Text

Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors
	ML20031E699
Person / Time
Site:	LaSalle
Issue date:	01/31/2020
From:	Demetrius Murray; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-20-011
	Download: ML20031E699 (82)
	v • d • e

Exelon Generation 4300 Winfield Road Warrenville, IL 60555 www.exeloncorp.com 10 CFR 50.90 10 CFR 50.69 RS-20-011 January 31, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 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

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 (EGC) is requesting an amendment to the license to Renewed Facility Operating License Nos. NPF-11 and NPF-18 for LaSalle County Station, Units 1 and 2.

The proposed amendment would modify 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.

The enclosure to this letter provides the basis for the proposed change to LaSalle County Station (LSCS), Units 1 and 2, Renewed Facility Operating Licenses. 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. 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 EGC submittal of the LAR dated January 31 , 2020 for Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) (RS 009). EGC 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 EGC and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action, as the

January 31, 2020 U.S. Nuclear Regulatory Commission Page 2 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.

EGC requests approval of the proposed license amendment by January 31, 2021, with the amendment being implemented within 60 days.

These proposed changes have been reviewed and approved by the LSCS Plant Operations Review Committee in accordance with the requirements of the EGC Quality Assurance Program.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (a)(1 ), the analysis about the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is being provided to the Commission.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), EGC is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

There are no regulatory commitments contained within this letter.

Should you have any questions concerning this letter, please contact Ryan Sprengel at (630) 657-2814.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 31 51 day of January 2020.

Respectfully,

Enclosure:

Evaluation of the Proposed Change cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector - LaSalle County Station Illinois Emergency Management Agency - Division of Nuclear Safety NRR Project Manager, LaSalle County Station

Enclosure Evaluation of the Proposed Change Table of Contents 1

SUMMARY

DESCRIPTION ............... ..... ................ ..... ................ ..... ..................... ..... .......... 1 2 DETAILED DESCRIPTION ................................................................................................... 1 2 .1 CURRENT REGULATORY REQUIREMENTS .... .. .. .. .. .. .. .. .. .. ................ .. .. .. .. .. .. .. .. .. .. .. 1 2.2 REASON FOR PROPOSED CHANGE ........................................................................ 1 2 .3 DESCRIPTION OF THE PROPOSED CHANGE .. .. .. .. ................ .. .. .. .. .. .. .. .. .. .. ............ . 2 3 TECHNICAL EVALUATION .................................................................................................. 4 3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(b)(2)(i)) ................... .4 3.1.1 Overall Categorization Process ... .. .. .. ... .. .. .. .. .. .. .. .. ... .. .. .. .. .. .. .. ... .. .. .. .. .. .. .. ... .. .. . 4 3.1.2 Passive Categorization Process ....................................................... ..... ....... 11 3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii)) ............................ 11 3.2 .1 Internal Events and Internal Flooding .... .. .. .. .. .. .. .. .. .. ................ .. .. .. .. .. .. .. .. .. .. . 12 3.2 .2 Fire Hazards .. ................... .. ............... .. ............... .. ............... .. ................. ..... . 12 3.2.3 Seismic Hazards .................... ..... ................ ..... ................ ..... ....................... 12 3.2.4 Other External Hazards ......... ................. ................. ................. ................... . 20 3.2 .5 Low Power & Shutdown .......... .. .. .. .. .. .. .. .. .. .................. .. .. .. .. .. .. .. .. .. .............. . 20 3.2.6 PRA Maintenance and Updates ................................................................... 20 3.2.7 PRA Uncertainty Evaluations ....................................................................... 20 3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii)) ................ .. .. .. .. .. .. .. .. . 22 3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv)) ......................................................... 23 3.5 FEEDBACK AND ADJUSTMENT PROCESS ...... ..... ..................... ..... ................ ....... 23 4 REGULATORY EVALUATION ................ .. ............... .. ............... .. ............... .. ................. ..... 26 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA .. .. .. .. .. .. .................. .. .. .. 26 4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS ................................... 26

4.3 CONCLUSION

S ......................................................................................................... 28 5 ENVIRONMENTAL CONSIDERATION .... ................... .. ............... .. ............... .. ............... .. .. 29 6 REFERENCES ................................................................................................................... 30

Enclosure Evaluation of the Proposed Change LIST OF ATTACHMENTS : List of Categorization Prerequisites .................................................................... 35 : Description of PRA Models Used in Categorization ............................................ 36 : Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items ..................................................................................................... 37 : External Hazards Screening ......................................... ..................... ..... ............ .44 : Progressive Screening Approach for Addressing External Hazards ............................................................................................ 60 : Disposition of Key Assumptions/Sources of Uncertainty ... .. .. ... .. .. .. .. .. .. .. ... .. .. .. .. .. 62 : Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-1111, CC-11/111, and CC-1/11/111 .................................................................................... 78

Enclosure Evaluation of the Proposed Change 1

SUMMARY

DESCRIPTION The proposed amendment modifies the licensing basis to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50 .69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance (LSS), alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance (HSS), requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

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

This deterministic approach establishes requirements for engineering margin and quality assurance in design, manufacture, and construction. In addition, it assumes that adverse conditions can exist (e.g., equipment failures and human errors) and establishes a specific set of design basis events (DBEs). The deterministic approach then requires that the facility include safety systems capable of preventing or mitigating the consequences of those DB Es 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.

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 Page 1

Enclosure Evaluation of the Proposed Change 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-i nformed approach to treatment that provides reasonable, though reduced, level of confidence that these SSCs will satisfy functional requirements.

Implementation of 10 CFR 50 .69 will allow Exelon Generation Company, LLC (EGC) to improve focus on equipment that has safety significance resulting in improved plant safety.

2.3 DESCRIPTION

OF THE PROPOSED CHANGE EGC proposes the addition of the following condition to the renewed operating license of LSCS, Units 1and2, to document the NRC's approval of the use 10 CFR 50 .69.

Exelon Generation Company, LLC (EGC) 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 Page 2

Enclosure Evaluation of the Proposed Change shutdown risk; the Arkansas Nuclear One, Unit 2 (AN0-2) passive categorization method to assess passive component risk for Class 2 and Class 3 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 the EGC submittal letter dated January 31, 2020, 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).

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Enclosure Evaluation of the Proposed Change 3 TECHNICAL EVALUATION 10 CFR 50 .69 specifies the information to be provided by a licensee requesting adoption of the regulation . This request conforms to the requirements of 10 CFR 50 .69(b )(2), which states:

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

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

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

(iii) Results of the PRA review process conducted to meet§ 50 .69(c)(1 )(i).

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

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

The PRA models described within this LAR are the same as those described within the EGC submittal of the LAR dated January 31, 2020 for Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) (RS-20-009). EGC 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 EGC 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 EGC 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 Page 4

Enclosure Evaluation of the Proposed Change 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 3002012988 [3] 1 approach for seismic Tier 2 sites, which includes LSCS, to assess seismic hazard risk for 50.69. Inclusion of additional process steps discussed below to address seismic considerations will ensure that reasonable confidence in the evaluations required by 10 CFR 50 .69(c)(1 )(iv) is achieved . RG 1.201 states that "the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence" and that "all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1 )(iv)." However, neither RG 1.201 nor NEI 00-04 prescribe a particular sequence or order for each of the elements to be completed.

Therefore, the order in which each of the elements of the categorization process (listed below) is completed is flexible and as long as they are all complete they may even be performed in parallel. Note that NEI 00-04 only requires Item 3 to be completed for components/functions categorized as Low Safety Significant (LSS) by all other elements. Similarly, NEI 00-04 only requires Item 4 to be completed for safety-related active components/functions categorized as LSS by all other elements.

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

2. non-PRA approaches (e.g., Fire Safe Shutdown Equipment List, Seismic Safe Shutdown Equipment List, other external events screening, and shutdown assessment)

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

4. the defense-in-depth assessment

5. the passive categorization methodology 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:

1 Updates to EPRI 3002012988 report [3] are incorporated by reference into this LSCS submittal. These updates are cited in Attachment 2 of the EGC RAI response dated July 19, 2019 for Calvert Cliffs' 10 CFR 50.69 LAR (ML19200A216) [60] .

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Enclosure Evaluation of the Proposed Change Figure 3-1 : Categorization Process Overview Define System Boundaries Define System Functions and Assign Components to Functions Identify Seismi c 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 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., HSS or 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 RISC category can be assigned.

The IDP may direct and approve detailed categorization of components in accordance with NEI 00-04 Section 10.2. The IDP may always elect to change a preliminary LSS component or function to HSS, however the ability to change component categorization from preliminary HSS to LSS is limited. This ability is only available to the IDP for select process steps as described in NEI 00-04 and endorsed by RG 1.201. Table 3-1 summarizes these IDP limitations in NEI 00-04. The steps of the process are performed at either the function level, component Page 6

Enclosure Evaluation of the Proposed Change 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 Drives Categorization Step Element Evaluation Level Change Associated

- NEI 00-04 Section HSS to LSS Functions Internal Events Base Case- Not Allowed Yes Section 5.1 Fire, Seismic and Other External Allowable No Risk (PRA Events Base Case Component Modeled)

PRA Sensitivity Allowable No Studies Integral PRA Assessment - Not Allowed Yes Section 5.6 Fire and Other Component Not Allowed No External Hazards -

Risk (Non-modeled) Seismic- Function/Component Allowed 2 No Shutdown - Section Function/Component Not Allowed No 5.5 Core Damage -

Function/Component Not Allowed Yes Defense-in- Section 6.1 Depth Containment -

Component Not Allowed Yes Section 6.2 Qualitative Considerations -

Function Allowable 1 N/A Criteria Section 9.2 Passive Passive - Section 4 Segment/Component Not Allowed No Notes:

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

however the final assessments of the seven considerations are the direct responsibility of the IDP.

The seven considerations are addressed preliminarily by the 50.69 categorization team for at least the system functions that are not found to be HSS due to any other categorization step. Each of the seven considerations requires a supporting justification for confirming (true response) or not confirming (false response) that consideration. If the 50.69 categorization team determines that one or more of the seven considerations Page 7

Enclosure Evaluation of the Proposed Change 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 50.69 team (i.e. all considerations for all functions are reviewed). The IDP may confirm the preliminary function risk and associated justification or may direct that it be changed based upon their expert knowledge. Because the Qualitative Criteria are the direct responsibility of the IDP, changes may be made from preliminary HSS to LSS or from preliminary LSS to HSS at the discretion of the IDP. If the IDP determines any of the seven considerations cannot be confirmed (false response) for a function, then the final categorization of that function is HSS.

