JAFP-21-0069, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

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

ML21211A078
Person / Time
Site:	FitzPatrick
Issue date:	07/30/2021
From:	David Gudger Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
JAFP-21-0069
Download: ML21211A078 (67)

Text

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com JAFP-21-0069 10 CFR 50.90 10 CFR 50.69 July 30, 2021 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 James A. FitzPatrick Nuclear Power Plant Renewed Facility Operating License No. DPR-59 NRC Docket No. 50-333

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, Exelon Generation Company, LLC (Exelon) is requesting an amendment to the license of James A.

FitzPatrick (JAF) Nuclear Power Plant.

The proposed amendment would modify the JAF 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 JAF 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. 50-333 July 30, 2021 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 Exelon submittal of the LAR dated July 30, 2021, "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF505, Revision 2, 'Provide RiskInformed Extended Completion Times - RITSTF Initiative 4b,' " (ML21211A053). Exelon 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 Exelon 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.

Exelon requests approval of the proposed license amendment by July 30, 2022, with the amendment being implemented within 60 days following NRC approval.

In accordance with 10 CFR 50.9, 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 (610) 765-5247.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 30th day of July 2021.

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

Enclosure:

Evaluation of the Proposed Change

License Amendment Request Adopt 10 CFR 50.69 Docket No. 50-333 July 30, 2021 Page 3 cc: USNRC Region I, Regional Administrator w/ attachments USNRC Project Manager, FitzPatrick "

USNRC Senior Resident Inspector, FitzPatrick "

A. L. Peterson, NYSERDA "

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 1 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 .................................................................................................. 5 3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(b)(2)(i)) ................... 6 3.1.1 Overall Categorization Process ................................................................... 6 3.1.2 Passive Categorization Process ................................................................ 13 3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii)) .......................... 14 3.2.1 Internal Events and Internal Flooding ....................................................... 14 3.2.2 Fire Hazards ................................................................................................. 15 3.2.3 Seismic Hazards .......................................................................................... 15 3.2.4 Other External Hazards ............................................................................... 24 3.2.5 Low Power & Shutdown.............................................................................. 24 3.2.6 PRA Maintenance and Updates ................................................................. 25 3.2.7 PRA Uncertainty Evaluations ..................................................................... 25 3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii)) ................................ 26 3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv)) ......................................................... 27 3.5 FEEDBACK AND ADJUSTMENT PROCESS ............................................................ 28 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 .............................................................................. 32 6 REFERENCES ..................................................................................................................... 33

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 2 LIST OF ATTACHMENTS : List of Categorization Prerequisites ............................................................. 42 : Description of PRA Models Used in Categorization .................................... 43 : Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items.................................................................................................... 44 : External Hazards Screening............................................................................ 45 : Progressive Screening Approach for Addressing External Hazards ........ 71 : Disposition of Key Assumptions/Sources of Uncertainty ......................... 73

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 3 1

SUMMARY

DESCRIPTION The proposed amendment modifies the 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.

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. 50-333 Page 4 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. 50-333 Page 5 Implementation of 10 CFR 50.69 will allow Exelon to improve focus on equipment that has safety significance resulting in improved plant safety.

2.3 DESCRIPTION

OF THE PROPOSED CHANGE Exelon proposes the addition of the following condition to the renewed operating license of JAF to document the NRC's approval of the use 10 CFR 50.69.

Exelon 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; 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 Exelon's submittal letter dated July 30, 2021, 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).

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.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 6 (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 Exelon submittal of the LAR dated July 30, 2021, "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,' " (ML21211A053).

Exelon 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 Exelon 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.

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

3.1.1 Overall Categorization Process Exelon 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 seismic hazard, which will use the EPRI 3002017583 (Reference [3]) approach for seismic Tier 2 sites, which includes JAF, to assess seismic hazard risk for 10 CFR 50.69. Inclusion of additional process steps discussed below to

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 7 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 completed, they may even be performed in parallel. Note that NEI 00-04 only requires Item 3 to be completed for components/functions categorized as 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 (SSEL), seismic safe shutdown equipment list (SSEL), 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 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:

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 8 Figure 3-1: Categorization Process Overview Define System Boundaries Define System Functions and Assign Components to Functions Identify Seismic Insights Risk Characterization Defense in Depth Characterization Passive Characterization Qualitative Characterization Non-PRA Modeled PRA Modeled Core Damage Containment Evaluation Evaluation Evaluation Evaluation Cumulative Risk Sensitivity Study Preliminary Component Categorization LSS or Can be HSS and can Overturned not be Overturned Review Seismic Insights IDP Review Component Categorization 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., High Safety Significant (HSS) or Low Safety Significant (LSS) that is presented to the Integrated Decision-Making Panel (IDP)). Note: the term "preliminary HSS or LSS" is synonymous with the NEI 00-04 term "candidate HSS or LSS." A component or function is preliminarily categorized as HSS if any element of the process results in a preliminary HSS determination in accordance with Table 3-1 below. The safety significance determination of each element, identified above, is independent of each other and therefore the sequence of the elements does not impact the resulting preliminary categorization of each component or function. Consistent with NEI 00-04, the categorization of a component or function will only be "preliminary" until it has been confirmed by the IDP. Once the IDP confirms that the categorization process was followed appropriately, the final Risk-Informed Safety Class (RISC) category can be assigned.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 9 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 IDP Categorization Drives Change Element Step - NEI 00-04 Evaluation Level Associated HSS to Section Functions LSS Internal Events Not Base Case - Yes Allowed 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 - Yes Allowed Section 5.6 Fire and Other Not External Hazards Component No Allowed Risk -

(Non-modeled) Seismic Function/Component Allowed1 No Shutdown - Not Function/Component No Section 5.5 Allowed Core Damage - Not Function/Component Yes Section 6.1 Allowed Defense-in-Depth Containment - Not Component Yes Section 6.2 Allowed Qualitative Considerations -

Function Allowable2 N/A Criteria Section 9.2 Passive - Not Passive Segment/Component No Section 4 Allowed

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 10 Notes:

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

2 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.

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

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 11 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 JAF is a seismic Tier 2 (moderate 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 Integrated Decision-Making Panel (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 Exelon procedures.

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

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

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.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 12

• 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 SER (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, Exelon will not take credit for alternate means unless the alternate means are proceduralized and included in Licensed Operator training.

• JAF 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 2 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 July 30, 2021, (ML21211A053) (Refer to Attachment 2).

• Fire Risks: Fire PRA model, as submitted to the NRC for TSTF 505 dated July 30, 2021, (ML21211A053) (Refer to Attachment 2).

• Seismic Risks: EPRI Alternative Approach in EPRI 3002017583 for Tier 2 plants with the markups provided in Attachment 2 of References [5] and [6] and additional considerations discussed in Section 3.2.3 of this license amendment request (LAR).

• Other External Risks (e.g., tornados, external floods): Using the IPEEE screening process as approved by NRC SE dated September 21, 2000, (TAC No. M83622) (Reference [7]).

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"

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 13 (Reference [8]), 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 [9]

(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, if categorized, 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.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 14 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 JAF for 10 CFR 50.69 SSC categorization.

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

The following sections demonstrate that the quality and level of detail of the processes used in categorization of SSCs are adequate. The PRA models described below have been 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 Exelon submittal of the LAR dated July 30, 2021, "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,'" (ML21211A053) with routine maintenance updates applied.

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

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 15 3.2.2 Fire Hazards The JAF 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 Exelon risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for JAF. 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 JAF seismic hazard assessment, Exelon 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 Reference [3]1 with the EPRI markups provided in Attachment 2 of References [5] and [6] and includes additional considerations that are discussed in this section.

(Note: The discussion below pertaining to Reference [3] includes the markups provided in Attachment 2 of References [5] and [6]).

The proposed categorization approach for JAF is a risk-informed graded approach that is demonstrated to produce categorization insights equivalent to a seismic PRA. This approach relies on the insights gained from the seismic PRAs examined in Reference [3] and plant specific insights considering seismic correlation effects and seismic interactions. Following the criteria in Reference [3], the JAF site is considered a Tier 2 site because the site Ground Motion Response Spectrum (GMRS) to SSE [Safe Shutdown Earthquake] comparison is above the Tier 1 threshold but not high enough that the NRC required the plant to perform an SPRA 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 [74]) 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 [75]).

(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 [76]), 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 Exelons proposed approach for the 50.69 Seismic Alternative Tier 1.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 16 to respond to Recommendation 2.1 of the Near Term Task Force 50.54(f) letter (Reference [10]). Reference [3] also demonstrates that seismic risk is adequately addressed for Tier 2 sites by the results of additional qualitative assessments discussed in this section and existing elements of the §50.69 categorization process specified in NEI 00-04.

The trial studies in Reference [3], as amended by their RAI responses and amendments (References [11], [12], [13], [14], [15], [16], [17]), [18], and [19]) 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 2 classification and resulting criteria 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 Reference [3].

"At Tier 2 sites, there may be a limited number of unique seismic insights, most likely attributed to the possibility of seismically correlated failures, appropriate for consideration in determining HSS SSCs. The special seismic risk evaluation process recommended using a Common Cause impact approach in the FPIE PRA can identify the appropriate seismic insights to be considered with the other categorization insights by the Integrated Decision-making Panel for the final HSS determinations."

At sites with moderate seismic demands (i.e., Tier 2 range) such as JAF, there is no need to perform more detailed evaluations to demonstrate the inherent seismic capacities documented in industry sources such as Reference [20]. Tier 2 seismic demand sites have a lower likelihood of seismically induced failures and less challenges to plant systems than trial study plants. This, therefore, provides the technical basis for allowing use of a graded approach for addressing seismic hazards at JAF.

