RS-24-056, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors

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

ML24149A261
Person / Time
Site:	Dresden
Issue date:	05/28/2024
From:	Humphrey M Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RS-24-056
Download: ML24149A261 (1)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office

RS-24-056 10 CFR 50.69 10 CFR 50.90 May 28, 2024

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

Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249

Subject:

Application to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of S tructures, Systems, and C omponents for Nuclear Power Reactors"

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests amendments to Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS) Units 2 and 3.

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

The enclosure to this letter provides the basis for the proposed change to the D NPS, Units 2 and 3 Facility Operating Licenses. The categorization process being implemented through this change is consistent with NEI 00- 04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, dated July 2005, which was endorsed by the NRC in Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1, dated May 2006.

provides a list of categorization prerequisites. Use of the categorization process on a plant system will only occur after these prerequisites are met.

May 28, 2024 U.S. Nuclear Regulatory Commission Page 2

The PRA models described within this license amendment request (LAR) are the same as those described within the CEG submittal, dated May 8, 2024, "RS 003 - License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b,' and TSTF-591, 'Revise Risk Informed Completion Time (RICT) Program,'"

(i.e., Accession No. ML24129A135).

CEG requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the application currently in-process. This would enhance the efficiency of CEG and NRC resources used for the review of the applications.

These requests should not be considered linked licensing actions, as the details of the PRA models in each LAR are complete which will allow the NRC to independently review and approve each LAR on its own merits without regard to the results from the review of the other.

CEG requests approval of the proposed change by May 28, 2025. The amendment shall be implemented within 60 days of approval.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation, "

Paragraph (a)(1), the analysis about the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is being provided to the NRC.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

Paragraph (b), CEG is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

May 28, 2024 U.S. Nuclear Regulatory Commission Page 3

There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (779) 231-5765.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 28th day of May 202 4.

Respectfully,

Mark D. Humphrey Sr. Manager - Licensing Constellation Energy Generation, LLC

Enclosure:

Application to Adopt 10 CFR 50.69

Attachments to the

Enclosure:

1. List of Categorization Prerequisites

2. Description of PRA Models Used in Categorization

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

4. External Hazards Screening

5. Progressive Screening Approach for Addressing External Hazards

6. Disposition of Key Assumptions/Sources of Uncertainty

cc: Regional Administrator - NRC Region III NRC Senior Resident Inspector - DNPS NRC Project Manager, NRR - DNPS Illinois Emergency Management Agency and Department of Homeland Security -

Division of Nuclear Safety RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 1 of 74

TABLE OF CONTENTS

1

SUMMARY

DESCRIPTION................................................................................................... 3

2 DETAILED DESCRIPTION................................................................................................... 3 2.1 Current Regulatory Requirements.............................................................................. 3 2.2 Reason For Proposed Change................................................................................... 3 2.3 Description of the Proposed Change.......................................................................... 4

3 TECHNICAL EVALUATION.................................................................................................. 5 3.1 Categorization Process Description (10 CFR 50.69(b)(2)(i))...................................... 6 3.1.1 Overall Categorization Process.......................................................................... 6 3.1.2 Passive Categorization Process.......................................................................11 3.2 Technical Adequacy Evaluation (10 CFR 50.69(b)(2)(ii)).......................................... 12 3.2.1 Internal Events and Internal Flooding................................................................12 3.2.2 Fire Hazards.....................................................................................................12 3.2.3 Seismic Hazards...............................................................................................12 3.2.4 Other External Hazards....................................................................................13 3.2.5 Low Power and Shutdown................................................................................20 3.2.6 PRA Maintenance and Updates........................................................................20 3.2.7 PRA Uncertainty Evaluations............................................................................20 3.3 PRA Review Process Results (10 CFR 50.69(b)(2)(iii)).............................................21 3.4 Risk Evaluations (10 CFR 50.69(b)(2)(iv))................................................................. 23 3.5 Feedback and Adjustment Process...........................................................................24

4 REGULATORY EVALUATION............................................................................................ 25 4.1 Applicable Regulatory Requirements/Criteria............................................................25 4.2 No Significant Hazards Consideration Analysis.........................................................25 4.3 Conclusions..............................................................................................................27

5 ENVIRONMENTAL CONSIDERATION............................................................................... 27

6 REFERENCES.................................................................................................................... 27

RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 2 of 74

LIST OF ATTACHMENTS

List of Categorization Prerequisites...................................................... 33

Description of PRA Models Used in Categorization.............................. 34

Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items............................................................... 35

External Hazards Screening.................................................................. 36

Progressive Screening Approach for Addressing External Hazards................................................................................... 60

Disposition of Key Assumptions/Sources of Uncertainty...................... 61

RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 3 of 74

1

SUMMARY

DESCRIPTION

The proposed amendment modifies the Dresden Nuclear Power Station (DNPS), Units 2 and 3 licensing bases, by the addition of License Conditions, for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, " Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance (LSS), alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance (HSS),

requirements will either not be changed or will be enhanced. This allows improved focus on equipment that has high safety significance resulting in improved plant safety.

2 DETAILED DESCRIPTION

2.1 Current Regulatory Requirements

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

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

"safety-related," "important to safety," or "basic component." The terms "safety-related" and "basic component" are defined in the regulations, while " important to safety," used principally in the general design criteria (GDC) of Appendix A to 10 CFR Part 50, is not explicitly defined.

2.2 Reason For Proposed Change

A probabilistic approach to regulation enhances and extends the traditional deterministic approach by allowing consideration of a broader set of potential challenges to safety, providing a logical means for prioritizing these challenges based on safety significance, and allowing RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 4 of 74

consideration of a broader set of resources to defend against these challenges. In contrast to the deterministic approach, Probabilistic Risk Assessments (PRAs) address credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for common cause failures. The probabilistic approach to regulation is an extension and enhancement of traditional regulation by considering risk in a comprehensive manner.

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

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

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

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

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

2.3 Description of the Proposed Change

Constellation Energy Generation, LLC (CEG) proposes the addition of the following condition to the renewed operating licenses of DNPS, Units 2 and 3 to document the NRC's approval of the use 10 CFR 50.69.

RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 5 of 74

CEG is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC) -1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire, seismic events; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non -Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except external flooding and extreme winds and tornadoes; an external flood safe shutdown equipment list for external floods; and a high wind safe shutdown equipment list for extreme winds and tornadoes.

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 probabilistic risk assessment approach to seismic margins approach).

3 TECHNICAL EVALUATION

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

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

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

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

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

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

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

RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 6 of 74

The PRA models described within this license amendment request (LAR) are the same as those described within the CEG submittal, dated May 8, 2024, "RS 003 - License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b,' and TSTF-591, 'Revise Risk Informed Completion Time (RICT) Program,'"

(i.e., Accession No. ML24129A135).

CEG requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the TSTF -505 and TSTF-591 (RICT Program) application currently in-progress. Concurrent review will enhance efficiencies for the use of CEG and NRC resources necessary to complete the review of the separate applications. Th ese requests should not be considered linked requested licensing actions (RLAs), since each LAR independently captures t he complete details of the PRA models which would allow the NRC to independently review and approve each LAR on its own merits without regard to the results from the review of the other.