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

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 IDP's consideration of the seismic insights applicable to the system being categorized. Therefore, if an HSS component is mapped to an LSS function, that component will remain HSS. If an LSS component is mapped to an HSS function, that component may be driven HSS based on Table 3-1 above or may remain LSS. For the seismic hazard, given that LSCS 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 .

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Enclosure Evaluation of the Proposed Change The following are clarifications to be applied to the NEI 00-04 categorization process:

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

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

The decision criteria for the IDP for categorizing SSCs as safety significant or low safety-significant pursuant to§ 50 .69(f)(1) will be documented in EGC 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 safety significant and LSS.

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

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

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

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

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

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

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Enclosure Evaluation of the Proposed Change

LSCS 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 3002012988 (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 January 31, 2020 (RS-20-009) (Refer to Attachment 2).

Fire Risks: Fire PRA model, as submitted to the NRC for TSTF-505 dated January 31, 2020 (RS-20-009) (Refer to Attachment 2).

Seismic Risks: EPRI Alternative Approach in EPRI 3002012988 (Reference [3]) for Tier 2 plants with the additional considerations discussed in Section 3.2 .3 of this LAR.

Other External Risks (e.g., tornados, external floods): Using the IPEEE screening process as approved by NRC SE dated December 8, 2000 (Reference [5]). The other external hazards were determined to be insignificant contributors to plant risk.

Low Power and Shutdown Risks: Qualitative defense-in-depth (DID) shutdown model for shutdown Configuration Risk Management (CRM) based on the framework for DID provided in NUMARC 91-06, "Guidance for Industry Actions to Assess Shutdown Management" (Reference [6]), 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 Page 10

Enclosure Evaluation of the Proposed Change 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 (Rl-RRA) methodology contained in Reference [7]

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

The Rl-RRA methodology is a risk-informed safety classification and treatment program for repair/replacement activities (Rl-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., defense in depth, safety margins) in determining safety significance.

Component supports are assigned the same safety significance as the highest passively ranked component within the bounds of the associated analytical pipe stress model. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

The use of this method was previously approved to be used for a 10 CFR 50.69 application by NRC in the final Safety Evaluation for Vogtle dated December 17, 2014 (Reference [4]) . The Rl-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 Rl-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 AN02-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 1.14 7, 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 LSCS 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 credited in this request are the same PRA models credited in the TSTF-505 application dated January 31, 2020, (RS-20-009).

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

3.2.2 Fire Hazards The LSCS 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 EGC risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for LSCS. 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 LSCS seismic hazard assessment, EGC Nuclear 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] and includes additional considerations that are discussed in this section.

The proposed categorization approach for LSCS 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 LSCS site is considered a Tier 2 site because the site GMRS to SSE comparison is above the Tier 1 threshold but not high enough that the NRC required the plant to perform an SPRA to respond to Recommendation 2.1 of the Near Term Task Force 50.54(f) letter (Reference [8]). 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.

For example, the 50 .69 categorization process as defined in NEI 00-04 includes an Integral Assessment that weighs the hazard-specific relative importance of a component (e.g .,

internal events, internal fire, seismic) by the fraction of the total Core Damage Frequency (CDF) contributed by that hazard. The risk from an external hazard can be reduced from the default condition of HSS if the results of the integral assessment meets the importance measure criteria for LSS. In applying the EPRI 3002012988 (Reference [3]) process to the 10 CFR 50.69 categorization process, the Integrated Decision-making Panel (IDP) will be provided with the rationale for applying the EPRI 3002012988 guidance and informed of plant SSC-specific seismic insights for their consideration in the HSS/LSS deliberations.

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Enclosure Evaluation of the Proposed Change The trial studies in Reference [3], as amended by their RAI responses and amendments (References [9] , [1 O] , [11], [12], and [13]), 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 LSCS, there is no need to perform more detailed evaluations to demonstrate the inherent seismic capacities documented in industry sources such as Reference [14]. Tier 2 seismic demand sites have a lower likelihood of seismically induced failures and less challenges to plant systems. This, therefore, provides the technical basis for allowing use of a graded approach for addressing seismic hazards at LSCS .

Test cases described in Section 3 of Reference [3], as amended by their RAI responses and amendments (References [9] , [1 O], [11], [12], and [13]) , showed 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 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.

EGC 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 Study A (Reference [15]), Case Study C (Reference [16]), and Case Study D (Reference [17]) as well as, RAI responses and amendments (References [9] , [1 O], [11], [12],

and [13]), clarifying aspects these case studies.

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

Page 13

Enclosure Evaluation of the Proposed Change 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 1.0Hz 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 [8]).

As shown in Figure 2 in Section 2.5, comparing the LSCS GMRS (derived from the seismic hazard) to the SSE (i .e. seismic design basis capability), the GMRS is largely below the SSE up through 6 Hz and exceeds the SSE above 6 Hz. (Reference [18)). The NRC screened out LSCS from performing an SPRA in response to the NTTF 2.1 50.54(f) letter (Reference [19)).

As such , it is appropriate that LSCS is considered a Tier 2 plant. The basis for LSCS 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 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 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 [14))

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.

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Enclosure Evaluation of the Proposed Change In applying the Reference [3] process for Tier 2 sites to the LSCS 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 LSCS 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 LSCS 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 LSCS) are calculated per Section 5.6 of NEI 00-04, and components for which these measures exceed the specified criteria are preliminary HSS which cannot be changed to LSS. For HSS SSCs uniquely identified by the LSCS 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 LSCS 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. Determination of seismic insights will make use of the full power internal events PRA model supplemented by focused seismic Page 15

Enclosure Evaluation of the Proposed Change walkdowns. To determine the importance of SSCs for mitigating seismic events the following process will be utilized on a system basis:

o Identify SSCs within the system to be categorized o Group SSCs into equipment classes according to Reference [14] and separate-out distribution systems such as cable tray, piping, and HVAC ductwork.

o Refine the list and screen:

Inherently rugged components like check valves, manual valves, and valves (AOV and MOV) not required to change state.

Screen if the component is not used in safety functions that support mitigation of core damage or containment performance

Screen if the component is already identified as a HSS component from the Internal Events PRA or Integrated assessment

Do a seismic walkdown on SSCs screened IN to look for correlation and spatial interaction concerns

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- set the probability of failure to 1E-04 or justify based on the hazard o Quantify the FPIE model for LOOP and Small LOCA (SLOCA) accident sequences setting the LOOP initiating event frequency to 1.0/yr and the SLOCA initiating event frequency to 1E-02/yr o Utilize the Importance Measures from this sensitivity study to identify appropriate SSCs that should be HSS due to correlation or seismic interactions Page 16

Enclosure Evaluation of the Proposed Change Figure 3-2: Seismic Correlated Failure Assessment for Tier 2 Plants 2

]I Iden ify 55Cs Group ~SCs inro in svstem to be equ pment classes categonred 4 ~r<<<led C>\lt fro furth ~ sei mi<:

eonside tittion Step 3 ~"ce11s rnn l:>a pr::r/Ofmr:d in' Qnr order j------------ ---~---------------------------

r-------------------------

~---- No Sa Sb ldi:nrtv SSC~

l dcnti y 5c-ismk tons.idcrcd sci~mitilllY :- - - - . .14 - - ----t intc-rac *on f.ai lurcs correlated

>-~~__,v*~~~~~ ......~-- 1 1Step 5 eval!Mltlons are

perjnrmed as part of a

~et.1mk: wal down

' 7 J Add ~ismi' surTOgate common cause . 'fe o r interactioo /

gro\J ps to FPI E PRA fa ilures / ~

9 S 0.1Ja nt1 I). LOOP /SBO and SLOCA eve nts ail d meet r.v or RAW

~ab1la te FV and RAW thr~ l ~ for lrnport.arKe Meas.ures HS5 2 Reproduced from Reference [3] Figure 2-3 Page 17

Enclosure Evaluation of the Proposed Change Such 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 can challenge any candidate HSS recommendation for any SSC from a seismic perspective if they believe there is a basis . 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.

In the unlikely event that the LSCS seismic hazard changes from medium risk (i.e., Tier 2) at some future time , EGC will follow its categorization review and adjustment process procedures to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e).

Historical Seismic References for LSCS County Generating Station The LSCS SSE and GMRS curves from the seismic hazard and screening response are shown in Section 2.4 and 3.1 of Reference [20] . The LSCS SSE and GMRS curves from the seismic hazard and screening response are shown in Figure A4-1 of Attachment 4. The NRC's Staff assessment of the LSCS seismic hazard and screening response is documented in Reference [19] . In the Staff Confirmatory Analysis (Section 3.3.3) on page 10 of Reference

[19], the NRC concluded that the methodology used by EGC in determining the GMRS was acceptable and that the GMRS determined by EGC adequately characterizes the reevaluated hazard for the LSCS 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 LSCS, 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 [19] and [20]).

2. NTTF Recommendation 2.1 spent fuel pool assessment (References [21] and [22]).

3. NTTF Recommendation 2.3 seismic walkdowns (References [23] and [24]).

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

(References [25] and [26]).

The following additional post-Fukushima seismic reviews were performed for LSCS :

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

(References [27], [28])

6. NTTF Recommendation 2.1 seismic High Frequency Evaluation (References [29]

and [30]

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Enclosure Evaluation of the Proposed Change Summary Based on the above, the Summary from Section 2.3.3 of Reference [3] applies to LSCS; namely, LSCS 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. 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).

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Enclosure Evaluation of the Proposed Change 3.2.4 Other External Hazards All external hazards, except for seismic, were screened for applicability to LSCS per a plant-specific evaluation in accordance with GL 88-20 (Reference [31]) 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 LSCS categorization process will use the shutdown safety management plan described in NUMARC 91-06 for evaluation of safety significance related to low power and shutdown conditions . The overall process for addressing shutdown risk is illustrated in Figure 5-7 of NEI 00-04.

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

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 EGC risk management process ensures that the applicable PRA models used in this application continues to reflect the as-built and as-operated plant for LSCS . 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, EGC 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.

Page 20

Enclosure Evaluation of the Proposed Change 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 of NEI 00-04.

In the overall risk sensitivity studies, EGC 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, EGC will perform both an initial sensitivity study and a cumulative sensitivity study. The initial sensitivity study applies to the system that is being categorized. In the cumulative sensitivity study, the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in all identified PRA models for all systems that have been categorized are increased by a factor of 3. This sensitivity study together with the periodic review process assures that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed 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 [32]) . The process in these references was mostly developed to evaluate the uncertainties associated with the internal events PRA model; however, the approach can be applied to other types of hazard groups.

