W3F1-2022-0009, Supplement to Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Reactors

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

ML22115A062
Person / Time
Site:	Waterford
Issue date:	04/25/2022
From:	Couture P Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2022-0009
Download: ML22115A062 (94)

Text

Phil Couture Senior Manager Fleet Regulatory Assurance - Licensing 601-368-5102 W3F1-2022-0009 10 CFR 50.90 April 25, 2022 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Supplement to Application to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors" Waterford Steam Electric Station, Unit 3 NRC Docket No. 50-382 Renewed Facility Operating License No. NPF-38 By letter dated December 18, 2020 (Reference 1), Entergy Operations, Inc. (Entergy) requested an amendment to Renewed Facility Operating License (FOL) No. NPF-38 for Waterford Steam Electric Station, Unit 3 (Waterford 3). 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."

On June 25, 2021, the U.S. Nuclear Regulatory Commission (NRC) provided a draft request for additional information (RAI) to Entergy by electronic mail (email) regarding the Reference 1 license amendment request (LAR). A conference call was held on July 7, 2021 between the NRC and Entergy for clarification of the previously issued draft RAI questions. By email dated July 26, 2021 (Reference 2), the NRC issued the final RAI.

The responses to the NRC RAI were provided in a letter dated October 1, 2021 (Reference 3).

Clarification calls between the NRC and Entergy were held on November 4, 2021 and January 18, 2022 to discuss additional questions raised concerning the LAR, the previous RAI responses, and the acceptability of the Waterford 3 application of the alternate Seismic Tier 1 categorization process. At the conclusion of these clarification calls, there was consensus agreement that in order to move forward with the LAR review and approval process, Entergy would have to alter its approach to addressing the seismic risk by pursuing the alternate Seismic Tier 2 categorization process.

Entergy has decided to pursue the alternate Seismic Tier 2 categorization process based on the guidance provided in Electric Power Research Institute (EPRI) Report 3002017583, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," dated Entergy Operations, Inc., 1340 Echelon Parkway, Jackson, MS 39213

W3F1-2022-0009 Page 2 of 4 February 2020 (Reference 4). This is the same EPRI Report that served as the basis for the alternate Seismic Tier 1 categorization process information provided in the Reference 3 RAI response.

This letter is a supplement to the Reference 1 LAR and Reference 3 RAI response to support the transition to the alternate Seismic Tier 2 approach. Enclosure 1 of this supplement letter contains the "Waterford 3 10 CFR 50.69 Tier 2 License Amendment Request Supplement Report" which provides the information needed to justify the application of the alternate Seismic Tier 2 categorization process for application of the alternate Seismic Tier 2 categorization process for implementation of the risk-informed categorization and treatment of the structures, systems, and components (SSCs) for Waterford 3 in accordance with the requirements of 10 CFR 50.69. The information is identified by the Sections in the Reference 1 LAR Enclosure that discussed the alternate Seismic Tier 1 approach and supersedes that information in its entirety. Revision bars have not been included in the Enclosure 1 Tier 2 report since it is essentially a re-write of the affected Sections.

The Reference 3 RAI response letter included five commitments. Enclosure 2 of this supplement letter provides an updated commitment list. The delay resulting from the transition to the alternate Seismic Tier 2 approach necessitated extensions of the scheduled completion dates for two of the commitments. One of the commitments has been completed. The changes are indicated with revision bars. There are no new regulatory commitments included in this supplement letter. to this supplement letter provides a "Revised List of Categorization Prerequisites" and Attachment 2 provides a "Revised Section 6.0 Reference List." These revised lists supersede the lists previously provided as Attachment 1 and Section 6.0 of the Reference 1 LAR Enclosure, respectively. Revision bars are provided to indicate the changes.

The transition to the alternate Seismic Tier 2 approach also required revisions to Attachment 3, "Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items,"

and Attachment 6, "Disposition of Key Assumptions/Sources of Uncertainty," of the Reference 1 LAR. This supplement letter provides revised Attachments 3 and 6 which supersede the corresponding Attachments 3 and 6 provided in the Reference 1 LAR in their entirety. Revision bars are provided to indicate the changes to these Attachments. Note that no technical changes were made to Tables A3-3 or A3-4 of Attachment 3. These Tables are included in this supplement only for completeness. Also note that Attachments 2, 4, and 5 of the Reference 1 LAR are unchanged by this supplement and remain as previously submitted.

In addition, the transition to the alternate Seismic Tier 2 approach required minor changes to the Waterford 3 FOL License Condition wording that was proposed in Section 2.3 of the Reference 1 LAR Enclosure. Section 2.3 of the "Waterford 3 10 CFR 50.69 Tier 2 License Amendment Request Supplement Report" in Enclosure 1 of this supplement letter describes the change to the FOL License Condition, and Enclosures 3 and 4 contain the FOL markup pages and revised clean typed pages, respectively.

Entergy has reviewed the information supporting the No Significant Hazards Consideration and the Environmental Consideration that was previously provided to the NRC in the Reference 1 LAR. The information in this supplement to the LAR and Reference 3 RAI response does not alter the conclusion that the proposed change presents no significant hazards consideration and no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

W3F1-2022-0009 Page 3 of 4 Entergy requests approval of the Reference 1 LAR as supplemented by June 24, 2022 with 120 days allowed for implementation of the amendment.

In accordance with 10 CFR 50.91(b)(1), "Notice for public comment; State consultation," a copy of this letter is being provided to the designated State Official.

Should you have any questions or require additional information, please contact John Lewis, Regulatory Assurance Manager, Waterford 3, at 504-739-6028.

I declare under penalty of perjury, that the foregoing is true and correct. Executed on April 25, 2022.

Respectfully, Philip Digitally signed by Philip Couture Couture Date: 2022.04.25 06:43:49 -05'00' Phil Couture PC/cdm/ajh

Enclosures:

1. Waterford 3 10 CFR 50.69 Tier 2 License Amendment Request Supplement

2. List of Regulatory Commitments (revised status and completion dates for regulatory commitments provided in Enclosure 3 of Reference 3)

3. Renewed Facility Operating License - Markup Pages

4. Renewed Facility Operating License - Revised Clean Pages Attachments: 1. Revised List of Categorization Prerequisites (supersedes Attachment 1 to Enclosure of Reference 1)

2. Revised Section 6.0 Reference List (supersedes Section 6.0 of Enclosure of Reference 1)

3. Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items (supersedes Attachment 3 to Enclosure of Reference 1)

6. Disposition of Key Assumptions/Sources of Uncertainty (supersedes Attachment 6 to Enclosure of Reference 1)

W3F1-2022-0009 Page 4 of 4

References:

1) Entergy Operations, Inc. (Entergy) letter to U.S. Nuclear Regulatory Commission (NRC), "Application to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, System [sic], and Components (SSCs) for Nuclear Power Reactors," (ADAMS Accession No. ML20353A433), dated December 18, 2020

2) NRC electronic mail (email) message to Entergy, "Final RAIs to Entergy Operations, Waterford Steam Electric Station, Unit 3 - LAR to Adopt 10 CFR 50.69 (EPID L 2020 LLA-0279)," (ADAMS Accession No. ML21218A040), dated July 26, 2021

3) Entergy letter to NRC, "Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69," (ADAMS Accession No. ML21274A876), dated October 1, 2021

4) Electric Power Research Institute (EPRI) Technical Report 3003017583, "Alternate Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," dated February 2020 cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - Waterford Steam Electric Station, Unit 3 NRC Project Manager Waterford Steam Electric Station, Unit 3 Louisiana Department of Environmental Quality

Enclosure 1 W3F1-2022-0009 Waterford 3 10 CFR 50.69 Tier 2 License Amendment Request Supplement

W3F1-2022-0009 Page 1 of 25 WATERFORD 3 10 CFR 50.69 TIER 2 LICENSE AMENDMENT REQUEST SUPPLEMENT1 In Reference [1], Entergy Operations, Inc. (Entergy) requested an amendment to the Renewed Facility Operating License Nos. NPF-38 for Waterford Steam Electric Station, Unit 3 (Waterford 3).

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 (SSCs) for Nuclear Power Plants." 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.

By letter dated April 14, 2021 (Reference [2]), the U.S. Nuclear Regulatory Commission (NRC) notified Entergy of their intent to conduct a regulatory virtual audit from approximately April 14, 2021, through June 11, 2021, with formal audit meetings from June 1 to through June 3, 2021. On May 21, 2021 (Reference [3]), the NRC provided the agenda and audit questions. During the audit, questions were raised by the NRC concerning the acceptability of the application of the Seismic Tier 1 categorization process to Waterford 3. On October 1, 2021 (Reference [4]), Waterford 3 responded to the NRC's request for additional information (RAI) to allow NRC staff to complete their review. Subsequently, the NRC informed Waterford 3 that the seismic Core Damage Frequency (CDF) and seismic Large Early Release Frequency (LERF) values were not low enough relative to the total CDF and LERF metrics such that the Tier 1 approach would not be approved. Therefore, Entergy has decided to pursue the alternative Seismic Tier 2 categorization process provided in Section 2.3 of the same Electric Power Research Institute (EPRI) Report (Reference [5]) in order to move forward with the review of this LAR.

Therefore, this letter is a supplement to the Reference [1] license amendment request (LAR).

This attachment provides the information needed to justify the application of the seismic Tier 2 alternative categorization process for implementation of the risk-informed categorization and treatment of structures, systems, and components (SSCs) for Waterford 3 in accordance with the requirements of 10 CFR 50.69. The information is identified by the Sections within the original Reference [1] LAR that discussed the alternative seismic approach and supersedes that information in its entirety.

Also, because of submitting this supplement to transition to the alternative Seismic Tier 2 approach, a minor change is required to the wording of the operating license of Waterford 3 that was proposed in the original Reference [1] LAR.

1 The Reference numbering applies to this Enclosure only and does not match the original LAR application listed as Reference [1] in this Enclosure. The original LAR Reference list has been revised and is provided in Attachment 2 to this supplement letter.

W3F1-2022-0009 Page 2 of 25

2.3 DESCRIPTION

OF THE PROPOSED CHANGE The description of the proposed change in Reference [1] currently states:

"Entergy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports. The evaluation of impact of the seismic hazard uses the approach for seismic Tier 1 sites documented in Electric Power Research Institute (EPRI)

Technical Update 3002017583, 'Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization,' which includes Waterford 3. Entergy/Waterford 3 will use a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; 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)."

The license condition is proposed to be revised as follows:

"Entergy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Entergy's submittal letter dated December 18, 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)."

Enclosures 3 and 4 of this supplement letter contain the markup pages and revised clean typed pages, respectively, for the Waterford 3 Renewed Facility Operating License.

W3F1-2022-0009 Page 3 of 25

3.0 TECHNICAL EVALUATION

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

3.1.1 Overall Categorization Process (Note: This Section replaces Section 3.1.1 of Reference [1] in its entirety.)

Waterford 3 will implement the risk categorization process in accordance with the Nuclear Energy Institute (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 [6]). NEI 00-04 Section 1.5 states, "Due to the varying levels of uncertainty and degrees of conservatism in the spectrum of risk contributors, the risk significance of SSCs is assessed separately from each of five risk perspectives and used to identify SSCs that are potentially safety-significant." A separate evaluation is appropriate to avoid reliance on a combined result that may mask the results of individual risk contributors.

The process to categorize each system will be consistent with the guidance in NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," as endorsed by RG 1.201, with the exception of the evaluation of impact of the seismic hazard, which will use the EPRI 3002017583 (Reference

[5]) approach for seismic Tier 2 sites to assess seismic hazard risk for 10 CFR 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 10 CFR 50.69(c)(l)(iv)." However, neither RG 1.201 nor NEI 00-04 prescribe a particular sequence or order for each of the elements to be completed. Therefore, the order in which each of the elements of the categorization process (listed below) is completed is flexible and as long as they are all completed, they may even be performed in parallel. Note that NEI 00-04 only requires Item 3 to be completed for components/functions categorized as 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 (Fire SSEL), Seismic Safe Shutdown Equipment List (Seismic SSEL), other external events screening, and shutdown assessment)

3) Seven qualitative criteria in Section 9.2.2 of NEI 00-04

W3F1-2022-0009 Page 4 of 25

4) The defense-in-depth assessment

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

Figure 1 Categorization Process Overview2 Define System Boundaries Define System Functions and Assign Components to Functions Identify Seismic Insights Risk Characterization Defense in Depth Characterization Passive Characterization Qualitative Characterization NonPRA Modeled PRA Modeled Core Damage Containment Evaluation Evaluation Evaluation Evaluation Cumulative Risk Sensitivity Study Preliminary Component Categorization LSS or Can be HSS and can Overturned not be Overturned Review Seismic Insights IDP Review Component Categorization Categorization of SSCs will be completed per the NEI 00-04 process, as endorsed by RG 1.201, which includes the determination of safety significance through the various elements identified above. The results of these elements are used as inputs to arrive at a preliminary component categorization (i.e., High Safety Significant (HSS) or 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 2 This Figure is shown as Figure 3-1 in the original LAR [1] but is renumbered here.

W3F1-2022-0009 Page 5 of 25 as HSS if any element of the process results in a preliminary HSS determination in accordance with Table 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 1 summarizes these IDP limitations in NEI 00-04. The steps of the process are performed at either the function level, component level, or both. This is also summarized in the Table 1. A component is assigned its final RISC category upon approval by the IDP.

Table 1: Categorization Evaluation Summary3 IDP Categorization Drives Change Element Step - NEI 00-04 Evaluation Level Associated HSS to Section Functions LSS 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 Allowed Risk (Non-modeled) Seismic Function/Component (See Note 1) No Shutdown -

Function/Component Not Allowed No Section 5.5 3

This Table is shown as Table 3-1 in the original LAR [1] but is renumbered here.

W3F1-2022-0009 Page 6 of 25 Table 1: Categorization Evaluation Summary (Continued)

IDP Categorization Drives Change Element Step - NEI 00-04 Evaluation Level Associated HSS to Section Functions LSS Core Damage -

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

Component Not Allowed Yes Section 6.2 Considerations - Allowable Qualitative Criteria Function N/A Section 9.2 (See Note 2)

Passive - Section Passive Segment/Component Not Allowed No 4

Notes:

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

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

The seven considerations are addressed preliminarily by the 10 CFR 50.69 categorization team for at least the system functions that are not found to be HSS due to any other categorization step. Each of the seven considerations requires a supporting justification for confirming (true response) or not confirming (false response) that consideration. If the 10 CFR 50.69 categorization team determines that one or more of the seven considerations cannot be confirmed, then that function is presented to the IDP as preliminary HSS. Conversely, if all the seven considerations are confirmed, then the function is presented to the IDP as preliminary LSS.

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

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

The mapping of components to system functions is used in some categorization process steps to facilitate preliminary categorization of components. Specifically, functions with mapped components that are determined to be HSS by the PRA-based assessment (i.e., Internal events PRA or Integral PRA assessment) or defense-in-depth evaluation will be initially treated as HSS. However, NEI 00-04 Section 10.2 allows detailed categorization which can result in some components mapped to HSS functions being treated as LSS; and Section 4.0 discusses additional functions that may be identified (e.g., fill and drain) to group and consider potentially LSS components that may have been initially associated with an HSS function but which do not support the critical attributes of that HSS function. Note that certain steps of the categorization

W3F1-2022-0009 Page 7 of 25 process are performed at a component level (e.g., Passive, Non-PRA-modeled hazards - see Table 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 a 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 1 above or may remain LSS. For the seismic hazard, seismic considerations are not required to drive an HSS determination at the component level, but the IDP will consider available seismic information pertinent to the components being categorized and can, at its discretion, determine that a component should be HSS based on that information.

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

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

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

The decision criteria for the IDP for categorizing SSCs as safety significant or low safety-significant pursuant to 10 CFR 50.69(f)(1) will be documented in Entergy 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 HSS 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 risk evaluation 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 [7]) which states "if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of

W3F1-2022-0009 Page 8 of 25 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/Off-Normal Operating Procedures, Waterford 3 will not take credit for alternate means unless the alternate means are proceduralized and included in Licensed Operator training.

Waterford 3 proposes to apply an alternative seismic approach to those listed in NEI 00-04 Sections 1.5 and 5.3. This approach is specified in EPRI 3002017583 (Reference [5]) 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 model Revision 6 (refer to Attachment 24).

Fire Risks: Fire PRA model Revision 6 (refer to Attachment 24).

Seismic Risks: EPRI Alternative Approach in EPRI 3002017583 (Reference [5]) for Tier 2 plants with the additional considerations discussed in Section 3.2.3 of this letter.

Other External Risks (e.g., tornados, external floods, etc.): These were determined to be insignificant contributors to plant risk. A 2017 analysis (PSA-WF3-07 Waterford 3 Re-Examination of External Events Evaluation in the IPEEE - Reference [8]) was valid and to account for updated events/consequence data and plant design. This evaluation of external hazards was performed using Part 6 of the ASME/ANS PRA Standard (Reference [9]).

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 [10]), which provides guidance for assessing and enhancing safety during shutdown operations.

4 This refers to Attachment 2 of Reference [1] (i.e., the original application).

W3F1-2022-0009 Page 9 of 25 A change to the categorization process that is outside the bounds specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach) will not be used without prior NRC approval. The SSC categorization process documentation will include the following elements:

1) Program procedures used in the categorization

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

3) Mapping of components to support function(s)

4) PRA model results, including sensitivity studies

5) Hazards analyses, as applicable

6) Passive categorization results and bases

7) Categorization results including all associated bases and RISC classifications

8) Component critical attributes for HSS SSCs

9) Results of periodic reviews and SSC performance evaluations

10) IDP meeting minutes and qualification/training records for the IDP members 3.2 Technical Adequacy Evaluation (10 CFR 50.69(b)(2)(ii))

3.2.3 Seismic Hazards (Note: This Section replaces Section 3.2.3 of Reference [1] in its entirety.)

The use of PRA to assess risk from internal events is required by 10 CFR 50.69(c)(1). For other risk hazards, such as seismic, 10 CFR 50.69(b)(2) allows, and NEI 00-04 (Reference [6])

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 Waterford 3 seismic hazard assessment, Entergy 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]5 with the EPRI markups provided in Attachment 2 of References [11] and [12], and includes additional considerations that are discussed in this Section.

(Note: The discussion below pertaining to Reference [5] includes the markups provided in Attachment 2 of References [11] and [12]).

The proposed categorization approach for Waterford 3 is a risk informed graded approach that is demonstrated to produce categorization insights equivalent to a seismic PRA. This approach 5 EPRI 3002017583 is an update to EPRI 3002012988, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," July 2018 (Reference [40]) which was referenced in the NRC issued amendment and SE for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, to implement 10 CFR 50.69 as noted below:

(1) Calvert Cliffs Nuclear Power Plant, Units 1 and 2, "Issuance of Amendment Nos. 332 and 310 Re: Risk-Informed Categorization and Treatment of Systems, Structures, and Components (EPID L-2018-LLA-0482)," February 28, 2020.

(ADAMS Accession No. ML19330D909) (Reference [40]).

(2) This license amendment request incorporates by Reference the Clinton Power Station, Unit 1 response to request for additional information letter of November 24, 2020 (ML20329A433) (Reference [42]), in particular, the response to the question regarding the differences between the initial EPRI report 3002012988 and the current EPRI report 3002017583.

W3F1-2022-0009 Page 10 of 25 relies on the insights gained from the seismic PRAs examined in Reference [5] and plant specific insights considering seismic correlation effects and seismic interactions. During the NRC Audit conducted in June 2021 for this LAR (Reference [1]), questions were raised by the NRC concerning the acceptability of the application of the alternative Seismic Tier 1 approach.

As a result, Entergy is submitting Waterford 3 as a Tier 2 site.

Per Reference [5], for Tier 2 sites, the site Ground Motion Response Spectrum (GMRS) to Safe Shutdown Earthquake (SSE) comparison is above the Tier 1 threshold but not high enough that the NRC required the plant to perform a Seismic PRA (SPRA) to respond to Recommendation 2.1 of the Near-Term Task Force 50.54(f) letter (Reference [13]). Reference [5] 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 10 CFR 50.69 categorization process specified in NEI 00-04.

The trial studies in Reference [5], as amended by their RAI responses and amendments (References [14], [15], [16], [17], [18], [19], [20], [21], [22]) 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 seismic insights to the categorization process. That is the basis for the following statements in Table 4-1 of Reference [5].