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

3.1.1 Overall Categorization Process

CEG will implement the risk categorization process at DNPS in accordance with NEI 00- 04, Revision 0, "10 CFR 50.69 SSC Categorization Guideline, " as endorsed by Regulatory Guide (RG) 1.201, " Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance" (Reference [2]). NEI 00-04 Section 1.5 states " Due to the varying levels of uncertainty and degrees of conservatism in the spectrum of risk contributors, the risk significance of SSCs is assessed separately from each of five risk perspectives and used to identify SSCs that are potentially safety-significant." A separate evaluation is appropriate to avoid reliance on a combined result that may mask the results of individual risk contributors.

The process to categorize each system will be consistent with the guidance in NEI 00 04, "10 CFR 50.69 SSC Categorization Guideline," as endorsed by RG 1.201. RG 1.201 states that "the implementation of all processes described in NEI 00 04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence" and that "all aspects of NEI 00 04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1)(iv)."

However, neither RG 1.201 nor NEI 00 04 prescribe a particular sequence or order for each of the elements to be completed. Therefore, the order in which each of the elements of the categorization process (listed below) is completed is flexible and as long as they are all complete, they may even be performed in parallel. Note that NEI 00 04 only requires Item 3 to be completed for components/functions categorized as Low Safety Significant (LSS) by all other elements. Similarly, NEI 00 04 only requires Item 4 to be completed for safety related active components/functions categorized as LSS by all other elements.

RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 7 of 74

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

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

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

4. Defense-in-depth assessment

5. Passive categorization methodology

Categorization of SSCs at DNPS will be completed in accordance with the NEI 00-04 process, as endorsed by RG 1.201, which includes the determination of safety significance through the various elements identified above. The results of these elements are used as inputs to arrive at a preliminary component categorization (i.e., HSS or LSS that is presented to the Integrated Decision-Making Panel (IDP). Note that the term " preliminary HSS or LSS" used in this application is synonymous with the NEI 00-04 term " candidate HSS or LSS." A component or function is preliminarily categorized as HSS if any element of the process results in a preliminary HSS determination in accordance with Table 3-1 below. The safety significance determination of each element, identified above, is independent of each other and therefore the sequence of the elements does not impact the resulting preliminary categorization of each component or function. Consistent with NEI 00- 04, the categorization of a component or function will only be " preliminary" until it has been confirmed by the IDP. Once the IDP confirms that the categorization process was followed appropriately, the final RISC category can be assigned.

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

Table 3-1: Categorization E valuation Summary Categorization IDP Drives Element Step - NEI 00- 04 Evaluation Level Change Associated Section HSS to Functions LSS Internal Events Not Base Case - Allowed Yes Section 5.1 Risk (PRA Fire, Seismic and Modeled) Other External Component Allowable No Events Base Case PRA Sensitivity Allowable No Studies RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 8 of 74

Categorization IDP Drives Element Step - NEI 00- 04 Evaluation Level Change Associated Section HSS to Functions LSS Integral PRA Not Assessment - Allowed Yes Section 5.6 Other External Risk (Non-PRA Hazards Component Not Allowed No Modeled) Shutdown - Not Section 5.5 Function/Component Allowed No Core Damage - Function/Component Not Yes Defense-in-Depth Section 6.1 Allowed Containment - Component Not Yes Section 6.2 Allowed Qualitative Considerations - Function Allowable1 N/A Criteria Section 9.2 Passive Passive - Section 4 Segment/Component Not Allowed No

Notes:

1 The assessments of the qualitative considerations are agreed upon by the 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 by 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

RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 9 of 74

seven considerations cannot be confirmed (false response) for a function, then the final categorization of that function is HSS.

The mapping of components to system functions is used in some categorization process steps to facilitate preliminary categorization of components. Specifically, functions with mapped components that are determined to be HSS by the PRA-based assessment (i.e., Internal Events PRA or Integral PRA assessment) or defense-in-depth evaluation will be initially treated as HSS. However, NEI 00- 04 Section 10.2 allows detailed categorization which can result in some components mapped to HSS functions being treated as LSS; and Section 4.0 discusses additional functions that may be identified (e.g., fill and drain) to group and consider potentially LSS components that may have been initially associated with a HSS function but which do not support the critical attributes of that HSS function. Note that certain steps of the categorization process are performed at a component level (e.g., Passive, Non-PRA-modeled hazards - see Table 3-1). These components from the component level assessments will remain HSS (IDP cannot override) regardless of the significance of the functions to which they are mapped.

Therefore, if an HSS component is mapped to an LSS function, that component will remain HSS. If an LSS component is mapped to an HSS function, that component may be driven HSS based on Table 3-1 above or may remain LSS.

The following clarifications are 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, (1) the purpose of the categorization; (2) present treatment requirements for SSCs including requirements for design basis events; (3) PRA fundamentals; (4) 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 (5) the defense-in-depth philosophy and requirements to maintain this philosophy.

• The decision criteria for the IDP to categorize SSCs as safety significant or LSS in accordance with § 50.69(f)(1) will be documented in CEG procedures. Decisions of the IDP will be arrived at by consensus. Differing opinions will be documented and resolved, if possible. However, a simple majority of the panel is sufficient for final decisions regarding safety significant and LSS.

• Passive catego rization 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.

RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 10 of 74

• An unreliability factor of 3 will be used for the sensitivity studies described in Section 8 of NEI 00- 04. The factor of 3 was chosen as 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 [5]) which states "if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00- 04) or the defense-in-depth assessment (Section 6), the associated system function(s) would be identified as HSS."

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

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

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

• Internal Event Risks: Internal events including internal flooding PRA model, as submitted to the NRC for TSTF-505/TSTF -591, dated May 8, 2024 (RS-24-003) (Refer to Attachment 2).

• Fire Risks: Fire PRA model, as submitted to the NRC for TSTF -505/TSTF -591, dated May 8, 2024 (RS 003) - (Refer to Attachment 2).

• Seismic Risks: Seismic PRA model (Refer to Attachment 2).

• External Hazard Flooding Risks: an external flood safe shutdown equipment list.

• External Hazard Extreme Winds and Tornado Risk: a high wind safe shutdown equipment list.

• Other External Hazard Risks: Using the IPEEE screening process as approved by the NRC in its Safety Evaluation dated September 28, 2001 (TAC Nos. M83616 and M83617). The other external hazards (besides external flooding and extreme winds and tornadoes) were determined to be insignificant contributors to plant risk.

• Low Power and Shutdown Risks: Qualitative defense-in-depth (DID) shutdown model for shutdown Configuration Risk Management (CRM) based on the framework for DID provided in NUMARC 91-06, "Guidance for Industry Actions to Assess Shutdown RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 11 of 74

Management" (Reference [4]), which provides guidance for assessing and enhancing safety during shutdown operations.

A change to the categorization process that is outside the bounds specified above will not be used without prior NRC approval. The SSC categorization process documentation will include the following elements:

1. Program procedures used in the categorization

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

3. Mapping of components to support function(s)

4. PRA model results, including sensitivity studies

5. Hazards analyses, as applicable

6. Passive categorization results and bases

7. Categorization results including all associated bases and RISC classifications

8. Component critical attributes for HSS SSCs

9. Results of periodic reviews and SSC performance evaluations

10. IDP meeting minutes and qualification/training records for the IDP members

3.1.2 Passive Categorization Process

For the purposes of 10 CFR 50.69 categorization, passive components are those components that have a pressure retaining function. Passive components and the passive function of active components will be evaluated using the Arkansas Nuclear One (ANO) Risk-Informed Repair/Replacement Activities (RI-RRA) methodology contained in Reference [8] consistent with the related Safety Evaluation (SE) issued by N RC.

