.90 10 CFR 50.69 NMP1L3484

December 15, 2022

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

Nine Mile Point Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-63 NRC Docket No. 50-220

SUBJECT:

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 10 CFR 50.69 and 10 CFR 50.90, Constellation Energy Generation, LLC (CEG) is requesting an amendment to the license of Nine Mile Point Nuclear Station, Unit 1 (NMP1).

The proposed amendment would modify the NMP1 licensing basis, by the addition of a License Condition, 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, 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.

The enclosure to this letter provides the basis for the proposed change to the NMP1 Operating License. The categorization process being implemented through this change is consistent with NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, dated July 2005, which was endorsed by the 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.

License Amendment Request Adopt 10 CFR 50.69 Docket No. 50-220 December 15, 2022 Page 2 of the enclosure provides a list of categorization prerequisites. Use of the categorization process on a plant system will only occur after these prerequisites are met.

The PRA models described within this license amendment request (LAR ) are the same as those described within the CEG submittal of the LAR dated, December 15, 2022, "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF505, Revision 2, 'Provide RiskInformed Extended Completion Times - RITSTF Initiative 4b,' " (ML22349A108). CEG requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the application currently in-process. This would reduce the number of CEG and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action (RLA), as the details of the PRA models in each LAR are complete which will allow the NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

CEG requests approval of the proposed license amendment by December 1 5, 2023, with the amendment being implemented within 60 days following NRC approval.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated New York State Official.

Should you have any questions concerning this submittal, please contact Ron Reynolds at ronnie.reynolds@constellation.com.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 15th day of December 2022.

Respectfully,

David T. Gudger Senior Manager - Licensing Constellation Generation Company, LLC

License Amendment Request Adopt 10 CFR 50.69 Docket No. 50-220 December 15, 2022 Page 3

Enclosure:

Evaluation of the Proposed Change

cc: USNRC Region I, Regional Administrator w/ attachments USNRC Project Manager, NMP "

USNRC Senior Resident Inspector, NMP "

A. L. Peterson, NYSERDA "

B. Frymire, NYSPSC "

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 1 of 86

Enclosure Evaluation of the Proposed Change Table of Contents

1

SUMMARY

DESCRIPTION................................................................................................... 3 2 DETAILED DESCRIPTION.................................................................................................... 3 2.1 CURRENT REGULATORY REQUIREMENTS............................................................. 3 2.2 REASON FOR PROPOSED CHANGE......................................................................... 4

2.3 DESCRIPTION

OF THE PROPOSED CHANGE......................................................... 5 3 TECHNICAL EVALUATION.................................................................................................. 6 3.1 CATEGORIZATION PROCESS DESCRIPTION ( 10 CFR 50.69(b)(2)(i))................... 7 3.1.1 Overall Categorization Process................................................................... 7 3.1.2 Passive Categorization Process................................................................ 13 3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii)).......................... 15 3.2.1 Internal Events and Internal Flooding....................................................... 15 3.2.2 Fire Hazards................................................................................................. 15 3.2.3 Seismic Hazards.......................................................................................... 15 3.2.4 Other External Hazards............................................................................... 21 3.2.5 Low Power & Shutdown.............................................................................. 22 3.2.6 PRA Maintenance and Updates................................................................. 23 3.2.7 PRA Uncertainty Evaluations..................................................................... 23 3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii))................................ 25 3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv))......................................................... 27 3.5 FEEDBACK AND ADJUSTMENT PROCESS............................................................ 27 4 REGULATORY EVALUATION............................................................................................ 30 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA................................... 30 4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS.................................. 30

4.3 CONCLUSION

S........................................................................................................... 32 5 ENVIRONMENTAL CONSIDERATION.............................................................................. 33 6 REFERENCES..................................................................................................................... 34

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 2 of 86

LIST OF ATTACHMENTS

List of Categorization Prerequisites............................................................. 41

Description of PRA Models Used in Categorization.................................... 42

Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items......................................................................... 43

External Hazards Screening............................................................................ 49

Progressive Screening Approach for Addressing External Hazards........ 71

Disposition of Key Assumptions/Sources of Uncertainty......................... 72

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 3 of 86

1

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 DETAILED DESCRIPTION

2.1 CURRENT REGULATORY REQUIREMENTS

The Nuclear Regulatory Commission (NRC) 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 NRC 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 NRC 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.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 4 of 86

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, Probabilistic Risk Assessments (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 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 low safety significance, 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 NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline" (Reference [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.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 5 of 86

Implementation of 10 CFR 50.69 will allow CEG to improve focus on equipment that has safety significance resulting in improved plant safety.

2.3 DESCRIPTION

OF THE PROPOSED CHANGE

CEG proposes the addition of the following condition to the renewed operating license of NMP1 to document the NRC's approval of the use 10 CFR 50.69.

CEG is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC )-1, RISC-2, RISC-3, and RISC -4 Structures, Systems, and Components (SSCs ) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; high wind safe shutdown equipment list to evaluate high wind / tornado missile events; 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 and Class 3 and non -Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in CEG's submittal letter dated, December 15, 2022, and all its subsequent associated supplements 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).

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 6 of 86

3 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 § 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 § 50.69(c)(1)(i).

(iv) A description of, and basis for acceptability of, the evaluations to be conducted to satisfy § 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.

The PRA models described within this license amendment request (LAR) are the same as those described within the CEG submittal of the LAR dated, December 15, 2022, "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk -Informed Extended Completion Times - RITSTF Initiative 4b,' " (ML22349A108).

CEG requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the application currently in-process. This would reduce the number of CEG and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action (RLA),

as the details of the PRA models in each LAR are complete which will allow the NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 7 of 86

3.1 CATEGORIZATION PROCESS DESCRIPTION ( 10 CFR 50.69(b)(2)(i))

3.1.1 Overall Categorization Process

CEG 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" (Reference [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," as endorsed by RG 1.201, with the exception of the evaluation of impact of the extreme wind or tornado hazard, which will use a high wind safe shutdown equipment list; and the seismic hazard, which will use the EPRI 3002017583 (Reference [3]) approach for seismic Tier 1 sites, which includes NMP1, to assess seismic hazard risk for 50.69. Inclusion of additional process steps discussed below to address seismic considerations will ensure that reasonable confidence in the evaluations required by 10 CFR 50.69(c)(1)(iv) is achieved. RG 1.201 states that "the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence" and that "all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1)(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 complete, they may even be performed in parallel. Note that NEI 00-04 only requires Item 3 to be completed for components/functions categorized as Low Safety Significant (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, and fire PRAs)

2. non-PRA approaches (e.g., Fire Safe Shutdown Equipment List, Seismic Safe Shutdown Equipment List, 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

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 8 of 86

Figure 3-1 is an example of the major steps of the categorization process described in NEI 00-04; two steps (represented by four blocks on the figure) have been included to highlight review of seismic insights as pertains to this application, as explained further in Section 3.2.3:

Figure 3-1: Categorization Process Overview

Categorization of SSCs will be completed per the NEI 00-04 process, as endorsed by RG 1.201, 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., HSS or LS S 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 only be

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 9 of 86

"preliminary" until it has been confirmed by the IDP. Once the IDP confirms that the categorization process was followed appropriately, the final 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. 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. This is also summarized in the Table 3-1. A component is assigned its final RISC category upon approval by the IDP.

Table 3-1: Categorization Evaluation Summary

Categorization IDP Drives Element Step - NEI 00-04 Evaluation Level Change Associated Section HSS to Functions LSS Internal Events Not Base Case - Allowed Yes Section 5.1 Fire, Seismic and Other External Allowable No Risk (PRA Events Base Component Modeled) Case PRA Sensitivity Allowable No Studies Integral PRA Not Assessment - Allowed Yes Section 5.6 Fire and Other Not External Hazards Component Allowed No Risk (Non-modeled) Seismic - Function/Component Allowed 2 No

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

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Categorization IDP Drives Element Step - NEI 00-04 Evaluation Level Change Associated Section HSS to Functions LSS Containment - Component Not Yes Section 6.2 Allowed Qualitative Considerations - Function Allowable1 N/A Criteria Section 9.2 Passive Passive - Section Segment/Component Not No 4 Allowed

Notes:

1 The assessments of the qualitative considerations are agreed upon by the IDP in accordance with Section 9.2. In some cases, a 10 CFR 50.69 categorization team may provide preliminary assessments of the seven considerations for the IDPs consideration, however the final assessments of the seven considerations are the direct responsibility of the IDP.

The seven considerations are addressed preliminarily by the 10 CFR 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 10 CFR 50.69 categorization team determines that one or more of the seven considerations cannot be confirmed, then that function is presented to the IDP as preliminary HSS. Conversely, if all the seven considerations are confirmed, then the function is presented to the IDP as preliminary LSS.

The System Categorization Document, including the justifications provided for the qualitative considerations, is reviewed by the IDP. The IDP is responsible for reviewing the preliminary assessment to the same level of detail as the 10 CFR 50.69 team (i.e.,

all considerations for all functions are reviewed). The IDP 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 IDP, changes may be made from preliminary HSS to LSS or from preliminary LSS to HSS at the discretion of the IDP. If the IDP determines any of the seven considerations cannot be confirmed (false response ) for a function, then the final categorization of that function is HSS.

2 IDP consideration of seismic insights can also result in an LSS to HSS determination.

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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 00-04 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 a 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). Except for seismic, 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. Components having seismic functions may be HSS or LSS based on the IDPs consideration of the seismic insights applicable to the system being categorized. 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. For the seismic hazard, given that NMP1 is a seismic Tier 1 (low seismic hazard) plant as defined in Reference [3], seismic considerations are not required to drive an HSS determination at the component level, but the IDP will consider available seismic information pertinent to the components being categorized and can, at its discretion, determine that a component should be HSS based on that information.

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

• The IDP will be composed of a group of at least five experts who collectively have expertise in plant operation, design (mechanical and electrical) 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 § 50.69(f)(1) will be documented in CEG procedures.

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• 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 safety significant and LSS.

• Passive characterization will be performed using the processes described in Section 3.1.2.

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

• An unreliability factor of 3 will be used for the sensitivity studies described in Section 8 of NEI 00-04. The factor of 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 SE (Reference [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 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 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, CEG will not take credit for alternate means unless the alternate means are proceduralized and included in Licensed Operator training.

• NMP1 proposes to apply an alternative seismic approach to those listed in NEI 00-04 Sections 1.5 and 5.3. This approach is specified in EPRI 3002017583 (Reference [3]) for Tier 1 plants and is discussed in Section 3.2.3.

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

• Internal Event Risks: Internal events including internal flooding PRA, as submitted to the NRC for TSTF 505 dated December 15, 2022, (ML22349A108) (Refer to Attachment 2).

• Fire Risks: Fire PRA model, as submitted to the NRC for TSTF 505 dated December 15, 2022, (ML22349A108) (Refer to Attachment 2).

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• Seismic Risks: EPRI Alternative Approach in EPRI 3002017583 (Reference [3]) for Tier 1 plants with the additional considerations discussed in Section 3.2.3 of this LAR.

• Extreme Wind or Tornado: High Wind Safe Shutdown Equipment List as discussed in Section 3.2.4 of this LAR.

• Other External Risks (e.g., external floods): Using the IPEEE screening process as approved by NRC SE dated July 18, 2000 (TAC No. M83645). 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" (Reference [5]), 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 probabilistic risk assessment 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 support 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

3.1.2 Passive Categorization Process

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 contained in Reference [6]

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(ML090930246) consistent with the related Safety Evaluation (SE) issued by the Office of Nuclear Reactor Regulation.

The RI-RRA methodology is a risk-informed safety classification and treatment program for repair/replacement activities (RI-RRA methodology) 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. Component supports are assigned 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.

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 (Reference [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. 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 1.147, Revision 15.