"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), there is no need to perform more detailed evaluations to demonstrate the inherent seismic capacities documented in industry sources such as Reference [23]. Tier 2 seismic demand sites have a lower likelihood of seismically induced failures and less challenges to plant systems than trial study plants. This, therefore, provides the technical basis for allowing use of a graded approach for addressing seismic hazards at Waterford 3.

Test cases described in Section 3 of Reference [5], as amended by their RAI responses and amendments (References [14], [15], [16], [17], [18], [19], [20], [21], [22]) showed that there are very few, if any, SSCs that would be designated HSS for seismic unique reasons. The test cases identified that the unique seismic insights were typically associated with seismically correlated failures and led to unique HSS SSCs. While it would be unusual even for moderate hazard plants to exhibit any unique seismic insights, it is prudent and recommended by Reference [5] to perform additional evaluations to identify the conditions where correlated failures and seismic interactions may occur and determine their impact in the 10 CFR 50.69 categorization process. The special sensitivity study recommended in Reference [5] uses common cause failures, similar to the approach taken in a Full Power Internal Events (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.

W3F1-2022-0009 Page 11 of 25 Entergy is using test case information from Reference [5] developed by other licensees. The test case information is being incorporated by Reference into this application; specifically, Case Study A (Reference [14], [15], [16]), Case Study C (Reference [24], Case Study D (Reference

[25]), as well as RAI responses and amendments (References [14], [15], [16], [17], [18], [19],

[20], [21], [22]) that clarify aspects of these case studies.

Basis for Waterford 3 being a Tier 2 Plant Entergy considers the Waterford site to be a Tier 1 site in accordance with the EPRI (Reference [5]) criteria; however, as previously mentioned, Entergy is submitting Waterford 3 as a Tier 2 site. The EPRI (Reference [5]) Tier 2 criteria are as follows:

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

Note: Reference [5] applies to the Tier 2 sites in its entirety except for Sections 2.2 (Tier 1 sites) and 2.4 (Tier 3 sites).

For comparison, Tier 1 plants are defined as having a GMRS peak acceleration at or below approximately 0.2g or where the GMRS is below or approximately equal to the SSE between 1.0 Hz and 10 Hz. Tier 3 plants are defined where the GMRS to SSE comparison between 1.0 Hz and 10 Hz is high enough that the NRC required the plant to perform an SPRA to respond to the Fukushima 50.54(f) letter (Reference [13]).

As shown in Figure 2, comparing the Waterford 3 GMRS (derived from the seismic hazard) to the SSE (i.e., seismic design basis capability), the GMRS peak acceleration is at or below approximately 0.2g (as well as the GRMS is below the SSE in the 1.0 Hz to 10 Hz range),

meeting the Tier 1 criterion in Table 4-1 of Reference [5]. The NRC screened out Waterford 3 from performing an SPRA in response to the NTTF 2.1 50.54(f) letter (Reference [26]. As such, it is appropriate that Waterford 3 is considered a Tier 1 plant. However, given NRC concerns regarding the acceptability of Tier 1 for Waterford due to the relatively high (i.e., in comparison to Waterford 3 total calculated risk) estimated seismic LERF, Entergy has decided to pursue the Tier 2 approach. The basis for Waterford 3 being Tier 2 will be documented and presented to the IDP for each system categorized.

W3F1-2022-0009 Page 12 of 25 Figure 2: GMRS and SSE Response Spectra for Waterford 3 (From Reference [27])

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 [5] recommends a risk informed graded approach for addressing the seismic hazard in the 10 CFR 50.69 categorization process. There are a number of seismic fragility fundamental concepts that support a graded approach and there are important characteristics about the comparison of the seismic design basis (represented by the SSE) to the site-specific seismic hazard (represented by the GMRS) that support the selected thresholds between the three evaluation Tiers in the report. The coupling of these concepts with the categorization process in NEI 00-04 are the key elements of the approach defined in Reference [5] 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 [23])

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.

W3F1-2022-0009 Page 13 of 25 There are some plant features such as equipment anchorage that have seismic capacities more closely associated with the site-specific seismic demand since those specific features are specifically designed to meet that demand. However, even for these features, the design basis criteria have intended conservatisms that result in significant seismic margins within SSCs.

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

In applying the Reference [5] process for Tier 2 sites to the Waterford 3 10 CFR 50.69 categorization process, the IDP will be provided with the rationale for applying the Reference [5]

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 [5] study and the bases for Waterford 3 being considered a Tier 2 plant. The discussion of the Tier 2 bases will include such factors as:

The seismic hazard for the plant, The definition of Tier 2 in the EPRI study, and The basis for concluding Waterford 3 as a Tier 2 plant.

At several steps of the categorization process, (e.g., as noted in Figure 1 and Table 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 Waterford 3) 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 Waterford 3 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.

As part of the preparation steps for Tier 2 analyses, the categorization team will review available Waterford 3 plant specific seismic reviews and other resources such as those identified above and will pursue this information review in a manner that has been proposed in past Tier 1 LAR applications (Reference [28]). 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

W3F1-2022-0009 Page 14 of 25 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)

The above review of existing information will be used to inform the Tier 2 system walkdowns and analyses. For each system categorized, the categorization team will evaluate as part of the Tier 2 process correlated seismic failures and seismic interactions between SSCs. This process is detailed in Reference [5] Section 2.3.1 and is summarized in Figure 3.

Determination of seismic insights will make use of the FPIE PRA model (as a quantification vehicle to develop seismically biased risk estimates for system components) supplemented by focused seismic walkdowns. An overview of the process to determine the importance of SSCs for mitigating seismic events follows and is utilized on a system basis:

o Identify SSCs within the system to be categorized o Group SSCs within the system into the classes of equipment and distributed systems used for SPRAs o Refine the list and screen out the following SSCs from consideration of functional correlated seismic failures:

Inherently rugged components Components not used in safety functions that support mitigation of core damage or containment performance Components already identified as HSS components from the Internal Events PRA or Integrated assessment o Perform a seismic walkdown:

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

Screen out IF SSCs have high seismic capacity AND not included in seismically correlated groups or correlated interaction groups

W3F1-2022-0009 Page 15 of 25 o Add surrogate basic events to the FPIE PRA logic model that simulate spatial interaction or Correlation (for the system being categorized) - set the probability of failure to 1E-04 for each surrogate basic event o Quantify the FPIE model (for the system being categorized) for Loss of Offsite Power (LOOP) and Small Loss of Coolant Accident (SLOCA) initiated accident sequences setting (1) the LOOP initiating event frequency to 1.0/yr, (2) the SLOCA initiating event frequency to 1E-02/yr, and (3) the initiating event frequency for all initiators other than LOOP and SLOCA initiators to 0 (zero), and also removing credit for restoration of offsite power in LOOP/Station Blackout (SBO) accident sequences as well as other functional recoveries o Utilize the Importance Measures from this sensitivity study to identify appropriate SSCs (in the system being categorized) that should be HSS due to correlation or seismic interactions Seismic impacts would be compiled on an SSC basis. As each system is categorized, the system specific seismic insights will be documented in the categorization report and provided to the IDP for consideration as part of the IDP review process (e.g., Figure 1). The IDP cannot challenge any candidate HSS recommendation for any SSC from a seismic perspective if they believe there is a basis, except for certain conditions identified in Step 10 of Section 2.3.1 of Reference [5]. Any decision by the IDP to downgrade preliminary HSS components to LSS will consider the applicable seismic insights in that decision. SSCs identified from the Fire PRA as candidate HSS, which are not HSS from the internal events PRA or integrated importance measure assessment, will be reviewed for their design basis function during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events.

These insights will provide the IDP a means to consider potential impacts of seismic events in the categorization process.

If the seismic hazard or industry guidance changes at some future time, prior NRC approval, under 10 CFR 50.90, will be requested if Waterford 3's feedback process determines that a process different from the proposed alternative seismic approach is warranted for seismic risk consideration in categorization under 10 CFR 50.69. After receiving NRC approval, Entergy will follow its categorization review and adjustment process to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e) and the EPRI 3002017583 SSC categorization criteria for the updated Tier. This includes use of the Entergy corrective action process (CAP).

If the seismic hazard is further reduced or industry guidance changes such that EPRI 3002017583 Tier 1 processes are judged to be more appropriate for Waterford 3, Entergy will implement the following process:

a) For previously completed system categorizations, Entergy may review the categorization results to determine if use of the criteria in EPRI 3002017583 Section 2.2, "Low Seismic Hazard / High Seismic Margin Sites," would lead to categorization changes. If changes

W3F1-2022-0009 Page 16 of 25 are warranted, they will be implemented through the Entergy design control and corrective action programs and NEI 00-04, Section 12.

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

If the seismic hazard increases to the degree that an SPRA becomes necessary (i.e., EPRI 3002017583 Tier 3) for the 10 CFR 50.69 SSC categorization process, Entergy will implement the following process following completion of the SPRA, including adequate closure of Peer Review Findings and Observations:

a) For previously completed system categorizations, Entergy will review the categorization results using the SPRA insights as prescribed in NEI 00-04 Section 5.3, Seismic Assessment and Section 5.6, "Integral Assessment". If changes are warranted, they will be implemented through the Entergy design control and corrective action programs and NEI 00-04 Section 12.

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

Historical Seismic References for Waterford 3 The Waterford 3 GMRS and SSE curves from the seismic hazard and screening response are shown in Section 2.4 and 3.1, respectively, in the seismic hazard and screening report of Reference [27]. The Waterford 3 SSE and GMRS curves from Reference [27] are shown in Figure 2. The NRC's Staff assessment of the Waterford 3 seismic hazard and screening response is documented in Reference [26].

In the Staff Confirmatory Analysis (Section 3.4 of Reference [26]), the NRC concluded that the methodology used by Entergy in determining the GMRS was acceptable and that the GMRS determined by Entergy adequately characterizes the reevaluated hazard for the Waterford 3 site. Section 1.1.3 of Reference [5] cites various post Fukushima seismic reviews performed for the fleet of nuclear power plants. For Waterford 3, 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 [27], [26])

2. NTTF Recommendation 2.3 seismic walkdowns (References [29], [30], [31])

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

(References [32], [33])

The following additional post Fukushima seismic reviews were performed for Waterford 3:

4. NTTF Recommendation 2.1 seismic High Frequency Evaluation (References [34], [35])

W3F1-2022-0009 Page 17 of 25 Summary Based on the above, the Summary from Section 2.3.3 of Reference [5] applies to Waterford 3; namely, Waterford 3 meets the Tier 1 criteria described in Reference [5] but is pursuing 10 CFR 50.69 SSC categorization as a Tier 2 plant for which there may be a limited number of unique seismic insights, most likely attributed to the possibility of seismically correlated failures, appropriate for consideration in determining HSS SSCs. References [11] [12], and [36]6 are incorporated into this LAR, as applicable to Waterford 3, since they provide additional supporting bases for Tier 2 plants. In addition, References [37], [38], and [39] are incorporated into this LAR, as applicable, since they provide additional supporting bases for Tier 1 plants that are also used for Tier 2 plants. The Reference [4] recommended Tier 2 sensitivity study 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 IDP for the final HSS determinations. Use of the EPRI approach outlined in Reference [5] to assess seismic hazard risk for 10 CFR 50.69 with the additional reviews discussed above will provide a process for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs that satisfies the requirements of 10 CFR 50.69(c).

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

W3F1-2022-0009 Page 18 of 25 Figure 3: Seismic Correlated Failure Assessment for Tier 2 Plants7 7 Reproduced from Reference [5] Figure 2-3 including the markups provided in Attachment 2 of References [9] and [10].

W3F1-2022-0009 Page 19 of 25 3.5 Feedback and Adjustment Process (Note: This Section replaces Section 3.5 in Reference [1] in its entirety.)

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 Waterford 3 Tier 2 approach discussed in Section 3.2.3, implementation of the Entergy/Waterford 3 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 will be described in the Entergy/Waterford 3 10 CFR 50.69 program documents. The program requires that the periodic review assess changes that could impact the categorization results and provides the IDP with an opportunity to recommend categorization and treatment adjustments. Station personnel from engineering, operations, risk management, regulatory assurance, and others have responsibilities for preparing and conducting various performance monitoring tasks that feed into this process. The intent of the performance monitoring reviews is to discover trends in component reliability, to identify and reverse negative performance trends, and take corrective action if necessary. The Entergy 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 will include a checklist of configuration activities to recognize those systems that have been categorized in accordance with 10 CFR 50.69 to ensure that any physical change to the plant or change to plant documents is evaluated prior to implementing those changes.

The checklist includes:

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

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

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

Waterford 3 uses Entergys 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 Entergy/Waterford 3 10 CFR 50.69 program requires that SCDs cannot be approved by the IDP until the panels comments have been resolved to the satisfaction of the IDP. This includes

W3F1-2022-0009 Page 20 of 25 issues related to system-specific seismic insights considered by the IDP during categorization.

Scheduled periodic reviews at least once every other refueling outage 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 upgraded, a review of the SSC categorization will be performed.

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

W3F1-2022-0009 Page 21 of 25

[1] Entergy Operations, Inc. (Entergy) letter to U.S. Nuclear Regulatory Commission (NRC),

"Application to Adopt 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, System [sic], and Components (SSCs) for Nuclear Power Reactors,'" (ADAMS Accession No. ML20353A433), dated December 18, 2020

[2] NRC letter to Entergy, "Waterford Steam Electric Station, Unit 3 - Regulatory Audit In Support of Review of Application to Adopt 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, System, and Components for Nuclear Power Reactors,'

(EPID L-2020-LLA-0279)," (ADAMS Accession No. ML21099A002), dated April 14, 2021

[3] NRC letter to Entergy, "Waterford Steam Electric Station, Unit 3 - Audit Meeting Agenda and Audit Questions for License Amendment Request to Adopt 10 CFR 50.69,

'Risk-Informed Categorization and Treatment of Structures, System [sic], and Components for Nuclear Power Reactors,' (EPID L-2020-LLA-0279)," (ADAMS Accession No. ML21134A212), dated May 21, 2021

[4] Entergy letter to NRC, "Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Waterford Steam Electric Station, Unit 3," (ADAMS Accession No. ML21274A876), dated October 1, 2021

[5] Electric Power Research Institute (EPRI) 3002017583, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," February 2020

[6] Nuclear Energy Institute (NEI) Report 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, (ADAMS Accession No. ML052910035), dated July 2005

[7] NRC letter to Southern Nuclear Operating Company, Inc. (SNC), "Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments Re: Use of 10 CFR 50.69 (TAC NOS. ME9472 AND ME9473)," (ADAMS Accession No. ML14237A034), dated December 17, 2014

[8] Entergy Report, PSA-WF3-07-01, "Waterford 3 Re-Examination of External Events Evaluation in the IPEEE," November 2017

[9] American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) RA-S-2009, Addenda to ASME/ANS RA-S-2008, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications,"

February 2009

[10] Nuclear Management and Resources Council, Inc. (NUMARC) Report 91-06, "Guidelines for Industry Actions to Assess Shutdown Management," (ADAMS Accession No. ML14365A203), dated December 1991

[11] Exelon Generation Company, LLC. (EGC) letter to NRC, LaSalle County Station, Units 1 and 2, Renewed Facility Operating License Nos. NPF-11 and NPF-18, NRC Docket Nos.

50-373 and 50-374, "Response to Request for Additional Information Regarding 'LaSalle License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors,' (EPID L-2020-LLA-0017)," (ADAMS Accession No. ML20290A791), dated October 16, 2020

W3F1-2022-0009 Page 22 of 25

[12] EGC letter to NRC, LaSalle County Station, Units 1 and 2, Renewed Facility Operating License Nos. NPF-11 and NPF-18, NRC Docket Nos. 50-373 and 50-374, "Response to Request for Additional Information Regarding the License Amendment Request to Adopt 10 CFR 50.69,' (EPID L-2020-LLA-0017)," (ADAMS Accession No. ML21022A130), dated January 22, 2021

[13] NRC letter to all Power Reactor Licensees and Holders of Construction Permits in Active or Deferred Status, "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," (ADAMS Accession No ML12053A340), dated March 12, 2012

[14] EGC letter to NRC, Peach Bottom Atomic Power Station, Units 2 and 3, "Seismic Probabilistic Risk Assessment Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," (ADAMS Accession No. ML18240A065), dated August 28, 2018

[15] NRC letter to EGC, "Peach Bottom Atomic Power Station, Units 2 and 3 - Staff Review of Seismic Probabilistic Risk Assessment Associated with Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic (EPID NO.

L-2018-JLD-0010)," (ADAMS Accession No. ML19053A469), dated June 10, 2019

[16] NRC letter to EGC, "Peach Bottom Atomic Power Station, Units 2 and 3 - Correction Regarding Staff Review of Seismic Probabilistic Risk Assessment Associated With Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic (EPID NO. L-2018-JLD-0010)," (ADAMS Accession No. ML19248C756), dated October 8, 2019

[17] SNC letter to NRC, "Vogtle Electric Generating Plant, Units 1 and 2, License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process," (ADAMS Accession No. ML17173A875), dated June 22, 2017

[18] NRC letter to SNC, "Vogtle Electric Generating Plant], Units 1 and 2 [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),"

(ADAMS Accession No. ML18180A062), dated August 10, 2018

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

[20] TVA letter to NRC, "Tennessee Valley Authority (TVA) - Watts Bar [Plant D] Nuclear Plant Seismic Probabilistic Risk Assessment Supplemental Information," (ADAMS Accession No. ML18100A966), dated April 10, 2018

[21] NRC letter to TVA, "Watts Bar Nuclear Plant, Units 1 and 2 [Plant D] - Staff Review of Seismic Probabilistic Risk Assessment Associated With Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic (CAC NOS.

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

dated July 10, 2018

W3F1-2022-0009 Page 23 of 25

[22] NRC letter to TVA, "Watts Bar Nuclear Plant, Units 1 and 2 [Plant D] - Issuance of Amendment Nos. 134 And 38 Regarding Adoption of Title 10 of the Code of Federal Regulations Section, 'Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants' (EPID L-2018-LLA-0493)," (ADAMS Accession No. ML20076A194), dated April 30, 2020

[23] EPRI NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin", Revision 1, August 1991

[24] SNC letter to NRC, "Vogtle Electric Generating Plant, Units 1 & 2, License Amendment Request to Incorporate Seismic Probabilistic Risk Assessment into the 10 CFR 50.69 Categorization Process, Response to Request for Additional Information (RAIs 4-11),"

(ADAMS Accession No. ML18052B342), dated February 21, 2018

[25] TVA letter to NRC, "Watts Bar Nuclear Plant, Units 1 and 2 [Plant D], Application to Adopt 10 CFR 50.69, 'Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors' (WBN-TS-17-24)," (ADAMS Accession No. ML18334A363), dated November 29, 2018

[26] NRC letter to Entergy, "Waterford Steam Electric Station, Unit 3 - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulation 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 NO.