The RI-RRA methodology is a risk-informed safety classification and treatment program for repair/replacement activities for pressure retaining items and their associated supports. In this method, the component failure is assumed with a probability of 1.0 and only the consequence evaluation is performed. It additionally applies deterministic considerations (e.g., DID, safety margins) in determining safety significance. Component supports, if categorized, are assigned the same safety significance as the highest passively ranked component within the bounds of the associated analytical pipe stress model. Consistent with NEI 00- 04, an HSS determination by the passive categorization process cannot be changed by the IDP.

The use of this method for a 10 CFR 50.69 application was previously approved in the final SE for Vogtle dated December 17, 2014 (Reference [5] ). The RI-RRA method as approved for use at Vogtle for 10 CFR 50.69 does not have any plant specific aspects and is therefore generic. It relies on the conditional core damage and large early release probabilities associated with postulated ruptures. Safety significance is generally measured by the frequency and the consequence of the event. However, this RI-RRA process categorizes components solely based on consequence, which measures the safety significance of the passive component given that it ruptures. This approach is conservative compared to including the rupture frequency in the categorization since this approach will not allow the categorization of SSCs to RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 12 of 74

be affected by any changes in frequency due to changes in treatment. The passive categorization process is intended to apply the same risk-informed process accepted by the NRC in the ANO2-R&R- 004 for the passive categorization of Class 2, 3, and non-class components. This is the same passive SSC scope the NRC has conditionally endorsed in ASME Code Cases N-660 and N-662 as published in RG 1.147, Revision 15 (Reference [9]).

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

3.2 Technical Adequacy Evaluation (10 CFR 50.69(b)(2)(ii ))

The following sections demonstrate that the quality and level of detail of the processes used in categorization of SSCs are adequate. The PRA models described below have been peer reviewed and there are no PRA upgrades that have not been peer reviewed. T he PRA models described within this license amendment request (LAR) are the same as those described within the CEG submittal, dated May 8, 2024, "RS 003 - License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2,

'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b,' and TSTF-591,

'Revise Risk Informed Completion Time (RICT) Program,'" (i.e., Accession No. ML24129A135).

3.2.1 Internal Events and Internal Flooding

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

3.2.2 Fire Hazards

The DNPS categorization process for fire hazards will use a peer reviewed plant-specific fire PRA model. The internal Fire PRA model was developed consistent with NUREG/CR-6850 (Reference [10]) and only utilizes methods previously accepted by the NRC. The CEG risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for DNPS. Attachment 2 of this enclosure identifies the applicable Fire PRA model.

3.2.3 Seismic Hazards

The DNPS categorization process for seismic hazards will use a peer reviewed plant-specific seismic PRA model. The CEG risk management process ensures that the PRA model used in this application reflects the as built and as operated plant for DNPS. Industry standard methods were utilized in the development of the seismic hazards for the SPRA. Updates to seismic hazard curves will be reflected in the PRA used for the categorization in accordance with the RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 13 of 74

PRA model maintenance process. Attachment 2 of this enclosure identifies the applicable SPRA model.

3.2.4 Other External Hazards

An analysis of other external hazards is documented in Reference [6]. The analysis in this Section is also taken from Reference [6].

3.2.4.1 External Flooding

DNPS External Hazards analysis report (Reference [6]) describes two external flood mechanisms not bounded by the current design basis: Local Intense Precipitation and the Combined Effects Flood Mechanism from riverine, dam failure and wind/wave action. The impact of these flood mechanisms on the10 CFR 50.69 application are discussed below.

Local Intense Precipitation (LIP)

DNPS utilized the Integrated Assessment (Reference [7]) to demonstrate that the site has adequate protection from the LIP event; particularly, water ingress through normally closed exterior doors would not accumulate enough volume to impact any SSCs. A bounding assessment concluded that approximately 9.78 inches of water will accumulate in the Torus basement where there are no SSCs to be impacted. The Torus level sensors are approximately 2 feet off the floor and the Diverse and Flexible Coping Strategies (FLEX) pumps are mounted on 12 inch high plates. The analysis demonstrates physical protection from a flooding event and the ability to deploy FLEX equipment and strategies during the event. The NRC concluded in its Staff Assessment of the Integrated Assessment (Reference [8]) that " available physical margin and reliable flood protection features exist for the LIP flooding mechanism."

The DNPS Integrated Assessment confirmed that the site has adequate protection from the LIP flooding mechanism with no direct external flooding impacts to the plant due to protection from ten normally closed exterior doors. Therefore, the ten doors listed in Table 3-2 below are credited for screening the LIP flood mechanism and thus will be categorized as (High Safety Significant or HSS) during the10 CFR 50.69 categorization process.

Table 3-1: Doors Credited with Screening LIP Scenario Door/Bay No.

Bay 124 Doors 125 - 126 Bay 127 RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 14 of 74

Table 3-1: Doors Credited with Screening LIP Scenario Door 128 Bay 129 Doors 130 - 131 Bay 132 Doors 133 - 139 Bay 140 Door 141

Combined Effects Flood Mechanism

The combined effects flooding mechanism could not be screened from the10CFR 50.69 program for floods that exceed plant grade at 517.5. The probabilistic flood hazard analysis (PFHA) describing the analysis is provided in Reference[9]. The screening evaluation for this mechanism is described in Reference [6].

Floods below plant grade do not have any impacts to key SSCs responsible for maintaining the plant in a safe stable condition during an external flooding event. Once water exceeds plant grade, it is conservatively assumed that all mitigation capabilities are lost and no cooling capabilities are available. This is conservative given the documented strategies in the USFAR to provide alternate source of core cooling during a flood event.

The PFHA was used to create the flood frequency curve shown in Figure A41. The data -

presented in Table 9 of the PFHA report (Reference[9] ) shows an annual exceedance frequency of 2E-5/yr for floods exceeding plant grade. Further evaluation was performed breaking down the combined effects flooding mechanism into two scenarios. One was for combined effects floods producing water surface elevations (WSELs) above plant grade and the second was for WSELs below plant grade. For the below plant grade scenario, there are no impacts to key SSCs responsible for maintaining the plant in a safe stable state during an external flood.

For combined effects floods above plant grade, however, water can enter the plant and impact key SSCs used for flood mitigation. Therefore, DNPS proposes to use an External Flood Safe Shutdown Equipment List (XFSSEL) comprising those key SSCs needed to achieve and maintain safe shutdown of the reactor for external floods higher than 517.5.

These key SSCs would be classified as high safety significant (HSS) during the 10 CFR 50.69 categorization process.

NEI 00-04, Section 5.4 allows the safety significance of SSCs categorized under 10 CFR 50.69 to be determined using either an external hazards PRA (e.g., external flood

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PRA) or the process shown in NEI 00-04 Figure 5-6 for screened external hazards. Since DNPS does not have an external flood PRA, and the external flood hazard was not screened for combined effects external floods above plant grade, use of the NEI 00- 04 Figure 5-6 safety significance process for plants with screened external hazards does not apply.