Test cases described in Section 3 of Reference [3], as amended by their RAI responses and amendments (References [11], [12], [13], [14], [15], [16], [17], [18], and [19]), showed that there are very few, if any, SSCs that would be designated HSS for seismic unique reasons. The test cases identified that the unique seismic insights were typically associated with seismically correlated failures and led to unique HSS SSCs. While it would be unusual even for moderate hazard plants to exhibit any unique seismic insights, it is prudent and recommended by Reference [3] to perform additional evaluations to identify the conditions where correlated failures and seismic interactions may occur and determine their impact in the 10 CFR 50.69 categorization process. The special sensitivity study recommended in Reference [3] uses common cause failures, similar to the approach taken in a FPIE PRA and can identify the appropriate seismic insights to be considered with the other categorization insights by the IDP for the final HSS determinations.

Exelon is using test case information from Reference [3], developed by other licensees. The test case information is being incorporated by Reference into this application, specifically Case

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 17 Study A (Reference [21]), Case Study C (Reference [22]), and Case Study D (Reference [23])

as well as, RAI responses and amendments (References [11], [12], [13], [14], [15], [16], [17],

[18] and [19]), clarifying aspects these case studies.

Basis for JAF being a Tier 2 Plant As defined in Reference [3], JAF meets the Tier 2 criteria for a "Moderate Seismic Hazard /

Moderate Seismic Margin" site. The Tier 2 criteria are as follows:

"Tier 2: Plants where the GMRS [Ground Motion Response Spectrum] to SSE [Safe Shutdown Earthquake] comparison between 1.0 Hz and 10 Hz is greater than in Tier 1 but not high enough to be treated as Tier 3. At these sites, the unique seismic categorization insights are expected to be limited."

Note: Reference [3] applies to the Tier 2 sites in its entirety except for Sections 2.2 (Tier 1 sites) and 2.4 (Tier 3 sites).

For comparison, Tier 1 plants are defined as having a GMRS peak acceleration at or below approximately 0.2g or where the GMRS is below or approximately equal to the SSE between 1.0 Hz and 10 Hz. Tier 3 plants are defined where the GMRS to SSE comparison between 1.0 Hz and 10 Hz is high enough that the NRC required the plant to perform an SPRA to respond to the Fukushima 50.54(f) letter (Reference [10]).

As shown in Figure A4-1, comparing the JAF GMRS (derived from the seismic hazard) to the SSE (i.e., seismic design basis capability), the GMRS is below the SSE up to approximately 6 Hz and exceeds the SSE above 6 Hz and then drops back below at approximately 50 Hz (Reference [24]). The NRC screened out JAF from performing an SPRA in response to the NTTF 2.1 50.54(f) letter (Reference [25]). As such, it is appropriate that JAF is considered a Tier 2 plant. The basis for JAF being Tier 2 will be documented and presented to the IDP for each system categorized.

The following paragraphs describe additional background and the process to be utilized for the graded approach to categorize the seismic hazard for a Tier 2 plant.

Implementation of the Recommended Process Reference [3] recommends a risk-informed graded approach for addressing the seismic hazard in the 10 CFR 50.69 categorization process. There 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 the GMRS) that support the selected thresholds between the three evaluation Tiers in the report. The coupling of these concepts with the categorization

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 18 process in NEI 00-04 are the key elements of the approach defined in Reference [3] for identifying unique seismic insights.

The seismic fragility of an SSC is a function of the margin between an SSC's seismic capacity and the site-specific seismic demand. References such as EPRI NP-6041 (Reference [20])

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.

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.

In applying the Reference [3] process for Tier 2 sites to the JAF 10 CFR 50.69 categorization process, the IDP will be provided with the rationale for applying the Reference [3] guidance and informed of plant SSC-specific seismic insights that the IDP may choose to consider in their HSS/LSS deliberations. As part of the categorization team's preparation of the System Categorization document (SCD) that is presented to the IDP, a section will be included that provides identified plant seismic insights as well as the basis for applicability of the Reference [3] study and the bases for JAF being a Tier 2 plant. The discussion of the Tier 2 bases will include such factors as:

• The moderate seismic hazard for the plant,

• The definition of Tier 2 in the EPRI study, and

• The basis for concluding JAF is a Tier 2 plant.

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 JAF) 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

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 19 SSCs uniquely identified by the JAF PRA models but having design-basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events, these 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 JAF plant-specific seismic reviews and other resources such as those identified above. The objective of the seismic review is to identify plant-specific seismic insights 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 tanks and heat exchangers

• Any known structural or anchorage issues with a particular SSC

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

For each system categorized, the categorization team will evaluate correlated seismic failures and seismic interactions between SSCs. This process is detailed in Reference [3]

Section 2.3.1 and is summarized below in Figure 3-2.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 20 Figure 3-2: Seismic Correlated Failure Assessment for Tier 2 Plants 2 2

Reproduced from Reference [3] Figure 2-3 including the markups provided in Attachment 2 of References [5] and [6].

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 21 Determination of seismic insights will make use of the full power internal events PRA model supplemented by focused seismic walkdowns. An overview of the process to determine the importance of SSCs for mitigating seismic events follows and is utilized on a system basis:

o Identify SSCs within the system to be categorized o Group SSCs within the system into the classes of equipment and distributed systems used for SPRAs.

o Refine the list and screen out the following SSCs from consideration of functional correlated seismic failures:

• Inherently rugged components

• Components not used in safety functions that support mitigation of core damage or containment performance

• Components already identified as HSS components from the Internal Events PRA or Integrated assessment o Perform a seismic walkdown:

• For SSCs screened IN look for correlation

• For SSCs screened IN or OUT assess for spatial interaction concerns that could fail multiple components in the system, or could fail a single component in the system due to either seismic interaction or direct component failure modes, that result in total loss of a multi-train system and where there is not another system that independently provides the same function o Based on the seismic walkdown:

• Screen out IF SSCs have high seismic capacity AND not included in seismically correlated groups or correlated interaction groups o Add surrogate events to the FPIE model that simulate spatial interaction or Correlation (for the system being categorized) - set the probability of failure to 1E-04.

o Quantify the FPIE model (for the system being categorized) for LOOP and Small LOCA (SLOCA) initiated accident sequences setting (1) the LOOP initiating event frequency to 1.0/yr, (2) the SLOCA initiating event frequency to 1E-02/yr, and (3) the initiating event frequency for all initiators other than LOOP and small LOCA initiators to 0 (zero), and also removing credit for restoration of offsite power in LOOP/SBO accident sequences as well as other functional recoveries o Utilize the Importance Measures from this sensitivity study to identify appropriate SSCs (in the system being categorized) that should be HSS due to correlation or seismic interactions

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 22 Seismic impacts would be compiled on an SSC basis. As each system is categorized, the system-specific seismic insights will be documented in the categorization report and provided to the IDP for consideration as part of the IDP review process (e.g., Figure 3-1). The IDP cannot challenge any candidate HSS recommendation for any SSC from a seismic perspective if they believe there is a basis, except for certain conditions identified in Step 10 of Section 2.3.1 of Reference [3]. Any decision by the IDP to downgrade preliminary HSS components to LSS will consider the applicable seismic insights in that decision. SSCs identified from the Fire PRA as candidate HSS, which are not HSS from the internal events PRA or integrated importance measure assessment, will be reviewed for their design basis function during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events. These insights will provide the IDP a means to consider potential impacts of seismic events in the categorization process.

If the JAF seismic hazard changes from medium risk (i.e., Tier 2) at some future time, prior NRC approval, under 10 CFR 50.90, will be requested if JAF's feedback process determines that a process different from the proposed alternative seismic approach is warranted for seismic risk consideration in categorization under 10 CFR 50.69. After receiving NRC approval, Exelon will follow its categorization review and adjustment process to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e) and the EPRI 3002017583 SSC categorization criteria for the updated Tier.

This includes use of the Exelon corrective action process (CAP).

If the seismic hazard is reduced such that it meets the criteria for Tier 1 in EPRI 3002017583, Exelon will implement the following process.

a) For previously completed system categorizations, Exelon may review the categorization results to determine if use of the criteria in EPRI 3002017583 Section 2.2, "Low Seismic Hazard / High Seismic Margin Sites" would lead to categorization changes. If changes are warranted, they will be implemented through the Exelon design control and corrective action programs and NEI 00-04, Section 12.

b) Seismic considerations for subsequent system categorization activities will be performed in accordance with the guidance in EPRI 3002017583 Section 2.2, "Low Seismic Hazard / High Seismic Margin Sites."

If the seismic hazard increases to the degree that a seismic probabilistic risk assessment (SPRA) becomes necessary to demonstrate adequate seismic safety, Exelon will implement the following process following completion of the SPRA, including adequate closure of Peer Review Findings and Observations.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 23 a) For previously completed system categorizations, Exelon will review the categorization results using the SPRA insights as prescribed in NEI 00-04 Section 5.3, Seismic Assessment and Section 5.6, "Integral Assessment". If changes are warranted, they will be implemented through the Exelon design control and corrective action programs and NEI 00-04 Section 12.

b) Seismic considerations for subsequent system categorization activities will follow the guidance in NEI 00-04, as recommended in EPRI 3002017583 Section 2.4, "High Seismic Hazard / Low Seismic Margin Sites".

Historical Seismic References for JAF The JAF GMRS and SSE curves from the seismic hazard and screening response are shown in Section 2.4 and 3.1, respectively, in the seismic hazard and screening report of Reference [26]. The JAF Safe Shutdown Earthquake (SSE) and GMRS curves from Reference [26] are shown in Figure A4-1 in Attachment 4. The NRC's Staff assessment of the JAF seismic hazard and screening response is documented in Reference [25]. In the Staff Confirmatory Analysis (Section 3.3.3) of Reference [25], the NRC concluded that the methodology used by Exelon in determining the GMRS was acceptable and that the GMRS determined by Exelon adequately characterizes the reevaluated hazard for the JAF 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 JAF, the specific seismic reviews prepared by the licensee and the NRC's 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 [26] and [25]).