MF3712)," (ADAMS Accession No. ML15335A050), dated December 15, 2015

[27] Entergy letter to NRC, "Entergy Seismic Hazard and Screening Report (CEUS Sites),

Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, Waterford Steam Electric Station, Unit 3 (Waterford 3),"

(ADAMS Accession No. ML14086A427), dated March 27, 2014

[28] EGC letter to NRC, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-53 and DPR-69, NRC Docket Nos. 50-317 and 50-318, "Revised submittal to Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors,"

(ADAMS Accession No. ML19130A180), dated May 10, 2019

[29] Entergy letter to NRC, "Seismic Walkdown Report - Entergys Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, Waterford Steam Electric Station, Unit 3 (Waterford 3),"

(ADAMS Accession No. ML12333A276), dated November 26, 2012

[30] Entergy letter to NRC, "Seismic Walkdown Report Revision 1 - Planned Update to Entergy's Response to NRC Request for Information Pursuant to 10 CFR 50.54(f)

Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, Waterford Steam Electric Station, Unit 3 (Waterford 3)," (ADAMS Accession No. ML13120A460), dated April 25, 2013

[31] NRC letter to Entergy, "Waterford Steam Electric Station, Unit 3 - Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force

W3F1-2022-0009 Page 24 of 25 Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident (TAC NO. MF0191," (ADAMS Accession No. ML14087A181), dated May 8, 2014

[32] Entergy letter to NRC, "NEI 12-06, Appendix H, Revision 2, H.4.2 Path 2: GMRS SSE with High Frequency Exceedances, Mitigating Strategies Assessment (MSA) report for the New Seismic Hazard Information, Waterford Steam Electric Station, Unit 3 (Waterford 3),"

(ADAMS Accession No. ML16235A337), dated August 22, 2016

[33] NRC letter to Entergy, "Waterford Steam Electric Station, Unit 3 - Staff Review of Mitigation Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed In Response to the March 12, 2012, 50.54(f) Letter," (ADAMS Accession No. ML16245A890), dated September 19, 2016

[34] Entergy letter to NRC, "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, Waterford Steam Electric Station, Unit 3 (Waterford),"

(ADAMS Accession No. ML15350A389), dated December 16, 2015

[35] NRC letter to the Power Reactor Licensees on the Enclosed List, "Staff Review of High Frequency Confirmation Associated With Reevaluated Seismic Hazard in Response to March 12, 2012 50.54(f) Request for Information," (ADAMS Accession No. ML15364A544), dated February 18, 2016

[36] EGC letter to NRC, LaSalle County Station, Units 1 and 2, Renewed Facility Operating License Nos. NPF-11 and NPF-18, NRC Docket Nos. 50-373 and 50-374, "Response to Request for Additional Information regarding LaSalle License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors' (EPID L-2020-LLA-0017)," (ADAMS Accession No. ML20275A292), dated October 1, 2020

[37] EGC letter to NRC, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, "Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69,

'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors,'" (ADAMS Accession No. ML19183A012), dated July 1, 2019

[38] EGC letter to NRC, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, "Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69,

'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors,'" (ADAMS Accession No. ML19200A216), dated July 19, 2019

[39] EGC letter to NRC, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, "Revised Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69,

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

[40] NRC letter to EGC, "Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 332 and 310 RE: Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, (EPID L-2018-LLA-0482)," (ADAMS Accession No. ML19330D909), February 28, 2020

W3F1-2022-0009 Page 25 of 25

[41] EPRI Technical Update 3002012988, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," July 2018

[42] EGC letter to NRC, Clinton Power Station, Unit 1, "Response to Request for Additional Information Regarding License Amendment Requests to Adopt TSTF-505, Revision 2, and 10 CFR 50.69," (ADAMS Accession No. ML20329A433), dated November 24, 2020

Enclosure 2 W3F1-2022-0009 List of Regulatory Commitments*

• Revised status and completion dates for regulatory commitments previously provided in Enclosure 3 of response to NRC request for additional information regarding license amendment to adopt 10 CFR 50.69 (see Reference 3 of cover for this supplement letter).

W3F1-2022-0009 Page 1 of 2 The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

Type (check one) Scheduled Commitment One-Time Continuing Completion Action Compliance Date The systematic review of key assumptions and sources of uncertainty in PSA-WF3-08-06 will be updated to include more thorough documentation of the process used prior to categorization of any SSCs. This update will include a review of plant specific assumptions to verify that all key assumptions and sources of uncertainty are captured for disposition for consideration in this application. The screening process documentation will be updated to list the specific screening criteria used in determining if assumptions and Complete sources of uncertainty that are "key" for their respective hazards are also potentially "key" for the application. The specific criteria to be used are those listed in the response to APLA 2.c. Additionally, the documentation update will include the aspects of NUREG 1855 that are employed.

This update will denote the FLEX system as a key source of uncertainty for this application.

The next Fire PRA model revision will include an update to ignition frequencies. December 2023 If any of the systems (EFW, DC, ID, SSD, 4KV) for which the sensitivity study shows changes in the safety significance in the fire PRA model are selected for categorization prior to the update of the ignition frequencies to the industry consensus approach (currently December 2023 NUREG 2169), the results will be shared with IDP members during review to ensure they are both aware of the model limitations, but also that these limitations are related to a limited subset of components.

W3F1-2022-0009 Page 2 of 2 Type (check one) Scheduled Commitment One-Time Continuing Completion Action Compliance Date The results of the FLEX equipment sensitivity study will be shared with the 10 CFR 50.69 Integrated Decision-making Panel (IDP) for categorization of systems shown to have changes in safety significance (EDG, EFW, and FLEX) to ensure they are aware of the May 2022 impact of FLEX modeling when making decisions. The IDP will have the list of impacted components and systems to allow for informed decision-making for all 10 CFR 50.69 categorization efforts.

In the section "Monitoring of Inputs to and Outcome of Proposed Alternative Seismic Approach" of the CCNPP SE, the configuration control program for CCNPP had been updated to include a checklist of configuration activities to recognize those systems that have been categorized in accordance with 10 CFR 50.69, to ensure that any physical change to the plant or change to plant documents is evaluated prior to May 2022 implementing those changes. This checklist is the same as what is included in Section 3.5 of the Waterford 3 LAR except for "Review of impact to seismic loading and SSE seismic requirements, as well as the method of combining seismic components." This checklist item will also be included in the Entergy configuration control program.

Enclosure 3 W3F1-2022-0009 Renewed Facility Operating License Markup Pages (4 pages follow)

the NRC of any action by equity investors or successors in interest to Entergy Louisiana, LLC that may have an effect on the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

1. Maximum Power Level EOI is authorized to operate the facility at reactor core power levels not in excess of 3716 megawatts thermal (100% power) in accordance with the conditions specified herein.

2. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. XXX, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. Antitrust Conditions (a) Entergy Louisiana, LLC shall comply with the antitrust license conditions in Appendix C to this renewed license.

(b) Entergy Louisiana, LLC is responsible and accountable for the actions of its agents to the extent said agent's actions contravene the antitrust license conditions in Appendix C to this renewed license.

AMENDMENT NO. 258, 259, 261, 262, XXX

(a) The information in the FSAR supplement, submitted pursuant to 10 CFR 54.21(d) and as revised during the license renewal application review process, and licensee commitments as listed in Appendix A of the "Safety Evaluation Report Related to the License Renewal of Waterford Steam Electric Station Unit 3," are collectively the "License Renewal FSAR Supplement." This Supplement is henceforth part of the FSAR, which will be updated in accordance with 10 CFR 50.71(e). As such, EOI may make changes to the programs, activities, and commitments described in this Supplement, provided the EOI evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59, "Changes, Tests, and Experiments," and otherwise complies with the requirements in that section.

(b) The License Renewal FSAR Supplement, as defined in license condition 21(a) above, describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).

(1) EOI shall implement those new programs and enhancements to existing programs no later than 6 months before the PEO.

(2) EOI shall complete those activities by the 6 month date prior to the PEO or to the end of the last refueling outage before the PEO, whichever occurs later.

(3) EOI shall notify the NRC in writing within 30 days after having accomplished item (b)(1) above and include the status of those FOL activities that have been or remain to be completed in item (b)(2)

INSERT above.

D. The facility requires an exemption from certain requirements of Appendices E and J to 10 CFR Part 50. These exemptions are described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 10 (Section 6.1.2) and Supplement No. 8 (Section 6.2.6), respectively. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. These exemptions are, therefore, hereby granted pursuant to 10 CFR 50.12. With the granting of these exemptions, the facility will Will operate, to the extent authorized herein, in conformity with the application, as roll to amended, the provisions of the Act, and the rules and regulations of the next Commission.

page E. EOI shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plan, which contains Safeguards Information protected under 10 CFR 73.21, is entitled:

"Physical Security, Safeguards Contingency and Training & Qualification Plan," and was submitted on October 4, 2004.

AMENDMENT NO. XXX

EOI shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The EOI CSP was approved by License Amendment No. 234 and supplemented by a change approved by Amendment Nos. 239, 241, and 247.

F. Except as otherwise provided in the Technical Specifications or the Environmental Protection Plan, EOI shall report any violations of the requirements contained in Section 2.C of this renewed license in the following manner. Initial notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC Operations Center via the Emergency Notification System with written follow-up within 30 days in accordance with the procedures described in 10 CFR 50.73(b), (c) and (e).

G. Entergy Louisiana, LLC shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

H. This renewed license is effective as the date of issuance and shall expire at midnight on December 18, 2044.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Ho K. Nieh, Director Office of Nuclear Reactor Regulation

Enclosures:

1. (DELETED)

2. Attachment 2

3. Appendix A (Technical Specifications) (NUREG-1117)

4. Appendix B (Environmental Protection Plan)

5. Appendix C (Antitrust Conditions)

Date of Issuance: December 27, 2018 AMENDMENT NO. XXX

FOL INSERT

22. 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants Entergy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Entergy's submittal letter dated December 18, 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).

Enclosure 4 W3F1-2022-0009 Renewed Facility Operating License Revised Clean Pages (3 pages follow)

the NRC of any action by equity investors or successors in interest to Entergy Louisiana, LLC that may have an effect on the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

1. Maximum Power Level EOI is authorized to operate the facility at reactor core power levels not in excess of 3716 megawatts thermal (100% power) in accordance with the conditions specified herein.

2. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. XXX, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. Antitrust Conditions (a) Entergy Louisiana, LLC shall comply with the antitrust license conditions in Appendix C to this renewed license.

(b) Entergy Louisiana, LLC is responsible and accountable for the actions of its agents to the extent said agent's actions contravene the antitrust license conditions in Appendix C to this renewed license.

AMENDMENT NO. 258, 259, 261, 262, XXX

(c) The information in the FSAR supplement, submitted pursuant to 10 CFR 54.21(d) and as revised during the license renewal application review process, and licensee commitments as listed in Appendix A of the "Safety Evaluation Report Related to the License Renewal of Waterford Steam Electric Station Unit 3," are collectively the "License Renewal FSAR Supplement." This Supplement is henceforth part of the FSAR, which will be updated in accordance with 10 CFR 50.71(e). As such, EOI may make changes to the programs, activities, and commitments described in this Supplement, provided the EOI evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59, "Changes, Tests, and Experiments," and otherwise complies with the requirements in that section.

(d) The License Renewal FSAR Supplement, as defined in license condition 21(a) above, describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).

(4) EOI shall implement those new programs and enhancements to existing programs no later than 6 months before the PEO.

(5) EOI shall complete those activities by the 6 month date prior to the PEO or to the end of the last refueling outage before the PEO, whichever occurs later.

(6) EOI shall notify the NRC in writing within 30 days after having accomplished item (b)(1) above and include the status of those activities that have been or remain to be completed in item (b)(2) above.

22. 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants Entergy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Entergy's submittal letter dated December 18, 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).

AMENDMENT NO. XXX

D. The facility requires an exemption from certain requirements of Appendices E and J to 10 CFR Part 50. These exemptions are described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 10 (Section 6.1.2) and Supplement No. 8 (Section 6.2.6), respectively. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. These exemptions are, therefore, hereby granted pursuant to 10 CFR 50.12. With the granting of these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

E. EOI shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and10 CFR 50.54(p). The plan, which contains Safeguards Information protected under 10 CFR 73.21, is entitled:

"Physical Security, Safeguards Contingency and Training & Qualification Plan," and was submitted on October 4, 2004.

EOI shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The EOI CSP was approved by License Amendment No. 234 and supplemented by a change approved by Amendment Nos. 239, 241, and 247.

F. Except as otherwise provided in the Technical Specifications or the Environmental Protection Plan, EOI shall report any violations of the requirements contained in Section 2.C of this renewed license in the following manner. Initial notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC Operations Center via the Emergency Notification System with written follow-up within 30 days in accordance with the procedures described in 10 CFR 50.73(b), (c) and (e).

G. Entergy Louisiana, LLC shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

H. This renewed license is effective as the date of issuance and shall expire at midnight on December 18, 2044.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Ho K. Nieh, Director Office of Nuclear Reactor Regulation

Enclosures:

1. (DELETED)

2. Attachment 2

3. Appendix A (Technical Specifications) (NUREG-1117)

4. Appendix B (Environmental Protection Plan)

5. Appendix C (Antitrust Conditions)

Date of Issuance: December 27, 2018 AMENDMENT NO. XXX

Attachment 1 W3F1-2022-0009 Revised List of Categorization Prerequisites*

• Supersedes list previously provided as Attachment 1 to Enclosure of original application to adopt 10 CFR 50.69 (see Reference 1 of cover for this supplement letter).

W3F1-2022-0009 Page 1 of 1 REVISED LIST OF CATEGORIZATION PREREQUISITES Entergy will develop fleet level procedures to outline the process for categorization of plant systems. The Entergy fleet procedures will contain the elements/steps listed below for categorizing systems at Waterford 3.

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 of the Waterford 3 10 CFR 50.69 Tier 2 License Amendment Request Supplement Report [this Attachment]). Any component supporting an HSS function is categorized as preliminary HSS. Components supporting an LSS function are categorized as preliminary LSS.

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

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

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

Risk sensitivity study. For PRA-modeled components, an overall risk sensitivity study is used to confirm that the population of preliminary LSS components results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF) and meets the acceptance guidelines of Regulatory Guide 1.174 (Reference 20).

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 Waterford 3 10 CFR 50.69 Tier 2 License Amendment Request Supplement Report (this Attachment).

Attachment 2 W3F1-2022-0009 Revised Section 6.0 Reference List*

• Supersedes list previously provided in Section 6.0 of Enclosure of original application to adopt 10 CFR 50.69 (see Reference 1 of cover for this supplement letter).

W3F1-2022-0009 Page 1 of 5 REVISED SECTION 6.0 REFERENCE LIST

6.0 REFERENCES

1. Nuclear Energy Institute (NEI) Report NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, (ADAMS Accession No. ML052910035), dated July 2005

2. U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1, (ADAMS Accession No. ML061090627), dated May 2006

3. Electric Power Research Institute (EPRI) Technical Update 3002017583, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," dated February 11, 2020

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

5. American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS),

"Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," RA-Sa-2009, Addendum A to RA-S-2008, dated February 2009

6. Nuclear Management and Resources Council, Inc. (NUMARC) Report 91-06, "Guidelines for Industry Actions to Assess Shutdown Management," (ADAMS Accession No. ML14365A203), dated December 1991

7. NRC letter to Entergy Operations, Inc. (Entergy), "Arkansas Nuclear One, Unit 2 -

Approval of Request for Alternative ANO2-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)," (ADAMS Accession No. ML090930246), dated April 22, 2009

8. EPRI Technical Report NP-6041-SL, "A Methodology for Assessment of Nuclear Plant Seismic Margin (Revision 1)," dated August 1, 1991

9. ASME, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," RA-Sb-2005, Addendum B to ASME RA-S-2002, dated December 30, 2005

10. NRC letter to All Power Reactor Licensees and Holders of Construction Permits in Active or Deferred Status, "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident,"

(ADAMS Accession No. ML12053A340), dated March 12, 2012

W3F1-2022-0009 Page 2 of 5

11. Entergy letter to NRC, "Entergy Seismic Hazard and Screening Report (CEUS Sites),

Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," W3F1-2014-0023, (ADAMS Accession No. ML14086A427), dated March 27, 2014

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

13. NRC letter to Entergy, "Waterford Steam Electric Station, Unit 3 - 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 No. MF3712)," (ADAMS Accession No. ML15335A050), dated December 15, 2015

14. Entergy letter to NRC, "Seismic Walkdown Report Revision 2 - Planned Update to Entergys Response to NRC Request for Information Pursuant to 10 CFR 50.54(f)

Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," (transmits Engineering Report WF3-CS-12-00003, Rev. 2), (ADAMS Accession No. ML14189A696), dated July 8, 2014

15. NRC to All Licensees Holding Operating Licenses and Construction Permits for Nuclear Power Reactor Facilities, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f)," Generic Letter No. 88-20, Supplement 4, (ADAMS Accession No. ML031150485), dated June 28, 1991

16. EPRI Technical Report TR-1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," dated December 2008

17. NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 19, (ADAMS Accession No. ML19128A244), dated October 2019

18. NRC Record of Review, "Dispositions to Waterford 3 Internal Events PRA Facts and Observations (F&Os)," (ADAMS Accession No. ML15363A374), dated October 6, 2015

19. NEI letter to NRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," (ADAMS Accession No. ML17086A431), dated February 21, 2017

20. 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, (ADAMS Accession No. ML17317A256), dated January 2018

21. NRC Report NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," Revision 1, (ADAMS Accession No. ML17062A466), dated March 2017

W3F1-2022-0009 Page 3 of 5

22. EPRI Technical Report 1016737, "Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments," dated December 19, 2008

23. EPRI Technical Update 1026511, "Practical Guidance on the use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty," dated December 4, 2012

24. NRC Report NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines," (ADAMS Accession No. ML12216A104), dated July 2012

25. NRC Report NUREG/CR-6850, EPRI 1011989, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," (ADAMS Accession Nos. ML052580075, ML052580118),

dated September 2005

26. Entergy Report, PSA-WF3-07-01, "Waterford 3 Re-Examination of External Events Evaluation in the IPEEE," November 2017

27. Entergy Report, PSA-WF3-04-01, "Waterford 3 Seismic Risk Bounding Evaluation,"

dated June 6, 2018

28. Entergy Report, PSA-WF3-08-06, Revision 0 - "PRA Technical Adequacy to Support Risk-Informed Applications," dated November 11, 2020

29. Entergy Report, PSA-WF3-08-02, "Waterford 3 Finding Level F&O Independent Technical Review," dated December 11, 2018

30. Entergy Report, PSA-WF3-08-01, "Waterford 3 PRA Peer Review Gap Assessment to 2009 PRA Standard," dated July 30, 2015

31. Entergy Report, PSA-WF3-01-QU, Revision 3, "WF3 PSA At-Power Level 1 Integration and Quantification Analysis," dated March 27, 2019

32. Entergy Report, PSA-WF3-01-QU-01, Revision 2, "WF3 PSA Uncertainty and Sensitivity Analysis," dated February 27, 2019

33. Entergy Report, PSA-WF3-01-IF-SOU, "Waterford 3 Internal Flooding Sources of Uncertainty," dated May 6, 2020

34. Entergy Report, PSA-WF3-03-UNC-01, "WF3 Fire PRA Sensitivity and Uncertainty Report," dated December 23, 2019

35. Entergy Report, PSA-WF3-03-UNC-02, "Fire PRA Parametric Uncertainty Analysis (UnCert)," dated December 23, 2019

36. Entergy Report, PSA-WF3-03-FQ-01, "Waterford 3 Fire PRA Quantification Report,"

dated December 19, 2019

37. Entergy Report, PSA-WF3-03-ES-01, "Fire PRA Equipment and Cable Selection Notebook," dated December 19, 2019

W3F1-2022-0009 Page 4 of 5

38. Entergy Report, PSA-WF3-03-ES-02, "Fire PRA Circuit Analysis and Failure Probability Development," dated December 19, 2019

39. Entergy Report, PSA-WF3-03-PRM, "Fire PRA Quantification Model Preparation and Database Development," dated December 19, 2019

40. Entergy Report, PSA-WF3-03-FSS-02, "Waterford 3 PRA Fixed Ignition Source Zone of Influence Methods," dated December 19, 2019

41. Entergy Report, PSA-WF3-03-FSS-03, "Waterford 3 Transient Fire Scenario Report,"

dated December 19, 2019

42. Entergy Report, PSA-WF3-03-FSS-06, "Development of Fire Non-suppression Factors for WF3 Fire PRA Scenarios," dated December 23, 2019

43. Entergy Report, PRA-W3-01-IF-QU, "Waterford 3 Internal Flooding Quantification Report," dated May 6, 2020

44. Westinghouse Report, LTR-RAM-II-09-39, "RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements For The Waterford Steam Electric Station, Unit 3 Probabilistic Risk Assessment," dated August 17, 2009

45. Westinghouse Report, LTR-RAM-II-11-003 "Fire PRA Peer Review of Waterford Steam Electric Station Unit 3 Fire Probabilistic Risk Assessment Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard," dated February 23, 2011

46. NRC letter to Entergy, "Waterford Steam Electric Station, Unit 3 - Issuance of Amendment Re: Adoption of TSTF-425, Revision 3 'Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b' (CAC NO. MF6366)," (ADAMS Accession No. ML16159A419), dated July 26, 2016

47. NRC letter to Entergy, "Waterford Steam Electric Station, Unit 3 - Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c) (CAC NO. ME7602)," (ADAMS Accession No. ML16126A033), dated June 27, 2016

48. Entergy Report, PSA-WF3-01-HR, Revision 3, "WF3 At-Power Human Reliability Analysis," dated March 27, 2019

49. Entergy Report, PSA-WF3-01-LE, Revision 3, "WF3 PSA Large Early Release Frequency (LERF) model," dated May 12, 2020

50. NRC Regulatory Issue Summary (RIS) 2007-06, "Regulating Guide 1.200 Implementation," dated March 22, 2007

51. NRC Letter to NEI, "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, AND 12-13, Close-Out of Facts and Observations (F&Os)," (ADAMS No. ML17079A427), dated May 3, 2017

W3F1-2022-0009 Page 5 of 5

52. Entergy Report, PSA-WF3-08-03, "Waterford 3 PRA Focused Scope Peer Review (IF &

LERF)," dated March 23, 2020

53. Entergy Report, PSA-WF3-08-04, "Focused Scope Fire PRA Peer Review of Waterford Steam Electric Station Unit 3 Fire Probabilistic Risk Assessment Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard," dated September 2012

54. NUREG/CR-6850 Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," dated September 2010

55. Entergy Report, PSA-WF3-08-05, "2nd Focused Scope Fire PRA Peer Review of Waterford Steam Electric Station Unit 3 Fire Probabilistic Risk Assessment Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard," dated May 2013

56. Entergy Report, PSA-WF3-08-14, "Waterford 3 PRA Focused Scope Peer Review -

HRA," dated December 2021

57. Entergy Report, PSA-WF3-08-09, "Waterford 3 - 10 CFR 50.69 Program Support Sensitivity Assessments," dated August 16, 2021

58. Entergy letter to NRC, "Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69,"

(ADAMS Accession No. ML21274A876), dated October 1, 2021

Attachment 3 W3F1-2021-0009 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items*

• Supersedes Attachment 3 to Enclosure of original application to adopt 10 CFR 50.69 (see Reference 1 of cover for this supplement letter).