Therefore, the DNPS 10 CFR 50.69 categorization process will use an alternate approach to identify high safety significant (HSS) SSCs for combined effects flooding above 517.5'.

Specifically, the external flood (XF) safety significance process will use an External Flood Safe Shutdown Equipment List (XFSSEL) comprising those SSCs needed to achieve and maintain safe shutdown of the reactor for external floods higher than 517.5.

The proposed approach is conservative since all SSCs and their associated functions that are needed to achieve and maintain safe shutdown of the reactor are considered to be high safety significant (HSS) regardless of their flood damage susceptibility or frequency of challenge. There is no reliance on operator actions in the determination of whether or not an SSC should be assigned to the XFSSEL. There are no PRA importance measures used in determining safety significance of SSCs related to the XF. As stated in NEI 00- 04, an SSC identified as HSS by a non-PRA method for external events "may not be re-categorized by the IDP." This approach ensures the SSCs that are credited to achieve and maintain the capability for safe shutdown for XF hazards are retained as safety significant.

During categorization of systems, the NEI 00- 04 component to function mapping process will be applied to the safe shutdown function of (1) Decay Heat Removal; (2) Reactivity Control; (3) Inventory Control; (4) Power Availability, and (5) Reactor Pressure Control. The SSCs that fulfill the XF safe shutdown functions for floods greater than 517.5, as well as XF barriers that are credited with protecting equipment that fulfills an XF function, will be identified as candidate high safety significant (HSS). The safety significance process for the unscreened Scenario 2 XF hazard is shown below in Figure 3-1.

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Se le c t S SC

I s t he S SC on t he Do es the SS C supp ort Ca ndida t e L ow XF S SE L? No a n XFS SEL Func t ion? No Sa f e ty Si gnif ica nt

Ye s

Ca ndida t e H igh Sa f e ty Ye s Si gnif ica nt

I d en tif y Sa f e ty Si gnif ica nt Att ribut e s of SS C

Figure 3-1: Safety Significance Process for SSCs for the External Flood Hazard

Plant SSCs on the XFSSEL will be chosen based on DNPS site design engineering input for the external flooding safe shutdown strategy and will use a screening process with criteria to identify SSCs whose failure would have no significant risk impact given an external flooding scenario that exceeds 517.5 or are not otherwise credited for mitigating the effects of the external flooding scenario. The criteria below will be used to screen out as Low Safety Significant (LSS) Scenario 2 SSCs (i.e., Combined Effects External Floods above plant grade):

1. SSCs not powered by emergency onsite AC sources. The rationale for this step is that the external flooding event is assumed to cause a LOOP without credit for offsite power recovery. Therefore, if SSCs do not have emergency power sources, they are screened.

2. SSCs not required to function during or after a loss of offsite AC power event.

Examples for this screening criterion include SSCs needed to generate a reactor trip signal, since station procedure DOA-0010- 04, Revision 59 (Reference [10]) requires that Units 2 and 3 be scrammed with crib house intake canal level being greater than 510.5 ft.

3. SSCs in systems that are assumed unavailable following an external flooding event.

4. SSCs outside a Category I structure not protected against external floods and/or not credited for mitigation of external floods. Unless designed for external flooding, SSCs

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outside Category I structures (either unprotected or in a non-Category I structure) will be assumed to fail during an external flooding event and are not credited in mitigation.

5. SSCs that only perform a passive safety function during an external flooding event and are protected inside Category I structures (e.g., normally closed valves whose external flooding function is to remain closed).

6. SSCs determined by DNPS operations to not be part of the external flooding mitigation strategy for safe shutdown.

The remaining SSCs not screened out will be identified as the risk significant (RISC 1 or 2 SSCs) SSCs on the XFSSEL.

3.2.4.2 Extreme High Winds and Tornadoes

As documented in DR-LAR-008 (Reference [6] ), neither straight winds nor tornado/tornado missiles screen. Since HW/TM do not screen from consideration for 10 CFR 50.69, a HW Safe Shutdown Equipment List (HWSSEL) will be used for 10 CFR 50.69 to identify those high risk-significant SSCs during system categorization.

NEI 00-04, Section 5.4 allows the safety significance of SSCs categorized under 10 CFR 50.69 to be determined using either an external hazards PRA (e.g., high winds PRA) or the process shown in NEI 00-04 Figure 5-6 for screened external hazards. Since DNPS does not have a high winds PRA, and the extreme winds and tornado hazard was not screened, use of the NEI 00- 04 Figure 5-6 safety significance process for plants with screened external hazards does not apply. Therefore, the DNPS 10 CFR 50.69 categorization process will use an alternate approach to identify high safety significant (HSS)

SSCs for the extreme winds and tornado hazard, including tornado missiles (HW/TM).

Specifically, the HW/TM safety significance process for categorization will use a High Wind Safe Shutdown Equipment List (HWSSEL) comprising those SSCs needed to achieve and maintain safe shutdown of the reactor for extreme winds and tornado events.

In response to RIS 2015-06, Tornado Missile Protection (Reference [1 1]), Dresden developed a Tornado Safe Shutdown Equipment List (TSSEL). The purpose of tornado and tornado missile protection is to ensure that the plant can be safely shutdown and cooled down and be maintained in a cold shutdown condition in the event of a tornado causing a loss of offsite power. This requires that tornado and tornado missile protection be provided consistent with the licensing basis for the necessary structures, systems and components, including necessary support equipment to achieve safe shutdown, cool down and maintain cold shutdown without offsite power. The method used to develop the TSSEL is provided in Reference [12].

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The following functions were addressed in developing the TSSEL:

• Reactivity control

• Inventory control

• Pressure control

• Decay heat removal capability

The TSSEL was developed assuming a loss of offsite power; therefore, the DGs are relied upon to provide electric power in support of the safe shutdown strategy. Since the straight and tornado wind components of the HW/TM risk are driven mainly by wind pressure and nominal/random failures of the three DGs, the list of equipment in the TSSEL can also be used for the HWSSEL. For purposes of categorization under 10 CFR 50.69, the TSSEL developed by Dresden (Reference [12] ) will be referred to as the high wind safe shutdown equipment list (HWSSEL) to indicate that the list includes equipment for mitigation of all high wind events (straight winds as well as tornados).

The proposed approach to use the HWSSEL during system categorization is conservative since all SSCs and their associated functions that are needed to achieve and maintain safe shutdown of the reactor during HW/TM events are considered to be high safety significant (HSS) regardless of their HW/TM damage susceptibility or frequency of challenge. There was no reliance on operator actions in the determination of whether or not an SSC should be assigned to the HWSSEL. There are no PRA importance measures used in determining safety significance of SSCs related to HW/TM hazards. As stated in NEI 00- 04, an SSC identified as HSS by a non-PRA method for external events "may not be re-categorized by the IDP." This approach ensures the SSCs that are credited to achieve and maintain the capability for safe shutdown for HW/TM hazards are retained as safety significant. During categorization, the safety significance process for the HW/TM events is shown in Figure 3-2.

RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 19 of 74

Se le c t S SC

I s t he S SC on t he Do es the SS C supp ort Ca ndida t e L ow HW S SE L? No a HWSS EL Func ti on? No Sa f e ty Si gnif ica nt

Ye s

Ca ndida t e H igh Sa f e ty Ye s Si gnif ica nt

I d en tif y Sa f e ty Si gnif ica nt Att ribut e s of SS C

Figure 3-2: Safety Significance Process for SSCs for the Extreme Winds and Tornado Hazard

As part of the safety significance determination during system categorization, a function-level evaluation for SSCs on the HWSSEL will be performed using the flowchart in Figure 3-2.