2. NTTF Recommendation 2.1 spent fuel pool assessment (References [27] and [28]).

3. NTTF Recommendation 2.3 seismic walkdowns (References [29] and [30]).

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

(References [31] and [32]).

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 24 The following additional post-Fukushima seismic reviews were performed for JAF:

5. NTTF Recommendation 2.1 Expedited Seismic Evaluation Process (ESEP)

(References [33] and [34]).

6. NTTF Recommendation 2.1 seismic High Frequency Evaluation (References [35] and

[36].

Summary Based on the above, the Summary from Section 2.3.3 of Reference [3] applies to JAF; namely, JAF is a Tier 2 plant for which there may be a limited number of unique seismic insights, most likely attributed to the possibility of seismically correlated failures, appropriate for consideration in determining HSS SSCs. References [5], [6], and [37]3 are incorporated into this LAR as they provide additional supporting bases for Tier 2 plants. In addition, References

[38], [39], and [40] are incorporated into this LAR as they provide additional supporting bases for Tier 1 plants that is also used for Tier 2 plants. The special sensitivity study recommended using common cause failures, similar to the approach taken in a FPIE PRA, can identify the appropriate seismic insights to be considered with the other categorization insights by the Integrated Decision-making Panel (IDP) for the final HSS determinations. Use of the EPRI approach outlined in Reference [3] to assess seismic hazard risk for §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).

3.2.4 Other External Hazards All external hazards, except for seismic, were screened for applicability to JAF per a plant-specific evaluation in accordance with GL 88-20 (Reference [41]) 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 JAF 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.

3 Excludes RAI APLC 50.69-RAI No. 12 that addresses a non-seismic topic (external events).

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 25 NUMARC 91-06 specifies that a defense-in-depth approach should be used with respect to each defined shutdown key safety function. The key safety functions defined in NUMARC 91-06 are evaluated for categorization of SSCs.

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

3.2.6 PRA Maintenance and Updates The Exelon risk management process ensures that the applicable PRA models used in this application continues to reflect the as-built and as-operated plant for JAF. 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, Exelon 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 those changes in the PRA model used for categorization.

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

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

In the overall risk sensitivity studies, Exelon 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, Exelon will perform both an initial sensitivity study and

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 26 a cumulative sensitivity study. The initial sensitivity study applies to the system that is being categorized. In the cumulative sensitivity study, the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in all identified PRA models for all systems that have been categorized are increased by a factor of 3. This sensitivity study together with the periodic review process assures that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed 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 [42]). 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.

Each PRA element notebook was reviewed for assumptions and sources of uncertainties. The characterization of assumptions and sources of uncertainties are based on whether the assumption and/or source of uncertainty is key to the 10 CFR 50.69 application in accordance with RG 1.200 Revision 2.

Key JAF 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 JAF PRA model specific assumptions or sources of uncertainty.

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 [43]), 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 [44)) as accepted by NRC in the letter dated May 3, 2017 (Reference [45]).

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

Full Power Internal Events and Internal Flooding PRA Model The JAF FPIE PRA model was peer reviewed in September 2009 using the NEI 05-04 process (Reference [46]). the PRA Standard (ASME/ANS RA-Sa-2009) (Reference [47], and Regulatory Guide 1.200, Revision 2 (Reference [43]). This Peer Review (Reference [48]) was a full-scope

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 27 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.

In November 2019, an F&O Closure Review (Reference [49]) was conducted for JAF. The scope of the review included the Internal Events and Internal Flooding PRA model. The F&O Independent Assessment Team closed twenty-three of twenty-four Finding level F&Os and the re-assessments for linked SRs were deferred to a concurrent focused-scope Peer Review. The concurrent focused-scope Peer Review (Reference [50]) was performed in November 2019 and the review team determined there were six Finding level F&Os resulting in three Not-Met SRs.

An additional F&O Closure by Independent Assessment (Reference [51]) was held in July 2020 to review close out of Finding level F&Os from the three prior PRA Peer Reviews against the PRA Standard. All seven Finding level F&Os from the FPIE PRA were assessed as closed during the review. Currently, there are no open Finding level F&Os against the FPIE PRA model.

Fire PRA Model The JAF Fire PRA (FPRA) Peer Review (Reference [52] was performed in January 2020 using the NEI 07-12 Fire PRA peer review process (Reference [53]), the ASME PRA Standard, ASME/ANS RA-Sa-2009 (Reference [47]), and Regulatory Guide 1.200, Rev. 2 (Reference

[43]). The FPRA Peer Review was a full-scope review of all technical elements of the JAF at-power FPRA against all technical elements in Part 4 of the ASME/ANS PRA Standard, including the referenced Internal Events Supporting Requirements (SRs) in Part 2.

The Fire PRA Peer Review team determined there were thirty-nine Finding level F&Os resulting in twenty-three Not-Met SRs. In July 2020, an F&O Closure by Independent Assessment (Reference [51]) was conducted for JAF. The scope of the review included FPIE and Fire PRA Peer Review Finding level F&Os. Thirty-three of the FPRA Finding level F&Os were assessed as closed during the review. In March 2021, a follow-on F&O Closure by Independent Assessment (Reference [54]) was conducted and the remaining open Findings were assessed as closed. Currently, there are no open Finding level F&Os against the FPRA model.

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

The JAF 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

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 28 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 JAF Tier 2 approach discussed in section 3.2.3, implementation of the Exelon 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 Exelons 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 Exelon 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 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.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 29

• 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.

Exelon 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 Exelon 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 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.

• 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

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 30 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.

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 3, January 2018.

• 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 Exelon 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.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 31 Exelon 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?

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.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 32

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin.

The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

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

Based on the above, Exelon 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.

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. 50-333 Page 33 6 REFERENCES

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

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[22] Southern Nuclear Operating Company, Inc. letter to NRC, "Vogtle Electric Generating Plant, Units 1 & 2, "License Amendment Request to Incorporate Seismic Probabilistic Risk Assessment into 10 CFR 50.69, Response to Request for Additional Information (RAIs 4-11)," February 21, 2018, ADAMS Accession No. ML18052B342.

[23] 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," November 29, 2018 ADAMS Accession No. ML18334A363.

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MF3725) February 18, 2016, ADAMS Accession No. ML16043A411.

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[26] Entergy letter to NRC, Seismic Hazard and Screening Report (Central Eastern United States (CEUS) Sites, "Response [to] NRC Request for Information (RFI) Pursuant to 10 CFR 50.54(f), Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF)

Review of Insights from the Fukushima Dai-ichi Accident," March 31, 2014, ADAMS Accession No. ML14090A243.

[27] Entergy Letter to NRC, Spent Fuel Pool Evaluation, 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," December 22, 2016 Adams Accession No. ML16357A786.

[28] NRC Letter to Entergy, James A. FitzPatrick Nuclear Power Plant, "Staff Review of Spent Fuel Pool Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1," (CAC NO. MF3725) March 22, 2017, Adams Accession No. ML17072A342.

[29] Entergy Letter to NRC, "Seismic Walkdown Report - Entergy's Response to NRC Request for Information (RFI) Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima-Dai-ichi Accident," November 27, 2012, ADAMS Accession Number ML123420188.

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(TAC NO. MF0125) April 1, 2014, ADAMS Accession No. ML14073A155.

[31] Exelon Generation Company, LLC, letter to NRC, James A. FitzPatrick Nuclear Power Plant, Renewed Facility Operating License No. DPR-59, NRC Docket No. 50-333, "Seismic Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - Nuclear Energy Institute (NEI) 12-06, Revision 4, Appendix H, H.4.4 Path 4: GMRS 2x SSE," December 15, 2017, ADAMS Accession No. ML17349A028.

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Letter," (CAC NO. MF7829; EPID L-2016-JLD-0006), April 30, 2018, ADAMS Accession No. ML18115A508.

License Amendment Request Enclosure Adopt 10 CFR 50.69 Docket No. 50-333 Page 37

[33] Entergy letter to NRC, "Entergys Expedited Seismic Evaluation Process Report (CEUS Sites), Response [to] NRC Request for Information Pursuant to 10 CFR 50.54(f)

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[35] Exelon Generation Company LLC letter to NRC, James A FitzPatrick Nuclear Power Plant, Renewed Facility Operating License No. DPR-059, NRC Docket No. 50-333, "High Frequency Confirmation Report for March 12, 2012, Information Request Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1, Seismic," August 30, 2017, ADAMS Accession No. ML17242A263.

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[37] Exelon Generation Company, LLC. Letter to NRC, LaSalle County Station, Units 1 and 2, Renewed Facility Operating License Nos. NPF-11 and NPF-18, NRC Docket Nos. 50-373 and 50-374, "Response to Request for Additional Information Regarding the License Amendment Request to Adopt 10 CFR 50.69 (EPID L-2020-LLA-0017)," October 1, 2020, ADAMS Accession Number ML20275A292.

[38] 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'," July 1, 2019, ADAMS Accession No. ML19183A012.

[39] 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'," July 19, 2019, ADAMS Accession No. ML19200A216.

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[40] 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,' July 19, 2019," August 5, 2019, ADAMS Accession No. ML19217A143.

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[50] James A FitzPatrick NPP, "Focused PRA Peer Review Report Using ASME/ANS PRA Standard Requirements," April 2020.