W3F1-2022-0009 Page 1 of 23 DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS The Waterford 3 Internal Events Probabilistic Risk Assessment (PRA) model (including internal flood) was peer reviewed in 2009 by the PWR Owners Group (PWROG). The review was conducted in May 2009 and the report was issued in August 2009 (Reference 44). The Internal Events PRA technical adequacy (including the 2009 peer review and follow up self-assessment results) has previously been reviewed by the U.S. Nuclear Regulatory Commission (NRC) in previous license amendment requests (LARs) associated with Technical Specification Task Force (TSTF)-425 (surveillance frequency control program) and National Fire Protection Association (NFPA) 805 license amendments (References 46 and 47).

The 2009 Peer Review was completed using the previous revisions of industry guidance.

Revision 1 of Regulatory Guide 1.200 was applied, and the review was conducted using the 2005 issued PRA standard (Reference 9). In development of the NFPA 805 LAR, a gap assessment was performed due to revisions to the guidance and standard. The goal of this assessment was to both identify the differences in the two documents, and to identify the impact the differences potentially have on existing Waterford 3 peer review findings. The results of this gap assessment are documented in PSA-WF3-08-01 (Reference 30). Subsequent assessments were made relative to the 2009 PRA Standard (American Society of Mechanical Engineers [ASME]/American Nuclear Society [ANS] RA-Sa-2009 - Reference 5) and Revision 2 of Regulatory Guide 1.200 (Reference 12).

The Fire PRA model was subject to a self-assessment and a full-scope peer review was completed in 2011. The review was in November 2010 and the resulting report was issued in February 2011 (Reference 45). Following a 2012 NRC Audit of the Waterford 3 NFPA 805 LAR and supporting documents, Waterford 3 revised the Fire PRA including several method changes. Waterford 3 completed two focused scope peer reviews (September 2012 and May 2013) to ensure proper evaluation of the revised methods.

The results of those peer reviews (for internal events and for the fire model) were the basis for the NFPA 805 LAR and safety evaluation (SE) and the TSTF-425 LAR and SE.

In the time following those reviews, the Waterford 3 PRA models have been through several updates (with some technical upgrades) as well as peer reviews and formal Facts and Observation Close-out reviews. Findings for both the full power Internal Events and Fire PRA models (as of October 2017) were reviewed and closed using the process documented in Appendix X to Nuclear Energy Institute (NEI) 05-04, NEI 07-12, and NEI 12-13, "Close-out of Facts and Observations (F&Os)," as accepted by NRC in the staff memorandum dated May 3, 2017 (Reference 51). The results of this review have been documented (Reference 29) and are available for NRC audit. The closure review concluded that the resolution of one finding resulted in application of a modeling upgrade. This one PRA upgrade as defined by the ASME PRA Standard RA-Sa-2009 (Reference 5) has occurred to the Internal Events PRA model since conduct of the PWROG peer review in 2009. That Upgrade is associated with cooling tower success criteria. It was reviewed during the F&O closure review in October 2017 (as an imbedded focused peer review).

Following the closure review, both the internal events model and fire PRA model were revised to address findings that were not closed during the closure review (in addition to a routine model update). During this update, the Large Early Release Frequency (LERF) model was revised

W3F1-2022-0009 Page 2 of 23 with model upgrades to resolve issues and more thoroughly meet the PRA Standard. The LERF model was subject to a peer review in August 2019 as a result of the upgraded methods used. No unreviewed PRA upgrades (except those noted for Flood and LERF) were included in the internal events and internal fire PRA revision efforts.

The Closure Review conducted excluded open findings associated with the internal flood PRA model. The model had not been thoroughly updated since the original findings. The Waterford 3 internal flood PRA model was revised in 2019. The flood PRA model update also included methodology upgrades. A peer review was necessary due to the method upgrades and the time gap from the previous flood model update (nearly ten years). The updated flood model was subject to a peer review in August 2019 for all PRA Standard elements relevant to internal flooding PRA models. Following the peer review, the flood PRA was updated in 2020 to address the peer review findings.

In December 2021, a focused scope peer review was performed on the Human Reliability Analysis (HRA) in response to concerns from the NRC regarding whether the transition to HRA Calculator (HRAC) was considered an upgrade. All of the supporting requirements relevant to the calculation tool were re-assessed (all supporting requirements for high level requirements (HLR) HR-B, -C, -D, -F, -G, -H, and -I). Previous findings for these HLRs are no longer carried forward as they are superseded by this review.

Table A3-1 lists the Peer Review efforts conducted in the past several years including dates and descriptions.

Table A3 Waterford 3 PRA Peer Reviews Review Description Review Document RG 1.200 PRA Peer Review Against the ASME PRA Standard LTR-RAM-II-09-39 Requirements for the Waterford Steam Electric Station, Unit 3 Probabilistic (Reference 44)

Risk Assessment (Westinghouse Owners Group August 2009)

Fire PRA Peer Review of Waterford Steam Electric Station Unit 3 Fire LTR-RAM-II-11-003 Probabilistic Risk Assessment Against the Fire PRA Standard Supporting (Reference 45)

Requirements from Section 4 of the ASME/ANS Standard (Westinghouse Owners Group February 2011)

Focused Scope Fire PRA Peer Review of Waterford Steam Electric Station PSA-WF3-08-04 Unit 3 Fire Probabilistic Risk Assessment Against the Fire PRA Standard (Reference 53)

Supporting Requirements from Section 4 of the ASME/ANS Standard (Robert Brady - URS Corporation - September 2012) 2nd Focused Scope Fire PRA Peer Review of Waterford Steam Electric PSA-WF3-08-05 Station Unit 3 Fire Probabilistic Risk Assessment Against the Fire PRA (Reference 55)

Standard Supporting Requirements from Section 4 of the ASME/ANS Standard (Robert Brady - URS Corporation - May 2013)

Waterford 3 Finding Level F&O Independent Technical Review (October PSA-WF3-08-02 2017) (Reference 29)

Waterford 3 Internal Flood/LERF Focused Scope Peer Review (August PSA-WF3-08-03 2019) (Reference 52)

Waterford 3 PRA Focused Scope Peer Review - HRA (December 2021) PSA-WF3-08-14 (Reference 56)

W3F1-2022-0009 Page 3 of 23

• Note - All of the Findings from the first four entries in the table were evaluated in the F&O Closure Review conducted in 2017. The only exception to this is the internal flooding and LERF related findings. All PRA Standard Elements for internal flooding and LERF were evaluated in the August 2019 flood/LERF peer review that was conducted on the updated flood model and LERF model.

Tables A3-2 through A3-4 provide a summary of the remaining findings and open items, including:

Table A3-2 below lists the open Findings associated with the internal events PRA model and provides a disposition associated with the risk-informed applications for each open technical issue. This table includes open issues remaining from the 2017 closure review as well as the 2019 LERF related peer review. Open findings related to the 2021 Focused Scope HRA peer review replaced the open items from previous peer reviews for the related HRA SRs.

Table A3-3 below lists the open Findings associated with the August 2019 internal flood PRA model peer review.

Table A3-4 below lists the open Findings associated with the internal Fire PRA model.

This table addresses the findings that remain open following 2017 F&O Closure Review.

All three tables provide a disposition for each entry with regard to a risk-informed 10 CFR 50.69 program (and LAR).

The content of Tables A3-1, A3-2, A3-3, and A3-4 is contained in Waterford 3 PRA report PSA-WF3-08-06 (Reference 28). This includes listed references.

PRA Credit for FLEX The Waterford 3 PRA model credits FLEX equipment and strategies. The credit for FLEX related equipment is limited to specific extended loss of offsite power scenarios and is limited to primarily permanently installed FLEX equipment. The one portable component that is credited from the FLEX strategy is the FLEX Diesel Fuel Transfer Pump. The model changes to incorporate FLEX equipment and strategies referenced site procedures and is a direct representation of the as-built, as-operated plant. The specific changes to add the FLEX equipment and strategies were not judged to be PRA upgrades, as existing modeling methods and techniques were used to update the model (no new or unique methods were applied).

The Waterford 3 PRA model credits a FLEX diesel generator (DG) to provide power to battery chargers (given an extended loss of offsite power). This diesel unit is installed in the Reactor Auxiliary Building (RAB). Use of this equipment and actions necessary to start and align it are included in site procedures for loss of offsite power, and all necessary equipment (cables, panels, keys, etc.) is pre-staged. Existing model failure data type codes were used for diesel generator and circuit breaker failure data. As the equipment is permanently installed and procedurally controlled, generic failure data was judged applicable.

The FLEX Diesel Fuel Transfer Pump is used to transfer fuel oil from the emergency diesel generator (EDG) fuel oil day tanks to the FLEX DG fuel tank. The deployment of this component and its associated equipment is done according to the FLEX guideline procedures and is in close proximity to its storage location. Additionally, this component is only required later in the event after all of the other credited components have been deployed. Based on commitments made in Waterford 3 letter W3F1-2021-0050, "Response to U.S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment

W3F1-2022-0009 Page 4 of 23 Request to Adopt 10 CFR 50.69," dated October 1, 2021 (Reference 58), PWROG industry data will be used for this portable component.

The FLEX Core Cooling Pump (FCCP) is a permanently installed pump that can be aligned to either provide Feedwater to the Steam Generators (SGs), makeup to the Reactor Coolant System (RCS), or backup cooling to the Spent Fuel Pool. The pump can be powered from a charging pump breaker, supplied by emergency power, or powered from the FLEX Diesel. For the Waterford 3 PRA model, only alternate Feedwater to the Steam Generators is credited. As with the FLEX diesel, all equipment and actions necessary to align and operate the pump for this function are driven by site procedures. Existing model failure data type codes were used for the FCCP. As the equipment is permanently installed and procedurally controlled, generic failure data was judged applicable.

The human actions added to the PRA model for FLEX deployment followed the same Human Error Probability (HEP) development methods as all other modeled actions. Credited actions are all procedure driven actions. Peer reviewed HEP methodology was applied to the added actions. Live timed field trials were used to support timing inputs for HEP development. The credited operator actions are procedure driven actions and are similar to other operator actions evaluated using approaches consistent with the ASME/ANS PRA Standard as endorsed by Regulatory Guide 1.200, Revision 2.

The modeling of the credited FLEX equipment and actions:

Does not represent any new methods; Does not change the scope of the model given that the equipment, dependencies, and type of accident sequences remain the same; Does not represent a change in capability of the PRA model given the original and updated models can both evaluate the risk associated with loss-of-offsite power and station blackout.

The changes implemented for the incorporation of the FLEX modifications were within the framework of the existing peer reviewed PRA model structure.

W3F1-2022-0009 Page 5 of 23 Table A3 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC)

AS-B3-01 AS-B3 Not Met The AS report (PRA-W3-01-001S01 Revision 1) includes This was graded by the 2017 F&O closure review team as partially discussion of the phenomenological impacts of heating of resolved (with open documentation issues).

the containment sump water (failure of HPSI recirculation due to loss of required NPSH and pump cavitation) and WCAP-16679-P - Accident Sequence Phenomena was reviewed to large containment rupture (loss of safety injection due to determine if any additional phenomena needed to be addressed in the the rapid depressurization, flashing of hot water in the current Waterford 3 AS analysis. All other phenomena have been sump, and loss of net positive suction head to the HPSI addressed in the accident sequence and the system analyses, as pumps) that can occur due to inadequate containment necessary. The effects of steam line and feed line breaks are heat removal. However, some events such as steamline evaluated in the initiating event document.

breaks and feedwater line breaks can result in harsh environments (especially steam and high temperature) Waterford 3 completed a review of the phenomenological where mitigating equipment are located. considerations in the AS report immediately following the original peer review (in 2010). That analysis and the results of it were not included Phenomenological impacts must be considered in order to in (or referenced) in the model Revision 5 AS report reviewed during ensure risk results are not underestimated. the closure review. The considerations are included in the model - the updated documentation was noted to be insufficient. This disconnect Consider and document the phenomenological conditions in documentation was the basis for grading this finding as partially from the entire range of initiating events, especially high resolved with open documentation issues.

energy line break.

The Revision 6 update has a more thorough documentation of such phenomena and their treatment. The Revision 6 Accident Sequence and Success Criteria documentation contain the necessary details to satisfy AS-B3 (though a formal closure review has not been completed).

This Finding has no impact on quantified results and no impact on implementation of a 10 CFR 50.69 program.

W3F1-2022-0009 Page 6 of 23 Table A3 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Finding Supporting Capability Number Requirement(s) Category Description Disposition for 10 CFR 50.69 (CC)

HR 1-2 HR-C2 CC II-III There is no documentation to indicate that a review of No additional failure modes were identified in the analysis, however, plant specific or generic data was performed to search for this was not documented in the HRA notebook. This documentation any additional failure mechanisms. See discussions in issue has no impact on the 50.69 application.

HR-I1, I2, and I3.

HR 7-1 HR-G6 Not Met The reasonableness check discussion and results do not No documentation of the reasonableness check was included with the indicate that plant history and experience were considered HRA report, however, this is included in the cutset review process with in the check. the inclusion of operations in this review. This documentation issue will not impact the application.

HR 7-3 HR-I1 Not Met Population of the text fields in the HRAC leads to This Finding does not note any deficits in information, only the entry of confusion because notes are frequently not placed in the information in various fields in the HRA Calculator in an inconsistent associated text fields. way. This documentation issue will not impact the application.

Examples:

-MAAP references appear under the Procedure list for several HFEs. This leads to the source for timing of these events to be located in the Procedure section rather than the Timing section

-Timing bases are scattered over the individual reports.

PSA-WF3-06-01_000-1, FLEX notebook, is out of date with respect to BE naming.

W3F1-2022-0009 Page 7 of 23 Table A3 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Finding Supporting Capability Number Requirement(s) Category Description Disposition for 10 CFR 50.69 (CC)

HR 7-4 HR-I2 Not Met General: The items included in this Finding are specifically related to Some parts of the HRA analysis are not located in the documentation of the analysis that was performed and not to any HRA Notebook and are discussed in the Data Notebook. deficit in the analysis. In addition, the implementation of the use of the There is no note in the HRA Notebook pointing to the HRA Calculator is not adequately documented. This lack of discussions in the Data Notebook. documentation has no impact on the results calculated in the HRA.

This documentation issue will not impact the application.

Pre-initiators:

No documentation to demonstrate that a review of the pre-initiator HFEs to identify actions that could impact multiple trains or systems was performed, and there is no documentation to demonstrate that none were screened out.

There is no documentation to indicate that a review of plant specific or generic data was performed to search for any additional failure mechanisms.

Post-initiators:

The documentation in the HRA notebook is not consistent with the analysis in the HRAC. Specific examples are below.

There is no clear documentation trail from the timing values used in the HRAC back to their sources.

The documentation of the treatment of recovery is not described from the perspective of using the HRAC.

The quantification of post initiator events is not discussed from the perspective of using the HRAC.

The conduct of the dependency analysis does not describe how to perform the analysis in the HRAC. The

'floor' HEP is not discussed in this section (it does appear in Section 5, however).

Details for the use of the HRAC are generally missing from the documentation. The HRAC is mentioned in the documents but the use is not discussed.

W3F1-2022-0009 Page 8 of 23 Table A3 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Finding Supporting Capability Number Requirement(s) Category Description Disposition for 10 CFR 50.69 (CC)

SC-C1-02 SC-C1 Not Met Throughout the document there are a number of This was graded by the 2017 closure review team as partially resolved.

assumptions and statements made that directly impact the success criteria but do not have any references identified The review team concluded that, while the revised report did contain to justify their bases. Querying the PRA group determined some improvement in providing references/bases for assumptions and that most of the statements were based on valid SC treatments, it is still not sufficient to meet CC-II of the Standard.

references, but they were not identified in the success This includes some missing references, duplicate entries, and criteria documentation. The references need to be references to old/superseded documents.

specifically identified and included.

The PRA model Revision 6 update included an updated SC report with a detailed review of references to ensure a valid basis was provided for documented success criteria.

This has no impact on the 10 CFR 50.69 program. This is a documentation issue only and will have no impact on quantified results.

SY-A12b-01 SY-A12 Not Met Need to use the exclusion criteria in SY-A14 to justify This was graded by the 2017 closure review team as partially resolved.

(SY-13) excluding flow diversion pathways. Using the criteria 2 normally closed valves should be easily justified using The 1/3 exclusion criteria is noted as standard treatment for systems, SY-B14 criteria SY-A14(a). The criteria for excluding based on a unless otherwise noted. However, it is not uniformly applied to all (SY-15) 1 to 3 ratio between the primary piping and the potential systems. Systems with different modeling treatment and the basis for diversion piping needs to be backed up by pressure the different treatment are noted in the documentation. Additional flow differentials. This exclusion criteria is valid if the system diversion failures were credited following this finding for HPSI and LPSI pressures between the primary and potential diversion based on meeting 1/3 criteria but having high pressure differential, and piping is the same or similar. If the pressure differential is CCW Makeup for having a finite and limited volume. The CCW high, further analysis is required to justify exclusion. diversion would not fail system function but over time would reduce inventory below functional level before 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Overall, the assumptions used to exclude specific types of failures needs to be reevaluated and justification provided This element has been resolved in the Revision 6 PRA model on how the exclusion criteria is met. documentation but closure through a formal review is not yet complete.

This has no impact on the application of a 10 CFR 50.69 program.

LE2-3(1) LE-E1* Met The LERF model includes events reflecting conditional Continued operation of CCS and CSS under harsh probability of the Component Cooling System (CCS) and conditions/environment has an insignificant impact on overall LERF LE-C9 Not Met the Containment Spray (CSS) due to a harsh severe- results.

accident environment. Section 8.1.9 of the LERF Notebook (PSA-WF3-01-LE, Rev. 2) identifies events Vessel Rupture accounts for over 90% of total LERF. ISLOCA is the P_CCSFAILS and P_CSSFAILS and refers to the Level 2 second largest contributor accounting for over 50% of the non-Vessel

• Finding analysis for the development of the associated conditional Rupture LERF total. CCS and CSS have no impact on these two Reference this probabilities. The Level 2 report (PSA-WF3-01-L2-01, LERF contributors. The values applied to events P_CCSFAILS and SR but it is Rev. 0), in turn, refers to the Waterford 3 Individual Plant P_CSSFAILS do not significantly impact LERF results. The revised noted in the Examination (IPE, Waterford 3 Probabilistic Risk LERF report shows a sensitivity case with the applied CSS and CCS Report as Met Assessment Individual Plant Examination Submittal, values increased by a factor of 10. This change results in a less than August 1992) as the source. Table 4.6-2 of the IPE lists 1% increase in LERF.