NEI 00-04 requires that all functions for the system being categorized be identified. SSCs required to perform the HWSSEL functions will be risk significant SSCs per the guidance in NEI 00-04. During categorization of systems, NEI 00- 04 component to function mapping process will be applied to the safe shutdown functions of (1) Reactivity Control; (2) Inventory Control; (3) Pressure Control; and (4) Decay Heat Removal capability. The SSCs that fulfill the HW/TM safe shutdown functions, as well as any high wind or missile barriers that are credited with protecting equipment that fulfills a HWSSEL function, will be identified as candidate high safety significant (HSS) for the system being categorized regardless of their tornado damage susceptibility or frequency of challenge. This approach ensures the SSCs that are credited to achieve and maintain the capability for safe shutdown are retained as safety-significant.

3.2.4.3 All Other External Hazards

All other external hazards, except for extreme winds and tornadoes and external flooding, were screened for applicability to DNPS per a plant-specific evaluation in accordance with GL 88-20 (Reference [13]) and updated to use the criteria in ASME PRA Standard RA-Sa-2009. Attachment 4 provides a summary of the external hazards screening results.

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Attachment 5 provides a summary of the progressive screening approach for external hazards.

3.2.5 Low Power and Shutdown

Consistent with NEI 00- 04, the DNPS categorization process will use the shutdown safety management plan described in NUMARC 91- 06 (Reference [7]) for evaluation of safety significance related to low power and shutdown conditions. The overall process for addressing shutdown risk is illustrated in Figure 5-7 of NEI 00-04.

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

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

3.2.6 PRA Maintenance and Updates

The CEG risk management process ensures that the PRA models used in this application continue to reflect the as-built and as-operated plant for DNPS. The process delineates the responsibilities and guidelines for updating the PRA models and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, and industry operational experience) for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files.

The process will assess the impact of these changes on the plant PRA model in a timely manner but no longer than once every two refueling outages. If there is a significant impact on the PRA model, the SSC categorization will be re-evaluated.

In addition, CEG will implement a process that addresses the requirements in NEI 00- 04, Section 11, "Program Documentation and Change Control." The process will review the results of periodic and interim updates of the plant PRA that may affect the results of the categorization process. If the results are affected, adjustments will be made as necessary to the categorization or treatment processes to maintain the validity of the processes. In addition, any PRA model upgrades will be peer reviewed prior to implementing those changes in the PRA model used for categorization.

3.2.7 PRA Uncertainty Evaluations

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

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

In the overall risk sensitivity studies, CEG will utilize a factor of three ( 3) to increase the unavailability or unreliability of LSS components consistent with that approved for Vogtle in Reference [5]. Consistent with the NEI 00-04 guidance, CEG will perform both an initial sensitivity study and a cumulative sensitivity study. The initial sensitivity study applies to the system that is being categorized. In the cumulative sensitivity study, the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in all identified PRA models for all systems that have been categorized are increased by a factor of 3. This sensitivity study together with the periodic review process assures that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study.

The detailed process of identifying, characterizing, and qualitative screening of model uncertainties is found in Section 7 of NUREG 1855 and Section 3 of EPRI TR 1016737 (Reference [14]). The process in these References was mostly developed to evaluate the uncertainties associated with the internal events PRA model; however, the approach can be applied to other types of hazard groups.

Each PRA element notebook was reviewed for assumptions and sources of uncertainties. The characterization of assumptions and sources of uncertainties are based on whether the assumption and/or source of uncertainty is key to the10 CFR 50.69 application in accordance with RG 1.200, Revision 2. Key DNPS PRA model specific assumptions and sources of uncertainty for this application were identified and dispositioned in Attachment 6. The conclusion of this review is that no additional sensitivity analyses are required to address DNPS PRA model specific assumptions or sources of uncertainty.

3.3 PRA Review Process Results (10 CFR 50.69(b)(2)(iii))

The PRA models described in Section 3.2 have been assessed against RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2 (Reference [15] ), consistent with NRC Regulatory Information Summary (RIS) 2007-06 (Reference [16] ). Although Dresden will transition to RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, (Reference 59) going forward, the NRC finds it remains acceptable to refer to RG 1.200, Revision 2, to demonstrate technical acceptability of the DNPS PRA models.

Finding and Observation (F&O) closure reviews were conducted on the PRA models discussed in this section. Closed findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, "Close-out of Facts and Observations" (F&Os) (Reference [17] ) as accepted by NRC in the letter dated May 3, 2017 (Reference [18]).

The results of this review have been documented a re available for NRC audit.

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

A full-scope peer review of the DNPS Units 2 and 3 FPIE PRA model was conducted in 2016 (Reference [19]) with the Internal Flooding (IF) PRA model being peer reviewed in 2009 (Reference [20]). The reviews were performed using the NEI 05- 04 process [21], the ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009) (Reference [22]), and Regulatory Guide 1.200, Revision 2 (Reference [15]). Subsequent F&O closure reviews were conducted in 2017 (Reference [23]), in 2021(Reference [24]), and in 2023 (Reference [25]). In May 2023 (Reference [26]), a Focused-Scope Peer Review (FSPR) was performed on a new method with a follow-up F&O closure review conducted in June 2023 (Reference [27]). All Findings were closed with all applicable Supporting Requirements (SRs) meeting Capability Category (CC) II or greater of the ASME/ANS PRA Standard.

In conclusion, for the FPIE PRA model, all Finding-Level F&Os are closed and all applicable supporting requirements from the ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009) [12] are met with a capability category II or greater. CEG performed a self -assessment as to whether the resolution of each Finding constituted maintenance or upgrade of the PRA, as defined in the ASME/ANS PRA standard. The F&O Independent Assessment Team concurred with the CEG assessment that there were no PRA upgrades associated with the resolution of findings.

Given there are no partially resolved or open findings, the DNPS FPIE PRA is of adequate technical capability to support the 10 CFR 50.69 program.

Fire PRA (FPRA) Model

A FSPR FPRA report for DNPS was issued in December 2014 (Reference [28]) with a follow-on FSPR in 2016 (Reference [19]). To resolve the open Findings, the FPRA underwent three F&O closure reviews with two FSPRs to address new methods. The first F&O closure review was conducted in 2017 (Reference [29]), and a FSPR in 2021 (Reference [26]) concurrently with an F&O closure review (Reference [23]). In May 2023, a third F&O closure review was conducted (Reference [25]), and again later in May 2023 (Reference [26]) a FSPR of the high level requirement (HLR) FQ (Fire Quantification) was performed. A final F&O closure review was conducted in June 2023 (Reference [27]) closing all remaining open Findings. All of the reviews were performed using the NEI 07-12 process (Reference [30]), the ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009) (Reference [22]), and Regulatory Guide 1.200, Revision 2 (Reference [15]).

In conclusion, for the FPRA, all Finding-Level F&Os are closed and all applicable supporting requirements from the ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009) (Reference [22])

are met with a CC II or greater. CEG performed a self-assessment as to whether the resolution of each Finding constituted maintenance or upgrade of the PRA, as defined in the ASME/ANS

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PRA standard. Ultimately, the F&O Independent Assessment Team concurred with the CEG assessment that there were no PRA upgrades associated with the resolution of findings.