Both code cases employ a similar risk-informed safety classification of SSCs in order to change the repair/ replacement requirements of the affected LSS components. All ASME Code Class 1 SSCs with a pressure retaining function, as well as supports, will be assigned high safety-significant, HSS, for passive categorization which will result in HSS for its risk-informed safety classification and cannot be changed by the IDP. Therefore, this methodology and scope for passive categorization is acceptable and appropriate for use at NMP1 for 10 CFR 50.69 SSC categorization.

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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 peer reviewed and there are no PRA upgrades that have not been peer reviewed. The PRA models described within this license amendment request (LAR) are the same as those described within the CEG submittal of the LAR dated December 15, 2022, "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b,' "

(ML22349A108).

3.2.1 Internal Events and Internal Flooding

The NMP1 categorization process for the internal events and flooding hazard will use a peer reviewed plant-specific PRA model. The CEG risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for NMP1. of this enclosure identifies the applicable internal events and internal flooding PRA models.

3.2.2 Fire Hazards

The NMP1 categorization process for fire hazards will use a peer reviewed plant-specific fire PRA model. The internal Fire PRA model was developed consistent with NUREG/CR-6850 and only utilizes methods previously accepted by the NRC. The CEG risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for NMP1. Attachment 2 at the end of this enclosure identifies the applicable Fire PRA model.

3.2.3 Seismic Hazards

10 CFR 50.69(c)(1) requires the use of PRA to assess risk from internal events. For other risk hazards such as seismic, 10 CFR 50.69 (b)(2) allows, and NEI 00-04 (Reference [1])

summarizes, the use of other methods for determining SSC functional importance in the absence of a quantifiable PRA (such as Seismic Margin Analysis or IPEEE Screening) as part of an integrated, systematic process. For the NMP1 seismic hazard assessment, CEG proposes to use a risk informed graded approach that meets the requirements of 10 CFR 50.69 (b)(2) as an alternative to those listed in NEI 00-04 sections 1.5 and 5.3. This approach is specified in Electric Power Research Institute (EPRI) 3002017583, "Alternative Approaches for

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Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," Reference [3], and includes additional qualitative considerations that are discussed in this section1.

NMP1 meets the EPRI 30 02017583 Tier 1 criteria for a "Low Seismic Hazard/High Seismic Margin" site. The Tier 1 criteria are as follows:

"Tier 1: Plants where the GMRS [Ground Motion Response Spectrum] peak acceleration is at or below approximately 0.2g or where the GMRS is bel ow or approximately equal to the SSE [Safe Shutdown Earthquake] between 1.0 Hz and 10 Hz. Examples are shown in Figures 2-1 and 2-2. At these sites, the GMRS is either very low or within the range of the SSE such that unique seismic categorization insights are not expected."

Note: EPRI 3002017583 applies to the Tier 1 sites in its entirety except for sections 2.3 (Tier 2 sites), 2.4 (Tier 3 sites), Appendix A (seismic correlation),

and Appendix B (criteria for capacity-based screening).

The Tier 1 criterion (i.e., basis) in EPRI 3002017583 is a comparison of the ground motion response spectrum (GMRS, derived from the seismic hazard) to the safe shutdown earthquake (SSE, i.e., seismic design basis capability). U.S. nuclear power plants that utilize the 10 CFR 50.69 Seismic Alternative (EPRI 3002017583) will continue to compare GMRS to SSE.

The trial studies in EPRI 3002017583 show that seismic categorization insights are overlaid by other risk insights even at plants where the GMRS is far beyond the seismic design basis.

Therefore, the basis for the Tier 1 classification and resulting criteria is not that the design basis insights are adequate. Instead, it is that consideration of the full range of the seismic hazard produces limited unique insights to the categorization process. That is the basis for the following statements in Table 4-1 of the EPRI report.

"At Tier 1 sites, the likelihood of identifying a unique seismic condition that would cause an SSC to be designated HSS is very low.

Therefore, with little to no anticipated unique seismic insights, the 50.69 categorization process using the FPIE PRA and other risk evaluations along with the required Defense-

1 EPRI 3002017583 is an update to EPRI 3002012988, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," July 2018 (Reference [72]) which was referenced in the NRC issued amendment and SE for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, to implement 10 CFR 50.69 as noted below:

(1) Calvert Cliffs Nuclear Power Plant, Units 1 and 2, "Issuance of Amendment Nos. 332 and 310 Re: Risk-Informed Categorization and Treatment of Systems, Structures, and Components (EPID L-2018-LLA-0482)," February 28, 2020. (ADAMS Accession No. ML19330D909) (Reference

[73]).

(2) This license amendment request incorporates by Reference the Clinton Power Station, Unit 1 response to request for additional information letter of November 24, 2020 (ML20329A433)

(Reference [74]), in particular, the response to the question regarding the differences between the initial EPRI report 3002012988 and the current EPRI report 3002017583 as well as Constellations proposed approach for the 50.69 Seismic Alternative Tier 1.

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in-Depth and IDP qualitative considerations are expected to adequately identify the safety-significant functions and SSCs required for those functions and no additional seismic reviews are necessary for 10 CFR 50.69 categorization. "

The proposed categorization approach for NMP1 is a risk -informed graded approach that is demonstrated to produce categorization insight s equivalent to a seismic PRA. For Tier 1 plants, this approach relies on the insights gained from the seismic PRAs examined in Reference [3]

along with confirmation that the site GMRS is low. Reference [3] demonstrates that seismic risk is adequately addressed for Tier 1 sites by the results of additional qualitative assessments discussed in this section and existing elements of the 10 CFR 50.69 categorization process specified in NEI 00-04.

For example, the 10 CFR 50.69 categorization process as defined in NEI 00-04 includes an Integral Assessment that weighs the hazard -specific relative importance of a component (e.g.,

internal events, internal fire, seismic) by the fraction of the total Core Damage Frequency (CDF) contributed by that hazard. The risk from an external hazard can be reduced from the default condition of HSS if the integral assessment meets the importance measure criteria for LSS. For Tier 1 sites, the seismic risk (CDF/LERF) will be low such that seismic hazard risk is unlikely to influence an HSS decision. In applying the EPRI 3002017583 process for Tier 1 sites to the NMP1 10 CFR 50.69 categorization process, the IDP will be provided with the rationale for applying the EPRI 3002017583 gu idance and informed of plant SSC -specific seismic insights for their consideration in the HSS/LSS deliberations.

EPRI 3002017583 recommends a risk -informed graded approach for addressing the seismic hazard in the 10 CFR 50.69 categorization process. Ther e are a number of seismic fragility fundamental concepts that support a graded approach and there are important characteristics about the comparison of the seismic design basis (represented by the SSE) to the site -specific seismic hazard (represented by th e GMRS) that support the selected thresholds between the three evaluation Tiers in the EPRI report. The coupling of these concepts with the categorization process in NEI 00-04 are the key elements of the approach defined in EPRI 3002017583 for identifying unique seismic insights.

The seismic fragility of an SSC is a function of the margin between an SSCs seismic capacity and the site-specific seismic demand. References such as EPRI NP-6041 (Reference [7])

provide inherent seismic capacities for most SSCs that are not directly related to the site-specific seismic demand. This inherent seismic capacity is based on the non -seismic design loads (pressure, thermal, dead weight, etc.) and the required functions for the SSC. For example, a pump has a relatively high inherent seismic capacity based on its design and that same seismic capacity applies at a site with a very low demand and at a site with a very high demand. At sites with lower seismic demands such as NMP1, there is n o need to perform more detailed evaluations to demonstrate the inherent seismic capacities documented in industry sources such as Reference [7]. Low seismic demand sites have lower likelihood of seismically-induced failures and lesser challenges to plant systems. This, therefore, provides the technical basis for allowing use of a graded approach for addressing seismic hazard at NMP1.

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There are some plant features such as equipment anchorage that have seismic capacities more closely associated with the site-specific seismic demand since those specific features are specifically designed to meet that demand. However, even for these features, the design basis criteria have intended conservatisms that result in significant seismic margins within SSCs.

These conservatisms are reflected in key aspects of the seismic design process. The SSCs used in nuclear power plants are intentionally designed using conservative methods and criteria to ensure that they have margins well above the required design bases. Experience has shown that design practices result in margins to realistic seismic capacities of 1.5 or more.

The following provides the basis for establishing Tier 1 criteria in EPRI 3002017583.

a. SSCs for which the inherent seismic capacities are applicable, or which are designed to the plant SSE will have low probabilities of failure at sites where the peak spectral acceleration of the GMRS < 0.2g or where the GMRS < SSE between 1 and 10 Hz.

b. The low probabilities of failure of individual components would also apply to components considered to have correlated seismic failures.

c. These low probabilities of failure lead to low seismic CDF and LERF estimates, from an absolute risk perspective.

d. The low seismic CDF and LERF estimates lead to reasonable confidence that seismic risk contributions would allow reducing a HSS to LSS due to the 10 CFR 50.69 Integral Assessment if the equipment is HSS only due to seismic considerations.

Test cases described in Section 3 of Reference [3] showed that it would be unusual even for moderate hazard plants to exhibit any unique seismic insights, including due to correlated failures. The plant specific Reference [3] test case information CEG is using from the other licensees and being incorporated by reference into this application is described in Case Study A (References [8], [9], and [10]), Case Study C (References [11], [12]), and Case Study D (References [13], [14], [15], [16], and [17]). Hence, while it is prudent to perform additional evaluations to identify conditions where correlated failures may occur for Tier 2 sites, for Tier 1 sites such as NMP1, correlation studies would not lead to new seismic insights or affect the baseline seismic CDF in any significant way.

The Tier 1 to Tier 2 threshold as defined in EPRI 3002017583 provides a clear and traceable boundary that can be consistently applied plant site to plant site. Additionally, because the boundary is well defined, if new information is obtained on the site hazard, a sites location within a particular Tier can be readily confirmed. In the unlikely event that the NMP1 seismic hazard changes to medium risk (i.e., Tier 2) at some future time, NMP1 will follow its categorization review and adjustment process procedures to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e).

The following provides the basis for concluding that NMP1 meets the Tier 1 site criteria.

In response to the NRC 50.54(f) letter associated with post-Fukushima recommendations (Reference [18]), NMP1 submitted a seismic hazard screening report (Reference [19]) to the

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NRC. The GMRS for NMP1 is below the SSE between 1 Hz and 10 Hz and therefore meets the (second) Tier 1 criterion in Reference [3]. In addition, the maximum GMRS value for NMP1 in the 1-10 Hz range exceeds the 0.2g criterion by a small amount, and only above 7 Hz.

The NMP1 SSE and GMRS curves from the seismic hazard and screening response in Reference [19] are plotted in Figure 1 in Attachment 4. The NRCs staff assessment of the NMP1 seismic hazard and screening response is documented in Reference [20]. In Section 3.4 of Reference [20] the NRC concluded that the methodology used by Exelon (now CEG) in determining the GMRS was acceptable and that the GMRS determined by CEG adequately characterizes the reevaluated hazard for the NMP1 site.

Section 1.1.3 of Reference [3] cites various post-Fukushima seismic reviews performed for the U.S. fleet of nuclear power plants. For NMP1, the specific seismic reviews prepared by the licensee and the NRCs staff assessments are provided here. These licensee documents were submitted under oath and affirmation to the NRC.

1. NTTF Recommendation 2.1 seismic hazard screening (References [19], [20])

2. NTTF Recommendation 2.3 seismic walkdowns (References [21], [22], [23], [24], [25])

3. NTTF Recommendation 4.2 seismic mitigation strategy assessment (S-MSA)

(References [26], [27])

The following additional post-Fukushima seismic reviews were performed for NMP1.

4. NTTF Recommendation 2.1 seismic high frequency evaluation (References [28], [29])

The small percentage contribution of seismic to total plant risk makes it unlikely that an integral importance assessment for a component, as defined in NEI 00-04, would result in an overall HSS determination. Further, the low hazard relative to plant seismic capability makes it unlikely that any unique seismic condition would exist that would cause an SSC to be designated HSS for a Tier 1 site such as NMP1.