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

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

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Enclosure Evaluation of the Proposed Change 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 [33]), consistent with NRC RIS 2007-06.

The Internal Events PRA model received a formal industry peer review in April 2008. The FPIE Peer Review was performed using the NEI 05-04 process, the ASME PRA Standard ASME RA-Sc-2007, and Regulatory Guide 1.200, Rev. 1. The 2008 Internal Events PRA Peer Review findings were addressed in subsequent PRA updates and a F&O Closure Review was performed by an independent review team in June 2017 (Reference [34]) .

Eleven (11) F&Os associated with SRs assessed as less than Capability Category II (i.e., SRs assessed as "Not Met" or Capability Category I) were categorized as suggestions rather than findings. The resolution of these suggestion level F&Os was not previously independently reviewed, so a supplemental FPIE F&O Closure Review was performed in conjunction with the Fire PRA closure in October 2017 (Reference [35]).

Since the peer review of the Internal Events PRA model was performed prior to the publication of RG 1.200 Rev. 2, a self-assessment was conducted to assess the differences between RE 1.200 Rev. 2 and RG 1.200 Rev. 1 [36]. That assessment confirmed that the PRA model meets the requirements of RG 1.200 Rev. 2 and the results from that assessment are documented in .

The Fire PRA model received a formal industry peer review in December 2015. The Fire PRA Peer Review was performed using the NEI 07-12 Fire PRA peer review process, the ASME/ANS PRA Standard, ASME/ANS RA-Sa-2009, and Regulatory Guide 1.200, Rev. 2. The 2015 LSCS Fire PRA Peer Review was a full-scope review of the LSCS at-power Fire PRA against all technical elements in Part 4 of the ASME/ANS PRA Standard , including the referenced Internal Events Supporting Requirements (SRs).

The 2015 Fire PRA Peer Review findings were addressed in subsequent PRA updates and a F&O Closure Review was performed by an independent review team in October 2017 (Reference [35]). During the October 2017 F&O Closure Review, a Focused Scope Peer Review (FSPR) was conducted against the Fire Risk Quantification (FQ) Technical Element due to the large reduction in CDF and LERF as a result of the resolution to several technical F&Os and other model refinements.

In September 2019, another F&O Closure Review (Reference [37]) was conducted to independently review the remaining open F&Os (including the new F&Os from the FSPR).

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

[38]) as accepted by NRC in the letter dated May 3, 2017 (ML17079A427) (Reference [39]) .

The results of this review have been documented and are available for NRC audit. provides a summary of the remaining findings and open items, including:

Open items and disposition from the LSCS RG 1.200 self-assessment.

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Enclosure Evaluation of the Proposed Change

Open findings and disposition of the LSCS peer reviews.

The attachments identified above demonstrate that the PRA is of sufficient quality and level of detail to support the categorization process and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC as required 10 CFR 50.69(c)(1 )(i).

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

The LSCS 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 stud ies described in NEI 00-04 Section 8 will be used to confirm that the categorization process results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF). The failure rates for equipment and initiating event frequencies used in the PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, and human errors). Subsequent performance monitoring and PRA updates requ ired 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 LSCS Tier 2 approach discussed in section 3.2.3, implementation of the EGC 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 the EGC 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 responsibil ities 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 EGC 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 Page 23

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

A review of the impact on the System Categorization Document (SCD) for configuration changes that may impact a categorized system under 10 CFR 50 .69.

Steps to be performed if redundancy, diversity, or separation requirements are identified or affected. These steps include identifying any potential seismic interaction between added or modified components and new or existing safety related or safe shutdown components or structures. Review of impact to seismic loading, safe shutdown earthquake (SSE) seismic requirements, as well as the method of combining seismic components.

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

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

Scheduled periodic reviews no longer than once every two refueling outages will evaluate new insights resulting from available risk information (i.e ., PRA model or other analysis used in the categorization) changes, design changes, operational changes, and SSC performance. If it is determined that these changes have affected the risk information or other elements of the categorization process such that the categorization results are more than minimally affected, then the risk information and the categorization process will be updated . This scheduled review will include:

A review of plant modifications since the last review that could impact the SSC categorization.

A review of plant specific operating experience that could impact the SSC categorization .

A review of the impact of the updated risk information on the categorization process results.

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 updated, 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 Page 24

Enclosure Evaluation of the Proposed Change 50.69 periodic monitoring program includes immediate and periodic reviews, that include the requirements of the regulation, to ensure that all issues that could affect 10 CFR 50 .69 categorization are addressed. The periodic monitoring process also monitors the performance and condition of categorized SSCs to ensure that the assumptions for reliability in the categorization process are maintained.

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Enclosure Evaluation of the Proposed Change 4 REGULATORY EVALUATION 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA The following NRC requirements and guidance documents are applicable to the proposed change.

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

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

Revision 1, May 2006.

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, April 2015.

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

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

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS EGC 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.

EGC 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 Page 26

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

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

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

Response : No .

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

Under the proposed change, no additional plant equipment will be installed.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated .

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

Response: No.

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

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

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

Based on the above, EGC 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.

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Enclosure Evaluation of the Proposed Change

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

Page 28

Enclosure Evaluation of the Proposed Change 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.

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Enclosure Evaluation of the Proposed Change 6 REFERENCES

[1] NEI 00-04, "10 CFR 50 .69 SSC Categorization Guideline," Revision 0, Nuclear Energy Institute," July 2005.

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

[3] Electric Power Research Institute (EPRI) 3002012988, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization , July 2018.

[4] Vogtle Electric Generating Plant, Units 1 and 2 -Issuance of Amendments Re: Use of 10 CFR 50.69 (TAC NOS. ME9472 AND ME9473), December 17, 2014.

[5] LaSalle County Station, Units 1 and 2, "NRC Staff Evaluation of the Individual Plant Examination of External Events (IPEEE) Submittal," (TAC NOS . M83634 and M83635),

December 8, 2000.

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

December 1991.

[7] ANO SER Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative AN02-R&R-004, Revision 1, "Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems,"

(TAC NO. MD5250) (ML090930246), April 22, 2009.

[8] U.S . Nuclear Regulatory Commission, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50 .54(F) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident, March 12, 2012(ML12053A340).

[9] Peach Bottom Atomic Power Station Seismic Probabilistic Risk Assessment Report, "Response to NRC Request Regarding Recommendation 2.1 of the Near Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," August 28, 2018 (ML18240A065).

[10) Plant C, License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process, June 22, 2017 (ML17173A875).

[11) Plant C, "Issuance of Amendments Regarding Application of Seismic Probabilistic Risk Assessment Into the Previously Approved 10 CFR 50.69 Categorization Process (EPID L-2017-LLA-0248)," August 10, 2018(ML18180A062).

[12) Seismic Probabilistic Risk Assessment for Plant D Nuclear Plant, Units 1 and 2, "Response to NRC Request for Information Pursuant to 10 CFR 50 .54(f) Regarding Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident," June 30, 2017(ML1718A485).

[13) Plant D Nuclear Plant Seismic Probabilistic Risk Assessment Supplemental Information, April 10, 2018(ML18100A966).

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

[15] Exelon Generation Company, LLC, letter to U.S. Nuclear Regulatory Commission, "Peach Bottom Atomic Power Station , Units 2 and 3, RFOL Nos. DPR-44 and DPR-56, NRC Docket Nos. 50-277 and 50-278, Supplemental Information to Support Application to Adopt 10 CFR 50.69 Risk-Informed Categorization and Treatment of SSCs for NPPs," June 6, 2018 (ML18157A260).

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Enclosure Evaluation of the Proposed Change

[16] Vogtle Electric Generating Plant, Units 1 and 2, License Amendment Request to Incorporate Seismic Probabilistic Risk Assessment into 10CFR50 .69, February 21, 2018 (ML180528342).

[17] 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(ML18334A363).

[18] U.S. Nuclear Regulatory Commission, Support Document for Screening and Prioritization Results Regarding Seismic Hazard Re-Evaluations for Operating Reactors in the Central and Eastern United States, May 21, 2014 (ML14136A126).

[19] U.S. Nuclear Regulatory Commission, Lasalle County Station, Units 1 And 2 - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(F), "Seismic Hazard Reevaluations for Recommendation 2 .1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," (TAC Nos. MF3881 and MF3882) April 21, 2015(ML15013A132).

[20] Exelon Generation Company, LLC, Seismic Hazard and Screening Report (Central and Eastern United States (CEUS) Sites), Response to NRC Request for Information Pursuant to 10 CFR 50 .54(f), "Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, LaSalle County Station, Units 1 and 2, Facility Operating License Nos. NPF-11 and NPF-18," NRC Docket Nos. 50-373 and 50-374, March 31, 2014(ML14091A013).

[21] Exelon Generation Company, LLC, Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50 .54(f),

"Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, LaSalle County Station, Units 1 and 2," Facility Operating License Nos. NPF-11 and NPF-18, NRC Docket Nos. 50-373 and 50-374, August 31, 2016 (ML16244A802).

[22] U.S . Nuclear Regulatory Commission, LaSalle County Station, Units 1 and 2 - Staff Review of Spent Fuel Pool Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1, (CAC Nos. MF3881 and MF3882), September 13, 2016(ML16252A314).

[23] Exelon Generation Company, LLC's 180-day Response to NRC Request for Information Pursuant to 10 CFR 50.54(f), "Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, LaSalle County Station," License Nos. NPF-11 and NPF-18, NRC Docket Nos. 50-373 and 50-374, November 27, 2012(ML12346A030).

[24] U.S. Nuclear Regulatory Commission, LaSalle County Station, Units 1 and 2, "Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2 .3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident," (TAC Nos. MF0136 and MF0137), May 29, 2014(ML14128A334).

[25] Exelon Generation Company, LLC , Seismic Mitigating Strategies Assessment (MSA)

Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Appendix H, Revision 4, H.4.4 Path 4: GMRS < 2xSSE, LaSalle County Station, Units 1 and 2, Facility Operating License Nos. NPF-11 and NPF-18, NRC Docket Nos. 50-373 and 50-374 ,

August 22, 2017(ML17234A470).

[26] U.S . Nuclear Regulatory Commission, LaSalle County Station, Units 1 & 2 , "Staff Review of Mitigating Strategies Assessment Report of the Impact of the Reevaluated Seismic Page 31

Enclosure Evaluation of the Proposed Change Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter," (CAC Nos.