Given there are no partially resolved or open technical findings that may impact the 10 CFR 50.69 program, and the only remaining Finding is related to documentation, the DNPS FPRA is of acceptable technical capability to support the 10 CFR 50.69 program.

Seismic PRA (SPRA) Model

The SPRA Peer Review (Reference [31]) was performed in January 2019 using the NEI 12-13 process (Reference [32]) and the ASME/ANS PRA Standard (ASME/ANS RA-Sb-2013)

(Reference [33]). The 2019 DNPS SPRA Peer Review was a full-scope review of all the technical elements of the DNPS at-power SPRA against all technical elements in Section 5 of the PRA Standard. Both Addenda A and B of the PRA Standard were considered as well as EPRI Technical Report 1025287, "Seismic Evaluation Guidance Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near Term Task Force Recommendation 2.1: Seismic" (Reference [34]). A subsequent F&O closure review was performed in June 2022 (Reference [35]). All Finding-level F&Os were closed with all applicable SRs meeting CC II or greater from the ASME/ANS PRA Standard.

In conclusion, for the SPRA, all Finding-level F&Os are closed and all applicable supporting requirements from the ASME/ANS PRA Standard (ASME/ANS RA-Sb-2013) are met with a capability category II or greater. CEG performed a self-assessment as to whether the resolution of each Finding constituted maintenance or upgrade of the PRA, as defined in the ASME/ANS PRA standard. Ultimately, the F&O Independent Assessment Team concurred with the CEG assessment that there were no PRA upgrades associated with the resolution of findings.

Given there are no partially resolved or open findings, the DNPS Seismic PRA is of adequate technical capability to support the 10 CFR 50.69 program.

Conclusion

The above discussion demonstrates that the DNPS PRA models are of sufficient quality and level of detail to support the categorization process. Additionally, it is concluded that each model has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC as required by 10 CFR 50.69(c)(1)(i).

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

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

§50.69(b)(2)(iv). Sensitivity studies described in NEI 00- 04 Section 8 will be used to confirm that the categorization process results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF). The failure rates for equipment and initiating event frequencies used in the PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, and human errors). Subsequent performance monitoring and PRA updates required by the rule will RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 24 of 74

continue to capture this data and provide timely insights into the need to account for any important new degradation mechanisms.

3.5 Feedback and Adjustment Process

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

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

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

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

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

• A review of the importance measures used for screening in the categorization process.

• An update of the risk sensitivity study performed for the categorization.

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

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4 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

The following NRC requirements and guidance documents are applicable to the proposed change.

• 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors"

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

Revision 1, dated May 2006.

• Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, dated May 2011.

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

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

4.2 No Significant Hazards Consideration Analysis

In accordance with 10 CFR 50.90, " Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests an amendment to Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS), Units 2 and 3, respectively. Specifically, CEG proposes to modify the DNPS licensing basis to allow for the voluntary implementation of the provisions of 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements either will not be changed or will be enhanced. This allows improved focus on equipment that has high safety significance resulting in improved plant safety.

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According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

CEG has evaluated the proposed cha nge for DNPS, Units 2 and 3 by using the criteria in 10 CFR 50.92 and has determined that the change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

1. Does the proposed ch ange involve a si gnificant increase in the probability or consequences of an accident previously eva luated?

Response: No.

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

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed.

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

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

Response: No.

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

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

Based on the above, CEG concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.3 Conclusions

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

5 ENVIRONMENTAL CONSIDERATION

The proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

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6 REFERENCES

[1] Nuclear Energy Institute (NEI) 00- 04, "10 CFR 50.69 SSC Categorization Guideline,"

Revision 0, July 2005.

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

[3] NRC letter to Southern Nuclear Operating Company, Inc., "Vogtle Electric Generating Plant, Units 1 and 2 -Issuance of Amendments Re: Use of 10 CFR 50.69 (TAC NOS.

ME9472 AND ME9473)," (ADAMS Accession No. ML14237A034), dated December 17, 2014.

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

December 1991.

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

(TAC NO. MD5250), (ADAMS Accession No. ML090930246), dated April 22, 2009.

[6] DR-LAR-008, "External Hazards Assessment for Dresden Nuclear Power Station,"

Revision 2, May 2024.

[7] Exelon Letter to US NRC, "Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Integrated Assessment Submittal," dated September 8, 2017 (ADAMS Accession No. ML17251A365).

[8] NRC Letter to Exelon Generation Company, LLC, "Staff Assessment of Flood Hazard Integrated Assessment (CAC NOS. MG0221 AND MG0222; EPID L-2017-JLD-0050),"

(ADAMS Accession No. ML18138A385), dated March 6, 2019.

[9] Constellation Energy - Prepared by Aterra Solutions, Probabilistic Flood Hazard Assessment Report for the Illinois River - Dresden Nuclear Generating Station, April 8, 2023.

[10] Constellation Generation, Dresden DOA 0010- 04, FLOODS, Rev. 59, 12/07/2021.

[11] NRC Regulatory Issue Summary 2015- 06, "Tornado Missile Protection," June 10, 2015.

[12] DRE 19-0026, "Evaluation of Dresden's Tornado Missile Protection Design for Compliance with the Licensing Requirements," December 19, 2019.

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

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

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

[16] NRC Regulatory Issue Summary 2007- 06, "Regulatory Guide 1.200 Implementation,"

(ADAMS Accession No. ML070650428). dated March 22, 2007.

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

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

[19] Dresden Generating Station Units 2 & 3 PRA Peer Review Report Using ASME PRA Standard Requirements. BWR Owners Group, January 2017.

[20] Dresden Generating Station Internal Flood PRA Peer Review Report Using ASME PRA Standard Requirements, BWR Owners Group, August 2009.

[21] NEI 05-04, "Process for Performing PRA Peer Reviews Using the ASME PRA Standard (Internal Events)," (ADAMS Accession No. ML083430462), Revision 2, dated November 2008.

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

[23] Dresden Nuclear Power Plant, "Finding Level F&O Technical Review," Report: 032362-RPT-001, Revision 0, June 2018.

[24] Dresden Units 2 and 3, "PRA Finding Level Fact and Observation Independent Assessment," Report No. 32466-RPT-004, Revision 0, August 3, 2021.

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[25] Report DR-MISC-31, "Dresden Unit 2, PRA Finding Level Fact and Observation Independent Assessment," May 8, 2023.

[26] Report DR-MISC-32, "Dresden Unit 2 PRA Focused-Scope Peer Review," June 26, 2023.

[27] Report DR-MISC-45, "Dresden Unit 2 PRA Finding Level Fact and Observation Independent Assessment," July 2023.

[28] Dresden Generating Station Unit 2 Fire PRA Peer Review Report Using ASME PRA Standard Requirements. BWR Owners Group, December 2014.

[29] Exelon Risk Management, "2017 Risk Management (Dresden 2 and 3) Finding Level F&O Independent Technical Review," September 2017.

[30] NEI 07-12, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines,"

Revision 1, (ADAMS Accession No. ML102230070), dated June 2010.

[31] Dresden Generating Station, "Seismic PRA Peer Review Report Using ASME/ANS PRA Standard Requirements," Revision 0, March 2019.