W3F1-2022-0009 Page 9 of 23 Table A3 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Finding Supporting Capability Number Requirement(s) Category Description Disposition for 10 CFR 50.69 (CC) values for these events but provides no further analysis or Solving the fault tree for the loss of train A and B containment spray justification. system gate generates a 1.7E-03 result. The applied P_CSSFAIL is 5E-03. The LERF input for loss of CSS (originally from the IPEEE) is The SR requires that justification be provided for nearly a factor of 3 higher than the detailed fault tree logic. Similarly, continued operation of systems under adverse (severe- the CCS value is nearly a factor of 10 higher than calculated in the accident) environments. Although the potential for an fault tree.

elevated probability of failure has been incorporated for the CCS and CSS, no justification for the probabilities The systems in question are both designed to operate in a harsh selected is provided. environment, and most of the active components for the system are outside containment (and wont be impacted by a harsh containment Develop a justification for the continued operation of CSS environment).

and CCS under adverse conditions.

The model revision 6 LERF report (PSA-WF3-01-LE, Reference 49) has been updated to provide additional basis for the values used. The applied values are conservative and have no significant impact on LERF results. This finding has no impact on quantified results and will not impact risk informed applications.

LE2-4(1) LE-E2 Not Met The treatment of accident-progression phenomena relies The LERF analysis has been updated to include additional discussion almost entirely on predictions from MAAP calculations. relevant to the limitations of MAAP 4.0.6, as well as to include While this is appropriate for much of the LERF analysis, additional sensitivities regarding in-vessel melt retention issues (and LE-F2* Cat-I/II/III the reliance on MAAP for crediting that core debris several specific sensitivity cases related to the MAAP 4.0.6 limitation remains in the reactor pressure vessel for a associated with this phenomena). The updated documentation preponderance of scenarios is a potentially significant provides more detail on in-vessel retention and addresses the non-conservatism. It is difficult to determine what the uncertainty of the cases and computer codes used in the evaluation.

• Finding impact on LERF would be if these scenarios were References this permitted to lead to relatively early breaching of the The following is an excerpt from the revised LERF document:

SR but it is vessel. "Relative to the ability of MAAP 4.06 to model in-vessel retention, it is noted in the The reasonableness check did not address the important to be aware of a Trouble Report posted for MAAP 4.08 and Report as Met appropriateness of MAAP 4.0.6 for certain accident earlier releases. The error report indicates the following:

phenomena, including in-vessel melt retention.

Reference:

TR 810 Ex-vessel cooling code error (Mod. Package 833)

Investigate the potential for in-vessel retention in more As indicated in the reference above, there is a code error in MAAP4.06 detail, based on available literature and/or a tool that is related to ex-vessel cooling:

suitable for assessing this potential. Adjust the LERF analyses to reflect more realistic treatment of in-vessel If the reactor cavity is flooded, there is substantial debris in melt retention. the lower head, and ex-vessel cooling is enabled, it is possible to melt through the reactor wall heat sink (i.e.

Ensure that the treatment of in-vessel melt retention is mrvn(6,1) through mrvn(6,5) become 0), and the reactor considered during future checks of the reasonableness of vessel has not failed. If this occurs, the tabular output file LERF contributors. will show primary system energy imbalance, and RV failure will not have occurred for the MAAP sequence.

W3F1-2022-0009 Page 10 of 23 Table A3 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Finding Supporting Capability Number Requirement(s) Category Description Disposition for 10 CFR 50.69 (CC)

To address this error report, a sample of the existing MAAP cases indicating that vessel failure was prevented were reviewed to better understand if this error is occurring for those runs. As stated in the TR description above, the two vessel wall heat sinks should be examined to determine if the vessel had indeed failed, but that indication of breach was not acknowledged. In order to make this determination, those cases may need to be rerun to include plot variables associated with the heat sink mass as identified in the TR. The Modification package includes a work-around that can be added to the input file to correct this error. For the selected sample cases, the mass of the heat sinks did not become zero as discussed above and therefore the error was not a factor in the determination of vessel breach."

The LERF documentation has been updated to more thoroughly address the issues identified in the Finding. This finding has no impact on quantified results and will not impact risk informed applications.

LE4-4(1) LE-G5 Not Met Section 9.6 of PSA-WF3-01-LE, Rev. 2 lists the types of The LERF documentation has been updated to address this Finding PRA limitations to consider from the ASME/ANS PRA and detail the impact of LERF assumptions on PRA applications.

Standard but provides no specific discussion of the actual Assumptions related to model development can and do impact results.

limitations of the Waterford 3 LERF model nor how they However, the overall methods and process used limit those impacts.

may impact risk-informed applications. Attachment A Consensus modeling approaches are used in PRA model identifies the assumptions in the Waterford 3 LERF model, development. The LERF model is also peer reviewed against the PRA but does not identify impacts on applications from the Standard. Sensitivity cases for key sources of uncertainty have been treatment of the assumptions. developed to ensure the impact of assumptions and modeling Document the model limitations and potential impacts on decisions are known and documented. This thorough, peer reviewed, PRA applications for Waterford 3 LERF model. state-of-the-art approach to LERF modeling helps ensure the model and results maintain the technical adequacy requirements to support risk-informed applications.

The LERF documentation has been updated to more thoroughly address the issues identified in the Finding. This finding has no impact on quantified results and will not impact risk informed applications.

LE4-5(1) LE-F2* Met Cat An analysis using the MAAP 4.0.6 computer code based The LERF documentation has been updated to more thoroughly I/II/III on realistic inputs in PSA-WF3-01-LE, Rev. 2, calculates address the appropriateness of MAAP 4.0.6 as it applies to TISGTR.

LE-D6* the plant specific conditions for the TISGTR accident Met Cat III sequences. The TISGTR analysis relies solely on The updated reasonableness check is based on:

MAAP 4.0.6 analyses to make the determination of EPRI Perspective on Thermally-Induced Steam Generator Tube

• Finding whether an accident progression sequence is classified as Rupture Issues, NRC document ID: ML071340053, Marc Kenton.

References this TISGTR and does not apply split fractions based on SR but it is industry guidance. The capabilities of the MAAP 4.0.6 It is commonly believed that the TISGTR has a minor impact on the noted in the code should be validated for the appropriate use for this PRA results. Hot leg creep rupture has been found to either occur Report as Met determination. Plant procedures are incorporated into the prior to tube rupture or immediately after. In both instances, the operator actions governing the TISGTR response.

W3F1-2022-0009 Page 11 of 23 Table A3 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Finding Supporting Capability Number Requirement(s) Category Description Disposition for 10 CFR 50.69 (CC) dominant flow path will be from the reactor vessel to containment, with The reasonableness check did not address the a relatively small flow through the failed SG tube.

appropriateness of MAAP 4.0.6 for certain accident phenomena, including TISGTR. Based on the reference above, the MAAP code can accurately predict the peak temperature in the SG tubes as well as the hot leg.

The LERF report was revised to update the reasonableness argument (including addition of the noted reference) to address this Finding.

This Finding has no impact on quantifies LERF results or application of the model for the 10 CFR 50.69 program.

Note - The two SRs listed in the Finding are both noted in the report as met for Capability Category II or III. Even though the finding was issued the grading of the fining is that the analysis is sufficient for PRA applications as assessed (with a recommendation for improvement -

which was completed).

Note 1 - These findings are the result of a peer review of the LERF model conducted in August 2019. The other listed findings are the ones that remained open following an F&O Closure Review in 2017.

W3F1-2022-0009 Page 12 of 23 Table A3 Internal Flooding Open Peer Review Findings Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC)

FL1-3 IFSN-A15 Met CC-I/II/III Table B.3 of PSA-WF3-01-IF-WD lists piping segments The pipe segments excluded from the analysis have been (flood sources) that require further analysis as to whether identified and evaluated. The updated Flood PRA documents they could be screened. A Waterford 3 PRA Model account for the subject segments. Many screened but several Change Request (MCR #W3-6480) was written to track resulted in new or updated scenarios.

this issue.

The updated flood analysis properly treats the segments Complete the analysis of these piping segments to identified in this Finding. This issue has no impact on determine whether they can be screened per the criteria quantified results and will not impact risk-informed given in this SR. applications.

Though a Finding was issued, the peer review graded the SR as Met CC-I/II/III.

FL1-4 IFQU-B3* Met In Table A-2 of PSA-WF3-01-IF-SOU, item IF-C-9 dealing A detailed engineering evaluation of the subject door was with the structural analysis of doors may have a completed by Waterford 3/Entergy structural engineers. The

• Finding significant impact on internal flood CDF. The Flood PRA scenarios were updated based on this more References this configuration of doors within room 211 (vestibule area) thorough evaluation of flood propagation from room 211. The SR but it is noted may be such that the door leading to the outside environs Flood PRA was updated to explicitly address this Finding.

in the Report as may preferentially fail first, which could have a significant Resolution of this issue did result in a significant change in Met impact on the calculation of internal flood CDF (~50%). flood results. With the issue resolved, this Finding has no further impact on quantified results and will not impact risk-informed applications.

This Finding was issued even though IFQU-B3 was graded as "Met Cat 1-3" in the peer review report.

FL1-6 IFEV-A5* Met CC-I/II/III The frequency for event %FLD-TB_ALL used an older The Flood documentation has been updated. Updated data set (NUREG/CR-5750) that was inconsistent with initiating event frequencies have been developed (in part to

• Finding the data used throughout the rest of the plant. The EPRI address Finding IFEV-A6 listed below in this table). Additional References this reference that was primarily used for other event details on the development of the Turbine Building frequency SR but it is noted frequencies outside the Turbine Building can also be (%FLD-TB_ALL) have been added. Reasonableness checks in the Report as used to develop plant-level flood frequencies based on of resulting frequencies were completed to ensure compliance Met generic industry operating experience. with Standard SR IE-C12 (and IFEV-A5).

This SR refers to the requirements in Section 2-2.1 of the This Finding has been addressed in the updated Flood PRA Standard, which does involve SR IE-C12 that requires a documentation. This issue has no impact on quantified results comparison of results of the initiating event analysis with and will not impact risk-informed applications.

generic data sources to provide a reasonableness check of the results. The event frequency used from This Finding was issued even though IFEV-A5 was graded as NUREG/CR-5750 is an older data source that was not "Met Cat 1-3" in the peer review report.

compared with the more recent data set used for the rest of the plant.

Compare the event frequency used for the Turbine Building floods with the plant-level data found in Section 8

W3F1-2022-0009 Page 13 of 23 Table A3 Internal Flooding Open Peer Review Findings Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC) of EPRI 3002012997, which covers a larger time period than the period used in NUREG/CR-5750. This will confirm the "reasonableness" of the selected event frequency for this initiating event and whether it should be deemed conservative as stated in Section 2.2.146.

FL3-2 IFPP-B1* Met There are discrepancies between what was modeled in The Flood PRA documentation has been updated to resolve PSA-WF3-01-IF-AS and the information in Table 2 of discrepancies between the Accident Sequence and Flood Area PSA-WF3-01-IF-FA. For example, RAB-21-211 lists Development documents. The revised documents and tables

• Finding Door D11 to RAB21-Q as not having sufficient contain consistent technical information.

References this accumulation of water to fail, and PRA equipment SR but it is noted damage protected by D11 is listed as 'N/A'. The Flood PRA was updated to explicitly address this Finding.

in the Report as With the issue resolved, this Finding has no impact on Met Tables 1 and 2 of PSA-WF3-01-IF-FA should be updated quantified results and will not impact risk-informed to be consistent with the accident sequence analysis applications.

described in Section 4 of PSA-WF3-01-IF-AS.

This Finding was issued even though IFPP-B1 was graded as "Met Cat 1-3" in the peer review report.

FL3-4 IFEV-A6 Cat I EPRI TR-3002012997 R4 Section 4.5 requires EPRI has rescinded the TR-3002012997 R4 guidance consideration of age correction for significant scenarios. document. The flood PRA initiating events were updated to Aging factors from Table 4-19 were not applied. It could use EPRI TR-3002000079 R3 (which is still endorsed by be argued that age adjustment using EPRI). Following the TR-3002000079 R3 guidance does not EPRI TR-3002000079 R3 would not be applicable to require age correction due to the age of the Waterford 3 site Waterford 3 for another 2 to 3 years. and piping systems.

This SR requires consideration of material condition of The Flood PRA was updated to explicitly address this Finding.

fluid systems. EPRI TR-3002000079 R3 does not require With the issue resolved, this Finding has no impact on age correction for FP piping that has been in service less quantified results and will not impact risk-informed than 40 years. However, EPRI TR-3002012997 R4 applications.

Table 4-19 provides age correction factors ranging from 10 to 50 years of service. Frequencies calculated in PRA-WF3-IF-QU were based on nominal values from EPRI TR-3002012997 R4 without age correction.

FL3-5 IFEV-A7 Not Met Section 2.1.1 of PRA-WF3-01-IF-IE includes an invalid The flood documentation has been updated to correct the input to the equation used to calculate the frequency of a issue identified (including correction of the erroneous value) in maintenance induced flood event. The use of a this Finding. Human induced flood events are considered in maintenance unavailability value (or probability) was the Waterford 3 flood PRA analysis. Human induced flood erroneously used as a frequency in the justification for events have been screened, and a valid basis for the screening maintenance induced flooding events from the screening is included in the updated documentation.

analysis.

The Flood PRA was updated to explicitly address this Finding.

This SR requires consideration of human-induced floods With the issue resolved, this Finding has no impact on from maintenance. While consideration of maintenance

W3F1-2022-0009 Page 14 of 23 Table A3 Internal Flooding Open Peer Review Findings Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC) induced flooding, events was provided in Section 2.1.1 of quantified results and will not impact risk-informed PRA-WF3-01-IF-IE, the basis for screening these events applications.

is invalid.

Correct the qualitative argument to justify screening human induced flood events based on a review of generic industry data or apply the methodology given in Section 8 of EPRI 3002012997 R4.

FL3-7 IFQU-A10 Not Met Overlaying the internal flood logic and target sets onto the The Flood PRA was updated following the peer review. The LERF sequences produced erroneous cutsets relative to update included changes to key scenarios and initiating event the corresponding CDF cutsets. For example, the top frequencies. The update required updated quantification of LERF cutset from flooding scenario CDF and LERF. During the update, an existing modeling error

%FLD-RAB_21-211FP-L includes AB Switchgear (not explicitly related to flood) was found and corrected. CDF alignment events 'B_TO_AB' and 'HPIABISSTBY' while and LERF cutsets were reviewed for reasonableness and to the corresponding CDF cutset is more consistent with ensure mapping/overlaying the CDF results onto LERF was loss of equipment in the AB and A Switchgear Rooms. properly completed.

Justify the validity of the anomalous LERF cutsets Updated Flood LERF results were reviewed to check for generated by internal flood initiators or modify the LERF erroneous results. With the issue resolved, this Finding has no analysis as necessary to account for the unique flood- impact on quantified results and will not impact risk-informed induced scenarios in accordance with the applicable applications.

requirements described in 2-2.8.

Note - All Flood PRA findings are the result of an August 2019 Flood PRA peer review.

W3F1-2022-0009 Page 15 of 23 Table A3 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC)

CS-A3-01 CS-A3 Met The component and cable selection report, Rev. 0, The 2017 F&O Closure review judged this finding as Open.

reviewed looking for effects of interlocks and permissive.

ES-B4 Not Met There is discussion of the VCT in the MSO discussion, The Fire PRA model and documentation was updated in 2019 to but there does not appear to be any other interlock or incorporate the internal event (R6) model update and to address permissive discussion. These items would be tank level open Findings.

interlocks with valves and pumps, loss of one CCW, perhaps a flow switch that starts the pump in the other The mapping of all items on the SSEL was reexamined with train. Various interlocks that are associated with the attention paid to instruments to ensure that their consequential starting of a pump such as adequate lube oil pressure, impacts have been properly linked to the PRA model. The cooling water flow, etc. methodology and results of the analysis are documented in the Waterford 3 Component and Cable Selection Report The instruments and cables associated with permissives (PRA-W3-03-ES-01, Reference 37).

and interlocks do not appear to have been comprehensively addressed in the PRA. Starting Specific rationale of impacts on permissives and interlocks is interlocks for pumps and breaker closure or tank documented for several components and cables in the MSO interlocks that open or close valves or flow switches that Expert Panel discussions (PRA-W3-03-ES-01, Reference 37).

start pumps all could have fire effects that would adversely affect the success of various system functions. The updated documentation references the site Nuclear Safety Capability Assessment (NSCA). The NSCA contains a thorough, comprehensive review and treatment of interlocks and permissives. This finding has been addressed in the 2019 model update (though it has not been through formal closure review).

This finding has no impact on PRA results or risk informed applications.

CS-B1-01 CS-B1 CC-I Electrical coordination is addressed in the scenario The original finding was judged by the F&O closure teams as development report (R0247070001.06 Appendix E). Open.

Appendix E of R0247070001.06 provides information concerning electrical coordination. However, it is Breaker coordination is addressed in the updated Fire PRA incomplete because the supplemental coordination model documentation. An updated analysis that supplements the evaluation is missing from the document. Preliminary Component and Cable Selection report and the Plant Response coordination review has been performed and exists in an Model report has been developed. The analysis email (though not formally documented). (PSA-WF3-03-ES-02, Reference 38) provides a more thorough analysis of electrical coordination.

Appendix E of R0247070001.06 provides information concerning electrical coordination. However, it is The updated model and documentation provide a thorough incomplete because the supplemental coordination resolution to the breaker coordination issue. This finding has evaluation is missing. Information was provided by email been addressed in the 2019 model update. This finding has no and needs to be formally incorporated, (finding CS-B1- impact on PRA results or risk informed applications.

01) CC II/III when finished.

ES-A2-01 ES-A2 Not Met R0247070001.02, Rev. 0, reviewed. Three issues The original finding was judged by the 2017 F&O closure teams identified: 1. The loss of equipment due to a loss of room as Open.

cooling caused by a fire damper going closed could not be demonstrated. 2. Although the R0247070001.06

W3F1-2022-0009 Page 16 of 23 Table A3 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC) credited breaker coordination from the SSD analysis, the The first item identified in the Finding is no longer an issue.

PRA model does not appear to address the effect of a Room cooling has been removed from the model based on room fire interrupting the relay circuit that would inhibit the heatup evaluations that determined cooling was not needed to coordination and allow a fault to transfer upstream. support operation through the mission time. Room cooling is no

3. The loss of DC from a fire does not appear to be longer in the PRA model except for the Main Control Room.

addressed fully. For example, a fire induced loss of DC to the RCP breakers would inhibit the operator action to DC breaker impacts to the RCPs have been explicitly addressed trip the RCPs from the control room. Note: Items 1 and and a (ex-control room) Recovery Action was added to the model 3 have outstanding questions to Waterford 3. to address cases where fire damage could impact the ability of the operators to trip the pumps from the control room.

Breaker coordination (including DC breakers) is addressed in an analysis that supplements the Component and Cable Selection report and the Plant Response Model report. The analysis (PSA-W3-03-ES-02, Reference 38) provides a more thorough analysis of electrical coordination and describes modeling that was added to the PRA to correctly account for DC power requirements for overcurrent protection.

The updated model and documentation provide a thorough resolution to the breaker coordination issue. This finding has been addressed in the 2019 model update. This finding has no impact on PRA results or risk informed applications.

ES-A3-02 ES-A3 CC-I/II/III The loss of DC does not appear to be adequately The original finding was judged by the 2017 F&O closure teams addressed in the fire PRA. For example, a failure of DC as partially resolved.

CS-A3 Met to supply control power to the RCP breakers would inhibit the operator action to trip the Reactor Coolant Pumps DC breaker impacts to the RCPs have been explicitly addressed (RCPs) in a loss of seal cooling scenario. This was and a (ex-control room) Recovery Action was added to the model compensated for by a spurious start of the RCPs which to address cases where fire damage could impact the ability of would affect the same state in the model. Similarly, a the operators to trip the pumps from the control room.

loss of DC power could potentially transfer a fault due to inhibition of coordination. Breaker coordination (including DC breakers) is addressed in an analysis that supplements the Component and Cable Selection The plant has redundant DC supplies to the two breakers report and the Plant Response Model report. The analysis which makes this failure less probable. However, (PRA-W3-03-ES-02, Reference 38) provides a more thorough additional documentation is required to clarify the issue. analysis of electrical coordination and describes modeling that The fire effects on DC could adversely affect coordination was added to the PRA to correctly account for DC power as well as remote operation of breakers. requirements for overcurrent protection.