As an enhancement to the EPRI study results as they pertain to NMP1, the proposed NMP1 categorization approach for seismic hazards will include qualitative consideration of the mitigation capabilities of SSCs during seismically-induced events and seismic failure modes, based on insights obtained from prior seismic evaluations performed for NMP1. For example, as part of the categorization teams preparation of the System Categorization Document (SCD) that is presented to the IDP, a section will be included in the SCD that summarizes the identified plant seismic insights pertinent to the system being categorized, and will also state the basis for applicability of the EPRI 3002017583 study and the bases for NMP1 being a Tier 1 plant. The discussion of the Tier 1 bases will include such factors as:

• The low seismic hazard for the plant, which is subject to periodic reconsideration as new information becomes available through industry evaluations; and

• The definition of Tier 1 in the EPRI study.

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At several steps of the categorization process (e.g., as noted in Figure 3-1 and Table 3-1) the categorization team will consider the available seismic insights relative to the system being categorized and document their conclusions in the SCD. Integrated importance measures over all modeled hazards (i.e., internal events, including internal flooding, and internal fire for NMP1) are calculated per Section 5.6 of NEI 00-04, and components for which these measures exceed the specified criteria are preliminary HSS which cannot be changed to LSS.

For HSS SSCs uniquely identified by the NMP1 PRA models but having design -basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events, thes e will be addressed using non -PRA based qualitative assessments in conjunction with any seismic insights provided by the PRA.

For components that are HSS due to fire PRA but not HSS due to internal events PRA, the categorization team will review design-basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events and characterize these for presentation to the IDP as additional qualitative inputs, which will also be described in the SCD.

The categorization team will review available NMP1 plant -specific seismic reviews and other resources such as those identified above. The objective is to identify plant-specific seismic insights derived from the above sources, relevant to the components in the system being categorized, that might include potentially important impacts such as:

• Impact of relay chatter

• Implications related to potential seismic interactions such as with block walls

• Seismic failures of passive SSCs such as tan ks and heat exchangers

• Any known structural or anchorage issues with a particular SSC

• Components that are implicitly part of PRA -modeled functions (including relays)

• Components that may be subject to correlated failures Such impacts would be compiled on an SSC basis. As each system is categorized, the system -

specific seismic insights will be provided to the IDP for consideration as part of the IDP review process, as noted in Figure 3-1. As such, the IDP can challenge, from a seismic perspective, any candidate LSS recommendation for any SSC if they believe there is basis for doing so. Any decision by the IDP to downgrade preliminary HSS components to LSS will also consider the applicable seismic insights in that decision. These insights will provide the IDP a means to consider potential impacts of seismic events in the categorization process.

Use of the EPRI approach outlined in Reference [3] to assess seismic hazard risk for 10 CFR 50.69 with the additional reviews discussed above will provide a process for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs that satisfies the requirements of § 50.69(c).

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Based on the above, the Summary/Conclusion/Recommendation from Section 2.2.3 of Reference [3] applies to NMP1, i.e., NMP1 is a Tier 1 plant for which the GMRS is very low such that unique seismic categorization insights are expected to be minimal. As discussed in Reference [3] the likelihood of identifying a unique seismic insight that would cause an SSC to be designated HSS is very low. References [30], [31], and [32] are incorporated into this LAR as they provide additional supporting bases for Tier 1 plants. Therefore, with little to no anticipated unique seismic insights, the 10 CFR 50.69 categorization process using the FPIE PRA and other risk evaluations along with the defense-in-depth and qualitative assessment by the IDP adequately identify the safety-significant functions and SSCs.

3.2.4 Other External Hazards

Extreme High Winds and Tornados

The wind pressure and missile hazard from extreme high winds and tornados are not screened from further evaluation. Therefore, the NMP1 categorization process will use the safety significance process described below to determine the safety significance of SSCs for this hazard. The wind pressure / missile hazard is assumed to be present during a tornado-induced loss of offsite power event.

The wind pressure / missile hazard safety significance process uses a High Winds Safe Shutdown Equipment List (HWSSEL) of SSCs that will be developed from a list of SS Cs needed to achieve and maintain safe shutdown of the reactor assuming unavailability of offsite power. During categorization of systems, NEI 00-04 component to function mapping process will be applied to the safe shutdown function of (1) Decay Heat Removal; (2)

Reactivity Control; (3) Inventory Control; (4) Power Availability, and (5) Reactor Pressure Control. The SSCs that fulfill the wind pressure / missile hazard safe shutdown functions, as well as any high winds or missile barriers that are credited with protecting equipment that fulfills a HWSSEL function, will be identified as candidate high safety significant (HSS) for the system being categorized regardless of their tornado damage susceptibility or frequency of challenge. This approach ensures the SSCs that are credited to achieve and maintain the capability for safe shutdown are retained as safety-significant.

The safety significance process for the high winds hazard is shown in Figure 3-2. There are no importance measures used in determining safety significance of SSCs related to the wind pressure / missile hazard. As stated in NEI 00-04, an SSC identified as HSS by a non-PRA method for external events "may not be re-categorized by the IDP."

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Figure 3-2: Safety Significance Process for SSCs for the Extreme Winds or Tornados Hazard

All Other External Hazards

All other external hazards, except for seismic, were screened for applicability to NMP1 per a plant-specific evaluation in accordance with GL 88-20 (Reference [33]) and updated to use the criteria in ASME PRA Standard RA-Sa-2009. Attachment 4 provides a summary of the external hazards screening results. Attachment 5 provides a summary of the progressive screening approach for external hazards.

3.2.5 Low Power & Shutdown

Consistent with NEI 00-04, the NMP1 categorization process will use the shutdown safety management plan described in NUMARC 91-06 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.

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.

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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.

3.2.6 PRA Maintenance and Updates

The CEG risk management process ensures that the applicable PRA models used in this application continues to reflect the as-built and as-operated plant for NMP1. The process delineates the responsibilities and guidelines for updating the PRA models and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA models (e.g., due to changes in the plant, 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 process will assess 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.

In addition, CEG will implement a process that addresses the requirements in NEI 00-04, 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 upgrades will be peer reviewed prior to implementing th ose 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 in Section 5.

In the overall risk sensitivity studies, CEG will utilize a factor of 3 to increase the unavailability or unreliability of LSS components consistent with that approved for Vogtle in Reference [4].

Consistent with the NEI 00-04 guidance, CEG 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

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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 before reaching the rate assumed in the sensitivity study.

The detailed process of identifying, characterizing, and qualitative screening of model uncertainties is found in Section 5.3 of NUREG-1855 and Section 3.1.1 of EPRI TR-1016737 (Reference [34]). The process in these References was mostly developed to evaluate the uncertainties associated with the internal events PRA model; however, the approach can be applied to other types of hazard groups.

The list of assumptions and sources of uncertainty were reviewed to identify those which would be significant for the evaluation of this application. If the NMP1 PRA model used a non-conservative treatment, or methods that are not commonly accepted, the underlying assumption or source of uncertainty was reviewed to determine its impact on this application.

Only those assumptions or sources of uncertainty that could significantly impact the risk calculations were considered key for this application.

Key NMP1 PRA model specific assumptions and sources of uncertainty for this application were identified and dispositioned in Attachment 6. The conclusion of this review is that no additional sensitivity analyses are required to address NMP1 PRA model specific assumptions or sources of uncertainty.

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3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii))

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 (Reference [35]), consistent with NRC RIS 2007-06.

Finding and Observation (F&O) closure reviews were conducted on the PRA models discussed in this section. Closed findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, "Close-out of Facts and Observations" (F&Os) (Reference [36]) as accepted by NRC in the letter dated May 3, 2017 (Reference [37]).

The results of this review have been documented and are available for NRC audit.

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

The NMP1 FPIE PRA model was peer reviewed in 2008 using the NEI 05-04 process, the PRA Standard (ASME/ANS RA-Sa-2009) (Reference [38]), and Regulatory Guide 1.200, Revision 1 (Reference [39]). This Peer Review (Reference [40]) was a full-scope review of the technical elements of the internal events and internal flooding, at-power PRA. The findings from the peer review have been addressed in the internal events PRA model.

The NMP1 Full Power Internal Events (FPIE) and Fire Probabilistic Risk Assessment (FPRA)

Finding Level Fact and Observation (F&O) Independent Assessment was conducted in 2021 (Reference [41]). The purpose was to perform an independent assessment in accordance with NEI 17-07 (Reference [42]) to review close out of "Finding" level F&Os of record from prior PRA peer reviews against the ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 (Reference [38]). The review team closed all F&Os (two HRA related F&Os were superseded by a Focused-Scope Peer Review (FSPR) that was performed in parallel with the closure review (References [43] and [41]). Currently, there are no open findings against the FPIE PRA model (Reference [44]).

Fire PRA Model

A Fire PRA Peer Review for NMP1 was performed in 2011 and the Peer Review Report was issued (Reference [45]). The findings and observations (F&Os) that were identified in the 2011 peer review were addressed and an independent review was performed in November 2017 (Reference [46]) to evaluate and close, as appropriate, the Finding level F&Os of record from the prior PRA peer review against the ASME/ANS PRA Standard. In 2021 an F&O Independent Assessment and Closure Review was held, and the review team closed all F&Os (except one HR and one FQ related F&O were superseded by a Focused-Scope Peer Review (FSPR) that was performed in parallel with the closure review (References [43] and [41]). The Focused-Scope Peer Review (Reference [47]) was performed for upgrades to the fire PRA elements: (1)

HR, QU, and FQ dependency analyses using the EPRI HRA Calculator; (2) FSS for Obstructed

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Plume; (3) FSS for MCB Fire Scenarios; and (4) CF as the review team believed the Upgrades could impact this Element.

Note that the superseded HR and FQ related F&Os by FSPR as mentioned above were re-assessed in the PRA elements HR, QU, and FQ of the FSPR. As a result of the FSPR, a total of nine F&Os were issued, four were Findings and five were Suggestions. The four Findings issued from the Focused-Scope Peer Review (FSPR) are discussed Attachment 3.

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

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3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv))

The NMP1 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

§50.69(b)(2)(iv). Sensitivity studies described in NEI 00-04 Section 8 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 ADJUSTMENT PROCESS

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, an immediate 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.

To more specifically address the feedback and adjustment (i.e., performance monitoring) process as it pertains to the proposed NMP1 Tier 1 approach discussed in section 3.2.3, implementation of the CEG design control and corrective action programs will ensure the inputs for the qualitative determinations for seismic continue to remain valid to maintain compliance with the requirements of 10 CFR 50.69(e).

The performance monitoring process is described in CEGs 10 CFR 50.69 program documents.

The program requires that the periodic review assess changes that could impact the categorization results and provides the Integrated Decision-making Panel (IDP) with an opportunity to recommend categorization and treatment adjustments. Station personnel from engineering, operations, risk management, regulatory affairs, and others have responsibilities for preparing and conducting various performance monitoring tasks that feed into this process.

The intent of the performance monitoring reviews is to discover trends in component reliability; to help catch and reverse negative performance trends and take corrective action if necessary.

The CEG configuration control process ensures that changes to the plant, including a physical change to the plant and changes to documents, are evaluated to determine the impact to drawings, design bases, licensing documents, programs, procedures, and training. The configuration control program has been updated to include a checklist of configuration activities to recognize those systems that have been categorized in accordance with 10 CFR 50.69, to

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 28 of 86

ensure that any physical change to the plant or change to plant documents is evaluated prior to implementing those changes. The checklist includes:

• A review of the impact on the System Categorization Document (SCD) for configuration changes that may impact a categorized system under 10 CFR 50.69.

• Steps to be performed if redundancy, diversity, or separation requirements are identified or affected. These steps include identifying any potential seismic interaction between added or modified components and new or existing safety related or safe shutdown components or structures.

• Review of impact to seismic loading, safe shutdown earthquake (SSE) seismic requirements, as well as the method of combining seismic components.