MF7839 And MF7840 ; EPID No. L-2016-JLD-0006) , August 14, 2018(ML18207A854).

[27] Exelon Generation Company, LLC, Expedited Seismic Evaluation Process Report (CEUS Sites), "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, LaSalle County Station, Units 1 and 2," Facility Operating License Nos.

NPF-11 and NPF-18, NRC Docket Nos. 50-373 and 50-374, December 19, 2014 (ML14353A085).

[28] U.S. Nuclear Regulatory Commission, LaSalle County Station, Units 1 and 2, "Staff Review of Interim Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1," (TAC Nos. MF5247 and MF5248), June 16, 2015 (ML15160A168).

[29] Exelon Generation Company, LLC , High Frequency Supplement to Seismic Hazard Screening Report, "Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, LaSalle County Station," Units 1 and 2, Facility Operating License Nos. NPF-11 and NPF-18, NRC Docket Nos. 50-373 and 50-374, December 1, 2016(ML16336A810).

[30] U.S. Nuclear Regulatory Commission, LaSalle County Station, Units 1 and 2, "Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1," February 6, 2017 (ML17031A425).

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

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

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

[34] LaSalle County Generation Station Unit 2, PRA Facts and Observations Independent Assessment Report Using NEI 05-04/07-12/12-06 Appendix X, June 2017.

[35] LaSalle County Generating Station, PRA Fact and Observation Independent Assessment

& Focused-Scope Peer Review, Report# 032299RPT-09, Revision 0, March 2019.

[36] LaSalle County Generating Station Probabilistic Risk Assessment Self-Assessment of the LaSalle PRA Against the Combined ASME/ANS PRA Standard Requirements, LS-PSA-016, Rev.3, November2015.

[37] LaSalle Units 1 & 2, Fire PRA Finding & Suggestion Level Fact and Observation Closure by Independent Assessment, Report Number# 032362-RPT-01, Revision 0, November 2019.

[38] Nuclear Energy Institute (NEI) Letter to NRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," February 21, 2017, Accession Number ML17086A431 .

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

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Enclosure Evaluation of the Proposed Change

[40] LS-PSA-013, LaSalle County Generating Station Probabilistic Risk Analysis Summary Notebook, Revision 8, November 2015.

[41] LaSalle County Station (LSCS) Updated Final Safety Analysis Report (UFSAR),

Revision 23, April 2018.

[42] ER-AA-340 , "GL 89-13 Program Implementing Procedure," Revision 9.

[43] LaSalle Flood Hazard Reevaluation Report (FHRR), "NRC ADAMS Accession No. ML14079A425," March 12, 2014 .

[44] NRC Letter, LaSalle County Station , Units 1 and 2, Staff Assessment of Flooding Focused Evaluation, NRC ADAMS Accession No. ML17191A323, August 23, 2017.

[45] Commonwealth Edison, Individual Plant Examination and Individual Plant Examination (External Events) Submittal, LaSalle County Nuclear Power Station, April 28, 1994.

[46] LaSalle Design Analysis L-003414, Toxic Chemical Analysis of 2008 Offsite Chemical Survey Results.

[47] CRC, "Handbook of Chemistry and Physics," 49th Edition, 1969.

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

[49] LS-PSA-021.12, LaSalle Fire PRA Uncertainty and Sensitivity Analysis Notebook, Rev. 3, 2019.

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

[51] LS-MISC-046, Assessment of Key Assumptions and Sources of Uncertainty for Risk-Informed Applications, Revision 0, January 2020.

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

[53] NUREG/CR-6595, "An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events," Revision 1, October 2004.

[54] NUREG/CR-6850 (also EPRI 1011989), "Fire PRA Methodology for Nuclear Power Facilities," September 2005, with Supplement 1 (EPRI 1019259), September 2010.

[55] NUREG/CR-7150, Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE), October 2012.

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

[57] "Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE), Volume 1: Peak Heat Release Rates and Effect of Obstructed Plume",

NUREG-2178 Vol. 1/ EPRI 3002005578, U.S. NRC and Electric Power Research Institute, Draft Report for Comment, April 2015.

[58] "Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE), Volume 2: Expert Elicitation Exercise for Nuclear Power Plant Fire-Induced Electrical Circuit Failure", Final Report, NUREG/CR-7150, EPRI 3002001989, U.S. NRC and Electric Power Research Institute, May 2014 .

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Enclosure Evaluation of the Proposed Change

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

[60] Exelon Generation Company, LLC, letter to U.S. Nuclear Regulatory Commission, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Renewed Facility Operating License Nos.

DPR-53 adn 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(ML19200A216).

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Enclosure Evaluation of the Proposed Change Attachment 1 Attachment 1: List of Categorization Prerequisites Exelon Generation Company, LLC 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 IDP for review and approval. The IDP reviews the categorization results and makes the final determination on the safety significance of system functions and components.

Risk sensitivity study. For PRA-modeled components, an overall 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 Page 35

Enclosure Evaluation of the Proposed Change Attachment 2 Attachment 2: Description of PRA Models Used in Categorization Plant Units Model Baseline CDF Baseline LERF Comments 1 & 2 Full Power Internal 1.3E-06 (Unit 1) 1.3E-07 (Unit 1) Application Specific Events and Internal Model (ASM) to the Flooding PRA 1.3E-06 (Unit 2) 1.3E-07 (Unit 2) Full Power Internal Events (FPIE) with LS216C (Unit 1&2) Internal Flooding update which was based on the 2014 LaSalle PRA MORs 1&2 Fire PRA 1.0E-05 (Unit 1) 9.8E-07 (Unit 1) Application Specific Model (ASM) to the LS114AF3(Unit1) 7.8E-06 (Unit 2) 3.2E-07 (Unit 2) Fire Update which was based on the LS214AF3 (Unit 2) 2014 PRA MORs Page 36

Enclosure Evaluation of the Proposed Change Attachment 3 Attachment 3: Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items Finding Supporting Capability Description Disposition for 50.69 Number Requirement(s) Category (CC)

Internal Events F&Os IE-03-01 IE-03 Not Met CCII The Summary Notebook includes Additional documentation of (Finding) AS-C3 information that attempts to identify the LERF key sources of key sources of uncertainty in the initiating uncertainty including results and SC-C3 event analysis. However, with the important insights are needed to CY-C3 changes to eliminate "key" from the SR fully close out this Finding.

HR-13 definition, this SR cannot be considered However, this issue does not OA-03 met. impact 50.69 applications. The IF-F3 model sources of uncertainty, QU-E2 Section 4 of the LS-PSA-013 notebook both generic and plant-specific, LE-G4 (Reference [40]) discusses the industry as they impact this risk-

"key sources of uncertainty" per EPRI informed application are guidance. However, the current analysis specifically addressed in does not fully meet the requirements of Attachment 6 of this LAR.

RG 1.200, which requires a discussion of sources of model uncertainty and related assumptions. Also, there may be some plant-specific assumptions made that may not be fully captured by the generic list of potential sources of uncertainty.

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Enclosure Evaluation of the Proposed Change Attachment 3 Finding Supporting Capability Description Disposition for 50.69 Number Requirement(s) Category (CC)

SY-A4-01 SY-A4 Not Met CCII Enhance PRA technical capability. While it is judged that this (Suggestion) Finding has no impact on the Perform plant walkdowns with system PRA results and therefore, no engineers AND plant operators. Better impact on 50.69 document the walkdowns performed in implementation, this F&O will be support of the PRA and reference those resolved during a LSCS PRA walkdowns in each system notebook to update and system walkdowns achieve Capability Category II will be conducted and documented with System Engineers and Plant Operators prior to implementation of 50.69.

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Enclosure Evaluation of the Proposed Change Attachment 3 Finding Supporting Capability Description Disposition for 50.69 Number Requirement(s) Category (CC)

DA-C8-01 DA-C8 Not Met CCII Basic events used to model the standby While it is judged that this (Finding) status of various plant systems use a mix Finding has no impact on the of plant-specific operational data and PRA results and therefore, no engineering judgment. For the plant impact on 50.69 service water system and several other implementation, this F&O will be systems, standby estimates have been resolved during a LSCS PRA determined from procedures and update and plant-specific data operating data. For other components, reviews will be performed and assumptions are used (e.g., 50% documented prior to probability of either of two pumps in a implementation of 50.69.

system is in standby). So, overall the LSCS PRA has some Capability Category (CC) II attributes and some CC-I attributes.

Current approach of assuming standby time does not meet the requirements of the Supporting Requirement. The use of actual plant data could result in small changes in PRA results.

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Enclosure Evaluation of the Proposed Change Attachment 3 Finding Supporting Capability Description Disposition for 50.69 Number Requirement(s) Category (CC)

DA-C6-01 DA-C7 Not Met CCII LS-PSA-010, Component Data This issue has minimal impact (Suggestion) DA-C9 Notebook, Appendix C, states, "No actual on the 50 .69 application as the data or estimates for these parameters plant-specific data was updated DA-C10 are provided by system managers. Data during the 2011 and 2014 PRA from the MSPI basis document, scoping updates. Further, LSCS will be and performance criteria document, and updating the plant-specific data 2003 data notebook is used ." As the was during a PRA update before obtained from Maintenance Rule and implementation of 50.69.

MSPI sources, the techniques used to obtain this data are probably consistent with the guidance in this supporting requirement, but this cannot be positively determined. Similarly, for SR DA-C7, it is unable to be determined if surveillance tests, planned and unplanned maintenance activities were based on actual plant experience. For SR DA-C9, the reviewers were unable to conclude whether plant specific operational records were used to determine standby time.

Similarly, for DA-C10, it is not clear how surveillance tests were used.

This appears to be primarily a documentation issue, as it is expected that the assumptions used to collect data for Maintenance Rule and MSPI are similar to those required by the ASME standard. However, it is possible that some differences in methodology could exist between these programs and the PRA.

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Enclosure Evaluation of the Proposed Change Attachment 3 Finding Supporting Capability Description Disposition for 50.69 Number Requirement(s) Category (CC)

IF-C3b-01 IF-C3B Not Met CCII Address potential unavailability of This open issue has no impact (Suggestion) barriers that affect the propagation of on the 50 .69 implementation as water in order to meet the CC II it is a documentation issue.

requirements of the ASME Standard. However, this suggestion will be resolved during a PRA update.