The updated model and documentation provide a thorough resolution to the breaker coordination issue. This finding has been addressed in the 2019 model update. This finding has no impact on PRA results or risk informed applications.

W3F1-2022-0009 Page 17 of 23 Table A3 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC)

ES-C2-01 ES-C2 Not Met Component and Cable Selection Report The original finding was judged by the 2017 F&O closure teams R0247070001.02, Revision 0 in Section 2.6 states, "An as partially resolved.

instrumentation review was conducted using the simulator and operators to identify single instrument The Fire PRA specific operator interviews (Engineering Change reliance and single indication/instruments whose EC 46718) have the following statement: With respect to the malfunction would cause operators to take action that potential for undesired operation actions or errors of commission would result in un-recoverable states. No single (EOCs) in response to fire-induced instrument failures, the instrument vulnerabilities were identified." interviewed Senior Reactor Operators (SROs) indicated that it was extremely unlikely that any single instrument failure would However, there is no documentation or discussion of this cause an EOC if the alarm response procedure were activity. Engineering standard EN-FP-S-008-Multi has a implemented with verbatim compliance. The communications process for reviewing indication needs for post fire in the and conduct of operations protocols at Waterford 3 require simulator in 5.3.4, but this does not specifically address confirmation with redundant and/or diverse indications prior to spurious indications that cause unwanted actions. changing the state of any equipment. There were no indicators or alarms identified during the simulator control panel walk downs What is needed to meet Category II for this SR is to for which the operator would be expected to take an develop a process for how various indications are unrecoverable immediate action that would otherwise be reviewed and screened and then considered for inclusion undesired.

into the FPRA model. There is a sample process that the PWROG did for ERGs for Westinghouse sites, this The assessment of fire impacts on instruments combined with process is more detailed than required for meeting this the interviews of operators, and the simulator walkdowns SR for this application but does show the process. No represents a process for identifying vulnerabilities (including evidence other than statement that a simulator walkdown single instrument) that is sufficient to meet the Standard was performed. requirements for ES-C2. This finding has no impact on quantified results and will have no impact on application of a 10 CFR 50.69 program.

FQ-C1-01 FQ-C1 Not Met Dependencies on combinations of HFEs have been The original finding was judged by the 2017 F&O closure teams utilized from the internal events PRA, but not addressed as Open.

for specific combination on a scenario by scenario basis.

This needs to be done to ensure all combinations have The NFPA 805 LAR, RAIs, and SE had additional details that been addressed. were not in the PRA documentation. The Fire PRA HEPs and combinations have been properly developed following Provide additional justification for selection of the NUREG-1921. Any action with any instrument/control impact methodology use for combo events to address fire from the fire are failed (set to 1). Multipliers are applied to all impacts. other events/combos to account for increase failure rate during a fire. Any events were these are not applied were explicitly evacuated to document the basis for such treatment.

The update revision 6 Fire PRA documentation (Reference 39) has been revised to include more detail to ensure FQ-C1 is met with more thorough and detailed documentation.

W3F1-2022-0009 Page 18 of 23 Table A3 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC)

This finding has no impact on quantified results and will have no impact on application of a 10 CFR 50.69 program.

FSS-C1-01 FSS-C1 CC-I Two points or a range of heat release values were not The 2017 F&O Closure review judged this finding as Open.

assigned to the ignition sources. A lower overall CDF will likely be achieved by using a two-point analysis or Waterford 3 applied a single Heat Release Rate (HRR) Modeling additional fire modeling to represent HRR profiles from Treatment. This treatment offers a means for incorporation of fire ignition thru burnout and the corresponding probabilities modeling into the fire PRA in a manner that eliminates the need of damage. for separate scenario specific analyses which require significant effort for configuration control, review and update.

In reviewing Waterford 3 F&O resolution for the Fire PRA applications, the NRC graded this finding as follows: "The NRC staff finds that the resolution of the F&O, as described by the licensee in the LAR, would have a negligible effect on the evaluations relied upon to support fire risk evaluations and has no impact on the conclusions of the risk assessment and therefore the resolution of the F&O is acceptable or this application." (per NRC memo (Reference 18).

This modeling limitation would have no impact specific to the 10 CFR 50.69 application or program.

FSS-D7-01 FSS-D7 CC-I Fire detection and suppression system generic This was graded by the 2017 closure review team as partially unavailability values were used, and outlier behavior and resolved (with open documentation).

system unavailability were not specifically analyzed. To obtain a higher category, a specific WSES3 maintenance Plant specific suppression unavailability is applied to suppression history review to assess outlier behavior is to be data used in the Fire PRA. PSA-WF3-03-FSS-06 (Reference 42) documented. This capability assessment is the same as uses plant specific data and generic data to develop failure for the Model of Record (MOR). values used for each detection/suppression system credited.

Any plant specific outlier behavior is explicitly included in the To move from CC-I to CC-II, specific WSES maintenance values.

history review to assess outlier behavior is to be documented. The F&O closure team judged this finding a documentation issue only. The Fire PRA documentation has been updated to address this finding. This finding has no impact on quantified results and will have no impact on application of a 10 CFR 50.69 program.

FSS-E3-01 FSS-E3 CC-I Evaluate the validity of other parametric uncertainty The 2017 F&O Closure review judged this finding as Open.

probability distributions (other than Ignition Frequency, which is the only quantitative uncertainty addressed). Quantitative uncertainty intervals were generated and documented. The closure team determined that inputs to the Only qualitative discussions were provided with respect uncertainty calculation were overly simplistic (error factors of 5 to the uncertainty intervals for most fire modeling were applied to all ignition frequency inputs - actual error factor parameters. Provide quantitative uncertainty intervals.

W3F1-2022-0009 Page 19 of 23 Table A3 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC) data was not applied). The closure team evaluated the uncertainty completed as insufficient to close the finding.

The 2019 Fire PRA update included detailed quantitative uncertainty evaluation for Fire PRA parameters. This includes quantitative uncertainty intervals for fire frequencies, fire suppression factors, modeled circuit failure inputs, and credited recovery events. The updated uncertainty document (PSA-WF3-03-UNC-02, Reference 35) includes the necessary parametric uncertainty content to resolve this finding.

The documentation has been updated to address this finding.

This finding has no impact on quantified results and will have no impact on application of a 10 CFR 50.69 program.

FSS-F2-01 FSS-F2 CC-I Criteria for structural collapse or non-collapse was not This was graded by the 2017 closure review team as partially provided. Only judgment statements were provided, and resolved.

these statements do not appear to reflect reality.

The closure review team assessed that the non-suppression of Re-perform the analysis to address the situation where a the structural collapse scenario is not sufficiently evaluated.

turbine building collapse occurs due to a large turbine lube oil fire. Revise the documents to eliminate the In the 2019 Fire PRA update:

implication that failure of structural steel is not a credible event. The Turbine Generator Building (TGB) oil fire leading to collapse

- manual suppression is not credited, since it is a fast-growing oil fire, but auto suppression is credited, since there is a dedicated deluge system for the TG Oil skid, FPM-5. The failure probability is based on actual plant data. This scenario fails all of the FPRA targets in the Turbine Building (except those in the Turbine Building Switchgear Room).

The model and documentation have been updated to address this finding. This finding has no impact on quantified results and will have no impact on application of a 10 CFR 50.69 program.

FSS-H2-01 FSS-H2 CC-I Section 4.0 (scenarios report) details damage criteria This was graded by the closure review team as partially resolved used in the Fire Scenarios. No cases of where plant (with open documentation).

specific thresholds or damage mechanisms were used.

At the closure review, this finding remained open only due to Plant specific target damage evaluations that involve documentation. The assessment of location and quantities for combinations of thermoset and thermoplastic cable have thermoset cables was thoroughly addressed in the NFPA 805 not been made. Only generic damage mechanisms have LAR, RAIs, and SER. It was not explicitly documented in the been used. documents supporting the Fire PRA model. This limitation was noted by the review team and resulted in the finding only being partially closed.

W3F1-2022-0009 Page 20 of 23 Table A3 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC)

The 2019 Fire PRA documentation contains the necessary documentation to fully resolve this finding. PSA-WF3-03-FSS-02 (Reference 40) and PSA-WF3-03-FSS-03 (Reference 41) calculate damage thresholds for both thermoplastic and thermoset cables for various fire scenarios. Generic treatments are used for most scenarios. More detailed fire modeling is applied to the Relay Rooms. Different damage thresholds were also developed and applied for sensitive (solid state) electronics.

This has no impact on the 10 CFR 50.69 program. This is a documentation issue only and will have no impact on quantified results.

FSS-H3-01 FSS-H3 Not Met The basis for using the FDT tools over other fire This was graded by the 2017 closure review team as partially modeling tools has not been provided. It is clear that the resolved (with open documentation).

FDT tools have been V&V'd, (verification and validation) but their application to the specific scenarios involved in At the closure review, this finding remained open only due to the analysis must also be documented. documentation. Much of the details of the verification and validation for FDT is addressed in the NFPA 805 LAR, RAIs, and The basis for using the FDT tools over other fire SER. At the time of the F&O Closure review some of these modeling tools has not been provided. details were not explicitly in the documents supporting the Fire PRA model. This limitation was noted by the review team and resulted in the finding only being partially closed.

The necessary details have been added to the Fire PRA documentation as part of the 2019 Fire PRA model update.

PRA-W3-03-FSS-02 (Reference 40) , documents the verification and validation (V&V) of the fire modeling tools used in the Waterford 3 fire PRA including fixed ignition sources, transient ignition sources, multicompartment scenarios, and the main control room analysis through a comparison to published parameter values related to the various fire modeling tools.

This has no impact on the 10 CFR 50.69 program. This is a documentation issue only and will have no impact on quantified results.

HRA-A4-01 HRA-A4 Not Met WSES3 Fire Probabilistic Risk Assessment, This was graded by the closure review team as partially resolved Quantification Model Preparation and Database (with open documentation).

Development, R0247070001.03, Rev. 0, Appendix A, Single HFE Screening Process Results, has a "Cue At the time of the 2017 F&O Closure review, the report with the source/ Instrumentation" field that identifies for a number documented interviews was complete and available, the interview of records, the applicable procedure(s) that would be attachment was not included in the relevant Fire PRA report.

utilized to address the respective HFE. Additionally,

W3F1-2022-0009 Page 21 of 23 Table A3 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC)

Appendix D, "Detailed HRA for Selected HFEs," does Fire PRA specific interview (for operator actions during fire identify for certain HFEs, that there were limited STA events) have been conducted and documented. The interviews reviews of the certain aspects of the HFEs, namely the include discussion or procedures, cues, instrumentation, as well cues. However, documented interviews with Operations as the potential for fire impacts to the actions. These interviews to support the use of these cues/instruments were not are documented in Engineering Change EC 46718 found. Report PRAW3-01-001S03, Rev. 1, Operator (Documentation of Operator Interviews from Appendix E of PSA-Interview Sheets document the PRA Internal Events HRA W3-03-PRM).

events that were reviewed with Operations and includes the name of the Operator interviewed. No equivalent This has no impact on the 10 CFR 50.69 program. This was Operator Interview sheets were located to address the judged a documentation issue. The documentation has been impact of fire. resolved to address the issue. This Finding has no impact on quantified PRA results or application of a 10 CFR 50.69 program.

HRA-E1-01 HRA-E1 Not Met A part of the documentation uses future tense and This was graded by the 2017 closure review team as partially requires correction. The missing dependency analysis resolved (with open documentation).

needs to be added as well as the documentation for operator and training interfaces. The assessment of internal events PRA HRA actions used in the FPRA is discussed in the Plant Response Model report Documentation issues with R0247070001.03, Revision 0, (PSA-WF3-03-PRM,Reference 39). The impact of the fire on the Section 5: Page 5-8 has a paragraph explaining that ex- action (instruments, increased stress) is also evaluated. The MCR HFEs are set to true and then the risk significant current/updated document provides a clearer description of the HFEs are analyzed in more detail. If the HFE is set to applied methodology for HFEs.

true, then it would not show up in the cutsets. This paragraph needs a rewrite. Page 59 has a table The finding was judged partially closed due to issues concerning explaining various treatments of HFEs in the model. Two documentation of the HEP dependency treatment for the fire of the columns conflict, one recommends a course of PRA. The review team concluded through the referenced report, action and the resolution takes another course with no RAI responses, and review of the model recovery rule files that explanation of the differences. Also, there should be the treatment is acceptable. The open issue is that the details some discussion about thermal hydraulic analysis on any should all be in the report and review of RAI and recovery rule new sequences. files should not be necessary to make such judgments.

The Fire PRA update in 2019 provides a more thorough documentation of HFE treatment. The assessment of internal events PRA HRA actions used in the FPRA is discussed in the Plant Response Model (PSA-WF3-03-PRM, Reference 39). The impact of the fire on the action (instruments, increased stress) is also evaluated. The current document provides a clearer description of the applied methodology for HFEs.

The 2019 Fire PRA update provides more thorough documentation of the issue involved in this Finding. This finding has no impact on quantified PRA results or application of a 10 CFR 50.69 program.

W3F1-2022-0009 Page 22 of 23 Table A3 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC)

IGN-A10-01 IGN-A10 CC-I The Fire Frequency results presented in Table 4-1 are This was graded by the 2017 closure review team as Open.

mean values only. Previous uncertainties associated with the Bin frequencies and the results of the Bayesian Quantitative uncertainty intervals were generated and update are described. Uncertainties associated with the documented. The closure team determined that inputs to the application of the weighting factors which multiply the Bin uncertainty calculations were overly simplistic (error factors of 5 frequencies are not discussed anywhere. There is no were applied to all ignition frequency inputs - actual error factor discussion of uncertainties in this area in the final data was not applied). The closure team evaluated the quantification report. uncertainty completed as insufficient to close the finding.

Documentation of sources of uncertainty covers The 2019 Fire PRA update included detailed quantitative uncertainties associated with NUREG/CR-6850 Bin uncertainty evaluation for Fire PRA parameters. This includes elements, and uncertainties associated with the Bayesian quantitative uncertainty intervals for fire frequencies, fire update. No discussion on uncertainties associated with suppression factors, modeled circuit failure inputs, and credited partitioning and weighting factor applications. The recovery events. The updated parametric uncertainty document uncertainty analysis is incomplete. Provide either a (PSA-WF3-03-UNC-02, Reference 35) includes the necessary numerical uncertainty analysis or qualitative discussion of parametric uncertainty) includes the necessary parametric other sources of uncertainty as required by the standard. uncertainty content to resolve this finding.

A second uncertainty document, PSA-WF3-03-UNC-01 (Reference 34) documents sources of uncertainty and examines the impact the sources have on the model and results. The PSA-WF3-03-UNC-01 characterizes sources of uncertainty from all relevant Fire PRA tasks.

The documentation has been updated to address this finding.

This finding has no impact on quantified results and will have no impact on application of a 10 CFR 50.69 program.

UNC-A1-01 UNC-A1 Not Met QU-E1 and QU-E2 requires identification of sources of This was graded by the 2017 closure review team as Open.

model uncertainty and assumptions. In general, WSES had an assumption section in each report. However, a A detailed uncertainty evaluation for each task in the simple search on "assume" showed that there were many methodology is provided in the uncertainty report (PSA-WF3 more assumptions than were listed in the assumption UNC-01, Reference 34). The document provides qualitative and sections. At the CC-I level, QU-E3 requires estimation of quantitative uncertainties with explanations for the associated the uncertainty interval of the overall CDF results. WSES analyses.

does not provide an estimation of the uncertainty interval for CDF. QU-E4 requires that for each source of model The 2019 Fire PRA update includes two documents dedicated to uncertainty and related assumption identified in QU-E1 uncertainty. One document covers quantitative uncertainty and and QUE2, respectively; IDENTIFY how the PRA model includes a much more thorough assessment of parametric is affected. WSES only identifies how the model is uncertainty. Quantitative uncertainty intervals were generated impacted for some of the assumptions and sources of and documented.

uncertainty. The Uncertainty and Sensitivity Matrix in Appendix D of R0247070001.07. Per LE-F2, WSES The 2019 Fire PRA update includes quantitative uncertainty should review LERF contributors for reasonableness intervals for fire frequencies, fire suppression factors, modeled

W3F1-2022-0009 Page 23 of 23 Table A3 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Finding Supporting Capability Description Disposition for 10 CFR 50.69 Number Requirement(s) Category (CC)

(e.g., to assure excessive conservatisms have not circuit failure inputs, and credited recovery events. The updated skewed the results, level of plant specificity is appropriate parametric uncertainty document (PSA-WF3-03-UNC-02, for significant contributors, etc.). There is no evidence Reference 35) includes the necessary parametric uncertainty that such a review was performed. content to resolve this finding.

Additional explanation should be provided for the documented entries. The meaning of the confidence A second uncertainty document, PSA-WF3-03-UNC-01 intervals for the different values (mean, 5th/95th, and (Reference 34) documents sources of uncertainty and examines median) would not be obvious to most readers. It is the impact the sources have on the model and results. This suggested that this discussion be corrected in the next report (as the finding suggests) evaluates uncertainty topics from update of the documentation. The usefulness of the EPRI 1026511. The PSA-WF3-03-UNC-01 characterizes qualitative evaluation of modeling uncertainties could be sources of uncertainty from all relevant Fire PRA tasks and significantly enhanced by additional comments regarding examines the impact of assumptions made in those tasks.

their potential impacts. EPRI 1026511 provides a tabulation of sources of modeling uncertainty associated The documentation has been updated to address this finding.

with fire. There is no indication that this report was This finding has no impact on quantified results and will have no considered in identifying the sources discussed in the impact on application of a 10 CFR 50.69 program.

documentation. Because this has become a standard reference, and is a companion document to NUREG-1855, it would be worthwhile to check EPRI 1026511 to identify any additional sources that should be discussed.

Note - All findings in Table 2-4 are F&Os that remain open/partially open following a 2017 F&O Closure review. The original findings evaluated during the closure review were from a 2011 full scope fire peer review and two focused scope peer reviews conducted in September 2012 and May 2013.

Attachment 6 W3F1-2021-0009 Disposition of Key Assumptions/Sources of Uncertainty*

• Supersedes Attachment 6 to Enclosure of original application to adopt 10 CFR 50.69 (see Reference 1 of cover for this supplement letter).

W3F1-2022-0009 Page 1 of 19 DISPOSITION OF KEY ASSUMPTIONS/SOURCES OF UNCERTAINTY The Waterford 3 Probabilistic Risk Assessment (PRA) at power internal events model uncertainty document (Reference 32), internal flooding uncertainty document (Reference 33),

and fire PRA uncertainty document (Reference 34) each contain a detailed, thorough evaluation of uncertainty for the model of record. Each report follows Electric Power Research Institute (EPRI)/NUREG guidance and considers a variety of potential sources of uncertainty. Key sources of uncertainty are identified, and relevant sensitivity cases are documented to examine the key sources of uncertainty and their impact on results.

The purpose of this attachment is to disposition the impact of uncertainty in the PRA models for the 10 CFR 50.69 Risk-Informed Application. The baseline internal events PRA model, the internal flooding model, and Fire PRA model each document assumptions and sources of uncertainty. These models and the associated documentation have all been reviewed during the model peer reviews. The completeness uncertainty associated with scope and level of detail are documented in the models but are only considered for their impact on a specific application. No specific issues of PRA completeness have been identified in the results of the internal events PRA, internal flood, and fire PRA peer reviews. The approach for this application is, therefore, to review these documents to identify the key assumptions and sources of uncertainty which may be directly relevant to the 10 CFR 50.69 Program calculations or categorization efforts, and to identify needed sensitivity analyses where appropriate.

The Waterford 3 PRA models are continuously maintained and periodically updated under the guidance of fleet procedures. This includes continuous identification, documentation and tracking of open issues (the site maintains a Model Change Request (MCR) database for identified model issues). Entergy PRA guidance requires periodic model update as well as self-assessments and peer reviews. Model issues ranging from MCRs to open peer review Findings will be reviewed during 10 CFR 50.69 categorization efforts to ensure the open issues and the potential impact they may have on the program are understood. This may include additional 10 CFR 50.69 specific sensitivity cases if they are judged necessary.