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License Amendment Request Attachment 1 Adopt 10 CFR 50.69 Docket No. 50-333 Page 42 Attachment 1: List of Categorization Prerequisites Exelon 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 guidelines 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. 50-333 Page 43 Attachment 2: Description of PRA Models Used in Categorization Unit Model Baseline CDF Baseline LERF Comments Full Power Internal Events (FPIE) and Internal Flooding PRA Model JF117A 2020 FPIE 1 Peer Reviewed 3.2E-06 5.7E-07 Application Against RG 1.200 R2 Specific Model in September 2009 Fire (FPRA) Model JF2017CF4 2021 Fire PRA Peer Reviewed 1 1.9E-05 3.9E-06 Application Against RG 1.200 R2 in January 2020 Specific Model

License Amendment Request Attachment 3 Adopt 10 CFR 50.69 Docket No. 50-333 Page 44 Attachment 3: Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC)

There are no open peer review findings or self-assessment open items

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 45 Attachment 4: External Hazards Screening4 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria An aircraft (either a Acceptance criterion 1.A of portion of (e.g., missile) Standard Review Plan 3.5.1.6 or the entire aircraft) (Reference [55]) states the that collides either probability is considered to be less directly or indirectly than an order of magnitude of 10-7 (i.e., skidding impact per year by inspection if the with one or more plant-to-airport distance D is structures, systems, or between 5 and 10 statute miles, components (SSCs) at and the projected annual number or in the plants of operations is less than 500 D2, analyzed area causing or the plant-to-airport distance D is functional failure. greater than 10 statute miles, and the projected annual number of Secondary hazards operations is less than 1000 D2 resulting from an (PS2, PS4).

PS2 aircraft impact include, but are not necessarily The closest airport to the plant is Aircraft Impact PS4 limited to, fire. the Oswego County Airport, a small, public, general aviation C3 facility located approximately 11 miles south of the plant. According to the Federal Aviation Administrations Air Traffic Activity System (Reference [56]), 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 4

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 [77]).

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 46 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria Administrations Air Traffic Activity System (Reference [56]), the annual operations from this airport is less than ~70,000, which is less than the 1000 D2 criteria (PS2, PS4).

Per IPEEE 5.5.1.3 (Reference [57]), JAF is more than 5 statute miles from the edge of the nearest military training route (C3).

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

Rapid flow of a large JAF is located on the southeast mass of accumulated shore of Lake Ontario, which frozen precipitation and precludes the possibility of an other debris down a avalanche.

sloped surface resulting in dynamic loading of Based on this review, the Avalanche C3 SSCs at or in the plants Avalanche hazard can be analyzed area causing considered to be negligible.

functional failure or adverse impact on natural water supplies used for heat rejection.

Accumulation or This hazard is slow to develop and deposition of vegetation can be identified via monitoring or organisms (e.g., and managed via standard zebra mussels, clams, maintenance process. Actions fish, algae) on an intake committed to and completed by Biological Events C5 structure or internal to a JAF in response to Generic Letter system that uses raw 89-13 provide on-going control of cooling water from a biological hazards. These controls source of surface water, are described in Exelon procedure ER-AA-340, GL 89-13 Program

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 47 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria causing its functional Implementing Procedure failure. (Reference [58]). Additional actions to reduce potential intake cooling water blockages were performed in response to WANO SOER 07-2 (Reference [59]).

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

Removal of material The lake water Intake Structure is from a shoreline of a a reinforced concrete structure body of water (e.g., setting on the lake bottom at a river, lake, ocean) due distance of approximately 900 feet to surface processes from the shoreline in (e.g., wave action, tidal approximately 25 feet of water.

currents, wave currents, The main structure is anchored to drainage, or winds and the natural bedrock below the lake including river bed bottom by post-tensioned tendons scouring) that results in (Reference [59]).

damage to the foundation of SSCs at C1 No major structures are directly on Coastal Erosion or in the plants the coastline, so coastal erosion is analyzed area, causing C5 not a significant concern at JAF.

functional failure. JAF has not witnessed a change in the coastline or in the lakebed levels in the area of the intake structure (Reference [59]) (C1).

Additionally, coastal erosion would slowly develop, so any effects could be mitigated (C5).

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

A shortage of surface Drought is a slowly developing Drought water supplies due to a C5 hazard allowing time for orderly period of below-average

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 48 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria precipitation in a given plant reductions, including region, thereby shutdowns.

depleting the water supply needed for the Based on this review, the Drought various water-cooling hazard can be considered to be functions at the facility. negligible.

An excess of water The FHRR identified both LIP outside the plant stillwater and PMF maximum boundary that causes water surface elevation (WSE) of functional failure to 272.8 ft. This is greater than the plant SSCs. External existing CDB controlling flood flood causes include, elevation of 262 ft and is slightly but may not be limited above site grade, which is to, flooding due to dam nominally 272 ft. In-leakage from failure, high tide, exterior doors that are normally hurricane (tropical closed was evaluated in cyclone), ice cover, JAF-RPT-14-00035 local intense (Reference [60]) to quantify the precipitation, river volume of water that could diversion, river and possibly enter any buildings from stream overflow, the 0.8 ft of ponding during either seiche, storm surge, of these flood events. The External Flood and tsunami. C1 conclusions were that in-leakage will be minimal and interior drainage features would have more than enough capacity to mitigate the effects of any in-leakage from the normally closed exterior doors. Procedure AOP-13 (Reference [61]) directs operators to verify water intrusion is not occurring at building outer doors and to close doors as appropriate if sustained local intense precipitation is occurring.

Otherwise, no human actions are required to mitigate the effects of a flooding event at the station.

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 49 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria There are several doors whose failure to be in their normal position would result in an unscreened scenario. These doors are a mixture of personnel doors, roll up doors, and vertical hatches that are required to be closed. Therefore, these doors and barriers should be categorized as HSS in accordance with NRC-approved guidance since removal of any of these doors would result in an unscreened scenario.

With credit taken for the normally positioned personnel doors, external flooding mechanisms are screened utilizing Criterion C1 and will not impact 10 CFR 50.69 categorization of other SSCs.

Strong winds resulting Based on the plant design for in dynamic loading or tornado wind pressure and the missile impacts on very low frequency (<1E-7/yr) of SCCs causing occurrence of design basis functional failure.

tornadoes, a demonstrably conservative estimate of CDF Hazards that could associated with high wind hazard potentially result in high (other than wind-generated Extreme Winds wind include the PS4 missiles) is much less than and Tornadoes following:

1E-6/yr. Therefore, all non-missile high wind hazards can be

• hurricane - severe screened from consideration for winds developed the 10 CFR 50.69 application, from a tropical based on Criterion C of Supporting depression resulting Requirement EXT-C1 of in missiles or ASME/ANS RA-Sa-2009 dynamic loading on (Reference [47]).

SSCs. Secondary

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 50 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria hazards resulting Based on a plant-specific tornado from a hurricane, missile risk analysis for JAF include, but are not (Reference [62]), the CDF for necessarily limited to tornado missiles is conservatively tornado estimated to be less than 1E-6/yr.

Therefore, tornado missile

• straight wind - a hazards can be screened from strong wind resulting consideration for the in missiles or 10 CFR 50.69 application, based dynamic loading on on Criterion C of Supporting SSCs that is not Requirement EXT-C1 of associated with either ASME/ANS RA-Sa-2009 hurricanes or (Reference [47]). There are no tornadoes vulnerabilities to tornado missiles at JAF that would specifically

• tornado - a strong affect containment integrity and whirlwind that results large early release probability.

in missiles or There are no SSCs credited in the dynamic loading on screening determination of high SSCs Fog Low-lying winds and tornado missile water vapor in the hazards, including passive and/or form of a cloud or active components, other than obscuring haze of Seismic Category I structures atmospheric dust or which are already considered high smoke resulting in safety significant (HSS) for impeded visibility that 10 CFR 50.69 categorization.

could result in, for example, a transportation accident.

Low-lying water vapor The principal effects of such in the form of a cloud or events (such as freezing fog) obscuring haze of would be to cause a loss of off-site atmospheric dust or power, which is addressed in Fog smoke resulting in C4 weather-related LOOP scenarios impeded visibility that in the FPIE PRA model for JAF.

could result in, for Based on this review, the Fog example, a hazard can be considered to be transportation accident. negligible.

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 51 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria Direct (e.g., thermal The UFSAR Section 2.1.1 refers effects) and indirect to the "NMP-JAF" site since the effects (e.g., generation Nine Mile Point (NMP) and JAF of combustion products, plants are essentially transport of firebrand) of geographically co-located.

a forest fire outside the (Reference [63]). The JAF IPEEE plant boundary that and UFSAR do not discuss this causes functional failure hazard in any great detail; of plant SSCs. however, the hazard is discussed in sufficient detail and was Hazards that could screened in the NMP2 cause or be caused by 10 CFR 50.69 application a forest fire include, but (Reference ADAMS Accession may not be limited to, No. ML19360A145) and confirmed wildfires and grass fires. in the NMP2 NRC Safety Evaluation (ADAMS Accession No. ML20332A115).

C3 Forest Fire Per the NMP2 USAR Section C4 2.2.3.1.4 (Reference [64]), the site is sufficiently cleared in areas adjacent to the plant that forest or brush fires pose no safety hazards. (C3).

In addition, external fires (Forest or Range Fires) originating from outside the plant boundary have the potential to cause a loss of offsite power event, which is addressed for grid-related LOOP scenarios in the FPIE PRA model for JAF (C4).

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

A thin layer of ice The principal effects of such Frost C4 crystals that form on the events would be to cause a loss of

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 52 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria ground or the surface of off-site power, which is addressed an earthbound object for weather-related LOOP when the temperature scenarios in the FPIE PRA model of the ground or surface for JAF.

of the object falls below freezing. Based on this review, the Frost hazard can be considered to be negligible.