This is a suggestion since it is considered a documentation issue. The flood scenarios analyzed in detail are so large (i.e., typically involving draining the lake into the Turbine building until it fills) that structural analysis of non-flood doors and any difference in flood propagation will have no significant impact.

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Enclosure Evaluation of the Proposed Change Attachment 3 Finding Supporting Capability Description Disposition for 50.69 Number Requirement(s) Category (CC)

Fire Model F&Os 1-19 CS-A1 NOT MET CCII The peer review examined the cable This item has no impact on the (Suggestion) selection package for offsite power loss 50 .69 as it has been resolved, switchyard breaker (OCB 4-6). The just not reviewed and closed by circuit evaluation package includes two the Independent Assessment pages of notes regarding interlock Team .

evaluations and the notes and assumptions associated with the interlocks. For example, a note is made that "the interlock associated with trip and lockout of SAT 242. Cables that can cause relay to actuate are to be included with SAT 242". The FPRA development team indicated that this impact for SAT 242 is addressed by the FPRA, but that no systematic review of the circuit evaluation package notes was performed.

A review of circuit evaluation notes and assumptions is important to ensure that FPRA plant response model identifies cables whose fire-induced failure could adversely affect selected equipment and/or credited functions in the Fire PRA plant response model.

Page 42

Enclosure Evaluation of the Proposed Change Attachment 3 Finding Supporting Capability Description Disposition for 50.69 Number Requirement(s) Category (CC) 4-17 FSS-07 Not Met CCII There is no generic estimate or plant- The impact on the 50.69 LAR is (Finding) specific value assigned to the non- judged to be minimal. However, suppression probability. plant-specific data will be reviewed and refined data for The non-suppression values are only automatic detection and based on the NUREG/CR-6850 generic suppression systems will be values for unreliability with no account for incorporated into the FPRA unavailability. model during a Fire PRA update if necessary. This item will be resolved prior to 50.69 implementation .

6-11 CS-A1 Not Met CCII The cable selection work performed There is no impact on 50.69 (Finding) CS-A2 related to the cable data in the fire safe implementation as this issue will shutdown report pre-dates NEl-00-01 and be resolved prior to CS-A3 was done to the standards at that time. implementation.

No other information is currently available regarding the circuit analysis techniques used for the fire safe shutdown report. In general , the MSO circuit analysis work was performed using NEl-00-01, Revision 2 or Revision 3 {depending upon the particular package).

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Enclosure Evaluation of the Proposed Change Attachment 4 Attachment 4: External Hazards Screening Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

In the NRC Staff Evaluation of the IPEEE (Reference [5]), a probabilistic bounding analysis was performed for aircraft impact. The median frequency of CDF was calculated as 5E-7/year (PS4).

(PS2): From Section 3.5.1.6, Aircraft Hazards of the LSCS UFSAR, the airports and airways in the vicinity of the site are described in Subsection 2.2.2.5 of the UFSAR (Reference [41 ]).

a. There are no federal airways or airport approaches passing within 2 miles of the station . The closest airway corridor is 3 miles away from the station.

PS2 Aircraft Impact y b. There are no commercial airports existing within 10 miles of the site and PS4 there is only one private airstrip within 5 miles.

c. The projected landing and take-off operations out of those airports located within 10 miles of the site are far less than 500 d2 per year, where d is the distance in miles. The projected operations per year for airports located outside of 10 miles is less than 1000 d2 per year.

d. The only military facility within 10 miles of the site is the Illinois Army Reserve National Guard Training Facility. It is located approximately 1 mile northwest of LSCS cooling lake.

There are no airstrips at the Training Facility.

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Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

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

The mid-western location of LSCS station precludes the possibility of an avalanche.

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

Hazard is slow to develop and can be identified via monitoring and managed via standard maintenance process.

Actions committed to and completed by LSCS station in response to Generic Letter 89-13 provide on-going control of biological hazards. These Biological Event y cs controls are described in EGC procedure ER-AA-340, "GL 89-13 Program Implementing Procedure" (Reference [42]) .

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

The mid-western location of LSCS station precludes the possibility of coastal erosion.

Coastal Erosion y C3 Based on this review, the Coastal Erosion impact hazard can be considered to be negligible.

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

Drought y cs Based on this review, the Drought impact hazard can be considered to be neqliqible.

Page 4S

Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

The external flooding hazard at the site was recently updated as a result of the post-Fukushima 50 .54(f)

Request for Information. A flood hazard reevaluation report (FHRR) was submitted to NRC for review on March 12, 2014 [43]. The results indicate that flooding from all mechanisms except local intense precipitation (LIP) and probable maximum storm surge (PMSS) were bounded by the current licensing basis (CLB). Only LIP and PMSS require evaluation in a Focused Evaluation (FE) to determine if the plant's current design basis bounds the reevaluated flood parameters.

Further investigation was performed and the results of the FE were submitted to NRC for review and a External Flooding y C1 staff assessment was issued on August 23, 2017 [44]. The NRC acknowledged the results presented in the FE concluding that there were no impacts to SR SSCs from the LIP and PMSS events and the design basis of the plant is adequate to mitigate the effects from external flood causing mechanisms with sufficient margin.

In accordance with the external hazard screening process per Figure 5-6 of NEI 00-04, several flood doors integral to flood protection at LSCS were identified for categorization as High Safety Significant (HSS) SSCs should their associated systems be categorized .

Based on this review, the External Flooding impact hazard can be considered to be neolioible.

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Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

Based on the plant design for wind pressure and the low frequency (<1 E-7/yr) of design tornadoes , a demonstrably conservative estimate of CDF associated with high wind hazard (other than wind generated missiles) is much less than 1E-6/yr.

In addition, based on the plant design Extreme Wind or for tornado missiles, considering a Tornado y C1 limited set of SSCs vulnerable to tornado missiles, a demonstrably conservative estimate of CDF associated with tornado missiles is less than 1E-6/yr.

Based on this review, the Extreme Wind or Tornado impact hazard can be considered to be negligible.

The principal effects of such events (such as freezing fog) would be to cause a loss of off-site power and are addressed in the weather-related Loss Fog y of Offsite Power initiating event in the C4 internal events PRA model for LSCS .

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

Forest fires were screened in the IPEEE (Reference [45]). The site landscaping and lack of forestation prevent such fires from posing a threat Forest or Range Fire y C3 to LSCS station.

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

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Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

The principal effects of such events would be to cause a loss of off-site power and are addressed in the weather-related Loss of Offsite Power initiating event in the internal events Frost y C4 PRA model for LSCS.

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

The principal effects of such events would be to cause a loss of off-site power and are addressed in the weather-related Loss of Offsite Power initiating event in the internal events Hail y C4 PRA model for LSCS.

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

The plant is designed for this hazard (C1 ). The principal effects of such events would result in elevated lake temperatures which are monitored by station personnel. Should the ultimate heat sink temperature exceed the LSCS Technical Specification 3.7.3 temperature limit, an orderly shutdown High Summer C1 y would be initiated.

Temperature C4 In addition, plant trips due to this hazard are covered in the definition of another event in the PRA model (e.g .,

transients, loss of condenser) (C4) .

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

High Tide, Lake Level, C3 The mid-western location of LSCS or River Stage y station precludes the possibility of a cs high tide condition (C3) .

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Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

High lake effects would take place slowly allowing time for orderly plant reductions including shutdowns (CS) .

Based on this review, the High Tide, Lake Level, or River Stage impact hazard can be considered to be neolioible.

The mid-western location of LSCS station precludes the possibility of a hurricane.

Hurricane y C3 Based on this review, the Hurricane impact hazard can be considered to be negligible.

Per UFSAR 2.4.7 (Reference [41]),

essential for ice jam formation is a constriction to passage of flowing ice.

Such a constriction does not exist in the Illinois River near the site, since the river is approximately 800 feet wide and is kept navigable by dredging when required. The lake screen house is protected against icing in the lake by provision of warming lines near C1 Ice Cover y the screen house (C1).

C4 The principal effects of such events would be to cause a loss of off-site power and are addressed in the weather-related Loss of Offsite Power initiating events in the internal events PRA model for Lasalle (C4).

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

The only military facility within 10 miles Industrial or Military C1 is the Illinois Army Reserve National Facility Accident y Guard (ILARNG) Training Facility C3 within 1 mile northwest of LSCS Station and encompassino Page 49

Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a) approximately 2560 acres. There are no missile sites, bombing ranges or runways at the facility, but there are 5 firing ranges in the direction of north to northwest (C3).

Hazardous chemicals used and/or stored by manufacturers within five miles of the plant were also evaluated and determined to either screen from further evaluation or were determined to meet the acceptance criteria associated with Control Room operator protection as discussed in LSCS UFSAR, Section 2.2 .3 (Reference [41))

(C1)

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

The LSCS Internal Events PRA Internal Flooding includes evaluation of risk from internal NIA NIA flooding events.

The LSCS Internal Fire PRA includes evaluation of risk from internal fire Internal Fire NIA NIA events The mid-western location of LSCS station precludes the possibility of a landslide.

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

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Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

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

Lightning y Both events are incorporated into the C4 LSCS internal events model through the incorporation of generic and plant specific data.

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

These effects would take place slowly allowing time for orderly plant reductions, including shutdowns.

Low Lake Level or River Stage y C5 Based on this review, the Low Lake Level or River Stage impact hazard can be considered to be negligible.

The principal effects of such events would be to cause a loss of off-site power. These effects would take place slowly allowing time for orderly plant reductions, including shutdowns Low Winter C4 y (CS). At worst, the loss of off-site Temperature power events would be subsumed into C5 the base PRA model results (C4) .

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

The frequency of a meteor or satellite strike is judged to be so low as make the risk impact from such events Meteorite or Satellite y insignificant. This hazard also was Impact PS4 reviewed as part of the IPEEE submittal (Reference [45]) and screened based on low frequency of occurrence.

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Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

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

Per UFSAR Section 2.2.2.3 (Reference [41 ]), there are no tank farms or gas pipelines within 5 miles of the site. However, there are two natural gas pipelines between 5 and 7 miles from the site and two crude oil pipelines approximately 3 miles west Pipeline Accident y C1 of the plant. There is no significant hazard from toxic releases or explosions involving these pipelines that could interact with the plant.

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

The impact of releases of hazardous materials stored on-site was evaluated in the IPEEE submittal and updated in LSCS station's UFSAR.

UFSAR Section 2.2.3 (Reference [41])

discusses toxic gas. There is no onsite storage of chlorine; sodium hypochlorite/sodium bromide biocide system is used, thus eliminating an onsite chlorine hazard.