• Review of seismic dynamic qualification of components if the configuration change adds, relocates, or alters Seismic Category I mechanical or electrical components.

CEG has a comprehensive problem identification and corrective action program that ensures that issues are identified and resolved. Any issue that may impact the 10 CFR 50.69 categorization process will be identified and addressed through the problem identification and corrective action program, including seismic-related issues.

The CEG 10 CFR 50.69 program requires that SCDs cannot be approved by the IDP until the panels comments have been resolved to the satisfaction of the IDP. This includes issues related to system-specific seismic insights considered by the IDP during categorization.

Scheduled periodic reviews no longer than once every two refueling outages will evaluate new insights resulting from available risk information (i.e., PRA model or other analysis used in the categorization) changes, 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 scheduled review will include:

• 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.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 29 of 86

• 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.

The periodic monitoring requirements of the 10 CFR 50.69 process will ensure that these issues are captured and addressed at a frequency commensurate with the issue severity. The 10 CFR 50.69 periodic monitoring program includes immediate and periodic reviews, that include the requirements of the regulation, to ensure that all issues that could affect 10 CFR 50.69 categorization are addressed. The periodic monitoring process also monitors the performance and condition of categorized SSCs to ensure that the assumptions for reliability in the categorization process are maintained.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 30 of 86

4 REGULATORY EVALUATION

4.1 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.

• Regulatory Guide 1.174, "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.

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

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS

Constellation Energy Generation, LLC (CEG) 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.

CEG has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) 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?

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 31 of 86

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 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

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 32 of 86

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.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, CEG 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 CONCLUSION

S

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 Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 33 of 86

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.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 34 of 86

6 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," May 2006.

[3] Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, EPRI, Palo Alto, CA: 2020. 3002017583.

[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 ME9473)," (ADAMS Accession No. ML14237A034), dated December 17, 2014.

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

December 1991.

[6] ANO SER 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) (ML090930246), April 22, 2009.

[7] Electric Power Research Institute (EPRI) NP-6041-SL, "A Methodology for Assessment of Nuclear Plant Seismic Margin," Revision 1, August 1991.

[8] Peach Bottom Atomic Power Station Units 2 and 3, Seismic Probabilistic Risk Assessment Report, "Response to NRC Request Regarding Recommendation 2.1 of the Near Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," (RS-18-098),

(ADAMS Accession No. ML18240A065), dated August 28, 2018.

[9] Peach Bottom Atomic Power Station, Units 2 and 3 - 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-2018-JLD-0010),

(ADAMS Accession No. ML19053A469), dated June 10, 2019.

[10] Peach Bottom Atomic Power Station, Units 2 and 3 - Correction Regarding Staff Review of Seismic Probabilistic Risk Assessment, "Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic," (EPID NO. L-2018-JLD-0010),

(ADAMS Accession No. ML19248C756), dated October 8, 2019.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 35 of 86

[11] Plant C Nuclear Plant, Units 1 and 2, License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process, (ADAMS Accession No. ML17173A875), dated June 22, 2017.

[12] Plant C Nuclear Plant, Units 1 and 2, "Issuance of Amendments Regarding Application of Seismic Probabilistic Risk Assessment Into the Previously Approved 10 CFR 50.69 Categorization Process (EPID L-2017-LLA-0248)," (ADAMS Accession No. ML18180A062), dated August 10, 2018.

[13] Seismic Probabilistic Risk Assessment for Plant D Nuclear Plant, Units 1 and 2, "Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident," (ADAMS Accession No. ML17181A485), dated June 30, 2017.

[14] Plant D Nuclear Plant, Units 1 and 2, Seismic Probabilistic Risk Assessment Supplemental Information,, (ADAMS Accession No. ML18100A966), dated April 10, 2018.

[15] Plant D Nuclear Plant, Units 1 and 2 - Staff Review of Seismic Probabilistic Risk Assessment Associated With [...] NTTF Recommendation 2.1: Seismic, (CAC NOS.

MF9879 AND MF9880; EPID L-2017-JLD-0044), (ADAMS Accession No. ML18115A138),

dated July 10, 2018.

[16] Plant D 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," (ADAMS Accession No. ML18334A363), dated November 29, 2018.

[17] Plant D Nuclear Plant, Units 1 And 2 - Issuance of Amendment Nos. 134 And 38 Regarding, Adoption of 10 CFR 50.69, "Risk-Informed Categorization and Treatment Of Structures, Systems, and Components For Nuclear Power Plants (EPID L-2018-LLA-0493), (ADAMS Accession No. ML20076A194), dated April 30, 2020.

[18] U.S. Nuclear Regulatory Commission, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(F) Regarding Recommendations 2.1,2.3, And 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, (ADAMS Accession No. ML12053A340), dated March 12, 2012.

[19] Constellation Letter to NRC, Seismic Hazard and Screening Report (CEUS Sites),

"Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the NTTF Revi ew of the Fukushima Dai-ichi Accident,"

Attachment 3, Nine Mile Point Nuclear Station, Units 1 and 2, (ADAMS Accession No. ML14099A196), dated March 31, 2014.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 36 of 86

[20] NRC Letter to Exelon, Nine Mile Point Nuclear Station, Units 1 And 2, "Staff Assessment Of Information Provided Pursuant To [] 50.54(f), Seismic Hazard Reevaluations For Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident (TAC NOS. MF3973 and MF3974)," (ADAMS Accession No ML15153A660), dated June 16, 2015.

[21] Constellation Letter to NRC, Nine Mile Point Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-63 Docket No. 50-220, "Response to 10 CFR 50.54(f)

Request for Information, Recommendation 2.3, Seismic," (ADAMS Accession No ML12342A031), dated November 27, 2012.

[22] Constellation Letter to NRC, Nine Mile Point Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-63 Docket No. 50-220, "Supplemental Response to 10 CFR 50.54(f) Request for Information, Recommendation 2.3, Seismic," (ADAMS Accession No ML13197A222), dated July 12, 2013.

[23] Constellation Letter to NRC, Nine Mile Point Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-63 Docket No. 50-220, "Supplemental Response to 10 CFR 50.54(f) Request for Information, Recommendation 2.3, Seismic," (ADAMS Accession No ML13304B418), dated July 12, 2013.

[24] Constellation Letter to NRC, Response to Request for Additional Information Associated with NTTF Recommendation 2.3, Seismic Walkdowns, Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (response includes Ginna and Nine Mile Point), Renewed Facility Operating License Nos. DPR-53 and DPR-69, Docket Nos. 50-317 and 50-318, (ADAMS Accession No. ML13346A011), dated December 2, 2013.

[25] NRC Letter to Constellation, Calvert Cliffs Nu clear Power Plant, Unit Nos. 1 and 2, R.E.

Ginna Nuclear Power Plant, and Nine Mile Point Nuclear Station, Units 1 and 2, "Staff Assessment of Seismic Walkdown Reports Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident (TAC NOS. MF0104, MF0105, MF0127, MF0145, AND MF0146)," (ADAMS Accession No. ML14134A133), dated June 2, 2014.

[26] Exelon Letter to NRC, Renewed Facility Operating License No. DPR -63 NRC Docket No.

50-220, "Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Appendix H, Revision 2, H.4.2 Path 2: GMRS <SSE with High Frequency Exceedances," (ADAMS Accession No ML16307A017), dated November 2, 2016.

[27] NRC Letter to Exelon, Nine Mile Point Nuclear Station, Unit 1, "Staff Review of Mitigation Strategies Assessment Report of The Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) letter," (ADAMS Accession No ML17066A435), dated March 22, 2017.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 37 of 86

[28] Exelon Letter to NRC, Nine Mile Point Nuclear Station, Unit 1, "High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," (ADAMS Accession No 16307A018), dated November 2, 2016.

[29] NRC Letter to Exelon, "Nine Mile Point Nuclear Station, Unit 1 - Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1," (ADAMS Accession No 17031A156), dated February 6, 2017.

[30] Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-53 and DPR-69, Docket Nos. 50-317 and 50-318, "Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors'," (ADAMS Accession No. ML19183A012), dated July 1, 2019.

[31] Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-53 and DPR-69, Docket Nos. 50-317 and 50-318, "Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors'," (ADAMS Accession No. ML19200A216), dated July 19, 2019.

[32] Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-53 and DPR-69, Docket Nos. 50-317 and 50-318, "Revised Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69,

'Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors" letter dated July 19, 2019," (ADAMS Accession No. ML19217A143), dated August 5, 2019.

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

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

[35] NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," (ADAMS Accession No. ML090410014), Revision 2, March 2009.

[36] Nuclear Energy Institute (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)," (ADAMS Accession Number ML17086A431), dated February 21, 2017.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 38 of 86

[37] Nuclear Regulatory Commission (NRC) Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os)," (ADAMS Accession Number ML17079A427), dated May 3, 2017.

[38] ASME/ANS RA -Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum A to RAS-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.

[39] NRC Regulatory Guide 1.2 00, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1.

[40] Nine Mile Point 1 Nuclear Plant PRA Peer Review Report, May 2008.

[41] Nine Mile Point Unit 1 PRA Finding Level Fact and Observation Independent Assessment, Jensen Hughes Report 32466-RPT-011, Rev.0. January 2022.

[42] NEI 17-07, Revision 2, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standards, Nuclear Energy Institute, Washington, DC, August 2019.

[43] Nine Mile Point Unit 1 PRA Focused -Scope Peer Review, Jensen Hughes Report 32466-RPT-12, Rev.0. January 2022.

[44] Nine Mile Point Unit 1 Summary Notebook. N1-PRA-013, Rev. 1.

[45] Nine Mile Point Unit 1 Nuclear Plant Fire PRA Peer Review Report Using ASME PRA Standard Requirements, BWROG Final Report, January 2012.

[46] Nine Mile Point Nuclear Stat ion Unit 1 Fire PRA Finding Level Fact and Observation Technical Review, Jensen Hughes report 032299.004.035-RPT-01, April 2018.

[47] Nine Mile Point Unit 1 PRA Focused -Scope Peer Review. Jensen Hughes Report 32466-RPT-12, Rev.0. January 2022.

[48] NUREG -0800, "Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants," Chapter 3.5.1.6, "Aircraft Hazards," Revision 4, March 2010.

[49] ER-AA-340, "GL 89-13 Program Implementing Procedure," Revision 10.

[50] N1 -SOP-18.1, Special Operating Procedure, Service Water Failure/Low Intake Level, Revision 00600.

[51] N1-SOP-19, Special Operating Procedure, Intake Structure Icing, Revision 00500.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 39 of 86

[52] N1-OP-19, Operating Procedure, Circulating Water System, Revision 04200.

[53] N1-PRA-001, Initiating Events Notebook, 2021 PRA Update, Revision 3.01.

[54] Nine Mile Point Nuclear Power Station - Unit 1 Individual Plant Examination for External Events (IPEEE), August 1996.

[55] Nine Mile Point Nuclear Station, Unit 2 USAR, Revision 25.

[56] Constellation Energy Nuclear Group, LLC Letter to USNRC, Response to March 12, 2012 Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 2, Flooding Hazard Reevaluation Report, (ADAMS Accession No. ML13074A032), dated March 12, 2013.

[57] Exelon Generation Company, LLC Letter to USNRC, Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Focused Evaluation Summary Submittal, (ADAMS Accession No. ML17069A005), dated March 10, 2017.

[58] N1-OP-64, Meteorological Monitoring, Revision 02000.

[59] S0FLOODF002, NMP1 Flood Water Ingress from Probable Maximum Flood, Revision 9.

[60] Nine Mile Point Nuclear Station, Unit 1 - UFSAR, Revision 27.

[61] U.S. NRC, Nine Mile Point Unit 1 - Individual Plant Examination of External Events (TAC NO. M83645), "Staff Evaluation Report by the Office of Research, Individual Plant Examination of External Events (IPEEE) Submittal, Nine Mile Point Nuclear Station, Unit No. 1, Docket No. 50-220," July 18, 2000.