[32] NEI 12-13, "External Hazards PRA Peer Review Process Guidelines," August 2012.

[33] ASME/ANS RA-Sb-2013, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addenda B, 2013, American Society of Mechanical Engineers, New York, September 30, 2013.

[34] EPRI 1025287, "Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," Electric Power Research Institute, February 2013, ADAMS Accession # ML12333A170.

[35] Dresden Generating Station (Units 2&3) Report #: 32485-RPT-134-01, "Seismic PRA Fact And Observation Independent Assessment," August 10, 2022.

[36] Letter from J. M. Heffley (ComEd) to U. S. NRC, "Final Report - [Dresden] Individual Plant Examination of External Events (IPEEE) Generic Letter 88-20, Supplement 4," December 30,1997.

[37] Dresden Updated Final Safety Analysis Report, Revision 14, June 2021.

[38] NRC Standard Review Plan, NUREG-0800, Chapter 3.5.1.6, Revision 4, March 2010.

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

RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 31 of 74

[40] Procedure OP-DR-108-111-1004, "Cold Weather Strategy," Revision 1.

[41] NUREG-0823, "Integrated Plant Safety Assessment, Systematic Evaluation Program Dresden Nuclear Power Station, Unit 2," February 1983.

[42] U.S. NRC letter to Commonwealth Edison Company, "'SEP Topic ll-1.C, Potential Hazards Due to Nearby Transportation, Institutional, Industrial and Military Facilities - Dresden Unit 2," Docket No. 50- 237, dated August 20, 1982.

[43] Regulatory Guide 1.91, "Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants," February 1978 (ADAMS Accession No. ML003740286).

[44] Calculation No. DRE22-0002, "Dresden Buried Pipe Blast and Gas Leak Analysis,"

September 7, 2023, Revision 1.

[45] Control Room Habitability Study Update for Dresden Units 2 and 3, Enercon 2015.

[46] Exelon Letter to NRC, "Dresden Units 1 and 2 - Response to Request for Additional Information Regarding Fukushima Lessons Learned - Flooding Hazard Reevaluation Report," (ADAMS Accession No. ML15092A821), dated May 19, 2014.

[47] NRC letter to Exelon, "Dresden Nuclear Power Station, Units 2 and 3 - Staff Assessment of Response to 10 CFR 50.54(f) Information Request - Flood-Causing Mechanism Reevaluation (TAC NOS. MF1795 AND MF1796)," (ADAMS Accession No. ML15072A007), dated March 31, 2015.

[48] Regulatory Guide 1.78, "Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release," (ADAMS Accession No. ML003740298), June 1974.

[49] Control Room Habitability Study Update for Dresden Units 2 and 3, Commonwealth Edison Company, Bechtel Power Co., December 1981.

[50] Regulatory Guide 1.115, "Protection Against Turbine Missiles," U.S. Nuclear Regulatory Commission, Revision 2, (ADAMS Accession No. ML101650675), January 2012.

[51] MPR Letter 0958- 0147-LTR-001, "Impact of Increasing Test and Maintenance Intervals of Turbine Overspeed Protection System Components at Dresden," March 2019, Revision 0.

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

RS-24-056, DNPS Application to Adopt 10 CFR 50.69 Enclosure Page 32 of 74

[53] Electric Power Research Institute (EPRI) Technical Report TR-1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," December 2008.

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

[55] DR-LAR-009, "Assessment of Key Assumptions and Sources of Uncertainty for DNPS Nuclear Power Station," Revision 1, May 2024.

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

[57] Faulted Systems Recovery Experience, NSAC-161, March 25, 1991.

[58] EPRI 3002000709, "Seismic Probabilistic Risk Assessment Implementation Guide,"

Electric Power Research Institute, December 2013.

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

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Attachment 1: List of Categorization Prerequisites

CEG has established procedure(s) for use within the fleet for the use of the categorization process on a plant system. These CEG fleet procedures will be implemented at DNP S prior to the use of the categorization process at D NPS. The fleet procedures to be implemented at DNPS contain the elements/steps listed below.

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

• Qualitative assessment of system functions. System functions are qualitatively categorized as preliminary High Safety Significant (HSS) or Low Safety Significant (LSS) based on the seven criteria in Section 9 of NEI 00- 04 (see Section 3.2 of the enclosure). Any component supporti ng an HSS function is categorized as preliminary HSS. Components supporting an LSS function are categorized as preliminary LSS.

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

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

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

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

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

• Documentation requirements per Section 3.1.1 of the enclosure.

• External Flooding Safe Shutdown Equipment List (XFSSEL) of SSCs needed to achieve and maintain safe shutdown of the reactor as described in Section 3.2.4.1 of the enclosure.

• High Winds Safe Shutdown Equipment List (HWSSEL) of SSCs needed to achieve and maintain safe shutdown of the reactor as described in Section 3.2.4.2 of the enclosure.

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Description of PRA Models Used in Categorization

Units Model Baseline CDF Baseline LERF Comments

Full Power Internal Events &Internal Flooding PRA Model

DR221A (Unit 2) 3.1E-6/yr (Unit 2) 2.2E-7/yr (Unit 2) 2021 FPIE 2 and 3 Model of DR321A (Unit 3) 3.1E-6/yr (Unit 3) 2.3E-7/yr (Unit 3) Record

Fire Model

DR223A-F 2 and 3 (Unit 2) 3.3E-5/yr (Unit 2) 3.6E-6/yr (Unit 2) 2023 Fire PRA

DR323A-F 3.2E-5/yr (Unit 3) 4.7E-6/yr (Unit 3)

(Unit 3)

Seismic Model

DR217BS0 Model (Unit 2) 5.8E-6/yr (Unit 2) 2.9E-6/yr (Unit 2) 2019 Seismic 2 and 3 PRA DR317BS0 5.8E-6/yr (Unit 3) 2.8E-6/yr (Unit 3)

Model (Unit 3)

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

There are no Partially Resolved or Open Peer Review Findings or Self-Assessment Open Items for the DNPS Internal Events and Internal Flooding, Fire, and Seismic PRA models.

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Attachment 4: External Hazards Screening

Screening Result

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

Per IPEEE Report Section 5.2.3.7 (Reference [36]), A detailed NRC evaluation included a conservative calculation of aircraft impact probabilities projected into the early 1990s. The calculation gave a total projected probability of aircraft crashing into DNPS Unit 2 of 4.16E-7 per year and concluded aircraft operations in the vicinity of DNPS do not pose an undue risk to the plant.

PS2 An updated aircraft impact evaluation is documented in UFSAR Section 3.5.6 Aircraft Impact Y (Reference [37]). There are four federal airways that pass within 10 miles of the PS4 station (Fromm, Morris, Joliet, and Adelmann airports).