Release of Chemicals y Every 3 years a survey will be in Onsite Storage C1 conducted to re-evaluate the use of chlorine, within 5 miles of the control room, to ensure that a chlorine hazard does not exist. Every 6 years a survey will be conducted to re-evaluate the use of toxic chemicals, within 5 miles of the control room , to ensure that a toxic chemical hazard does not exist.

Based on this review, the Release of Chemicals in Onsite Storaoe impact Page 52

Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a) hazard can be considered to be negligible.

Per UFSAR Section 2.4 .9 (Reference

[41 ]), the Illinois River flows in the same general location as its predecessor of nearly a million years ago. Presence of navigation locks and dams over the entire length of the river has further stabilized the river course.

River Diversion y C1 Based on the available evidence, no change in the regime of the river is expected.

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

The mid-western location of LSCS station precludes the possibility of a sandstorm. More common wind-borne dirt can occur but poses no significant risk to LSCS station given the robust Sand or Dust Storm y C1 structures and protective features of the plant.

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

Flooding due to seiches is not relevant for LSCS station per Section 2.4.5 of the UFSAR [41).

Seiche y C3 Based on this review, the Seiche impact hazard can be considered to be negligible.

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

NIA NIA Page 53

Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

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

Potential flooding impacts covered Snow y cs under external flooding .

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

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

Consolidation y C1 Based on this review, the Soil Shrink-Swell Consolidation impact hazard can be considered to be negligible.

The mid-western location of LSCS station precludes the possibility of a sea level driven storm surge.

Storm Surge y C3 Based on this review, the Storm Surge impact hazard can be considered to be negligible.

UFSAR Section 2.2.3 (Reference [41])

discusses toxic gas. There is no onsite storage of chlorine; sodium hypochlorite/sodium bromide biocide system is used, thus eliminating an onsite chlorine hazard . In addition ,

there is no possibility of an accident Toxic Gas y C3 that could lead to the formation of flammable clouds in the vicinity of LSCS because ( 1) there is no chemical plant in the vicinity; (2) no gas pipeline passes the station ; and (3) no liquefied gases are transported in the vicinity.

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Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

Per the IPEEE, the bounding analysis showed that these accidents do not significantly contribute to the plant risk.

See also Transportation Accidents .

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

The impact of transportation accidents was evaluated in the IPEEE [45] and in UFSAR Section 2.2.3 [41]. In the IPEEE, an evaluation was conducted to demonstrate that the probability of a rail, land or waterway accident that resulted in release of toxic materials that could affect the site was less than 1E-6 /yr (PS4) .

Per the UFSAR:

Flammable Va12or Clouds (delayed ignition): There is no possibility of an C1 accident that could lead to the Transportation formation of flammable clouds in the Accident y C3 vicinity of LSCS because (1) there is no chemical plant in the vicinity; (2) no PS4 gas pipeline passes the station; and (3) no liquefied gases are transported in the vicinity (C3).

Trans12ortation of Toxic Chemicals:

The only transportation route carrying toxic chemicals which is within 5 miles of the station is the Illinois River. The toxic chemicals transported are chlorine and anhydrous ammonia . A toxic chemical analysis was performed (Reference [46]) which concluded that chlorine was an insignificant hazard to the station.

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Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

For anhydrous ammonia, redundant detectors have been added on each outside air intake of the control room area filtration system. These detectors will sense ammonia concentrations at the outside air intakes from near zero ppm and higher. On detection of ammonia in the outside air, a control room annunciator alarms. Within 2 minutes of detection of high ammonia concentration in the air intake, the Operator will align the control room envelope HVAC systems in recirculation mode and will don a self-contained breathing apparatus.

In accordance with the external hazard screening process per Figure 5-6 of NEI 00-04, the ammonia detectors and associated control room annunciators for ammonia were identified for categorization as High Safety Significant (HSS) SSCs should their associated systems be categorized .

The ammonia detectors and associated control room annunciators are considered HSS for 50 .69. (C1)

Ex12losions on the Highwa~t For explosions on the highway, the worst event would be an explosion from a truck carrying 43,000 pounds of TNT on County Highway 6 at the nearest location to the plant (2000 feet away).

If a 43,000-pound charge of TNT explodes at this distance, the structure will receive a peak reflected pressure of 1.5 psi. This magnitude is less than the tornado design pressure (C1).

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Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

Exglosions on the Waterway_: For explosions on the waterway, the volume of a maximum tank barge is about 1.8x105 ft3. Assuming the air mix ratio is adequate for an empty gasoline barge and a detonation takes place, the energy released will be on the order of 107 kcal (Reference [4 7]),

which is equivalent to an explosion of 10 tons of TNT. Since the Seismic Category I structures are located 4 miles away from the river, the peak reflected pressure on the structures will be less than 1 psi in case there is a detonation . Since the Seismic Category I structures have been designed for higher tornado wind pressures, the plant can withstand such a postulated explosion (C1).

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

The mid-western location of LSCS station precludes the possibility of a tsunami.

Tsunami y C3 Based on this review, the Tsunami impact hazard can be considered to be negligible.

Per the IPEEE [45] , the mean CDF for turbine-generated missiles was 1E-7/yr.

Turbine generated missiles are Turbine-Generated y C1 discussed in UFSAR Section 3.5.1.3 Missiles (Reference [41 ]. With the replacement of the Low Pressure (LP) rotors, all the turbine rotors are of the monoblock design. The monoblock rotors have very low stress level. Missile generation due to turbine failure is Page 57

Enclosure Evaluation of the Proposed Change Attachment 4 Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a) generally postulated to be caused by turbine overspeed. General Electric has established that the speed capability of these rotors is considerably higher than the maximum attainable speed of these turbine generator units. Consequently, the probability of missiles being generated is statistically insignificant.

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

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

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

Waves associated with adjacent large bodies of water are not applicable to the site (C3). Waves associated with external flooding are covered under C3 Waves y that hazard (C4).

C4 Based on this review, the Waves impact hazard can be considered to be negligible.

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

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Enclosure Evaluation of the Proposed Change Attachment 4 LaSalle Response Spectra 1.0 10.0 100.0 Spectral Frequency (Hz)

Figure A4-1 : GMRS and SSE Response Spectra for LSCS (From Reference [20] Sections 2.4 and 3.1 Page 59

Enclosure Evaluation of the Proposed Change Attachment 5 Attachment 5 Progressive Screening Approach for Addressing External Hazards Event Analysis Criterion Source Comments NUREG/CR-2300 C1 . Event damage potential is and ASME/ANS

< events for which plant is Standard designed.

RA-Sa-2009 C2 . Event has lower mean NUREG/CR-2300 frequency and no worse and ASME/ANS consequences than other Standard events analyzed . RA-Sa-2009 Initial Preliminary NUREG/CR-2300 Screening C3. Event cannot occur close and ASME/ANS enough to the plant to affect it. Standard RA-Sa-2009 NUREG/CR-2300 Not used to screen.

C4 . Event is included in the and ASME/ANS Used only to definition of another event. Standard include within RA-Sa-2009 another event.

C5. Event develops slowly, ASME/ANS allowing adequate time to Standard eliminate or mitigate the RA-Sa-2009 threat.

PS 1. Design basis hazard ASME/ANS cannot cause a core damage Standard accident. RA-Sa-2009 PS2. Design basis for the NUREG-1407 and event meets the criteria in the ASME/ANS Progressive NRC 1975 Standard Review Standard Screening Plan (SRP). RA-Sa-2009 NUREG-1407 as PS3. Design basis event modified in mean frequency is < 1E-5/y ASME/ANS and the mean conditional core Standard damage probability is < 0.1.

RA-Sa-2009 Page 60

Enclosure Evaluation of the Proposed Change Attachment 5 Event Analysis Criterion Source Comments NUREG-1407 and PS4 . Bounding mean CDF is ASME/ANS

< 1E-6/y. Standard RA-Sa-2009 Screening not successful. NUREG-1407 and PRA needs to meet ASME/ANS Detailed PRA requirements in the Standard ASME/ANS PRA Standard. RA-Sa-2009 Page 61

Enclosure Evaluation of the Proposed Change Attachment 6 Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty Internal Events I Internal Flooding PRA Model In order to identify key sources of uncertainty for the 50.69 Program application, an evaluation of Internal Events baseline PRA model uncertainty was performed , based on the guidance in NUREG-1855 (Reference [48]) and Electric Power Research Institute (EPRI) report 1016737 (Reference [32]). 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 LSCS County Generating Station (LSCS) baseline PRA model quantification (Reference [40]) and the Fire PRA uncertainty evaluation (Reference [49]).

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 LSCS Internal Events PRA technical elements are noted in the individual notebooks. The Internal Events PRA model uncertainties evaluation is documented in Reference

[40] and 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. EPRI compiled a listing of generic sources of modeling uncertainty to be considered for each PRA technical element (Reference [32]), and the evaluation performed for LSCS (Reference [40]) 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 (Reference [40]) . No specific issues of PRA completeness have been identified relative to the 50.69 application, based on the results of the Internal Events PRA and Fire PRA peer reviews .

Additionally, an evaluation of Level 2 internal events PRA model uncertainty was performed, based on the guidance in NUREG-1855 (Reference [48]) and Electric Power Research Institute (EPRI) report 1026511 (Reference [50]) . The potential sources of model uncertainty in the LSCS PRA model were evaluated for the 32 Level 2 PRA topics outlined in EPRI 1026511.

A detailed review of the generic and plant-specific sources of internal events model uncertainties are discussed in LS-MISC-046 (Reference [51]) and are therefore not repeated in this attachment.

The purpose of this attachment is to summarize the key sources of uncertainty that could potentially impact the 50 .69 application.

Based on following the methodology in EPRI 1016737, as supplemented by EPRI 1026511, the impact of key sources of uncertainty in the internal events PRA model on the 50.69 application is summarized in Table 6-1 . The key sources of uncertainty identified in Table 6-1 do not present a significant impact on the LSCS 50.69 application, and therefore, the internal events PRA model is capable of producing accurate 50.69 importance measure results.

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Enclosure Evaluation of the Proposed Change Attachment 6 Additionally, for the 50.69 program, the guidance in NEI 00-04 (Reference [1]) specifies that certain sensitivity studies be conducted for each PRA model to address key sources of uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask the SSC(s) importance. Regulatory Guide 1.174, Revision 3 (Reference [52)) cites NUREG-1855, Revision 1, as related guidance. In Section B of RG 1.174 (Reference [52)), the guidance acknowledges specific revisions of NUREG-1855 to include changes associated with expanding the discussion of uncertainties.