[62] NUREG -1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," June 1991.

[63] Niagara Mohawk Letter to NRC re: NMP1 Turbine Missile Analysis, dated August 14, 1984.

[64] NRC Letter to Niagara Mohawk Power Corporation re: NMP1, "Safety Evaluation, Turbine Missile Protection," May 29, 1985.

[65] Action Request (AR) 04356446, "NMP1 Turbine Risk Analysis," July 13, 2020.

[66] Electric Power Research Institute (EPR I)Technical Report TR-1026511, "Practical Guidance on the Use of PRA in Risk-Informed Applications with a Focus on the Treatment of Uncertainty," December 2012.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-220 Page 40 of 86

[67] NRC NUREG -1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," Revision 1, March 2017 (ML17062A466).

[68] N1 -PRA-014, Revision 2, "Nine Mile Point Nuclear Station Quantification Notebook," May 2022.

[69] NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, January 2018 (ML17317A256).

[70] "Nuclear Power Plant Fire Ignition Frequency and Non -Suppression Probability Estimation Using the Updated Fire Events Database, United States Fire Event Experience Through 2009, in NUREG-2169/EPRI 3002002936, U.S. NRC and Electric Power Research Institute, January 2015.

[71] NUREG -2178 / EPRI 3002005578, Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE), "Volume 2: Fire modeling guidance for electrical cabinets, electric motors, indoor dry transformers, and the main control board," U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research (RES), Rockville, MD, and Electric Power Research Institute (EPRI), Palo Alto, CA: 2019.

[72] EPRI 3002012988, Alternative Approaches for Addressing Seismic Risk in 10CFR 50.69 Risk-Informed Categorization, July 2018.

[73] Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Issuance Of Amendment Nos. 332 and 310, "Risk-Informed Section Categorization and Treatment of Structures, Systems, and Components For Nuclear Power Reactors," (EPID L-2018-LLA-0482), (ADAMS Accession No. ML19330D909), dated February 28, 2020.

[74] Clinton Power Station, Unit 1, "Response to Request for Additional Information Regarding License Amendment Requests to Adopt TSTF-505, Revision 2, and 10 CFR 50.69,"

(ADAMS Accession No. ML20329A433), dated November 24, 2020.

[75] NRC Regulatory Guide 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, December 2020.

License Amendment Request Attachment 1 Adopt 10 CFR 50.69 Docket No. 50-220 Page 41 of 86

Attachment 1: List of Categorization Prerequisites

Constellation 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 3.2). 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 lDP for review and approval. The lDP 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 risk sensitivity study is used to confirm that the population of preliminary LSS components results in acceptably small increases to core damage frequency (CDF ) and large early release frequency (LERF) and meets the acceptance guideline s of Regulatory Guide 1.174.

• Periodic reviews are performed to ensure continued categorization validity and acceptable performance for those SSCs that have been categorized.

• Documentation requirements per Section 3.1.1 of the enclosure.

License Amendment Request Attachment 2 Adopt 10 CFR 50.69 Docket No. 50-220 Page 42 of 86

Attachment 2: Description of PRA Models Used in Categorization

Baseline Unit Model Baseline CDF LERF Comments

Full Power Internal Events (FPIE) PRA Model

Model NM121A 2021 FPIE 1 Peer Reviewed 1.2E-06 1.3E-07 Model of Against RG 1.200 Record R1 in 2008 (MOR)

Fire (FPRA) Model

Model NM121A 2021 Fire 1 Peer Reviewed 2.6E-05 2.1E-06 PRA Model of Against RG 1.200 Record R2 in October 2011 (MOR)

License Amendment Request Attachment 3 Adopt 10 CFR 50.69 Docket No. 50-220 Page 43 of 86

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

Originating Capability Finding Supporting Category Finding Status Disposition Impact to ID Requirement (CC) 10 CFR 50.69 (SR)

1-2 HR-I2 Met The discussion in the HRA notebook Open In the special case of HFEs This is a on the HFEs for which a Joint HFE was associated with the alignment of documentation used rather than include them in the decay heat removal systems, a issue only dependency analysis with the other joint HEP which has been post-initiator actions does not provide a (ZQDHR_DEPOPERATO) is addressed and justification for why this was done. directly inserted into the FPIE fault so there will be (This F&O originated from SR HR -I2) tree logic to model the associated no impact on dependencies. 50.69 It appears the HRA notebook was applications.

PDF'd with tracked changes shown A number of operator action basic (e.g., page 250). events model specific actions related to containment heat On page 247, the last sentence of removal. These include operation Section 4.2.4.1 refers to "Table 4-12, of Containment Spray, Shutdown MCR Abandonment detailed analysis Cooling, Containment Venting, HFEs." This should instead be and Long-Term EC Shell makeup.

Table 4-13.

The dependency treatment of the HFEs associated with alignment of decay heat removal systems is described in Section 2.1.3.2 of the

License Amendment Request Attachment 3 Adopt 10 CFR 50.69 Docket No. 50-220 Page 44 of 86

Originating Capability Finding Supporting Category Finding Status Disposition Impact to ID Requirement (CC) 10 CFR 50.69 (SR)

internal events HRA notebook N1-PRA-004.

For the fire HRA dependency analysis of HFEs associated with alignment of decay heat removal systems, the joint HEP ZQDHR_DEPOPERATO is reconsidered in the fire context.

This joint HEP for the fire context is included in the FPRA using the joint HFE names:

• ZQDHRFDEPOPERATO for fires outside the MCR;

• ZQDHRADEPOPERATO for fires in the MCR that result in MCR abandonment; and

• ZQDHRIDEPOPERATO for fires in the MCR that do not result in MCR abandonment.

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Originating Capability Finding Supporting Category Finding Status Disposition Impact to ID Requirement (CC) 10 CFR 50.69 (SR)

Due to the long timeframe for these actions; the joint HEP is modeled as unaltered by fire and the internal events joint HEP is used for the fire-context HFEs (as shown in Table 4-18 of HRA notebook N1-HRA-F001).

The formatting (e.g., track changes) is an editorial issue that has been fixed. The table number error has been fixed, and no other instances of table numbering discrepancies were found.

1-3 HR-I2 Met Section 5.0 of the HRA Notebook Open Table 5-1 in Section 5.0 on This is a describes the recovery actions included Recovery Actions in the Fire HRA documentation in the NMP1 model in Table 5 -1. Notebook has been revised to issue only These actions are included directly in include the Fire PRA names. which has been the fault tree and are included in the addressed and dependency analysis with the post-so there will be initiator HFEs. There is confusion with no impact on Table 5-1and its introductory paragraph 50.69 as it uses the FPIE HFE names which applications.

are not used in the Fire PRA model.

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Originating Capability Finding Supporting Category Finding Status Disposition Impact to ID Requirement (CC) 10 CFR 50.69 (SR)

(This F&O originated from SR HR-I2)

2-3 FSS-D4 Met The control board definition is one of Open The Panels 1B-8B were The panel the primary inputs on which the entire reclassified as Bin 4, MCB panels. reclassification method is based. The MCB panels The FMDB and N1-FSS-F004 Rev has been defined for NMP1 include the full 6 were revised to reflect this. The performed prior horseshoe walk-through panel (Panels scenario frequencies are to F - N), the console (Panel E), and the calculated by the FMDB in the implementation front portion of a detached walkthrough same manner as all other MCBs, of the 50.69 panel (Panels 1A-8A). The rear portion and the MCB scenarios are program, so of the detached walkthrough panel evaluated in accordance with the there will be no (Panels 1B-8B) is not considered part guidance in Section 7 of Volume 2 impact on 50.69 of the MCB and is treated using Bin 15 of NUREG-2178. applications.

electrical panels. The back portion of the detached walkthrough panel should be included as part of the overall MCB evaluated in the MCB event tree method.

2-5 FSS-H4 Met There were several documentation Open The inconsistencies were This is a inconsistencies noted during the review corrected as follows: documentation where input values reported in the issue only notebooks reviewed did not match the 1. Removed specific values of which has been input values used in the analysis. Input availability factor, and instead, addressed and

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Originating Capability Finding Supporting Category Finding Status Disposition Impact to ID Requirement (CC) 10 CFR 50.69 (SR)

from the utility indicate the analysis is pointed to the IGN notebook for so there will be correct; therefore, these are the value. no impact on documentation issues. The following 50.69 were noted: 2. The value has been corrected. applications.

1. The plant availability factor is listed in N1-FSS-004, Rev. 5 as 0.88 3. Compartment T3B has been whereas it is listed as 0.92 in Section corrected in N1-FSS-F001 Rev 7.

2.1 of the N1-IGN-001, Rev. 5. The FMDB correctly uses 0.92. In addition, 4. NUREG-2178 Vol 2 was added because the V&V example presented to Section 4.1.23 in N1-FSS-F001 in Section 5.2 of N1-FSS-004, Rev. 5 Rev 7.

uses the older factor, the implied ignition frequency is not correct. 5. In Section 7.1.3.3, in N1-FSS-F001 Rev 7, text was revised to

2. The net Bin 4 ignition frequency for explain that the smoke detectors the MCBs is listed as 1.80E-3 in in the MCR have a failure of 0.05 Section 3.1 and 1.83E-3 in Table 4.2 of that includes unavailability and N1-FSS-004. The value used in the unreliability.

FMDB, which incorporates the correct availability factor, is 1.89E-3.

3. The compartments in which the obstructed plume model scenarios are applied re listed in Table AA-7 of N1-

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Originating Capability Finding Supporting Category Finding Status Disposition Impact to ID Requirement (CC) 10 CFR 50.69 (SR)

FSS-001, Rev. 6, but are incorrectly listed as T3A rather than T3B.

4. The reference in N1-FSS-001, Rev.

6 for NUREG-2178, Volume 2 is missing in Section 4.1.

5. The detection credit for the MCBs is based on a generic value of 0.05 as embedded in the NUREG -2178 Volume 2 methodology. This credit should be added to the detection discussion after Table 1 of N1-FSS-001, Rev. 6 with the unreliability factor of 0.05 noted. The unavailability discussion added to resolve previous F&O 2-30 should be extended to address these detection systems.

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Attachment 4: External Hazards Screening2

Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

Acceptance criterion 1.A of Standard Review Plan 3.5.1.6 (Reference [48])

states the probability is considered to be less than an order of magnitude of 10-7 per year by inspection if the plant-to-airport distance D is between 5 and 10 statute miles, and the projected annual number of operations is less than 500 D2, or the plant-to-airport distance D is greater than 10 statute miles, and the projected annual number of operations is less than 1000 D2. (PS2, PS4)

PS2 Aircraft Impact Y The closest airport to the plant is the PS4 Oswego County Airport, a small, public, general aviation facility located approximately 11 miles south of the plant. According to the Federal Aviation Administrations Air Traffic Activity System, the annual operations from this airport is less than 21,000, which is less than the 500 D2 criteria.

(PS2, PS4)

Syracuse International Airport, about 30 miles southwest of the plant, is the nearest airport with scheduled commercial air service. According to the Federal Aviation Administrations

2 The list of hazards and their potential impacts considered those items listed in Tables D-1 and D-2 in Appendix D of RG 1.200, Rev. 3 (Reference [75])

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

Air Traffic Activity System, the annual operations from this airport is less than 65,000, which is less than the 1000 D2 criteria. (PS2, PS4)

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

NMP1 is located on the southeast shore of Lake Ontario, which precludes Avalanche the possibility of an avalanche.

Y C3 Based on this review, the Avalanche hazard can be considered to be negligible.