A shower of ice or hard The principal effects of such snow that could result in events would be to cause a loss of transportation accidents off-site power, which is addressed or directly causes for weather-related LOOP dynamic loading or scenarios in the FPIE PRA model Hail C4 freezing conditions as a for JAF.

result of ice coverage.

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

Effects on SSC Per UFSAR Section 7.1.12 operation due to (Reference [63]), the plant is abnormally high designed for this hazard (C1).

ambient temperatures resulting from weather The principal effects of such phenomena. Secondary events would result in elevated hazards resulting from lake temperatures, which are C1 high ambient monitored by station personnel.

High Summer temperatures, include, Actions would be taken in C5 Temperature but are not necessarily response to elevated lake limited, to low lake or temperatures (C5).

C4 river water levels.

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).

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 53 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria Based on this review, the High Summer Temperature hazard can be considered to be negligible.

The periodic maximum The evaluation of the impact of the rise of sea level external flooding hazard at the site resulting from the was updated as a result of the combined effects of the NRC's post Fukushima 50.54(f) tidal gravitational forces Request for Information. The exerted by the Moon stations flood hazard reevaluation and Sun and the report (FHRR) was submitted to rotation of the Earth. NRC for review (Reference [65])

and NRC issued an interim staff response letter (Reference [66])

confirming the findings in the FHRR. The results indicate that all flood causing mechanisms, High Tide C1 except Local Intense Precipitation (LIP), flooding in streams and rivers (herein referred to as the PMF), and storm surge (herein referred to as Combined Effects flooding) are bounded by the current design basis (CDB) and do not pose a challenge to the plant.

Therefore, high tide, lake level, and river stage hazard impacts are considered negligible and can be screened (C1).

Flooding that results JAF is approximately 250 miles from the intense rain fall from the Atlantic Ocean. The risk from a hurricane associated with hurricane hazards (tropical cyclone). C3 do not need to be evaluated that Hurricane far inland (C3); see Supporting (Tropical Cyclone)

Secondary hazards C4 Requirement WHA-A2, Note (2) in resulting from a the ASME/ANS PRA Standard hurricane include, but (Reference [47]).

are not necessarily

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 54 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria limited to, dam failure, Wind and storm surge impacts on high tide, river and the plant due to hurricanes are stream overflow, bounded by the Extreme Wind /

seiche, storm surge, Tornado and External Flooding and waves. hazards (C4).

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

Flooding due to UFSAR Section 2.3.5 downstream blockages (Reference [63]) states that Lake of ice on a river. Ontario rarely, if ever, freezes over completely. However, ice floes Secondary hazards from the Niagara River and ball ice resulting from an ice and slush formed on the lake are blockage include, but driven by wind and frequently pack are not necessarily against the shoreline. Plant SSCs limited to, river and are designed for this hazard. The stream overflow. lake water Intake Structure is a reinforced concrete structure setting on the lake bottom at a distance of approximately 900 ft from the shoreline in C1 approximately 25 ft of water. The Ice Cover Screenhouse is approximately 150 C4 ft from the shore; this location 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 (Reference [59]). (C1)

A potential impact of ice cover on other SSCs is a loss of off-site power event, which is addressed for weather-related LOOP scenarios in the FPIE PRA model for JAF (C4).

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 55 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria Based on this review, the Ice Cover hazard can be considered to be negligible.

An accident at an offsite As described in IPEEE 5.2.3 and industrial or military 5.5.2.1, progressive screening facility that results in a steps were taken to assess release of toxic gases, hazardous chemical, a release of combustion transportation, and nearby products, a release of industrial facility incidents (there radioactivity, an are no military facilities within 5 explosion, or the miles of the site). No accident generation of missiles. scenarios were calculated to core damage frequency (CDF) exceeding 10-6/year (C3, PS4).

A similar analysis was performed to identify potential explosion incidents that could give rise to a Industrial or C3 1-psi overpressure at the plant. No Military Facility scenario resulted in an Accident PS4 overpressure greater than 1 psi or missile at the site, per Table 5.5.2.4 of the UFSAR (C3).

Table 2.1-4 of the UFSAR contains an updated list of facilities; these were either addressed in the IPEEE or are too far from the site to have an impact.

See also Toxic Gas.

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

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 56 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria Flooding that results The JAF Internal Events PRA from leaks or ruptures includes evaluation of risk from of liquid systems (e.g., internal flooding events.

Internal Flood tanks, pipes, valves, N/A pumps) originating inside the defined plant site boundary.

Dynamic loading of Plant site is located on level SSCs or impacts on terrain and is not subject to natural water supplies landslides.

used for heat rejection Landslide due to the movement of C3 Based on this review, the rock, soil, and mud Landslide hazard can be down a sloped surface considered to be negligible.

(does not include frozen precipitation).

Effects on SSCs due to Lightning strikes are not a sudden electrical uncommon in nuclear plant discharge from a cloud experience. They can result in to the ground or Earth- losses of off-site power or surges bound object. in instrumentation output if grounding is not fully effective.

The latter events often lead to reactor trips. Both events are Lightning C4 incorporated into the JAF 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.

A decrease in the water UFSAR 2.4.3.6 (Reference [63])

level of the lake or river defines the design minimum low Low Lake or River used for power C5 water level of Lake Ontario for JAF Water Level generation. as el. 236.5 ft. This is based on a minimum still water level of 240.6

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 57 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria and instantaneous lowering of 4.1 ft due to a maximum probable seiche on Lake Ontario. Historical low water levels are also provided; the lowest recorded since 1860 is el. 242.6 ft.

The UFSAR provides a description of the regulation of Lake Ontario by the International Lake Ontario-St. Lawrence River Board.

UFSAR Section 2.4.3.5 states the lake level is regulated within a range of approximately 242 to 249 ft. The basic water supply to the lake changes very slowly, permitting reasonably accurate forecasts and operating actions to maintain desired levels.

Section 2.4.3.6 of the UFSAR also describes the effect of the failure of the St. Lawrence Power Project, which consists of two dams and a hydroelectric power plant in the St.

Lawrence River (the outlet of Lake Ontario). In the unlikely event that all dams simultaneously fail, it is estimated that Lake Ontario water levels would decline gradually, with the full effect experienced about a year following the failure.

In this time the still water level could fall to a minimum at el.

240.6 ft.

Due to the normal regulation of lake level and the extended time available before minimum water

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 58 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria level in the event of dam failures downstream, this hazard can be screened based on Criterion C5.

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

Effects on SSC The principal effects of such operation due to events would be to cause a loss of abnormally low ambient off-site power. These effects temperatures resulting would take place slowly allowing from weather time for orderly plant reductions, phenomena. including shutdowns (C5). At C5 Low Winter worst, the loss of off-site power Temperature Secondary hazards events would be subsumed into C4 resulting from low the base PRA model results (C4).

ambient temperatures include, but are not Based on this review, the Low necessarily limited to, Winter Temperature hazard can frost, ice cover, and be considered to be negligible.

snow.

A release of energy due The frequency of a meteor or to the impact of a space satellite strike is judged to be so object such as a low as to make the risk impact meteoroid, comet, or from such events insignificant.

human-caused satellite falling within the Earths Based on this review, the atmosphere, a direct Meteorite or Satellite hazard can Meteorite/Satellite impact with the Earths be considered to be negligible.

PS4 Strikes surface, or a combination of these effects.

This hazard is analyzed with respect to direct impacts of an SSC and indirect impact effects

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 59 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria such as thermal effects (e.g., radiative heat transfer), overpressure effects, seismic effects, and the effects of ejecta resulting from a ground strike.

A release of hazardous IPEEE Section 5.5.2.2 material, a release of (Reference [57]) details screening combustion products, analysis of the natural gas pipeline an explosion, or the to the Independence Power generation of missiles Station, which is located about 3 due to an accident miles southwest of the plant.

involving the rupture of Explosions from this source would a pipeline carrying result in less than 1.0 psi hazardous materials. overpressure at the site (the screening damage criterion).

Pipeline Accident C3 Other natural gas pipelines, such as to the Alcan Rolled Products Company, are further removed and of smaller diameter and thus will also pose no major risk (Reference [57]).

See also Toxic Gas.

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

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 60 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria Flooding that results The FHRR identified local intense from local intense precipitation (LIP) stillwater precipitation.

maximum water surface elevation (WSE) of 272.8 ft. This is greater Secondary hazards than the existing CDB controlling resulting from local flood elevation of 262 ft and is intense precipitation, slightly above site grade, which is include, but are not nominally 272 ft. In-leakage from necessarily limited to, exterior doors that are normally dam failure and river closed was evaluated in and stream overflow.

JAF-RPT-14-00035 (Reference [60]) to quantify the volume of water that could possibly enter any buildings from the 0.8 ft of ponding during the LIP event.

The conclusions were that in-leakage will be minimal and Precipitation, interior drainage features would C1 Intense have more than enough capacity to mitigate the effects of any in-leakage from the normally closed exterior doors. Procedure AOP-13 (Reference [61]) directs operators to verify water intrusion is not occurring at building outer doors and to close doors as appropriate if sustained local intense precipitation is occurring.

Otherwise, no human actions are required to mitigate the effects of a flooding event at the station.

There are several doors whose failure to be in their normal position would result in an unscreened scenario. These doors are a mixture of personnel

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 61 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria doors, roll up doors, and vertical hatches that are required to be closed. Therefore, these doors and barriers should be categorized as HSS in accordance with NRC-approved guidance since removal of any of these doors would result in an unscreened scenario.

With credit taken for the normally positioned personnel doors, external flooding mechanisms are screened using Criterion C1 and will not impact 10 CFR 50.69 categorization of other SSCs.

See also External Flood.