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

Table 6-1 IE/ IF PRA IE/ IF PRA IE/ IF PRA Sources of Assumption/ Model Sensitivity and 50.69 Impact Uncertainty Disposition (50.69)

Core Melt Arrest Prior to Injection from these high For this source of model Vessel Failure capacity low pressure systems uncertainty, sensitivity analyses will preclude vessel failure if were performed assuming that the they are available following RPV ECCS is unavailable due to steam depressurization given core binding given failure to control damage occurs at high RPV containment venting .

pressure.

Operator actions related to ECCS survivability post containment venting remain the containment venting is treated top risk-significant operator probabilistically. Although the actions. The assumption is not treatment is realistic, there is the realistic and use of this bounding potential for a non-conservative failure probability would likely bias given the unknown mask key risk insights.

phenomenological events that could be associated with containment venting (e.g., Additionally, for the 50.69 hydrogen buildup in the Reactor program, the guidance in NEI 00-Buildings, harsh events due to 04 (Reference [1]) specifies that steam release, and other certain sensitivity studies be unknown consequences). conducted for each PRA model to address key sources of uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g ., human error, common cause failure, maintenance probabilities, and manual suppression probabilities for fire) do not mask the SSC(s) importance.

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Enclosure Evaluation of the Proposed Change Attachment 6 IE/ IF PRA IE/ IF PRA IE/ IF PRA Sources of Assumption/ Model Sensitivity and 50.69 Impact Uncertainty Disposition (50.69)

Therefore, the uncertainty associated with this model uncertainty is negligible within the 50.69 application.

Vapor Suppression Ex-vessel core melt progression Sensitivity analyses were Capabilities at Vessel overwhelms vapor suppression performed using the Failure noted as extremely unlikely for recommended upper bound low pressure RPV failures values from NUREG/CR-6595 modes and very unlikely for high (Reference [53)) for Mark II pressure failure modes based Containments as an alternate on reference to generic studies hypothesis (i .e., sensitivity and identification of plant- analysis uses upper bound values specific features. of 0.2 for low pressure scenarios and 0.3 for high pressure However, more recent MAAP scenarios).

results indicate that containment pressurization following vessel Operator actions related to failure for wet containment containment venting remain the conditions might be higher than top risk-significant operator what had previously been actions. The bounding sensitivity calculated or what was originally analysis utilizes the upper bound considered.

values, which is not a realistic assumption.

Additionally, for the 50.69 program , the guidance in NEI 00-04 (Reference [1]) specifies that certain sensitivity studies be conducted for each PRA model to address key sources of uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, maintenance probabilities, and manual suppression probabilities for fire) do not mask the SSC(s) importance.

Therefore, the uncertainty associated with this model uncertainty is negligible within the 50.69 application.

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Enclosure Evaluation of the Proposed Change Attachment 6 IE/ IF PRA IE/ IF PRA IE/ IF PRA Sources of Assumption/ Model Sensitivity and 50.69 Impact Uncertainty Disposition (50.69)

Digital Feedwater Controls There are model uncertainties The sensitivity analyses consisted associated with modeling digital of increasing the failure probability systems, such as those related associated with digital feedwater to determining the failure modes controls by a factor of 50 (i.e.,

of these systems and from 0.01 to 0.5).

components.

The results demonstrate that the The reliability values from the digital feedwater controls failure similar vendor study probability does not significantly demonstrating that the system impact the overall average-performance would result in less maintenance PRA results.

than 0.1 transients per year are used for the key components of the system. Due to the small impact demonstrated by the sensitivity The reliability analysis for cases, the uncertainty associated causing plant trips performed by with this model uncertainty is similar FW vendor studies is negligible.

assumed to be equally applicable to the reliability of the system post plant trips that are caused by other means that do not directly affect the feedwater availabilitv.

Water Hammer Pipe Water hammer is a potential The sensitivity analyses consisted Rupture failure mode of important of increasing the ECCS pipe systems and can also cause a rupture failure probabilities due to flood related event. water hammers by a factor of 100 (i.e., from 1E-3 to 1E-1 ).

ECCS system draindown scenarios are included in the The sensitivity analysis is only LSCS PRA model. Subsequent increasing the likelihood of pipe starting or restarting of these rupture due to water hammer systems causes a water events and additional equipment I hammer and system leak or accident mitigation strategies rupture.

remain unaffected (i .e., the sensitivity analysis does not postulate additional equipment being out-of-service I unavailable).

Due to the small impact demonstrated by the sensitivity cases, the uncertainty associated with this model uncertainty is nealiaible. This sensitivitv Page 65

Enclosure Evaluation of the Proposed Change Attachment 6 IE/ IF PRA IE/ IF PRA IE/ IF PRA Sources of Assumption/ Model Sensitivity and 50.69 Impact Uncertainty Disposition (50.69) analysis assumes the upper bound pipe rupture failure probability for ECCS, which would not be realistic and use of this bounding rupture failure probability would likely mask key risk insights.

Page 66

Enclosure Evaluation of the Proposed Change Attachment 6 Fire PRA Model The purpose of the following discussion is to address the epistemic uncertainty in the LSCS FPRA.

The LSCS 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 LSCS FPRA was guided by NUREG/CR-6850 (Reference [54] . The LSCS FPRA model used consensus models described in NUREG/CR-6850.

LSCS used guidance provided in NUREG/CR-6850 and NUREG-1855 (Reference [48]) to address uncertainties associated with FPRA for the 50.69 program application . As stated in Section 1.3 of NUREG-1855:

Although the guidance in this 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 defines:

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."

NUREG-1855 defines consensus model as:

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 NRG has utilized or accepted for the specific risk-informed application for which it is proposed."

The plant-specific assumptions in the LSCS FPRA (Reference [49]) and the 71 generic sources of uncertainty identified in EPRI 1026511 (Reference [50]) were evaluated for their potential impact Page 67

Enclosure Evaluation of the Proposed Change Attachment 6 on the 50.69 application. This guideline organizes the uncertainties in Topic Areas similar to those outlined in NUREG/CR-6850 and was used to evaluate the baseline FPRA epistemic uncertainty and evaluate the impact of this uncertainty on 50 .69 SSC component importance measures .

A detailed review of the generic and plant-specific sources of internal fire model uncertainties are discussed in LS-MISC-046 (Reference [51]) and are therefore not repeated in this attachment.

The purpose of this attachment is to summarize the key sources of uncertainty that could potentially impact the 50.69 application.

Table 6-2 summarizes the review for key sources of uncertainty in the internal fire PRA model for the 50 .69 application (organized by NUREG/CR-6850 tasks).

As noted above, the LSCS FPRA was developed using consensus methods outlined in NUREG/CR-6850 and interpretations of technical approaches as required by NRC. Fire PRA methods were based on NUREG/CR-6850, other more recent NUREGs, (e.g ., NUREG-7150 (Reference [55]), and published "frequently asked questions" (FAQs) for the Fire PRA.

The key sources of uncertainty identified in Table 6-2 do not present a significant impact on the LSCS 50 .69 application , and therefore, the fire PRA model is capable of producing accurate 50 .69 importance measure results.

Additionally, for the 50.69 program, the guidance in NEI 00-04 (Reference [1]) specifies that certain sensitivity studies be conducted for each PRA model to address key sources of uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g ., human error, common cause failure, maintenance probabilities , and manual suppression probabilities for fire) do not mask the SSC(s) importance. Regulatory Guide 1.174, Revision 3 (Reference [52]) cites NUREG-1855, Revision 1, as related guidance. In Section B of RG 1.174 (Reference [52]) , the guidance acknowledges specific revisions of NUREG-1855 to include changes associated with expanding the discussion of uncertainties.

Table 6-2 below describes the fire PRA sources of model uncertainty and their impact.

Table 6-2 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty Analysis Boundary This task establishes the overall Based on a review of the and Partitioning spatial scope of the analysis and assumptions and potential sources provides a framework for of sources of uncertainly organizing the data for the analysis. associated with this element, it is The partitioning features credited concluded that the methodology for are required to satisfy established the Analysis Boundary and industry standards. Partitioning task does not introduce any epistemic uncertainties that would affect the 50.69 program.

Therefore, this does not represent a key source of uncertainty for the LSCS 50.69 application.

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Enclosure Evaluation of the Proposed Change Attachment 6 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty Fire PRA Component This task involves the selection of The uncertainty associated with Selection components to be treated in the this task is related to the analysis in the context of initiating identification of all components that events and mitigation. The should be credited/linked in the potential sources of uncertainty FPRA. This source of uncertainty include those inherent in the is reduced as a result of multiple internal events PRA model as that overlapping tasks including the model provides the foundation for MSO expert panel , reviews of FPIE the FPRA. screened initiating events, screened containment penetrations, and screened ISLOCA scenarios. Additional internal reviews of analysis results further reduce the uncertainty associated with this task.

Based on a review of the assumptions and potential sources of uncertainty related to this element and the discussion above, it is concluded that the methodology for the Component Selection task does not introduce any epistemic uncertainties that would affect the 50.69 program.

Therefore, this does not represent a key source of uncertainty for the LSCS 50.69 application.

Cable Selection The selection of cables to be Additionally, as part of the Fire considered in the analysis is PRA, some components were identified using industry guidance conservatively assumed to be documents. The overall process is failed based on lack of cable data .

essentially the same as that used Components in this category are to perform the analyses to referred to as Unknown Location demonstrate compliance with 10 (UNL) components because CFR 50.48. specific cables were not identified for the components . Based on recent Fire PRA updates, the UNL components are mostly limited to Balance of Plant (BOP) systems.

A sensitivity analysis was performed to measure the risk associated with the assumption that these components fail in select fire scenarios. The sensitivity Page 69

Enclosure Evaluation of the Proposed Change Attachment 6 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty removed all UNL components from every fire scenario, as described in the Uncertainty & Sensitivity Analysis Notebook. Based on the results, the inclusion of the UNL components introduces moderate risk to both Fire CDF and LERF.

Although the sensitivity shows a moderate impact on Fire CDF and Fire LERF , complete removal of UNLs would not be considered realistic since those cables could be identified with detailed circuit analysis and those failures would exist in specific areas of the plant.

Also, the dominant fire scenarios are undeveloped full room burnouts that when refined with detailed fire modeling and fire scenario development would reduce the overall impact of the bounding sensitivity. Given that an informed approach was used to developing the assumed routing, the methodology employed by the Fire PRA is appropriate.