The hazard is slow to develop and can be identified via monitoring and managed via standard maintenance process. Actions committed to and completed by NMP1 in response to Generic Letter 89-13 provide on-going control of biological hazards. These C1 controls are described in Constellation Biological Event procedure ER-AA-340, GL 89-13 Y C4 Program Implementing Procedure (Reference [49]). (C5)

C5

In addition, there are several procedures implemented to maintain intake water, including N1-SOP-18.1, N1-OP-19, and N1-SOP-19 (References [50], [51], [52]). (C1)

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

Per the NMP1 Initiating Events Notebook (Reference [53]), loss of the plant intake (due to a variety of causes) is already identified as an IE category in the NMP1 PRA. (C4)

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

Per the IPEEE (Reference [54], a 1000ft rock dike constructed along the lake shoreline which is designed to protect the plant from wave action, ice accumulation, and soil erosion.

In addition, per NMP2 USAR Section 2.5.1.1 (Reference [55]), a dike was Coastal Erosion built extending from the existing dike in Y C1 front of Unit 1 on the west to a point where the ground rises naturally to el 80 m (263 ft). The dike prevents waves from reaching unit structures and thus eliminates the hazard of shoreline erosion at the site.

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

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

Drought is a slowly developing hazard allowing time for orderly plant Drought reductions, including shutdowns.

Y C5 Based on this review, the Drought hazard can be considered to be negligible.

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 flood hazard reevaluation report (FHRR) was submitted to NRC for review on March 12, 2013 (Reference [56]). The results indicate that all flood causing mechanisms, except Local Intense Precipitation (LIP), are bounded by the current licensing basis (CLB) and do not pose a challenge to the plant.

External Flooding Y C1 LIP was reevaluated and found to produce a maximum still water surface elevation (WSE) of 262.2 ft at NMP1, where finished floor elevation (FFE) is 261 ft (Reference [57]). Water will remain above 261 ft for approximately 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> during the event.

Temporary barriers are installed per plant procedures upon receiving warning of a consequential rainfall at the site. The amount of rain that initiates the procedure is conservatively set at 2 in/hr or 6-inches

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

in a 24 hr period (Reference [58]). The flood barriers take approximately 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to install at both units.

Given the uncertainty associated with forecasting large storms and the time required to install the flood barriers, NMP1 performed additional analysis in S0FLOODF002 (Reference [59]) to calculate the volume of water that will leak into the buildings during the LIP without any temporary flood barriers installed.

The study reviewed water intrusion through normally closed exterior doors that would ultimately accumulate on the 250 ft elevation to a maximum depth of 31 inches; as a result, no safety related equipment will be impacted.

Since the LIP mechanism requires several normally closed exterior doors to slow water intrusion for screening.

These doors will be considered high safety significant (HSS) since they are credited for screening the external flood hazard.

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

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

Extreme Wind or See Section 3.2.4 of this application.

Tornado N N/A

The principal effects of such events (such as freezing fog) would be to cause a loss of off-site power, which is Fog addressed in weather-related LOOP Y C4 scenarios in the FPIE PRA model for NMP1.

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

Per the IPEEE (Reference [54]), The site is sufficiently cleared in areas adjacent to the plant such that forest or brush fires pose no safety hazards.

(C3)

External fires (Forest or Range Fire)

Forest or Range Fire C3 originating from outside the plant Y boundary have the potential to cause a C4 loss of off-site power, which is addressed in grid-related LOOP scenarios in the FPIE PRA model for NMP1. (C4)

Based on this review, the Forest or Range Fire hazard can be considered to be negligible.

The principal effects of such events Frost would be to cause a loss of off-site Y C4 power, which is addressed for weather-related LOOP scenarios in the FPIE PRA model for NMP1.

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

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

The IPEEE (Reference [54]) states that "Principal structures and equipment which may serve either to prevent accidents or to mitigate their consequences are designed, fabricated and erected in accordance with applicable codes to withstand the Hail C1 most severe [] ice condition." (C1)

Y C4 In addition, the principal effects of such events would be to cause a loss of off-site power, which is addressed for weather-related LOOP scenarios in the FPIE PRA model for NMP1. (C4)

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

The NMP1 IPEEE (Reference [54])

discusses severe temperature transients (extreme heat) as being generally unimportant from a risk perspective due to it being a slow-High Summer C4 moving process allowing time for Temperature Y proper actions. (C5)

C5 In addition, plant trips due to this hazard are covered in the definition of another event in the PRA model (e.g.,

transients, loss of condenser). (C4)

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

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

The NMP2 UFSAR Section 2.4.1.2 (Reference [55]) states that tide magnitudes on Lake Ontario that could impact the Nine Mile Point site amount High Tide to less than 2.5 cm (1-inch).

Y C1 See also External Flooding.

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

Per the IPEEE (Reference [54]),

hurricanes are considered unlikely at NMP1 due to the geographic location, i.e., upstate New York. Tornadoes are considered to be the dominant wind Hurricane (Tropical hazard contributor to NMP1.

Cyclone) Y C3 See also Extreme Winds or Tornados, and External Flood.

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

Per the IPEEE (Reference [54]) and NMP1 UFSAR (Reference [60]), a Ice Cover Y C1 1000 ft rock dike constructed along the lake shoreline which is designed to protect the plant from wave action, ice accumulation, and soil erosion.

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

The lake water intake structure is a reinforced concrete structure setting on the lake bottom at a distance of approximately two-tenths of a mile from the shoreline at the bottom of the lake. The location of the intake structure was chosen in lieu of the conventional shoreline intake because of the large masses of ice which build up along the south shore of Lake Ontario every year.

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

Per the IPEEE (Reference [54]),

detailed analyses were performed for potential nearby industrial facility accidents, including explosions, flammable vapor clouds, toxic chemical release, fire, and collisions with intake and discharge structures.

The low severity of these potential Industrial or Military C1 accidents and/or their postulated Facility Accident Y distance from NMP1 is judged C3 adequate to make risk significance minimal. (C1, C3)

Per NMP2 USAR Section 2.2.1 (Reference [55]), there are no chemical plants, refineries, military bases, or underground gas storage facilities within 8 km (5 mi) of the plant.

(C3)

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

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

Internal Flooding The NMP1 Internal Events and Internal N/A N/A Flood PRA model addresses risk from internal Flood events.

Internal Fire N/A N/A The NMP1 Internal Fire PRA model addresses risk from internal fires.

Plant site is located on level terrain and is not subject to landslides.

Landslide Y C3 Based on this review, the Landslide hazard can be considered to be negligible.

Lightning strikes are not uncommon in nuclear plant experience. They can result in losses of off-site power or surges in instrumentation output if grounding is not fully effective. The latter events often lead to reactor trips.

Lightning Y C4 Both events are incorporated into the NMP1 internal events model through the incorporation of generic and plant-specific data.

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

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

In its review of the NMP1 IPEEE, (Reference [61]) the NRC noted low lake level was not addressed in the IPEEE but a related discussion was found in the NMP2 USAR stating that failure of two dams on the outlet of Lake Ontario would lead to gradual decline of lake level from an average of 242.7 ft to 240.6 ft, approximately 1 year after dam failure, allowing time for proper actions from the station. (C5)

Low Lake or River C1 Water Level Y Per UFAR Section 2.2 (Reference C5 [60]), water is admitted to the intake tunnel through a bellmouth-shaped inlet. The inlet is surmounted by a hexagonally shaped guard structure of concrete, the top of which is about 6 ft above the lake bottom and 14 ft below the lowest anticipated lake level. (C1)

See also External Flood.

Based on this review, the Low Tide, Lake Level, or River Stage hazard can be considered to be negligible.

The NMP1 IPEEE (Reference [54])

discusses severe temperature Low Winter C4 transients (extreme cold) as being Temperature Y generally unimportant from a risk C5 perspective due to it being a slow-moving process allowing time for proper actions. (C5)

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

In addition, plant trips due to this hazard are covered in the definition of another event in the PRA model (e.g.,

transients, loss of condenser). (C4)

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

Per the IPEEE (Reference [54]), the probability of a meteorite strike or a Meteorite or Satellite satellite fall is very small (<1E-09/yr).

Impact Y PS4 Based on this review, the Meteorite or Satellite hazard can be considered to be negligible.

Per the IPEEE (Reference [54]),

surveys were conducted of hazardous materials stored or used within 8 km of the site, including pipelines.

The Sithe Independence Power Station is located approximately 2 miles from the plant and includes a Pipeline Accident C1 natural gas pipeline in a remote area.

Y The IPEEE summarizes the C4 consequences of a postulated break in the natural gas pipeline assuming a bounding analysis and concludes that no critical structures would be damaged but a loss of offsite power event could occur. This event is included in the internal events PRA model. (C4)

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

In addition, per the NMP1 UFSAR Section 2.2.3.1.4 (Reference [60]), the nearest gas pipeline is over 3.2 km (2 mi) from the Nine Mile Point site. The production of high heat fluxes and smoke from fires in the site vicinity do not present a hazard to the safe operation of the plant due to the distance of these potential fires from the site. (C1)

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

Refer to External Flood.

LIP was reevaluated and found to produce a maximum still water surface elevation (WSE) of 262.2 ft at NMP1, where finished floor elevation (FFE) is 261 ft (Reference [57]). Water will remain above 261 ft for approximately 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> during the event.

Precipitation, Intense Y C1 Temporary barriers are installed per plant procedures upon receiving warning of a consequential rainfall at the site. The amount of rain that initiates the procedure is conservatively set at 2 in/hr or 6-inches in a 24 hr period (Reference [58]). The flood barriers take approximately 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to install at both units.

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

Given the uncertainty associated with forecasting large storms and the time required to install the flood barriers, NMP1 performed additional analysis in S0FLOODF002 (Reference [59]) to calculate the volume of water that will leak into the buildings during the LIP without any temporary flood barriers installed.

The study reviewed water intrusion through normally closed exterior doors that would ultimately accumulate on the 250 ft elevation to a maximum depth of 31 inches; as a result, no safety related equipment will be impacted.

Since the LIP mechanism requires several normally closed exterior doors to slow water intrusion for screening.

These doors will be considered high safety significant (HSS) since they are credited for screening the external flood hazard.

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

Per the IPEEE (Reference [54]) and Release of Chemicals UFSAR Section 2.2.1 (Reference [60]),

in Onsite Storage Y C1 no sources of potential toxic chemical hazards stored onsite were shown to have the potential to incapacitate control room operators.

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

See also Toxic Gas.

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

Per the IPEEE (Reference [54]), the principle body of water relating to NMP1 and NMP2 is Lake Ontario.

There are no major streams or rivers within the drainage area that contains River Diversion Y C3 the site. The location of NMP2 along Lake Ontario precludes the possibility of a river diversion.

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

The plant is designed for such events.

More common wind-borne dirt can occur but poses no significant risk to Sandstorm C1 NMP1 given the robust structures and Y protective features of the plant.

Based on this review, the Sand or Dust Storm hazard can be considered to be negligible.

Per the IPEEE (Reference [54]), the maximum lake level since the Army Seiche Y C1 Corps of Engineers began their current lake level management plan is 249.6 ft.

A maximum surge, seiche, and wave action induced runup is expected to

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

total 10 ft, which is an effective flood elevation below the ground elevation of NMP1.

See also External Flood.

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

Seismic Activity N/A N/A See Section 3.2.3 and Figure A4-1 in this Attachment.

This hazard is slow to develop and can be identified via monitoring and managed via normal plant processes.

Snow Y C5 Potential flooding impacts are covered under external flooding.

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

The potential for this hazard is low at C1 the site, the plant design considers this hazard (C1), and the hazard is slow to Soil Shrink-Swell Y develop and can be mitigated. (C5)

C5 Based on this review, the Soil Shrink-Swell Consolidation impact hazard can be considered to be negligible.

Per the IPEEE (Reference [54]), the Storm Surge maximum lake level since the Army Y C1 Corps of Engineers began their current lake level management plan is 249.6 ft.

A maximum surge, seiche, and wave

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

action induced runup is expected to total 10 ft, which is an effective flood elevation below the ground elevation of NMP1.

See also External Flood.

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

Per the IPEEE (Reference [54]),

sources of potential toxic chemical hazards include chemicals stored on site, as well as four stationary and two transportation sources within 8 km of the site.