A release of hazardous Per UFSAR Section 16.9.3.24 material including, but (Reference [63]), few chemicals not limited to liquids, which, when accidentally released, combustion products, or could pose safety hazard to the radioactivity. control room operators were evaluated using the VAPOR Such releases may be computer code. The results of this concurrent with or analysis have indicated that for the induce an explosion or conditions existing at the plant Release of the generation of C1 site, none of these chemicals Chemicals from missiles. could produce conditions which Onsite Storage PS4 would incapacitate control room In this context, an operators (C1). See also Toxic onsite release of Gas.

radioactivity is assumed to be associated with IPEEE Section 5.5.2.2 low-level radioactive (Reference [57]) details explosion waste. hazards from chemicals onsite, including hydrogen, propane, and carbon-dioxide. All the explosion hazards for chemicals stored

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 62 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria onsite are screened based on negligible damage impact from the explosion (C1), low frequency of occurrence (PS4), or low CDF (PS4).

See also Toxic Gas.

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

The redirection of all or Per UFSAR Section 2.4.1 a portion of river flow by (Reference [63]), there are no natural causes (e.g., a naturally occurring, perennial riverine embankment streams on the site, but drainage landslide) or ditches were constructed parallel intentionally (e.g., to the western and eastern power production, boundaries of the power block.

irrigation). Storm water runoff at the plant C3 River Diversion discharges to Lake Ontario via overland flow, intermittent streams, the drainage ditches, and/or the plants storm drain system.

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

Persistent heavy winds The plant is designed for such transporting sand or events. More common wind-borne dust that infiltrate SSCs dirt can occur but poses no at or in the plants significant risk to JAF given the analyzed area causing robust structures and protective Sandstorm C1 functional failure. features of the plant.

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

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 63 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria Flooding from water The evaluation of the impact of the displaced by an external flooding hazard at the site oscillation of the surface was updated as a result of the of a landlocked body of NRC's post Fukushima 50.54(f) water, such as a lake, Request for Information. The that can vary in period stations flood hazard reevaluation from minutes to several report (FHRR) was submitted to hours. NRC for review (Reference [65])

and NRC issued an interim staff response letter for the station (Reference [66]) confirming the Seiche C1 findings in the FHRR. The results indicate that all flood causing mechanisms, except LIP, PMF, and Combined Effects flooding are bounded by the current design basis (CDB) and do not pose a challenge to the plant.

Based on this review (full details in Section 2.2), the Seiche hazard can be considered to be negligible.

Sudden ground motion See Section 3.2.3 and Figure A4-1 or vibration of the Earth in this Attachment.

as produced by a rapid release of stored-up energy along an active fault.

Secondary hazards resulting from seismic Seismic Activity N/A activity include, but are not necessarily limited to, avalanche (both rock and snow), dam failure, industrial accidents, landslide, seiche, tsunami, and vehicle accidents.

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 64 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria The accumulation of This hazard is slow to develop and snow could result in can be identified via monitoring transportation accidents and managed via normal plant or directly cause processes. Potential flooding dynamic loading or impacts are accounted for in the Snow C5 freezing conditions as a External Flooding screening.

result of snow cover.

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

Dynamic forces on The potential for this hazard is low structures foundations at the site, the plant design due to the expansion considers this hazard (C1), and (swelling) and the hazard is slow to develop and C1 contraction (shrinking) can be mitigated (C5).

Soil Shrink-Swell of soil resulting from C5 changes in the soil Based on this review, the Soil moisture content. Shrink-Swell Consolidation impact hazard can be considered to be negligible.

Flooding that results The evaluation of the impact of the from an abnormal rise external flooding hazard at the site in sea level due to was updated as a result of the atmospheric pressure NRC's post Fukushima 50.54(f) changes and strong Request for Information. The wind generally stations flood hazard reevaluation accompanied by an report (FHRR) was submitted to intense storm. NRC for review (Reference [65])

and NRC issued an interim staff Storm Surge C1 Secondary hazards response letter (Reference [66])

resulting from a storm confirming the findings in the surge include, but are FHRR. The results indicate that not necessarily limited all flood causing mechanisms, to, high tide, river and except LIP, PMF, and Combined stream overflow, and Effects flooding are bounded by waves. the current design basis (CDB) and do not pose a challenge to the plant.

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 65 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria Based on this review (full details in Section 2.2), the Storm Surge can be considered to be negligible.

A release of hazardous IPEEE Sections 5.5.1.2 and toxic or asphyxiant 5.5.2.1 (Reference [57]) discuss gases. toxic chemical control room habitability studies. The toxic Such releases may be chemical release scenarios concurrent with or considered were from on-site induce an explosion or chemical storage and offsite the generation of industrial facilities. The only missiles. shipments of potentially hazardous material transported within five In this context, an miles of the site were those onsite release of shipped to or from facilities within radioactivity is assumed that distance, including the site.

to be associated with Toxic Gas C1 low-level radioactive The calculations performed for the waste. worst scenarios all concluded that the maximum predicted control room concentrations of toxic chemicals were all below the toxicity acceptance criteria.

Therefore, there are no toxic gas hazards that require either gas detectors or automatic isolation of the control room.

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

Accidents involving IPEEE Section 5.5.2.1 transportation resulting C3 (Reference [57]), shipments of in collision with SSCs, a hazardous material by boat, rail, Transportation release of hazardous C1 and highway (except to or from Accidents materials or combustion local facilities) are not sources of products, an explosion, C4 concern because the shipping or a generation of lane nearest to JAF is seven miles

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 66 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria missiles causing from the plant and serves the port functional failure of of Oswego (this port mostly SSCs. handles potash and urea; hazardous chemicals are not Hazards that could routinely handled at the port that potentially result in lies nine miles from the plant);

transportation accidents Conrail states that no hazardous include, for example, a chemicals are transported within vehicle, railcar or ship five miles of the plant; and (boat) accident that shipments of hazardous material involves a collision or by road, other than to or from local derailment, potentially facilities, must use interstate resulting in fire, highways (C3).

explosions, toxic releases, missiles, or Other releases of toxic gases other hazardous during transportation to local conditions. facilities are included in the discussion of the Toxic Gas hazard screening (C1, C4).

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

Flooding that results The location of JAF along Lake from a series of long- Ontario precludes the possibility of period sea waves that a tsunami.

displaces massive amounts of water as a result of an impulsive disturbance, such as a Tsunami C3 major submarine slides or landslide.

Secondary hazards resulting from a tsunami include, but are not necessarily limited to,

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 67 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria river and stream overflow.

Damage to safety- Per UFSAR Section 10.2.4 related SSCs from a (Reference [63]), the probability of missile generated turbine missile generation has internal or external to been evaluated and demonstrated the plant PRA boundary to be in conformance with the from rotating turbines or NRC acceptance criterion of other external sources 1x10-4/yr for favorably oriented (e.g., high-pressure gas turbines such as the machine at cylinders). JAF. For the monoblock rotors Damage may result used at JAF, missile generation from a falling missile or probability is based on the a missile ejected probability of a control system directly toward safety- failure that results in an overspeed related SSCs (i.e., low- event in which turbine speed trajectory missiles). exceeds 120% of the normal (rated) operating speed along with demonstration of margin to ductile failure limits for rotating Turbine-Generated PS4 components. The probability of Missiles exceeding 120% of normal operating speed was calculated to be 4.35x10-5/yr. Per Regulatory Guide 1.115 Reference [67], this meets the acceptance criteria because the probability of turbine missile being generated and ejected from the casing is

<1E-4/yr; since the probability of striking and damaging a safety-related (SR) component is 1E-3 (per Regulatory Guide 1.115), the frequency of damaging SR component is less than 1E-7/yr, so CDF is less than 1E-6/yr. Therefore, this hazard can be screened.

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 68 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria Based on this review, the Turbine-Generated Missiles hazard can be considered to be negligible.

Opening of Earths crust This hazard is not applicable to resulting in tephra (i.e., the site because of location (no rock fragments and active or dormant volcanoes particles ejected by located near plant site).

volcanic eruption), lava flows, lahars (i.e., mud Based on this review, the Volcanic flows down volcano Activity hazard can be considered slopes), volcanic gases, to be negligible.

pyroclastic flows (i.e.,

fast-moving flow of hot gas and volcanic matter moving down and away from a volcano), and Volcanic Activity landslides. C3 Indirect impacts include distant ash fallout (e.g.,

tens to potentially thousands of miles away).

Secondary hazards resulting from volcanic activity, include, but are not necessarily limited to, seismic activity and fire.

An area of moving The evaluation of the impact of the water that is raised external flooding hazard at the site above the main surface C1 was updated as a result of the Waves of a body of water as a NRC's post Fukushima 50.54(f) result of the wind C4 Request for Information. The blowing over an area of stations flood hazard reevaluation fluid surface. report (FHRR) was submitted to

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 69 Screening Hazard Definition Disposition for 10 CFR 50.69 Criteria NRC for review (Reference [65]),

and NRC issued an interim staff response letter (Reference [66])

confirming the findings in the FHRR. The results indicate that all flood causing mechanisms, except LIP, PMF, and Combined Effects flooding are bounded by the current design basis (CDB) and do not pose a challenge to the plant (C1). Waves are considered in screening of the External Flooding hazard (C4).

Based on this review, the impacts from waves hazard can be considered to be negligible and can be screened.

License Amendment Request Attachment 4 Adopt 10 CFR 50.69 Docket No. 50-333 Page 70 Figure A4-1: GMRS and SSE Response Spectra for JAF (From Reference [26]).