Based on a review of the assumptions and potential sources of uncertainty related to this element it is concluded that the methodology for the Cable Selection task does not introduce any epistemic uncertainties that would affect the 50.69 program.

Therefore , this does not represent a key source of uncertainty for the LSCS 50.69 application.

Qualitative Screening Qualitative screening was In the event a structure (location) performed ; however, some which could result in a plant trip structures (locations) were was incorrectly excluded, its eliminated from the global analysis contribution to CDF would be small boundary and ignition sources (with a CCDP commensurate with deemed to have no impact on the base risk). Such a location would FPRA (based on industry guidance have a negligible risk contribution and criteria) were excluded from to the overall FPRA.

the quantification based on Page 70

Enclosure Evaluation of the Proposed Change Attachment 6 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty qualitative screening criteria. The only criterion subject to uncertainty Based on a review of the is the potential for plant trip. assumptions and potential sources However, such locations would not of uncertainty related to this contain any features (equipment or element and the discussion above, cables identified in the prior two it is concluded that the tasks) and consequently are methodology for the Qualitative expected to have a low risk Screening task does not introduce contribution. any epistemic uncertainties that would affect the 50.69 program.

Therefore, this does not represent a key source of uncertainty for the LSCS 50.69 application.

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

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

PRA model includes structure to translate them into the appropriate induced initiator. Therefore, this does not represent a key source of uncertainty for the LSCS 50.69 application.

Fire Ignition Fire ignition frequency is an area The LSCS Fire PRA utilized the bin Frequencies with inherent uncertainty. Part of frequencies from NUREG/CR-2169 this uncertainty arises due to the (Reference [56]), which represents counting and related partitioning the most current approved source methodology. for bin frequencies. As such, some of the inherent conservatism However, the resulting frequency is associated with bin frequencies not particularly sensitive to from NUREG/CR-6850 was changes in ignition source counts. removed. A parametric uncertainty The primary source of uncertainty analysis using the Monte Carlo Page 71

Enclosure Evaluation of the Proposed Change Attachment 6 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty for this task is associated with the method is provided in the FPRA industry generic frequency values documentation.

used for the FPRA. This is Consensus approaches are because there is no specific employed in the model.

treatment for variability among plants along with some significant Based on a review of the conservatism in defining the assumptions and potential sources frequencies, and their associated of uncertainty related to this heat release rates. LSCS uses the element it is concluded that the ignition frequencies in NUREG-methodology for the Fire Ignition 2169 (Reference [56]) along with Frequency task does not introduce the revised heat release rates from any epistemic uncertainties that NUREG - 2178 (Reference [57]) .

would affect the 50 .69 program .

Therefore , this does not represent a key source of uncertainty for the LSCS 50.69 application.

Quantitative Other than screening out potentially Quantitative screening criteria was Screening risk significant scenarios (ignition defined for the LSCS Fire PRA as sources), this task is not a source the CDF I LERF contribution of of uncertainty. zero, such that all quantified fire scenarios are retained. All of the results were retained in the cumulative CDF I LERF, therefore, no uncertainty was introduced as a result of this task.

Based on the discussion above, it is concluded that the methodology for the Quantitative Screening task does not introduce any epistemic uncertainties that would affect the 50.69 program.

Therefore , this does not represent a key source of uncertainty for the LSCS 50.69 aoolication.

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

fire scenario development. These two tasks are Scoping Fire Modeling and Detailed Fire Modeling. The discussion of uncertainty for both tasks is provided in the discussion for Detailed Fire Modelino.

Page 72

Enclosure Evaluation of the Proposed Change Attachment 6 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty Detailed Circuit The circuit analysis is performed Circuit analysis was performed as Failure Analysis using standard electrical part of the deterministic post fire engineering principles. However, safe shutdown analysis.

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

are over-estimated.

Based on a review of the assumptions and potential sources of uncertainty related to this element and the discussion above, it is concluded that the methodology for the Detailed Circuit Failure Analysis task does not introduce any epistemic uncertainties that would affect the 50.69 program.

Therefore , this does not represent a key source of uncertainty for the LSCS 50.69 application.

Circuit Failure Model One of the failure modes for a The use of hot short failure Likelihood Analysis circuit (cable) given fire induced probability and duration probability failure is a hot short. A conditional is based on fire test data and probability and a hot short duration associated consensus probability are assigned using methodology published in NUREG-industry guidance published in 7150, Volume 2 (Reference [58]).

NUREG-7150, Volume 2 (Reference [58]). The uncertainty Based on a review of the values specified in NUREG-7150 ,

assumptions and potential sources Volume 2 are based on fire test of uncertainty related to this data.

element and the discussion above, it is concluded that the methodoloav for the Circuit Failure Page 73

Enclosure Evaluation of the Proposed Change Attachment 6 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty Mode Likelihood Analysis task does not introduce any epistemic uncertainties that would affect the 50.69 program.

Therefore , this does not represent a key source of uncertainty for the LSCS 50.69 application.

Detailed Fire The application of fire modeling Consensus modeling approach is Modeling technology is used in the FPRA to used for Detailed Fire Modeling translate a fire initiating event into a and it is concluded that the set of consequences (fire induced methodology for the Detailed Fire failures). The performance of the Modeling task does not introduce analysis requires a number of key any epistemic uncertainties that input parameters. These input would require sensitivity treatment.

parameters include the heat release rate (HRR) for the fire, the Therefore, this does not represent growth rate, the damage threshold a key source of uncertainty for the for the targets, and response of LSCS 50.69 application.

plant staff (detection, fire control, fire suppression).

The fire modeling methodology itself is largely empirical in some respects and consequently is another source of uncertainty. For a given set of input parameters, the fire modeling results (temperatures as a function of distance from the fire) are characterized as having some distribution (aleatory uncertainty). The epistemic uncertainty arises from the selection of the input parameters (specifically the HRR and growth rate) and how the parameters are related to the fire initiating event.

While industry guidance is available, that guidance is derived from laboratory tests and may not necessarily be representative of randomly occurring events.

The fire modeling results using these input parameters are used to identify a zone of influence (ZOI) for the fire and cables/equipment within that ZOI are assumed to be Page 74

Enclosure Evaluation of the Proposed Change Attachment 6 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertaintv damaged. In general , the guidance provided for the treatment of fires is conservative and the application of that guidance retains that conservatism . The resulting risk estimates are also conservative.

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

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

Additionally, for the 50.69 program, the guidance in NEI 00-04 (Reference [1]) specifies that certain sensitivity studies be conducted for each PRA model to address key sources of uncertainty.

The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g. , human error, common cause failure, maintenance probabilities, and manual suppression probabilities for fire) do not mask the SSC(s) importance.

It is concluded that the methodology for the Post-Fire Human Reliability Analysis task does not introduce any epistemic uncertainties that would require sensitivity treatment.

Therefore, this does not represent a key source of uncertainty for the LSCS 50.69 application.

Page 75

Enclosure Evaluation of the Proposed Change Attachment 6 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty Seismic-Fire Since this is a qualitative The qualitative assessment of Interactions evaluation, there is no quantitative seismic-induced fires should not be Assessment impact with respect to the a source of model uncertainty as it uncertainty of this task. is not expected to provide changes to the quantified Fire PRA model.

Based on the discussion above, it is concluded that the methodology for the Seismic-Fire Interactions Assessment task does not introduce any epistemic uncertainties that affect the 50.69 program .

Therefore , this does not represent a key source of uncertainty for the LSCS 50.69 application.

Fire Risk As the culmination of other tasks , The selected truncation was Quantification most of the uncertainty associated confirmed to be consistent with the with quantification has already requirements of the PRA Standard been addressed. The other source (Reference [59]).

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

Therefore, this does not represent a key source of uncertainty for the LSCS 50.69 application.

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

Additionally, for the 50.69 program, the guidance in NEI 00-04 (Reference [1]) specifies that certain sensitivity studies be Page 76

Enclosure Evaluation of the Proposed Change Attachment 6 Fire PRA Fire PRA Fire PRA Disposition Description Sources of Uncertainty conducted for each PRA model to address key sources of uncertainty.

The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, maintenance probabilities, and manual suppression probabilities for fire) do not mask the SSC(s) importance.

Based on the discussion above, it is concluded that the methodology for the Uncertainty and Sensitivity Analyses task does not introduce any epistemic uncertainties that would affect the 50.69 program.

Therefore , this does not represent a key source of uncertainty for the LSCS 50.69 application.

Fire PRA FPRA Documentation This task This task does not introduce any Documentation does not introduce any new new uncertainties to the fire risk as uncertainties to the fire risk. it outlines documentation requirements .

Based on the discussion above, it is concluded that the methodology for the Fire PRA documentation task does not introduce any epistemic uncertainties that would affect the 50.69 program.

Therefore, this does not represent a key source of uncertainty for the LSCS 50.69 aoolication.

Page 77

Enclosure Evaluation of the Proposed Change Attachment 7 : Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-I/II, CC-11/111, and CC-1/11/111 Capability Finding Supporting Disposition for Category Description Number Requirement(s) 50.69 (CC)

URE IFSO-A3 Not Met As part of the Self- Open LS2020- CCII Assessment performed 0001 IFSN-A7 during the 2014 FPIE These open issues PRA update, the have no impact on (Update IFQU-A3 following gaps to RG the 50.69 Requiring 1.200 (Rev. 2) and the implementation as Evaluation ASME/ANS PRA they are primarily Tracking ID) Standard were documentation identified : issues. However, these gaps will be

1. IFSO-A3 resolved during a Further documentation PRA update.

clarification is required for those flood locations that are screened out based on the quantitative screening criteria described in the PRA Standard.

2. IFSN-A7 Further documentation clarification is required for justification of crediting EQ limits for ensuring operability of instrumentation given spray-induced impacts.

3. IFQU-A3 Further documentation clarification is required for those flood locations that are screened out based on the quantitative screening criteria described in the PRA Standard.

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v t e Letter Sequence Request
EPID:L-2020-LLA-0018, TSTF-505 (Approved, Closed) EPID:L-2020-LLA-0017, TSTF-505, TSTF-505 (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance Administration Questions 29 September 2020 16 September 2020 1 March 2021 29 April 2021 Answers 29 October 2020 Results Approval, Approval, Approval, Approval Other: ML20090F585, ML20160A164, ML20248H494
MONTHYEAR2020-09-17 [Table View]

RS-20-011, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors

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RS-20-011, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors

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4.3 CONCLUSION

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4.3 CONCLUSION

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