The stationary sources include the James A. FitzPatrick plant, the Alcan Rolled Products Division, Oswego Wire Incorporated, and NMP2. One Toxic Gas Y C1 transportation source of possible hazardous materials is truck traffic along Route 104, which passes within 6.2 km of the site. The second transportation source is the railroad between Oswego and Mexico, N ew York.

Discussions with Conrail indicate that on an average, only one hazardous chemical shipment during an 18 -month period passes throughout the Oswego terminal. Traffic on a spur to the Site

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

is not frequent enough (<30 per year) to warrant consideration.

Also, per the NMP1 UFSAR (Reference [60]), Section 2.2.1, sources of potential toxic chemical hazards include chemicals stored onsite, as well as stationary and transportation sources within 8 km of the site. Analysis results indicated that none of the toxic chemicals evaluated have the potential to incapacitate the control room operators.

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

Per the IPEEE (Reference [54]),

transportation accidents were subject to detailed analysis per the methodology described in NUREG -

1407 (Reference [62]) and were C1 screened as insignificant to plant risk.

Transportation Chapter 2.2.1 of the NMP1 UFSAR Accident Y (Reference [60]) discusses C3 transportation accidents for toxic chemicals and concludes that none of the transportation sources have the potential to incapacitate the control room operators. (C1)

Chapter 2.2.3.1.1 of the NMP2 USAR (Reference [55]) states that due to the large separation distances to the Nine

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

Mile Point site, fires originating from transportation routes do not present a hazard to safe operation of the plant.

(C3)

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

Per the NMP1 UFSAR Section 2.0 (Reference [60], there is no record of wave activity such as tsunami of such a magnitude as to make inundation of the site likely.

Per the NMP2 USAR subsection 2.4.6 Tsunami Y C1 on hydrologic engineering (Reference [55]), tsunami flooding will not occur at the site.

See External Flooding.

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

NMP1 completed a turbine missile analysis as described in its August 14, 1984, letter to the NRC (Reference Turbine-Generated [63] that concluded that if a turbine Missiles Y C1 missile were generated at NMP1, the safe shutdown capability of the plant would be maintained, 2) the integrity of the reactor coolant boundary would be preserved and 3) no accidents which could result in off-site exposure greater

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

than a fraction of 10CFR100 limits are foreseen.

The NRC evaluated the response (Reference [64]) and concluded that the NMP1 analysis satisfied the intent of Regulatory Guide 1.115 (Protection Against Low Trajectory Turbine Missiles) and SRP 3.5.1.3. (Turbine Missiles).

Subsequent to the NRC evaluation, NMP1 had the main turbine rotors replaced as part of a LP monoblock upgrade project, and it continues to follow the General Electric (GE -

turbine manufacturer) inspection and maintenance recommendations (Reference [65]).

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

Per the IPEEE (Reference [54]), NMP1 is not located near a volcano.

Volcanic Activity Y C3 Based on this review, the Volcanic Activity hazard can be considered to be negligible.

Per the IPEEE (Reference [54], a Waves C1 1000ft rock dike constructed along the Y lake shoreline which is designed to protect the plant from wave action, ice accumulation, and soil erosion.

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Screening Result External Hazard Screened? Screening Comment (Y/N) Criterion (Note a)

In addition, a maximum surge, seiche, and wave action induced runup is expected to total 10ft, which is an effective flood elevation below the ground elevation of NMP1.

See also External Flood.

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

Note a - See Attachment 5 for descriptions of the screening criteria.

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Nine Mile Point Unit 1 SSE vs. GMRS 0.50

0.45 SSE

0.40 GMRS

0.35

0.30

0.25

0.20

0.15

0.10

0.05

0.00 0.1 1 10 100 Spectral Frequency (Hz)

Figure 1: SSE and GMRS Response Spectra for NMP1 (From Reference [19], Attachment 3, Figure 2.4-1 (GMRS) and Figure 3.1-1 (SSE)

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Attachment 5: Progressive Screening Approach for Addressing External Hazards

Event Analysis Criterion Source

Initial Preliminary C1. Event damage potential is < events NUREG/CR-2300 and Screening for 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 events analyzed. RA-Sa-2009

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

another event. ASME/ANS Standard RA-Sa-2009 C5. Event develops slowly, allowing ASME/ANS Standard

adequate time to eliminate or mitigate the RA-Sa-2009 threat.

Progressive PS1. Design basis hazard cannot cause ASME/ANS Standard Screening a core damage accident. RA-Sa-2009 PS2. Design basis for the event meets NUREG-1407 and ASME/ANS

the criteria in the NRC 1975 Standard Standard RA-Sa-2009 Review Plan (SRP).

PS3. Design basis event mean frequency NUREG-1407 as modified in is < 1E-5/y and the mean conditional ASME/ANS Standard core damage probability is < 0.1. 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. PRA needs to NUREG-1407 and ASME/ANS meet requirements in the ASME/ANS Standard RA-Sa-2009 PRA Standard.

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Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty

The NMP1 internal events and fire PRA models and documentation were reviewed for generic and plant-specific modeling assumptions and related sources of uncertainty, and the applicable lists of EPRI-identified generic sources of uncertainty per EPRI 1016737 (Reference [34]) and EPRI 1026511 (Reference [66]) were also reviewed.

Each PRA model includes an evaluation of the potential sources of uncertainty for the base models using the approach that is consistent with the ASME/ANS RA -Sa-2009 (Reference [38])

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. The process meets the intent of steps C-1 and E-1 of NUREG-1855, Revision 1 (Reference [67]).

These evaluations are documented in the internal events and internal flooding quantification report N1-PRA-014, Appendix I (Reference [68]). 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.

Additionally, an evaluation of Level 2 internal events PRA model uncertainty was performed, based on the guidance in NUREG-1855 (Reference [67]) and Electric Power Research Institute (EPRI) report 1026511 (Reference [66]). The potential sources of model uncertainty in the NMP1 FPRA model were evaluated for the 32 Level 2 PRA topics outlined in EPRI 1026511 (Reference [66]).

For the 10 CFR 50.69 Program, the guidance in NEI 00-04 (Reference [1]) specifies sensitivity studies to be conducted for each PRA model to address key sources of uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g.,

human error, common cause failure, and maintenance probabilities) do not mask the SSC(s) importance. Regulatory Guide 1.174, Revision 3 (Reference [69]) cites NUREG-1855, Revision 1, as related guidance. In Section B of RG 1.174, Revision 3, the guidance acknowledges specific revisions of NUREG-1855 to include changes associated with expanding the discussion of uncertainties. The results of the evaluation of PRA model sources of uncertainty as described above are evaluated relative to the 10 CFR 50.69 application in Attachment 6 to determine if additional sensitivity evaluations are needed.

Note: As part of the required 50.69 PRA categorization sensitivity cases directed by NEI 00-04, internal events / internal flood and fire PRA models human error and common cause basic events are increased to their 95th percentile and also decreased to their 5th percentile values. In addition, maintenance unavailability terms are set to 0.0.

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For the fire PRA model only, a sensitivity case is required to allow no credit for manual suppression. These results are capable of driving a component and respective functions HSS and therefore the uncertainty of the PRA modeled HEPs and CCFs are accounted for in the 50.69 application.

Disposition of IE/IF PRA Assumptions/Sources of Uncertainty

The table below describes the internal events / internal flooding (IE / IF) PRA sources of model uncertainty and their impact on this application.

IE / IF PRA IE / IF PRA IE / IF PRA Sources of Assumption/ 50.69 Impact Model Sensitivity and Disposition Uncertainty (50.69)

NUREG/CR-6890 is used to Scenarios related to offsite AC This approach provides a best develop the prior sources estimate assessment for the site.

distribution for the LOOP initiator frequency and This will not be a key source of incorporates four causal uncertainty for the NMP1 50.69 categories (plant centered, Application.

switchyard centered, grid related, and weather related). The priors utilize industry data for the plant centered, switchyard centered, and weather LOOP categories. A Bayesian update with plant specific data is utilized to obtain a posterior plant specific LOOP frequency.

Finally, the generic industry data in NUREG/CR-6890 for the failure to recovery probabilities are utilized directly for the applicable time frames in the model.

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IE / IF PRA IE / IF PRA IE / IF PRA Sources of Assumption/ 50.69 Impact Model Sensitivity and Disposition Uncertainty (50.69)

The consequential LOOP Scenarios related to offsite AC This approach provides a best failure probabilities are sources estimate assessment for the site.

derived consistent with NUREG/CR-6890 data and This will not be a key source of industry practice. uncertainty for the NMP1 50.69 Consequential LOOP Application.

probability is modeled and is an important contributor to risk.

Individual CCF groups per Scenarios for which ECCS Because suction strainer failures system are included in the systems are involved impact all ECCS systems as a model for the suppression common-mode failure, any pool suction strainers based potential extended unavailability on generic and strainer via RICT is not relevant.

plugging failure data and generic alpha factor data. As part of the required 10 CFR All BWRs have improved 50.69 PRA categorization their suppression pool sensitivity cases directed by NEI suction strainers to reduce 00 04, internal events / internal the potential for plugging. flood and fire PRA models human However, there is not a error and common cause basic consistent method for the events are increased to their 95th treatment of suppression percentile and also decreased to pool strainer performance. their 5th percentile values. These results are capable of driving a component and respective functions HSS and therefore the uncertainty of the PRA modeled CCFs are accounted for in the 10 CFR 50.69 application

This will not be a key source of uncertainty for the NMP1 50.69 Application.

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IE / IF PRA IE / IF PRA IE / IF PRA Sources of Assumption/ 50.69 Impact Model Sensitivity and Disposition Uncertainty (50.69)

Although ECCS pumps are Scenarios for which Because NMP1 ECCS pumps are designed to operate under containment heat removal not designed to operate under saturated conditions, systems are involved. saturated conditions, it would be uncontrolled venting as a optimistic to credit ECCS pumps cause of core damage is given uncontrolled venting.

not explicitly modeled However, core spray pumps could because and EOP/training potentially operate in such has operators control conditions. Modeling is judged to containment pressure in a best represent the as-designed band and not vent in an plant.

uncontrolled way (assumed to be very unlikely that both This will not be a key source of uncontrolled vent occurs uncertainty for the NMP1 50.69 and continues up to Application.

complete loss of ECCS suction).

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IE / IF PRA IE / IF PRA IE / IF PRA Sources of Assumption/ 50.69 Impact Model Sensitivity and Disposition Uncertainty (50.69)

Generally, credit for Scenarios for which ECCS Modeling is judged to be slightly operation of systems systems are involved conservative for HELB and beyond their design-basis ISLOCA, reasonable to environment is not taken. conservative for containment Some examples include the failure and realistic for room following: 1. HELB: HELB in cooling.

turbine building is assumed to fail BOP systems in this This will not be a key source of location due to uncertainty for the NMP1 50.69 environment. 2. ISLOCA: Application.

HELB and ISLOCA in reactor building are assumed to fail equipment based on local impacts and other systems in the building in the long term (uncertainty about impacts

- insufficient analysis for un-isolated breaks).

Modeling is judged to be slightly conservative for HELB and ISLOCA, reasonable to conservative for containment failure and realistic for room cooling.

Going beyond design basis does not guarantee failure and would likely impact mission time if severe enough. Failure early requires a local extreme impact, which is included in the evaluations.

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IE / IF PRA IE / IF PRA IE / IF PRA Sources of Assumption/ 50.69 Impact Model Sensitivity and Disposition Uncertainty (50.69)

The Internal flood analysis Scenarios involving internal Updated industry data is and initiating event flooding developed routinely where it is frequencies for spray, flood, common practice to implement and major flood scenarios this new data into the model developed consistent with during the next scheduled PRA the EPRI methodology. Update. The NMP1 PRA model The 2021 FPIE update to incorporated the new pipe internal flood uses a pipe rupture frequencies.

length approach per latest Considered an industry good revision of EPRI TR-practice, which has been used in 1013141. Peer Reviewed industry PRAs.