Note: As part of the required 10 CFR 50.69 PRA categorization, sensitivity cases directed by Nuclear Energy Institute (NEI) 00-04 (Reference 1), the Waterford 3 PRA model human error and common cause basic events are increased to their 95th percentile and also decreased to their 5th percentile values; and maintenance unavailability terms are set to 0.0. These and any other relevant sensitivity insights and results are capable of driving a component and respective functions to an HSS categorization. The uncertainty of the PRA modeled Human Error Probabilities (HEPs) and Common Cause Failures (CCFs) are thus explicitly accounted for in the 10 CFR 50.69 application. These sensitivity cases will be completed on all quantified models.

Internal Events and Internal Flood Uncertainty The process used to identify at power internal events related PRA model uncertainties and their impact is described in the Waterford 3 PRA uncertainty documentation (Reference 32). The internal flooding PRA model uncertainty document, PSA-WF3-01-IF-SOU (Reference 33),

contains details related to key assumptions and sources of uncertainty relevant to the flooding risk. In the uncertainty documentation, NUREG-1855 (Reference 21) and EPRI report 1016737 (Reference 22) were used to provide guidance for a structured process for addressing uncertainties in PRA inputs and results in the context of risk-informed decision-making.

W3F1-2022-0009 Page 2 of 19 Appendix A of EPRI 1016737 is used as a template to document plant-specific issue characterization and assessments to fully satisfy the related supporting requirements.

Uncertainty considerations for each model element (success criteria, human actions, data, etc.)

were also documented to ensure a comprehensive evaluation of PRA uncertainty was completed.

In addition to the items from EPRI 1016737 dispositioned in PSA-WF3-01-QU-01, a review of the plant specific assumptions was performed to ensure that the items that have the potential to be key for applications are retained for further discussion. That report screens items if they are a consensus approach, if the alternative choice is potentially non-conservative, if it is just a model choice (description of how as built plant is modeled), or if it is covered in the generic sources of uncertainty in EPRI 1016737. Any items that do not meet these criteria are potentially key for applications and retained for consideration in this attachment and included in Table 6-1. Items from the LERF uncertainty documented in PSA-WF3-01-LE are included in this report for discussion as well.

The parametric uncertainty analysis for the Waterford 3 at power internal events model is documented in Reference 32, and Reference 33 contains the parametric flood uncertainty evaluation. The parametric uncertainty analyses address the State-of Knowledge Correlation (SOKC) by the use of system level type codes for basic events. This applies the same variability of all components of that type within a system during the analysis. The parametric uncertainty analyses in the PRA quantification model and documentation demonstrate that the point estimate mean values provide a close representation of the propagated mean values reflecting SOKC and the propagated mean total Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) values were confirmed to meet Regulatory Guide 1.174, Revision 3 (Reference 20).

Table 6-1 evaluated key assumptions and sources of uncertainty from the baseline internal events and internal flooding models and assesses the impact of the item(s) on application of a 10 CFR 50.69 program. To determine whether these items should be considered key assumptions/ sources of uncertainty for the application, several criteria were applied as listed below. Items that do not meet these criteria are retained for consideration in the application.

This is consistent with the process in NUREG 1855 Stage E for identification of potential model uncertainties and related assumptions and determine their significance and identification of model uncertainties or related assumptions that are key to the decision.

A. The uncertainty or assumption will have no impact on the PRA results and therefore no impact on the decision of High Safety Significance (HSS) or Low Safety Significance (LSS) for any structures, systems, and components (SSCs).

B. There is no reasonable alternative to the assumption which would produce different results and/or there is no reasonable alternative that is at least as sound as the assumption being challenged. (Regulatory Guide 1.200, Rev 2)

C. The uncertainty or assumption implements a conservative bias in the PRA model, and that conservatism does not influence the results. These conservatisms are expected to be slight and only applied to minor contributors to the overall model. EPRI 1013491 uses the term "realistic conservatisms." Thus, uncertainties/assumptions that implement realistic (slight) conservativisms can be screened from further consideration.

W3F1-2022-0009 Attachment 6 Page 3 of 19 D. EPRI 1013491 elaborates on the definition of a consensus model to include those areas of the PRA where extensive historical precedence is available to establish a model that has been accepted and yields PRA results that are considered reasonable and realistic.

Thus, uncertainties/assumptions where there is extensive historical precedence that produces reasonable and realistic results can be screened from further consideration.

Table 6-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption / Uncertainty Discussion 10 CFR 50.69 Disposition The model has been updated to The uncertainty report documents a This represents a key source of include credit for FLEX sensitivity evaluation for the added uncertainty in the 50.69 application for equipment. The added FLEX FLEX changes. The sensitivity case specified systems modeling was the subject of a shows that FLEX has an impact and sensitivity evaluation to compare reduces Station Blackout (SBO) The Revision 6 Waterford Applications results with and without the added contribution. Uncertainty and Sensitivity Analysis FLEX equipment. contained a sensitivity on FLEX While this is documented in the equipment. This sensitivity examined the uncertainty and sensitivity impact the FLEX changes had on the documentation, it is not considered a model. While the current model reflects source of uncertainty. This was a the as-built, as-operated plant, due to sensitivity evaluation to gauge the regulatory concern over FLEX modeling, impact of a change to the model. This this is considered a source of uncertainty was a necessary change to accurately for the categorization of Emergency update the model to match plant Diesel Generator (EDG), Emergency response and the potential use of Feedwater (EFW), and FLEX SSCs FLEX equipment and strategies. based on the sensitivity performed in PSA-WF3-08-09.

Environmental impacts on Local environmental impacts can B. This was noted as a source of model initiating events (for example, increase or decrease the frequency of uncertainty for the Revision 6 model intake, offsite power). some initiators. An alignment for update but should not impact the severe weather is included for 10 CFR 50.69 application.

Waterford 3 that addresses the impact of severe weather on both Loss of The environmental impacts can impact Offsite Power (LOOP) frequency and the online risk monitor or influence recovery. changes to IE frequency. However, 10 CFR 50.69 risk evaluations will use The other environmental impact nominal conditions for assessing risk on applicable for Waterford 3 is extremely equipment and systems.

high temperatures. Waterford 3 has Technical Specifications that limit plant operation with excessive temperature in the ultimate heat sink. Forced shutdowns due to high temperature would be included in the reactor trip data.

Credit for repair and Recovery. Waterford 3 does not credit repair of C. No significant impact on the any equipment. Off-site power 10 CFR 50.69 application. Including This was noted as a source of recoveries are based on standard credit for equipment recovery would uncertainty in both the Success practices. reduce risk results. The current Criteria and System Analysis treatment is conservative. Conservative sections. Not crediting equipment recovery was treatment is more likely to result in noted as a source of uncertainty tied to additional HSS SSCs than to mask other Success Criteria. Applying recoveries results.

or completing more detailed off-site power recoveries could lower risk Lack of credit for potential equipment results. recovery can be a source of uncertainty

W3F1-2022-0009 Attachment 6 Page 4 of 19 Table 6-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption / Uncertainty Discussion 10 CFR 50.69 Disposition (the model not reflecting the real A sensitivity case for EDG recovery is response to an event). However, in this included in the quantification/sensitivity case the results bias in the direction of documentation. higher risk results and potentially more HSS SSCs.

Several bounding inputs and Bounding conditions for the Waterford B. Since the treatment of induced-SGTR conditions for LERF analysis were 3 LERF model include induced SGTRs and scrubbing are primarily a identified as potential sources of and credit for scrubbing. phenomenological uncertainty issue for uncertainty: LERF, this is not expected to have any Several sensitivity evaluations were impact on the 10 CFR 50.69 Use of generic value for SG included in the LERF documentation to categorization.

age/wear evaluate the impact the bounding and In-vessel core melt impact on potentially uncertain inputs have on All noted issues are treated TI-SGTR, and LERF results. For example, average conservatively in the LERF model and Credit for scrubbing. SG wear models were used even as are not expected to uniquely impact Waterford 3 recently replaced SGs. 10 CFR 50.69 categorization.

Sensitivity evaluations are documented in the LERF analysis for the impact of credited scrubbing, as well as in-vessel core melt on TI-SGTR.

Human actions credited during The Waterford 3 LERF model contains D. Treatment of post core damage severe accident conditions. three operator actions after core human actions is not expected to have damage: "bump the pump," late any impact unique to the 10 CFR 50.69 depressurization, and LOOP recovery. categorization.

Due to the uncertain nature of post- All credited human actions are already accident/post core damage conditions, subject to NEI guidance directed credit for these actions was identified sensitivity cases.

as a source of uncertainty. A sensitivity case is included in the documentation with no post core damage human actions credited.

Maintenance/operational activities Waterford 3 includes numerous B. This was noted in the PRA revision 6 (for example, switchyard work, alignment flags that can determine the documentation as a potential uncertainty system testing). impact of maintenance or operational candidate. However, 10 CFR 50.69 risk activities on risk. In addition, the risk evaluations will use nominal conditions associated with different types of for assessing risk on equipment and switchyard work can also be evaluated. systems.

Model may over-estimate Conservative criteria in LERF modeling C. No significant impact on a 10 CFR contribution of pressure induced may cause over estimation of LERF 50.69 application. Conservative SGTR (PI-SGTR) to LERF. risk. In modeling, induced SGTR - treatment is more likely to result in bounding conditions were applied. additional HSS SSCs than to mask other results.

The PI-SGTR and TI-SGTR values used the average wear models for the steam generator but the steam generators were recently replaced (the replacement Waterford 3 SGs have less wear than industry average).

W3F1-2022-0009 Attachment 6 Page 5 of 19 Table 6-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption / Uncertainty Discussion 10 CFR 50.69 Disposition For the Internal events analysis, it Uncertainty report disposition - A - Exclusion of negligible failure rates is extremely unlikely that the main redundant failures are several orders of will not impact the categorization of steam safety valves (MSSVs), magnitude below other failures and can systems for 10 CFR 50.69.

atmospheric dump valves (ADVs), be excluded.

and steam bypass valves will all fail to open and cause the required core heat removal criteria to not be met. Therefore, the steam relief portion of the function is not considered further.

For medium and large LOCAs Uncertainty report disposition - D - Requirement for hot leg injection is a simultaneous hot and cold leg conservative requirement of additional consensus approach of Large LOCAs.

injection ("hot leg injection") is systems for success. However, medium LOCAs could possibly assumed to be procedurally not have this concern depending on the required to be established within 3 size given that they are described by a hours in order to prevent boron range of break sizes. As such, the precipitation in the core that could approach is a simplification which will not disrupt core cooling and produce have a serious impact on categorization core damage. If "hot leg injection" as the hot leg injection implementation is is not established, core damage is dominated by operator action failure assumed to occur which will not impact categorization of SSCs for the 10 CFR 50.69 application.

For Steam Generator Tube Uncertainty report disposition - failure A - The transfer open failure and Ruptures (SGTRs), Waterford 3 rates several orders of magnitude lower subsequent failure to reclose is relies upon the ADVs to open to can be excluded. insignificant compared to other failure to relieve excess pressure. In this control pressure with ADVs. The situation, a random transfer open omission of this failure has a negligible of a MSSV is considered to be of impact on 10 CFR 50.69 categorization.

extremely low probability and is therefore not included.

Partial loss of feedwater events in Uncertainty report disposition - B - This assumption is appropriate given the Waterford 3 Licensee Event exclusion of insignificant failures. that these have historically led to reactor Report (LER) data were counted trips. Exclusion of these events would as reactor trip events. lead to potential exclusion of reactor trips.

As such, the categorization of systems in the 10 CFR 50.69 program will not be significantly impacted The ability to load shed allows the Uncertainty report disposition - C - this is a simplification that is only battery lifetime to be extended conservative treatment. conservative for FLEX related scenarios.

from 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> per Given the additional failures required for ER-W3-2002-0622. This provides FLEX scenarios, the impact of this additional time for loss of offsite assumption is minimal. As such, the power restoration. The 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> is categorization of systems in the conservative considering that the 10 CFR 50.69 program will not be use of FLEX load shedding could significantly impacted.

increase the battery life to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

W3F1-2022-0009 Attachment 6 Page 6 of 19 Table 6-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption / Uncertainty Discussion 10 CFR 50.69 Disposition Although Reference 44 (of Uncertainty report disposition - given C - this is a simplification that is only PSA-WF3-01-IE-01) is identified the additional load shed for FLEX, conservative for FLEX related scenarios.

as obsolete in the Entergy these estimates are conservative for Given the additional failures required for system, it was determined to still SBO scenarios. FLEX scenarios, the impact of this be applicable to the LOOP assumption is minimal. As such, the calculation done in support of the categorization of systems in the Waterford 3 PRA. This 10 CFR 50.69 program will not be conclusion is based on the fact significantly impacted.

that the Reference was performed on a snapshot of the DC loads at the time the study was completed.

A review of the current DC loads shows that the B and AB battery loads are essentially the same as the loads at the time of the study, and the Battery A loads have only increased by 3%. Since the PRA is based on best estimate calculations, the change in loads is very minor, and the battery depletion timing as used in the PRA analysis is not exact, the results of the calculation were determined to still be applicable for the LOOP calculation. The reference indicates that the battery lifetime is 2.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> without load shed. This could be extended if battery shedding were to be credited. This is considered conservative since the battery life could be extended with the use of FLEX equipment.

Those events that were caused by Uncertainty report disposition - these A - this assumption is appropriate given ice or salt spray were screened events have never occurred at the site that these events are or very unlikely. As from the Waterford 3 calculation and are extremely unlikely to occur. such, the categorization of systems in the since those types of LOOP events 10 CFR 50.69 program will not be are not likely to occur at significantly impacted.

Waterford 3.

Although the PSA model credits Uncertainty report disposition - C - Potentially conservative for SBO actions taken as part of the FLEX conservative treatment, more detailed scenarios only. FLEX is only credited strategy, the At-Power Loss of treatment will only drive down the when it is obvious that power will not be Off-Site Power Analysis did not results. restored within a few hours. Given that credit the use of FLEX equipment FLEX is already identified as a key source or strategies. This is conservative of uncertainty, any additional impacts since the use of FLEX equipment from this item are covered by that and strategies would result in consideration.

increased battery life and possibly a shorter time to recover site power.

W3F1-2022-0009 Attachment 6 Page 7 of 19 Table 6-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption / Uncertainty Discussion 10 CFR 50.69 Disposition Systems whose low-pressure Uncertainty report disposition - A - Exclusion of negligible failure rates will portions are isolated from RCS conservative as the product of the not impact the categorization of systems pressure by four or more normally failure rates is insignificant enough to in the 10 CFR 50.69 program.

closed valves or periodically leak- exclude.

tested check valves in series are eliminated from further consideration. Four or more valves in series was selected for Waterford 3 because using only three valves in series (as used in NUREG/CR-5744 can potentially eliminate lines from consideration that could have a potentially significant impact, especially if those valves have a long period of time between leak tests.

When the first valve fails, the Uncertainty report disposition - A - This assumption is consistent with failure can be detected and the conservative treatment. plant operation. Several redundant second or the third (if it is) valve failures are very unlikely and therefore will was not failed already, the not impact the categorization of systems mission time for the valve not in the 10 CFR 50.69 program.

failed downstream RCS interface is set to 7 days conservatively.

During the seven days the operators can identify failure of RCS interfacing valve and perform procedures to prevent ISLOCA only if there are procedures to identify the failure and allow the plant to enter a safe shutdown state.

Operator failure to close the motor Uncertainty report disposition - A - Exclusion of negligible failure rates will operated valves (MOVs) is exclusion of insignificant failures. not impact the categorization of systems considered insignificant for the in the 10 CFR 50.69 program.

MOVs in Shutdown Cooling (SDC) suction line since the MOVs are procedurally closed prior to startup, have close interlock and annunciation indicates the MOV position.

Inadvertent opening of the MOVs Uncertainty report disposition - A - Exclusion of negligible failure rates will in SDC suction line and in the Low exclusion of insignificant failures. not impact the categorization of systems Pressure Safety Injection (LPSI) in the 10 CFR 50.69 program.

injection line is considered insignificant since these valves status are indicated in the main control room (MCR) and checked every shift so wrong position would be corrected easily.

Consequently, the product of probability of Inadvertent opening of the MOV and the failure probability of not recovering is significantly low.

W3F1-2022-0009 Attachment 6 Page 8 of 19 Table 6-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption / Uncertainty Discussion 10 CFR 50.69 Disposition For SGTRs, it is assumed that Uncertainty report disposition - B - This is an appropriate estimate based operators have 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to initiate estimate based on industry guidance. on the guidance in the SGTR modeling cooldown, and shutdown entry WCAP and therefore the impact to the conditions will be reached in 20 results is negligible. As such, the hours. categorization of systems in the 10 CFR 50.69 program will not be significantly impacted.

The fuel oil transfer pump will start Uncertainty report disposition - value is A - Given that subsequent failures have and stop automatically as it refills an approximation, however it is based reduced failure rates that drop off the fuel oil day tank. The day tank on engineering judgement. exponentially after the first few starts, the holds enough fuel oil to maintain chosen approach is appropriate and likely at least 60 minutes of operation at reminiscent of the actual failure rate.

100 percent load plus 10 percent Given that this is an appropriate basic margin. Therefore, the fuel oil event value, categorization for the transfer pump is expected to start 10 CFR 50.69 application is not impacted.

multiple times over a 24-hour period. Six demands are assumed based on the capacity of the day tank and the reduced failure probability for subsequent pump starts. The fuel oil transfer pumps are included in the model.

The fuel oil transfer pumps Failure to Run (FR) during 1st hour is a generic value from NUREG-6928 (NRC Data Update issued in December 2016). The Failure to Start is a plant-specific value.

There will be no FR after 1st hour because these pumps work only for a short period of time, never longer than one hour.

Component Cooling valves Uncertainty report disposition - slightly C - Isolation of these valves impacts only CC-501 and CC-562 are not conservative modeling choice. the non-safety CCW header and will likely modeled because isolation valves Including these valves would decrease not impact the single pump operation CC-200A, CC-200B, CC-727, and the importance of Component Cooling concern. The failures that would be CC-563 are designed to provide Water (CCW). mitigated are not significant to the results.

isolation of both non-safety loops. As such, the categorization of systems in The AB header and the non- the 10 CFR 50.69 program will not be safety header are isolated by significantly impacted.

CC-501 and CC-562, respectively.

Not modeling these isolation valves is slightly conservative because this is equivalent to not crediting these valves for the case of failure of the modeled isolation valves.

W3F1-2022-0009 Attachment 6 Page 9 of 19 Table 6-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption / Uncertainty Discussion 10 CFR 50.69 Disposition Nitrogen (N2) system valves Uncertainty report disposition - as A - MCR Heating Ventilation and Air NG-ISV-609 and NG-ISV-610 described in assumption, additional Conditioning (HVAC) has a limited provide isolation of N2 detail not needed based on simple application in the model and is applicable accumulators I and II, which plant operation. only for transients which do not have support CCW and Auxiliary larger heat loads. Additionally, no credit Component Cooling Water is given in the model for opening doors (ACCW) to the essential chillers and using fans which would mitigate and the ACCW control valves much of the room cooling concern. The (ACC-126A and B). The only assumption is likely representative of the area in the Waterford 3 model actual conditions and therefore the impact where the Essential Chillers are to the results is negligible. As such, the credited in the model is for the categorization of systems in the 10 CFR Main Control Room (MCR). For 50.69 program will not be significantly the MCR, the room loads are impacted.

pretty constant so that an increase in loads might result in a 2 to 3 degree impact in the MCR but would not impact MCR habitability and if the MCR remains habitable, none of the MCR instrumentation would be close to the failure temperatures in the MCR. All other areas cooled by the Essential Chillers have calculational basis that show the equipment will not exceed the failure temperatures with a complete loss of HVAC and thus an increase or decrease in chiller loads would not impact equipment cooled by the chillers. Therefore, changes in Essential Chiller loads that would require throttling of the CCW and ACCW air operated valves (AOVs) do not impact any loads credited in the PRA and thus no Instrument Air or N2 dependence is needed for these valves.

W3F1-2022-0009 Attachment 6 Page 10 of 19 Table 6-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption / Uncertainty Discussion 10 CFR 50.69 Disposition The phase-mission success Uncertainty report disposition - A - This model simplification captures the criterion of the CCW system Wet modeling choice to avoid circular logic. risk significant failures of concern for the Cooling Towers (WCTs) and Dry highlighted scenarios. As such, the Cooling Towers (DCTs) for categorization of systems in the transient scenarios that evolve 10 CFR 50.69 program will not be into loss of coolant scenarios significantly impacted.

requires that the cooling necessary for success also change with this evolution. The success criteria for the WCT and DCT is based on the necessary loads. This changes between cases involving loss of coolant scenarios and those only requiring secondary-side cooling due to a need to address long term cooling for the High Pressure Safety Injection (HPSI) pumps. The logic model for the CCW system provides accurate assessment for loss of coolant and for loss of decay heat removal accident sequences. For transient-induced loss of coolant scenarios the model logic is correct for the point until the loss of coolant occurs since it applies the heat removal accident sequence success criterion for the CCW model.