License Amendment Request Attachment 5 Adopt 10 CFR 50.69 Docket No. 50-333 Page 71 Attachment 5: Progressive Screening Approach for Addressing External Hazards Event Analysis Criterion Source Comments Initial Preliminary C1. Event damage potential NUREG/CR-2300 and Screening is < events for which plant is ASME/ANS Standard designed. RA-Sa-2009 C2. Event has lower mean NUREG/CR-2300 and frequency and no worse ASME/ANS Standard consequences than other RA-Sa-2009 events analyzed.

C3. Event cannot occur NUREG/CR-2300 and close enough to the plant to ASME/ANS Standard affect it. RA-Sa-2009 C4. Event is included in the NUREG/CR-2300 and definition of another event. ASME/ANS Standard RA-Sa-2009 C5. Event develops slowly, ASME/ANS Standard allowing adequate time to RA-Sa-2009 eliminate or mitigate the threat.

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

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

License Amendment Request Attachment 5 Adopt 10 CFR 50.69 Docket No. 50-333 Page 72 Event Analysis Criterion Source Comments PS3. Design basis event NUREG-1407 as mean frequency is < 1E-5/y modified in ASME/ANS and the mean conditional Standard RA-Sa-2009 core damage probability is <

0.1.

PS4. Bounding mean CDF NUREG-1407 and is < 1E-6/y. ASME/ANS Standard RA-Sa-2009 Detailed PRA Screening not successful. NUREG-1407 and PRA needs to meet ASME/ANS Standard requirements in the RA-Sa-2009 ASME/ANS PRA Standard.

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 73 Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty Assessment of Internal Events PRA Epistemic Uncertainty Impacts In order to identify key sources of uncertainty, the Internal Events baseline PRA model uncertainty report was developed, based on the guidance in NUREG-1855 (Reference [68]) and EPRI 1016737 (Reference [42]). As described in NUREG-1855, sources of uncertainty include parametric uncertainties, modeling uncertainties, and completeness (or scope and level of detail) uncertainties.

Parametric uncertainty was addressed as part of the JAF baseline PRA model quantification.

The parametric uncertainty evaluation for the Internal Events PRA model is documented in Appendix B of the Summary Notebook (Reference [69]).

Modeling uncertainties are considered in both the base PRA and in specific risk-informed applications. Assumptions are made during the PRA development as a way to address a particular modeling uncertainty because there is not a single definitive approach. Plant-specific assumptions made for each of the JAF Internal Events PRA technical elements are noted in the Summary Notebook (Reference [69]). The Internal Events PRA model uncertainties evaluation considers the modeling uncertainties for the base PRA by identifying assumptions, determining if those assumptions are related to a source of modeling uncertainty and characterizing that uncertainty, as necessary. The Electric Power Research Institute (EPRI) compiled a listing of generic sources of modeling uncertainty to be considered for each PRA technical element (Reference [42]), and the evaluation performed for JAF considered each of the generic sources of modeling uncertainty as well as the plant-specific sources.

Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the PRA but are only considered for their impact on a specific application. No specific issues of PRA completeness have been identified relative to the 10 CFR 50.69 application, based on the results of the Internal Events PRA and Fire PRA peer reviews.

The impact of potential sources of uncertainty for the Internal Events model on the PRA or applications is discussed in the table below.

Note: As part of the required 10 CFR 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. 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 10 CFR 50.69 application.

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 74 Disposition of IE/IF PRA Key Assumptions/Sources of Uncertainty IE / IF PRA IE / IF PRA IE / IF PRA Sources of Assumption/ Model Sensitivity and 10 CFR 50.69 Impact Uncertainty Disposition (10 CFR 50.69)

Conditional probability that The ability to prevent melt- A sensitivity study was performed drywell shell melt-through through of the steel containment which demonstrated that the FPIE occurs despite the shell is dependent upon the and FPRA LERF (using the availability of water injection availability of water to cool the average maintenance PRA to the debris. core debris. For cases in which models) are not sensitive to the there is limited or no water failure probability associated with available to the debris, the drywell shell melt-through.

containment is modeled as likely to fail.

Relay Room Internal The operator action for The results demonstrate that FPIE Flooding scenarios Operator Fails to Perform CDF and LERF are highly conservatively assume Shutdown Outside Control sensitive to an increased complete failure of all Room for Relay Room Flood probability that operators fail to equipment if water level was both increased and perform plant shutdown following reaches 9. In reality, decreased by one decade in a Relay Room flood; however, this equipment located higher in separate quantifications. This increased probability would not be panels may survive and operator action is not highly realistic and would likely mask key support one or more uncertain, but was judged risk insights.

success paths. reasonable to use for the sensitivity because it is included In addition, as part of the required in various model logic elements 10 CFR 50.69 PRA categorization for Relay Room Internal sensitivity cases directed by NEI Flooding. 00 04, internal events / internal flood modeled human error basic events are increased to their 95th percentile and also decreased to their 5th percentile values. These results are capable of driving a component and respective functions HSS and therefore the uncertainty of the PRA modeled HEPs are accounted for in the 10 CFR 50.69 application.

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 75 IE / IF PRA IE / IF PRA IE / IF PRA Sources of Assumption/ Model Sensitivity and 10 CFR 50.69 Impact Uncertainty Disposition (10 CFR 50.69)

Conditional LOOP The probabilities that a The results demonstrate that the probabilities are based on a conditional LOOP occurs after a conditional LOOP probabilities do relatively few events with transient and after a LOCA are not significantly impact the overall some partial failure increased and reduced by a average maintenance FPIE and experience, which could be factor of 2 in separate FPRA results.

biased toward higher failure quantifications.

probabilities.

The likelihood of The probability that containment The results demonstrate that the containment failure below failure occurs below the DW probability of containment failure drywell (DW) head head elevation is both increased occurring below DW head elevation for CRD injection and reduced in separate elevation does not significantly capability is potentially non- quantifications. impact the overall average conservative given the maintenance PRA results.

unknown phenomenological events associated with containment venting that would affect equipment survivability.

Spurious opening of circuit Slight conservatism considering The results demonstrate that breakers is modeled as a potential impact of recovery of credit for equipment repair and credible means for de- spurious circuit breaker recovery does not impact the energization of credited operation. The failure rates for overall average maintenance PRA equipment; however, this spurious operation of circuit results.

failure mode is among the breakers were reduced by one potentially more decade to simulate a chance for recoverable because such successful recovery.

breakers often reclose successfully when manipulated by operators.

Little credit is taken for the The PRA currently takes little The results demonstrate that FPIE Reactor Building as an credit (e.g., failure rate = 0.99) and FPRA LERF are highly effective means of fission for secondary containment sensitive to changes in Reactor product retention due to the fission product retention Building effectiveness. However, reliance on the Standby capability to prevent a large and this sensitivity analysis assumes Gas Treatment System and early release. The probabilities the Reactor Building fails with a potential for hydrogen that the Reactor Building is low probability, which is not

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 76 IE / IF PRA IE / IF PRA IE / IF PRA Sources of Assumption/ Model Sensitivity and 10 CFR 50.69 Impact Uncertainty Disposition (10 CFR 50.69) deflagration during severe ineffective given DW or torus considered realistic in the PRA accidents. airspace failures, drywell shell due to significant challenges in melt-through, and torus below achieving Reactor Building fission water line failure were each product retention during a severe adjusted to 0.1. accident that would potentially lead to a large and early release.

There remains little justifiable basis to credit the Reactor Building for significant attenuation of releases.

The likelihood that SRVs The current model provides the The results demonstrate that the fail open in a severe most realistic assessment depressurization failure accident is credited defensible, but does rely on probabilities have a slight impact somewhat conservatively in judgment because experience on FPIE and FPRA LERF. The the PRA due to limitations with SRV response in severe lower bound probabilities for on experience with SRV accidents is limited and failure to depressurize are difficult response in severe phenomenological analysis is to assess as realistic due to accidents and likewise limited. A value of limited experience with SRV phenomenological analysis. 0.248 is used in the PRA and, response in severe accidents.

while realistic, is likely some degree of conservative. The As part of the required probabilities for failure to 10 CFR 50.69 PRA categorization depressurize the RPV in Class sensitivity cases directed by NEI IA and IBE sequences were 00 04, internal events / internal adjusted to 0.1. 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. 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 10 CFR 50.69 application

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 77 IE / IF PRA IE / IF PRA IE / IF PRA Sources of Assumption/ Model Sensitivity and 10 CFR 50.69 Impact Uncertainty Disposition (10 CFR 50.69)

FLEX equipment failure FLEX equipment is not yet A sensitivity study was performed probabilities are identified explicitly modeled in the JAF which demonstrated that FLEX as a candidate source of PRA; however, there are no equipment failure does not model uncertainty since industry sources to provide significantly impact the overall there are no industry- FLEX equipment reliability data. average maintenance PRA approved data sources for The PRA currently models the results.

FLEX equipment reliability. FLEX diesel generator with an HEP-based basic event based on human action being taken as the dominant contributor for FLEX generator failure probability.

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 78 Assessment of Supplementary FPRA Epistemic Uncertainty Impacts The purpose of the following discussion is to address the epistemic uncertainty in the JAF FPRA. The FPRA model includes various sources of uncertainty that exist because there is both inherent randomness in elements that comprise the FPRA, and because the state of knowledge in these elements continues to evolve. The development of the FPRA was guided by NUREG/CR-6850 (Reference [70]). The FPRA model used consensus models described in NUREG/CR-6850. In order to identify key sources of uncertainty for the 10 CFR 50.69 Program Application, an evaluation of Fire PRA model uncertainty was performed, based on the guidance in NUREG-1855 (Reference [68]) and Electric Power Research Institute (EPRI) report 1026511 (Reference [71]).