As such, this meets the intent of One of the most important, consensus model approach as and uncertain, inputs to an defined in Reg. Guide 1.200 and internal flooding analysis is is not required to be retained as the frequency of floods of a candidate modeling various magnitudes (e.g., uncertainty.

small, large, catastrophic) This will not be a key source of from various sources (e.g., uncertainty for the NMP1 50.69 clean water, untreated Application.

water, salt water, etc.).

EPRI has developed some data, but the NRC has not formally endorsed its use.

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IE / IF PRA IE / IF PRA IE / IF PRA Sources of Assumption/ 50.69 Impact Model Sensitivity and Disposition Uncertainty (50.69)

Detailed ISLOCA analysis Scenarios for which ECCS ISLOCA initiating event includes the relevant systems are involved frequency is implemented in the considerations listed in IE-model for each path individually.

C12 of the ASME/ANS PRA Probability of pressure boundary Standard (Reference [38]). rupture is included in the model.

Pipe rupture probability The approach for the ISLOCA given failure of two normally frequency determination is closed valves is explicitly considered an industry good included in the accident practice and probably a sequence model. The consensus model given the accident sequence numerous studies since WASH -

response model accounts 1400.

for CCF, but ISLOCA ISLOCA impacts are also initiators (failure of check required for containment valve and MOV) due to defense in depth assessments passive leak failures do not for 50.69 applications.

include common cause. This will not be a key source of ISLOCA is often a uncertainty for the NMP1 50.69 significant contributor to Application.

LERF. One key input to the ISLOCA analysis are the assumptions related to common cause rupture of isolation valves between the RCS/RPV and low pressure piping. There is no consensus approach to the data or treatment of this issue. Additionally, given an overpressure condition in low pressure piping, there is uncertainty surrounding the failure mode of the piping.

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Disposition of Key Assumptions/Sources of Uncertainty

The table below describes the fire PRA sources of model uncertainty and their impact.

Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty

Analysis Boundary This task establishes the overall Based on a review of the and Partitioning spatial scope of the analysis and assumptions and potential provides a framework for organizing sources of uncertainty associated the data for the analysis. The with this element it is concluded partitioning features credited are that the methodology for the required to satisfy established Analysis Boundary and industry standards. Partitioning task does not introduce any epistemic uncertainties that would affect the 50.69 application.

Component This task involves the selection of In the context of the FPRA, the Selection components to be treated in the uncertainty that is unique to the analysis in the context of initiating analysis is related to initiating events and mitigation. The event identification. However, that potential sources of uncertainty impact is minimized through use include those inherent in the of the BWROG Generic Multiple internal events PRA model as that Spurious Operation (MSO) list and model provides the foundation for the process used to identify and the FPRA. assess potential MSOs.

Based on the discussion of sources of uncertainty and the discussion above, it is concluded that the methodology for the Component Selection task does not introduce any epistemic uncertainties that would affect the 50.69 application.

Cable Selection The selection of cables to be Based on a review of the considered in the analysis is assumptions and potential identified using industry guidance sources of uncertainty related to documents. The overall process is this element it is concluded that essentially the same as that used to the methodology for the Cable perform the analyses to Selection task does not introduce

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demonstrate compliance with 10 any epistemic uncertainties that CFR 50.48. would affect the 50.69 calculation.

Qualitative Based on a review of the Screening Qualitative screening element is an assumptions and potential optional task whose objective is to sources of uncertainty related to identify physical analysis units this element, it is concluded that whose potential fire risk contribution the methodology for the can be judged negligible without Qualitative Screening task does quantitative analysis. Since not introduce any epistemic qualitative screening has not been uncertainties that would affect the performed for NMP1, no fire 50.69 calculation.

compartments are qualitatively screened and all compartments will be subjected to quantitative analysis.

Fire-Induced Risk The internal events PRA model was The identified source of Model updated to add fire specific initiating uncertainty could result in the event structure as well as additional over-estimation of fire risk. In system logic. The methodology general, the FPRA development used is consistent with that used for process would have reviewed the internal events PRA model significant fire initiating events and development and was subjected to performed supplemental industry Peer Review. assessments to address this possible source of uncertainty.

The developed model is applied in such a fashion that all postulated Based on a review of the fires are assumed to generate a assumptions and potential plant trip. This represents a source sources of uncertainty related to of uncertainty, as it is not this element and the discussion necessarily clear that fires would above, it is concluded that the result in a trip. In the event the fire methodology for the Fire-Induced results in damage to cables and/or Risk Model task does not equipment identified in Task 2, the introduce any epistemic PRA model includes structure to uncertainties that would affect the translate them into the appropriate 50.69 calculation.

induced initiator.

Fire Ignition Fire ignition frequency is an area Based on the discussion of Frequencies with inherent uncertainty. Part of sources of uncertainty, it is this uncertainty arises due to the concluded that the methodology for the Fire Ignition Frequency

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counting and related partitioning task does not introduce any methodology. epistemic uncertainties that would affect the 50.69 application.

However, the resulting frequency is Consensus approaches are not particularly sensitive to changes employed in the model.

in ignition source counts. The primary source of uncertainty for this task is associated with the industry generic frequency values used for the FPRA. This is because there is no specific treatment for variability among plants along with some significant conservatism in defining the frequencies, and their associated heat release rates.

NMP1 uses the ignition frequencies in NUREG-2169 (Reference [70]),

NUREG-2178 (Reference [71]), and NUREG-2230.

Quantitative The quantitative screening task is to Based on the discussion of source Screening use calculated annual core damage of uncertainty, it is concluded that frequencies (CDF) and annual large the methodology for the early release frequencies (LERF) to Quantitative Screening task does screen fire zones from detailed not introduce any epistemic quantitative analysis. The fire uncertainties that would affect the zones that are screened from 50.69 calculation.

detailed analysis will continue to be conservatively represented in the Fire PRA. That is, these fire zones will remain in the risk profile of the plant as full compartment burns, where every Fire PRA target in the zone is set to failure with total ignition frequency apportioned to the zone. In effect, there is no quantitatively screened fire zone at NMP-1.

Scoping Fire The framework of NUREG/CR-6850 See discussion for Detailed Fire Modeling includes two tasks related to fire Modeling.

scenario development. These two

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tasks are Scoping Fire Modeling and Detailed Fire Modeling (tasks 8 and 11, respectively). The discussion of uncertainty for both tasks is provided in the discussion for Detailed Fire Modeling (task 11).

Detailed Circuit The circuit analysis is performed Circuit analysis was performed as Analysis using standard electrical part of the deterministic post fire engineering principles. However, safe shutdown analysis.

the behavior of electrical insulation Refinements in the application of properties and the response of the circuit analysis results to the electrical circuits to fire induced FPRA were performed on a case-failures is a potential source of by-case basis where the scenario uncertainty. This uncertainty is risk quantification was large associated with the dynamics of fire enough to warrant further detailed and the inability to ascertain the analysis. Hot short probabilities relative timing of circuit failures. and hot short duration probabilities The analysis methodology assumes as defined in NUREG-7150, failures would occur in the worst Volume 2, based on actual fire possible configuration, or if multiple test data, were used in the NMP1 circuits are involved, at whatever Fire PRA. The uncertainty relative timing is required to cause a (conservatism) which may remain bounding worst-case outcome. This in the FPRA is associated with results in a skewing of the risk scenarios that do not contribute estimates such that they are over-significantly to the overall fire risk.

estimated.

Based on the discussion of sources of uncertainty and the discussion above, it is concluded that the methodology for the Detailed Circuit Analysis task does not introduce any epistemic uncertainties that would affect the 50.69 application

Circuit Failure One of the failure modes for a The use of hot short failure Model Likelihood circuit (cable) given fire induced probability and duration probability Analysis failure is a hot short. A conditional is based on fire test data and probability and a hot short duration associated consensus probability are assigned using methodology published in industry guidance published in NUREG-7150, Volume 2.

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NUREG-7150, Volume 2 The uncertainty values specified in Based on the discussion of NUREG-7150, Volume 2 are based sources of uncertainty and the on fire test data. discussion above, it is concluded that the methodology for the Circuit Failure Mode Likelihood Analysis task does not introduce any epistemic uncertainties that would affect the 50.69 application

Detailed Fire The application of fire modeling Consensus modeling approach is Modeling technology is used in the FPRA to used for the Detailed Fire translate a fire initiating event into a Modeling. The methodology for set of consequences (fire induced the Detailed Fire Modeling task failures). The performance of the does not introduce any epistemic analysis requires a number of key uncertainties that would affect the input parameters. These input 50.69 application parameters include the heat release rate (HRR) for the fire, the growth rate, the damage threshold for the targets, and response of plant staff (detection, fire control, fire suppression).

The fire modeling methodology itself is largely empirical in some respects and consequently is another source of uncertainty. For a given set of input parameters, the fire modeling results (temperatures as a function of distance from the fire) are characterized as having some distribution (aleatory uncertainty). The epistemic uncertainty arises from the selection of the input parameters (specifically the HRR and growth rate) and how the parameters are related to the fire initiating event. While industry guidance is available, that guidance is derived from laboratory tests and may not necessarily be

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representative of randomly occurring events.

The fire modeling results using these input parameters are used to identify a zone of influence (ZOI) for the fire and cables/equipment within that ZOI are assumed to be damaged. In general, the guidance provided for the treatment of fires is conservative and the application of that guidance retains that conservatism. The resulting risk estimates are also conservative.

Post-Fire Human The Human Error Probabilities The HEPs include the Reliability Analysis (HEPs) used in the FPRA were consideration of degradation or adjusted to consider the additional loss of necessary cues due to fire.

challenges that may be present The fire risk importance measures given a fire. The HEPs were indicate that the results are obtained using the EPRI HRAC and somewhat sensitive to HRA model included the consideration of and parameter values. The NMP1 degradation or loss of necessary FPRA model HRA is based on cues due to fire. Given the industry consensus modeling methodology used, the impact of approaches for its HEP any remaining uncertainties is calculations, so this is not expected to be small. considered a significant source of epistemic uncertainty.

Further, as directed by NEI 00-04, the fire model human error basic events are increased to their 95th percentile and also decreased to their 5th percentile values as part of the required 50.69 PRA categorization sensitivity cases.

Seismic-Fire Since this is a qualitative The qualitative assessment of Interactions evaluation, there is no quantitative seismic induced fires should not Assessment impact with respect to the be a source of model uncertainty uncertainty of this task. as it is not expected to provide

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changes to the quantified FPRA model.

Based on the discussion of sources of uncertainty and the discussion above, it is concluded that the methodology for the Seismic-Fire Interactions Assessment task does not introduce any epistemic uncertainties that would affect the 50.69 application

Fire Risk As the culmination of other tasks, The selected truncation was Quantification most of the uncertainty associated confirmed to be consistent with with quantification has already been the requirements of the PRA addressed. The other source of Standard.

uncertainty is the selection of the truncation limit. Based on a review of the assumptions and potential sources of uncertainty related to this element and the discussion above, it is concluded that the methodology for the Fire Risk Quantification task does not introduce any epistemic uncertainties that would affect the 50.69 calculation.

Uncertainty and This task does not introduce any This task does not introduce any Sensitivity Analyses new uncertainties. This task is new uncertainties. This task is intended to address how the fire intended to address how the fire risk assessment could be impacted risk assessment could be by the various sources of impacted by the various sources uncertainty. of uncertainty.

Based on the discussion of sources of uncertainty and the discussion above, it is concluded that the methodology for the Uncertainty and Sensitivity

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Analyses task does not introduce any epistemic uncertainties that would affect the 50.69 application

Fire PRA This task does not introduce any This task does not introduce any Documentation new uncertainties to the fire risk. new uncertainties to the fire risk as it outlines documentation requirements. The methodology for the FPRA documentation task does not introduce any epistemic uncertainties that would affect the 50.69 calculation.