However, should a loss of coolant event be generated (Reactor Coolant Pump (RCP) seal LOCA or safety relief valve failure to reclose, the logic does not require the cooling needed to address the post-LOCA system requirements including long term cooling of the HPSI pumps. To address this change the model logic has been adjusted to require operation of the associated train of ACCW for HPSI to be successful in recirculation. This change forces HPSI to fail on ACCW failure and simulates the change in state.

The Following Entries are Related to Internal Flooding:

The information shown on the Different pipe lengths would directly B - No significant impact on the walkdown data sheets in the impact the initiating event frequencies 10 CFR 50.69 application. Risk original internal flooding analysis which in turn would affect the core significant scenarios were checked and was assumed to be correct unless damage frequency. For risk-significant significant errors/discrepancies between changed in the Fire PRA flooding events, length data was based actual and documented pipe lengths are Walkdown Notebook. The legacy on drawing information or confirmed not expected.

information in the initial flooding during walkdowns. Therefore, the

W3F1-2022-0009 Attachment 6 Page 11 of 19 Table 6-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption / Uncertainty Discussion 10 CFR 50.69 Disposition walkdown was spot-checked to overall effect on the base model is verify accuracy. expected to be small.

For areas where fluid drains away This assumption is reasonable and D - This was noted in the document at a quickly, sealed penetrations are backed by regulatory requirements. It key assumption, but no sensitivity cases assumed to be effective at is also an industry consensus were developed as the treatment was preventing propagation between approach. Penetrations/barriers with a judged to be acceptable and not a source areas such that the propagation specific concern would be addressed of significant uncertainty.

would not directly impinge on on an individual basis.

equipment in the adjoining area This is a flood specific issue and is not and result in equipment failure. expected to impact application of the For example, the penetration 10 CFR 50.69 program.

seals will prevent spray and splash impacts from causing equipment failures in the adjoining area. Generic Letter 86-10 specifies that fire barriers must be capable of withstanding significant spray and splash after being exposed to fire. For areas where fluids drain away quickly, use of fire barriers is assumed to prevent propagation for the time needed for the fluid to drain away.

Flood-induced failure of AOVs Due to multiple potential flood-induced C/D - No significant impact on the involves the valve operators loss failure modes (e.g., valve shorts out, 10 CFR 50.69 application. Conservative of function but would also involve instrument air supply line fails, water treatment is more likely to result in the AOV failing to its designed fail affects valves diaphragm, etc.), it is not additional high safety SSCs than to mask position. However, if the fail-safe guaranteed that AOV will always fail to other results.

position is the desired position for its fail-safe position. Therefore, credit the PRA, it is assumed that the for fail-safe is not taken. This valve fails as-is. assumption is conservative, but reasonable. The overall effect on the base model is expected to be small.

It is considered improbable that The selection of 25 gpm is based on D - No significant impact on the flooding events with small flow engineering judgement, it is about the 10 CFR 50.69 application. Conservative rates will spray enough PRA same flow rate as three or four treatment is more likely to result in related equipment housed in the household garden hoses and is additional high safety SSCs than to mask affected zone to initiate a PRA considered reasonable. other results.

transient. Therefore, the smallest discharge rate from any break is This was identified in as a key assumed to be 25 gpm unless assumption. This assumption is noted otherwise (for specified reasonable and expected to be spray scenarios). conservative. The impact on the base model is expected to be small.

W3F1-2022-0009 Page 12 of 19 Fire PRA (FPRA) Model:

The process used to identify uncertainties and their impact is described for the fire PRA in the Waterford 3 FPRA uncertainty documentation (Reference 34).

The evaluation examines sources of uncertainty for each of the FPRA development follows NUREG-1855 (Reference 21). The Fire PRA Uncertainty Report contains discussions on topics outlined in EPRI report 1026511 Appendix B (Reference 23) arranged by tasks from NUREG/CR-6850 (Reference 25). Key sources of uncertainty are noted, and sensitivity cases are completed to evaluate them.

The Waterford 3 Fire PRA Sensitivity and Uncertainty Report is a thorough and comprehensive assessment of uncertainty. The report evaluates over seventy uncertainty topics and documents how each topic is addressed in the Waterford 3 Fire PRA model.

The parametric uncertainty analysis for the Waterford 3 Fire PRA is provided in the uncertainty documentation. The parametric uncertainty analysis addresses the State-of-Knowledge Correlation (SOKC) by the use of system level type codes for basic events, and additional SOKC factors included in quantitation. This applies the same variability of all components of that type within a system during the analysis. The parametric uncertainty analyses in the PRA quantification model/model documentation demonstrate that the point estimate mean values provide a close representation of the propagated mean values reflecting SOKC and the propagated mean total CDF and LERF values were confirmed to meet Regulatory Guide 1.174, Revision 3 (Reference 20).

The Fire PRA uncertainty document identifies key assumptions and sources of uncertainty relevant to development and quantification of the fire models. The items identified in that document are included in Table 6-2 below. The table includes items judged to be key assumptions and sources of uncertainty relevant to the fire PRA. Table 6-2 assesses whether the key fire PRA assumptions and uncertainty topics are specifically relevant to the use of that model for the 10 CFR 50.69 application.

As noted in the uncertainty/sensitivity documentation, the Waterford 3 Fire PRA analysis is believed to represent a somewhat conservative estimation of fire risk, within the constraints of the requirements for a model acceptable for the NFPA-805 program. As the model is somewhat conservative, its application for a 10 CFR 50.69 program will likely slightly bias results toward more HSS classification of components/systems. The evaluation of sources of uncertainty in the FPRA are documented in Table 2 below including consideration for impact on a 10 CFR 50.69 application.

The Fire PRA uncertainty document identifies key assumptions and sources of uncertainty relevant to development and quantification of the fire models. The items identified in that document are included in Table 6-2 below. The table includes items judged to be key assumptions and sources of uncertainty relevant to the fire PRA. Table 6-2 assesses whether the key fire PRA assumptions and uncertainty topics are specifically relevant to the use of that model for the 10 CFR 50.69 application. To determine whether these items should be considered key assumptions/ sources of uncertainty for the application, several criteria were applied as listed below. Items that do not meet these criteria are retained for consideration in the application. This is consistent with the process in NUREG 1855 Stage E for identification of potential model uncertainties and related assumptions and determine their significance and identification of model uncertainties or related assumptions that are key to the decision.

W3F1-2022-0009 Page 13 of 19 A. The uncertainty or assumption will have no impact on the PRA results and therefore no impact on the decision of HSS or LSS for any SSCs.

B. There is no reasonable alternative to the assumption which would produce different results and/or there is no reasonable alternative that is at least as sound as the assumption being challenged. (Regulatory Guide 1.200, Rev. 2).

C. The uncertainty or assumption implements a conservative bias in the PRA model, and that conservatism does not influence the results. These conservatisms are expected to be slight and only applied to minor contributors to the overall model. EPRI 1013491 uses the term "realistic conservatisms." Thus, uncertainties/assumptions that implement realistic [slight] conservativisms can be screened from further consideration.

D. EPRI 1013491 elaborates on the definition of a consensus model to include those areas of the PRA where extensive historical precedence is available to establish a model that has been accepted and yields PRA results that are considered reasonable and realistic.

Thus, uncertainties/assumptions where there is extensive historical precedence that produces reasonable and realistic results can be screened from further consideration.

W3F1-2022-0009 Attachment 6 Page 14 of 19 Table 6-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty 10 CFR 50.69 Disposition Scope of equipment credited in Lack of credit for some systems D - Categorization evaluations for the fire PRA model could mask the risk associated with Emergency Diesel Generator (EDG) those systems in some applications. system, if selected for categorization, will Additionally, that same lack of credit include sensitivity studies associated with could overestimate the importance of the TEDG.

other credited systems.

Besides the EDG/TEDG related systems, A sensitivity case was completed to this topic is not a source of uncertainty for evaluate this. The Temporary the 10 CFR 50.69 program (the sensitivity Emergency Diesel Generator (TEDG) case resulted in less than 1% change in credit was adjusted. The TEDG was fire CDF and LERF).

made permanent during a plant modification and was added to the PRA model in the last internal events update. There is some uncertainty with regard to how quickly the TEDG would be used in the event of a fire.

For this sensitivity, it is assumed the TEDG is not available.

Exclusion of certain systems due Lack of credit for some systems D - The current approach used (assume to lack of cable data (systems with limited cable data - all equipment lacking detailed cable data is assumed failed) could mask the risk failed) will result in conservative associated with those systems in evaluations. This conservativism would some applications. Additionally, that tend to result in additional SSCs being same lack of credit could categorized as HSS in the 10 CFR 50.69 overestimate the importance of other categorization process.

credited systems.

A sensitivity case was completed to evaluate this uncertainty. The model was evaluated with credit for Unlocated (assumed failed)

Equipment. In the case run, the same equipment was assumed available in all locations except Turbine Generator Building (TGB).

Development of fire frequencies Present NUREG/CR-6850 results in This represents a key source of uncertainty for each fire area and ignition different fire frequencies for the same in the 10 CFR 50.69 application for source equipment in different plants. For specified systems.

example, older BWRs with less equipment than a new PWR may Reference 27 performed a sensitivity for result in a factor of 2 higher fire fire ignition frequencies (using latest frequencies for pumps or electrical frequencies from NUREG-2169) and their equipment. This is a form of impact on importance measures. Due to parameter uncertainty. The the potential impact to categorization of Waterford 3 FPRA uses the ignition systems, this is considered a source of frequencies from EPRI Supplement 1 uncertainty for the categorization of DC, to NUREG/CR-6850. The source of EFW, 4KV, Inverters and Distribution (ID),

data for the ignition frequencies is a and Station Service Distribution (SSD) source of uncertainty. SSCs based on the sensitivity performed in PSA-WF3-08-09.

A sensitivity case was completed to evaluate this uncertainty. The model was run with ignition frequencies from different sources (NUREG/

CR-6850 vs. EPRI Supplement 1).

W3F1-2022-0009 Attachment 6 Page 15 of 19 Table 6-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty 10 CFR 50.69 Disposition Credit for fire wrap No credit is taken for fire wrap A - The inclusion/exclusion of fire wrap has (qualified 3M wrap) in the a negligible impact on the overall results.

Waterford 3 FPRA. Due to the small impact demonstrated by sensitivity case, the uncertainty sensitivity A sensitivity case was completed to cases associated with this model evaluate this uncertainty. Credit for uncertainty is negligible and not relevant Wrap in Risk Significant locations for the 10 CFR 50.69 application.

where wrap exists but is not credited was evaluated (it assumed the wrap prevents failure of FPRA targets that are wrapped).

Treatment of permissive signals, In the Waterford 3 FPRA cable B - The sensitivity cases were performed to interlocks, and associated logic selection, the majority of the circuit measure the risk associated with the analysis is performed using detailed treatment of these cables that in select fire and conservative safe shutdown scenarios. Although the sensitivity shows analysis. Any additional cable a small impact on Fire CDF and Fire LERF, selection is performed in a similar the uncertainty document concluded that manner. Automatic actuation logic the modeling treatment applied to these signals are modeled as separate cables is realistic and preferred to pseudo-components with their own alternative modeling treatments.

cable selection. Thus, associated circuits are accounted for in the cable This is not expected to have any impact on selection. Some component specific the 10 CFR 50.69 categorization for changes were made in fire Waterford 3 plant systems.

modeling/quantification (with documented basis for each decision).

Due to the special treatment applied to certain signal cables. This treatment was identified as a source of modeling uncertainty.

Two sensitivity cases were documented to examine this uncertainty. In one case, the cables with special treatment were assumed to always fail. In the other case, the same cables were assumed to never fail.

Circuit failure probabilities Failure probabilities utilized in circuit A - The use of hot short failure probability failure analysis could be a form of given fire induced failure probability is parameter uncertainty, but the choice based on fire test data and associated of the representative set of values is consensus methodology published in a form of model uncertainty. The NUREG-7150, Volume 2. Based on a Waterford 3 FPRA utilizes circuit review of the assumptions and potential failure probabilities from uncertainty related to this element, it is NUREG/CR-7150. The aggregate concluded that the methodology for the failure probabilities are used, which Circuit Failure Mode Likelihood Analysis combines intra and inter cable faults, task does not introduce any epistemic and is specific to the type of circuit uncertainties that would affect the and material type, e.g., single break 10 CFR 50.69 program.

control circuit, thermoset cables, ungrounded DC Solenoid Operated Valves (SOVs).

A sensitivity case was completed to examine this uncertainty. In the sensitivity, all Circuit Failure Mode

W3F1-2022-0009 Attachment 6 Page 16 of 19 Table 6-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty 10 CFR 50.69 Disposition Likelihood (CFMLA) values were set to 1.0, i.e., no credit for hot short probabilities.

Availability of power for spurious In the Waterford 3 FPRA, a B - This is a fire PRA specific issue and is operations after initial cable simplifying assumption is made that limited to the RCPs (components not likely failure the power is available to a cable to be included in 10 CFR 50.69 damaged by a fire and thus can categorization). This is not expected to spuriously operate, when in fact the have any impact on the 10 CFR 50.69 fire damage may cause the power categorization for Waterford 3 plant supply to be interrupted. Power is systems.

assumed to be initially available to allow spurious operation to occur.

This is a conservative assumption.

An exception to this modeling assumption is in the Turbine Building, if power is lost to the 6.9kV switchgears, the RCPs are assumed to be tripped and cannot spuriously start or fail to be tripped by the operators, despite a spurious failure of the RCP control cables.

A sensitivity case was completed to examine this uncertainty. The case removes this treatment for the RCPs (they can spuriously start/operate with loss of 6.9kV power).

Credit for detection and In the Waterford 3 FPRA, manual B - This is a fire detection/suppression suppression suppression is not credited for oil fire specific issue. This is not expected to scenarios, since it was determined have any impact on the 10 CFR 50.69 that oil fires damage targets quickly, categorization for Waterford 3 plant and it is assumed that target damage systems.

occurs prior to the fire brigade responding. For electrical and cable fires, a bounding time to damage is determined that is applicable to all such fire scenarios, and this time to damage is used to establish time available to credit manual suppression utilizing generic non-suppression probabilities, as provided in FAQ 08-0050 in Supplement 1 of NUREG/CR-6850 (Reference 54).

Several Sensitivity cases were run to examine the uncertainty and model sensitivity to credited detection and suppression. The following cases were completed.

Assume failure of all automatic suppression systems (no credit for automatic suppression.

W3F1-2022-0009 Attachment 6 Page 17 of 19 Table 6-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty 10 CFR 50.69 Disposition Credit automatic suppression in the general areas of the Turbine Building, PAU TGB.

A case that assumes failure of all automatic detection systems and thus manual suppression in all areas that are not continuously manned, such as the MCR.

A case was run to examine manual detection/suppression.

The case assumed failure of all manual suppression by the dedicated fire watch in hot work scenarios, as well as manual suppression in the MCR and other continuously manned locations.

Effectiveness of passive fire The fire barrier failure probabilities D - This is a fire barrier specific issue. This barriers between compartments are based on the partitioning is not expected to have any impact on the elements in the barrier. In the 10 CFR 50.69 categorization for Waterford 3 FPRA, the failure Waterford 3 plant systems.

probabilities for passive fire barrier partitioning elements are based on the generic values listed in NUREG/CR-6850 Appendix A. The barrier failure probability for the barrier is the sum of the partitioning elements applicable to that barrier.

A couple of sensitivity cases were run to examine uncertainty associated with credited barrier failure probabilities. In one case it was assumed all barriers fail with a probability of 1.0 (i.e., they are not effective due to a door being propped open, a penetration not being filled after maintenance, etc.).

In a second case, barrier failure probabilities were reduced. It was assumed all barriers are better by a factor of 10 (i.e., they are less conservatively modeled, unless the failure is currently 1.0).

Treatment of structural steel The potential impacts of structural A - This is a fire specific issue associated failures steel failures were considered and with large fires in the Turbine Building.

assessed at Waterford 3, and it was This is not expected to have any impact on determined that the TGB is the only the 10 CFR 50.69 categorization for PAU noted to contain potential Waterford 3 plant systems. It is also a high-hazard fire sources. All oil fire conservative treatment and any impacts scenarios in the TGB were evaluated, would slightly bias systems/components and those that could possibly impact more toward HSS classification.

a structural member were assumed to fail the entire TGB outside the

W3F1-2022-0009 Attachment 6 Page 18 of 19 Table 6-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty 10 CFR 50.69 Disposition Switchgear Room. This assumption could contain a large amount of conservatism.

A sensitivity case was completed to examine this uncertainty.

A case was run to reduce conservatism with structural collapse fires in the turbine building. Some hydrogen fires and oil fires were adjusted to only impact the source of the fire and nearby targets, but not collapse the Turbine Building.

Treatment of fire-induced For all operator actions credited in D - The Waterford 3 Fire PRA model is instrument failures the Waterford 3 FPRA, there is a based on industry consensus modeling discussion of the cues and/or approaches for its HEP calculations, so required instrumentation. Some this is not considered a significant source actions model the required of uncertainty. As directed by NEI 00-04, instrumentation. For other operator human error basic events are increased to actions, redundancy and diversity of their 95th percentile and also decreased to instruments is credited as a means to their 5th percentile values as part of the assume that sufficient indication required 10 CFR 50.69 PRA categorization exists. In this case, the assumption sensitivity cases. These results are that it does not need to be modeled capable of driving a component and becomes a source of model respective functions HSS and, therefore, uncertainty. the uncertainty of the Fire PRA modeled HEPs are accounted for in the A sensitivity case was completed to 10 CFR 50.69 application.

examine this uncertainty.

Indication for the Condensate Storage Pool (CSP) Makeup or Alternate EFW Suction are among the most important indications for operators. This is because the system window is several hours long, it is assumed alternate indications for CSP level could be used as an operator cue; however, if alternate indication is not available or provides false readings, these actions would be hindered. This sensitivity case assumes failure probabilities of CSP makeup and alternate CSP suction are increased by factor of 10 (this includes relevant combination HFEs).

HEP Methodology The basis for the HEP methodology D - As directed by NEI 00-04, human error utilized needs to be consistent with basic events are increased to their 95th the internal events PRA standard percentile and also decreased to their 5th requirements for HRA. Model percentile values as part of the required uncertainty exists on the actual 10 CFR 50.69 PRA categorization methodology utilized; this is sensitivity cases. These results are recognized as a generic source of capable of driving a component and model uncertainty. The Waterford 3 respective functions HSS and, therefore, FPRA utilized EPRI HRAC per the the uncertainty of the Fire PRA modeled guidance in NUREG-1921

W3F1-2022-0009 Page 19 of 19 Table 6-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty 10 CFR 50.69 Disposition (Reference 24) in the development of HEPs are accounted for in the fire-specific HEP values. 10 CFR 50.69 application.

Several Fire HEP related sensitivity cases were run to examine relevant uncertainties. The cases included:

A case examining - Manual Control of EFW Flow (Long term ex-MCR actions are assumed to have the same failure probability as the internal events model; however, if these actions would be hindered by the fire in some way, the failure probabilities would be higher; assume failure probabilities of these actions are increased by a factor of 10, including relevant combination HFEs.)

A case examining - Manual Actuation of EFAS if Auto-Actuation Fails (This is the most risk-significant single operator action; this is a simple action that is assumed to be a factor of 10 worse than the internal events probability, per the guidance in NUREG-1921; however, since this is the most important action and is a very simple action. The case assumes that the probability is equal to that calculated for internal events and reduces the relevant action and combination HFEs by a factor of 10.)

A case examining - Tripping of the RCPs Following Loss of Seal Cooling (Combination of tripping from the MCR, which is assumed to be a factor of 10 worse than the internal events value, and locally tripping the breaker in TGB, which has a detailed fire-specific calculation). This action is known by the operators to be an important action and is very simple.

This case reduces the failures associated with tripping the RCPs (and relevant HFE combination events) by a factor of 10.