As stated in Section 1.3 of NUREG-1855:

Although the guidance in the this [sic] report does not currently address all sources of uncertainty, the guidance provided on the uncertainty identification and characterization process and on the process of factoring the results into the decision making is generic and independent of the specific source of uncertainty. Consequently, the guidance is applicable for sources of uncertainty in PRAs that address at-power and low power and shutdown operating conditions, and both internal and external hazards.

NUREG-1855 also describes an approach for addressing sources of model uncertainty and related assumptions. It states:

A source of model uncertainty exists when (1) a credible assumption (decision or judgment) is made regarding the choice of the data, approach, or model used to address an issue because there is no consensus and (2) the choice of alternative data, approaches or models is known to have an impact on the PRA model and results. An impact on the PRA model could include the introduction of a new basic event, changes to basic event probabilities, change in success criteria, or introduction of a new initiating event. A credible assumption is one submitted by relevant experts and which has a sound technical basis. Relevant experts include those individuals with explicit knowledge and experience for the given issue. An example of an assumption related to a source of model uncertainty is battery depletion time. In calculating the depletion time, the analyst may not have any data on the time required to shed loads and thus may assume (based on analyses) that the operator is able to shed certain electrical loads in a specified time.

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 79 Section 2.1.3 of NUREG-1855 defines consensus model as:

A consensus model is a model that has a publicly available published basis and has been peer reviewed and widely adopted by an appropriate stakeholder group. In addition, widely accepted PRA practices may be regarded as consensus models.

Examples of the latter include the use of the constant probability of failure on demand model for standby components and the Poisson model for initiating events. For risk-informed regulatory decisions, the consensus model approach is one that the NRC has used or accepted for the specific risk-informed application for which it is proposed.

Modeling uncertainties are considered in both the base Fire PRA and in specific risk-informed applications. Assumptions are made during the Fire PRA development as a way to address a particular modeling uncertainty because there is not a single definitive approach. Plant-specific assumptions made for each of the JAF Fire PRA technical elements are noted in the Uncertainty and Sensitivity Analysis Notebook (Reference [72]).

Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the Fire PRA, but are only considered for their impact on a specific application (Reference [73]). No specific issues of PRA completeness have been identified relative to the 10 CFR 50.69 application, based on the results of the Internal Events PRA and Fire PRA peer reviews.

EPRI compiled a listing of generic sources of modeling uncertainty to be considered for each Fire PRA technical element in EPRI report 1026511 (Reference [71]). Based on following the methodology in EPRI 1026511 for a review of sources of uncertainty, the impact of potential sources of uncertainty on the PRA or applications was performed.

The table below describes the fire PRA sources of model uncertainty and their impact for the 10 CFR 50.69 Application relative to each task of NUREG/CR-6850 (Reference [70]).

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 80 Disposition of FPRA Key Assumptions/Sources of Uncertainty Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty Plant Boundary No plant specific items relative Consistent with NUREG/CR-6850.

Definition and to plant partitioning. The results of the plant partitioning Partitioning task are evaluated via fire scenarios and the multi-compartment analysis.

This does not represent a key source of uncertainty for the JAF 10 CFR 50.69 Application.

Equipment No plant specific items relative Uncertainty exists in the assumed Component to Equipment Selection. The plant trip and the scope of equipment Selection Fire PRA assumes at a credited in the Fire PRA. Sensitivity minimum a plant trip. This is evaluation is performed for consistent with accepted equipment or systems not credited in industry practice. The Fire the Fire PRA. There is no current PRA does not credit certain accepted industry guidance to FPIE equipment or systems. evaluate the plant trip probability.

Systems are included based This does not represent a key source on an iterative process to of uncertainty for the JAF include equipment that may be 10 CFR 50.69 Application.

significant to the fire risk.

Cable Selection No plant specific items relative Uncertainty exists for cables selected to Cable Selection. The cable specifically for the Fire PRA and have selection task for JAF includes missing cable routing location data.

accepted industry practice for Fire PRA assigns routing locations circuit analysis. Fire PRA based on known routing locations.

cables identified without known Therefore, this is not identified as a routing locations are assigned key source of uncertainty. Also, the routing locations based on the Fire PRA includes a conservative available data. Therefore, the treatment of the lack of coordination.

uncertainty related to these This treatment is refined as needed routings is small. The per fire scenario. This does not assumptions related to the represent a key source of uncertainty modeling of lack of coordination for the JAF 10 CFR 50.69 may be conservative. Lastly, Application.

certain equipment and systems were not credited in the Fire PRA.

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 81 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty Qualitative No plant specific items relative This does not represent a key source Screening to Qualitative Screening. The of uncertainty for the JAF plant partitioning, equipment 10 CFR 50.69 Application.

selection, and cable selection task are performed to ensure with confidence the Fire PRA does not screen plant locations that may contribute to the fire risk.

Plant Response No plant specific items relative The Internal Events PRA identifies Modell to Plant Response Model. The plant specific uncertainties which are Internal Events and Fire PRA also applicable to the Fire PRA and use the same modeling and fire should be included for applications, induced failures are added. as applicable. These uncertainties The Fire PRA uses the same include HRA and CCF treatment.

data used in the FPIE model. Other uncertainties may be Fire induced failures use applicable; for example, the amount accepted industry guidance. of credit for injection sources.

The Fire PRA models the Additionally, the uncertainties MCRAB sequence consistent associated with the basic event data with the equipment available for are identified in the JAF parametric remote shutdown and industry uncertainty analysis. This does not guidance for remote shutdown represent a key source of uncertainty operator actions. for the JAF 10 CFR 50.69 Application.

Fire Ignition No plant specific items relative The generic fire frequencies have Frequency to Fire Ignition Frequency. The been identified as the main source of key uncertainties related to fire uncertainty in this task. NUREG-ignition frequency remain 2169 has collected additional data related to the generic industry which has improved some of these data accepted for use. issues, as supplemented by subsequent guidance. However, the same uncertainty issues exist in NUREG-2169 and thus the characterization and treatment of uncertainty in the Fire PRA is the same as NUREG/CR-6850. In summary, the generic frequencies

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 82 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty are estimated and uncertainty bounds are provided. These uncertainty bounds account for the inherent randomness of the occurrence of fire events, and the variability among the plants. Fire ignition frequency distributions are propagated in the JAF parametric uncertainty analysis.

This does not represent a key source of uncertainty for the JAF 10 CFR 50.69 Application.

Quantitative No plant specific items relative Quantified Physical Analysis Units Screening to Quantitative Screening. scenarios were maintained. This does not represent a key source of uncertainty for the JAF 10 CFR 50.69 Application.

Fire Scoping No plant specific items relative This does not represent a key source Modeling to Fire Scoping Model. of uncertainty for the JAF 10 CFR 50.69 Application.

Detailed Circuit See Fire PRA Cable Selection See Fire PRA Cable Selection Failure Analysis Task of this table. Task of this table.

Circuit Failure No plant specific items relative The uncertainties are related to the Model Likelihood to Circuit Failure Mode specific configuration of the Analysis Likelihood Analysis. component cables being analyzed and the generic probabilities applied.

Circuit failure probabilities are propagated in the parametric uncertainty evaluation. This does not represent a key source of uncertainty for the JAF 10 CFR 50.69 Application.

Fire Scenario No plant specific items relative The fire scenario selection and fire Selection and Fire to Fire Scenario Selection. The modeling is iterative such that the fire Modeling fire scenario selection and fire risk contribution from fire ignition models use accepted industry sources and PAUs is reasonable and guidance. Fire scenarios are realistically characterized based on

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 83 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty generally developed in iterative the accepted guidance, methods, and stages. A conservative JAF specific characteristics. Plant definition is initially applied and specific inputs are used when the range of accepted fire available and otherwise modeling methods and tools supplemented by industry accepted are iteratively applied such that data. The fire scenario SF and NSP the resulting fire risk uncertainties are estimated in the contributors are reasonable parametric uncertainty evaluation.

based on the accepted This does not represent a key source methods. of uncertainty for the JAF 10 CFR 50.69 Application.

Human Reliability Plant specific key assumptions The Fire PRA includes conservative Analysis include conservative adjustments to the HFEs to account adjustments made to account for adverse impacts of fire events.

for fire events. Additionally, a The Fire PRA does not include credit JHEP floor value of 1E-6 is for all operator actions, including fire applied. Assumptions related response actions. The Fire PRA to fire event impacts on does not include credit for all of the operator actions are instrument cues that may be conservatively applied and available. A minimum joint HEP was refined as needed with operator applied for the HRA dependency training personnel input. A analysis. Applying a minimum joint JHEP floor value is used and is HEP may skew the results by considered to be good practice. artificially increasing the risk due to human actions. The HEPs are propagated in the parametric uncertainty evaluation based on the uncertainty parameters from the HRAC.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 basic events are increased to their 95th percentile and also decreased to their 5th percentile values. These results are capable of driving a component and respective functions HSS and therefore, the

License Amendment Request Attachment 6 Adopt 10 CFR 50.69 Docket No. 50-333 Page 84 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty uncertainty of the PRA modeled HEPs are accounted for in the 50.69 Application. This does not represent a key source of uncertainty for the JAF 10 CFR 50.69 Application.

Seismic-Fire No plant specific items relative This does not represent a key source Interactions to Seismic Fire. of uncertainty for the JAF Assessment 10 CFR 50.69 Application.

Fire Risk No plant specific items relative Technical reviews address the Quantification to Fire Quantification. The Fire uncertainties identified in each task to PRA is quantified using confirm adequacy for estimating the accepted industry codes and fire risk. Truncation study used to practices. The fire ensure risk contributors were not quantification task does identify inadvertently omitted from the results.

truncation limit as a source of This does not represent a key source uncertainty; however, this was of uncertainty for the JAF investigated in truncation 10 CFR 50.69 Application.

studies to ensure an appropriate truncation limit was used.