As such, in accordance with SRP Section 3.5.1.6 (Reference [38]), the probability per year of a commercial aircraft traveling along one of these airways and crashing into the station (PFA) was calculated using conservative assumptions to be 4.2E-07/yr as displayed in Table 3.5-6. (PS2, PS4)

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

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

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

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Screening Result

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

Per UFSAR Section 9.2.1.4 (Reference [37]), piping and heat exchanger intrusion by Corbicula (such as Asiatic Clams) has been identified as a potential hazard to the DNPS safety-related service water systems. This problem has been studied since the 1970s. DNPS has implemented a program to trend infestation characteristics. This information is used to ensure flow blockage will not occur in safety-related systems using river water. The program includes:

A. Periodic inspection and cleaning of the intake bays;

B. Periodic biocide injection into the intake bay or service water distribution header; Biological Event Y C1 C. Periodic flushing of infrequently used or stagnant lines in safety-related

service water systems;

D. Annual water and substrate sampling;

E. Periodic testing, inspection, and cleaning of safety-related heat exchangers; and

F. Periodic inspection of high-and low-flow service water piping for corrosion, erosion, silting, and biofouling.

Based on this review, the biological events hazard can be considered negligible.

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Screening Result

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

Coastal erosion is a slowly developing event and could be mitigated or adequately responded to (C5).

Also, per UFSAR Section 2.5.5 (Reference [37]), the only slopes at DNPS considered critical with regard to slope stability are those of the intake canal from the river to the crib house and of the discharge canal from the plant to the river.

A "sliding wedge" slope stability analysis under safe shutdown earthquake (0.2 g horizontal acceleration) loading indicates a minimum factor of safety of 1.5 C1 against failure of the intake or discharge canals. However, even if the Coastal Erosion Y overburdened slopes failed and this material moved into the canal, there would C5 still be an ample water supply in the intake canal for use in station operation.

The rock into which the canals are cut is sound and capable of maintaining a stable vertical cut under earthquakes or other events.

The rock, locally referred to as the Pottsville sandstone, is composed predominantly of cemented sub-angular fine-to-medium grains of quartz containing varying amounts of mica. No evidence of faulting exists in the sandstone at the site, but there are occasional vertical joints. Laboratory compressive strength tests on the sandstone indicate strengths of the rock in excess of 3000 psi.

Therefore, slope stability is not a safety concern for DNPS. (C1)

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Screening Result

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

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

Drought is a slowly developing hazard allowing time for orderly plant reductions, Drought Y C5 including shutdowns.

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

See Section 3.2.4.1 for the discussion of External Flooding.

The risk impact of the Local Intense Precipitation flood mechanism was screened by crediting ten normally closed exterior doors. The ten doors shown in Table 3-1 will be categorized as High Safety Significant during External Flooding N N/A the10 CFR 50.69 categorization process.

The Combined Effects Flood Mechanism does not screen when flood waters reach 517.5'. Therefore, DNPS will use an External Flood Safe Shutdown Equipment List (XFSSEL) comprising those key SSCs needed to achieve and maintain safe shutdown of the reactor for external floods higher than 517.5.Based on this review, the risk from e xternal flood hazard can be considered negligible and screened from further evaluation.

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Screening Result

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

Neither straight winds nor tornado/tornado missiles screen, as discussed in Section 3.2.4.2 of this report.

Extreme Wind or Tornado N N/A Since HW/TM do not screen from consideration for 10 CFR 50.69, a HW Safe Shutdown Equipment List (HWSSEL) will be developed for 10 CFR 50.69 to identify those high risk-significant SSCs during system categorization.

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

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

Per the IPEEE (Reference [36]), some wooded and grassy areas are present within the Exclusion Area, mainly along the Kankakee and Illinois Rivers.

Nevertheless, the site landscaping and lack of heavy forestation in and near the Protected Area are judged adequate to prevent such fires from posing a threat C3 to equipment with the Protected Area. Similarly, vegetation is limited in/near the Forest or Range high voltage switchyards and transmission line corridors that serve DNPS. (C3)

Fire Y C4 In addition, forest fires originating from outside the plant boundary may cause a loss of offsite power event, which is addressed for grid-related LOOP scenarios in the FPIE PRA model for DNPS (C4).

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

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Screening Result

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

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

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

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

Hail Y C4 Flooding impacts are covered under External Flooding/Intense Precipitation.

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

Per the IPEEE Table 5-1, (Reference [36]), the principal effects of such events would be to cause a loss of off -site power (C4). These effects would take place slowly, allowing time for orderly plant reactions including shutdowns. (C5)

C1 Technical Specification Surveillance Requirement (TSSR) 3.7.3.2 is to verify the High Summer C4 average water temperature of the ultimate heat sink (UHS) is 95°F.

Temperature Y Verification of the UHS temperature ensures that the heat removal capabilities of C5 the containment cooling service water (CCSW) and diesel generator cooling water (DGCW) systems are within the assumptions of the DBA analysis. (C1)

Plant trips due to this hazard are covered in the definition of another event in the PRA model (e.g., transients, loss of condenser) (C4).

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Screening Result

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

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

Per UFSAR Table 2.1-2 (Reference [37]), the site is located adjacent to the High Tide, Lake Illinois, Kankakee, and Des Plaines rivers. Thus, the high tide hazard does not Level, or River Y C3 apply.

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

Hurricanes are extreme tropical storms that originate offshore and as such do Hurricane Y C3 not reach DNPS due to the mid-western location of the site.

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

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Screening Result

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

The principal effect of ice cover events would be to cause a loss of off-site power event, which is addressed for weather-related LOOP scenarios in the FPIE PRA model for DNPS (C4).

Also, UFSAR Section 2.4.7 (Reference [37]) discusses ice effects. An 8 -foot C1 diameter deicing line connects the discharge canal headworks and the crib Ice Cover Y house forebay. A slide gate valve is used to isolate the deicing line when not in C4 use (C1).

UFSAR Section 2.5.6.1 states that the DNPS Lock and Dam were designed to withstand the large forces due to the mass movement of ice flows from the Des Plaines and Kankakee rivers. (C1)

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

IPEEE Sections 5.2.3.5 and 5.2.3.6 (Reference [36]) discuss industrial facilities and military facilities; respectively, and states that the initiator for these hazards screens based on Systematic Evaluation Program (SEP) reviews. (Note: The Industrial or C1 NRC initiated the SEP i n 1977 to review the designs of older operating nuclear Military Facility Y power plants.)

Accident C3 UFSAR Section 2.2.2.2 (Reference [37]) discusses six industries within five miles of the plant. None are licensed to store or use solid explosives.

Furthermore, separation distances are such that no hazard exists for plant

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Screening Result

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

safety except for the Mobil Chemical site. For the Mobil site, the standoff distance for one petroleum product slightly exceeds the available separation distances; however, the stored product has a high flash point, which eventually rules out the possibility of an explosion hazard from this source at a distance of 4.5 miles. (C3)

UFSAR Section 2.2.2.1.1 states that the Joliet Army Ammunition Plant, whose nearest boundary is approximately 4 miles east of DNPS, is used for storage of explosive materials from other installations, transported by way of the Atchison, Topeka, and Santa Fe Railroad (AT&SF). No explosive materials are stored within one mile of the eastern property line of the ammunition plant. Thus, the distance from DNPS to the storage area of the ammunition plant exceeds five miles. At this distance, an accidental explosion at the Joliet Army Ammunition Plant will not affect DNPS. (C3)

See also Toxic Gas.

Based on this review, the industrial or military facility accident hazard can be considered negligible.

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

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

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Screening Result

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

Plant site is located on level terrain and is not subject to landslides. Additionally, Landslide Y C3 the mid-western location of DNPS precludes the possibility of a landslide.

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

Lightning strikes are not uncommon in nuclear plant experience. They can result in losses of off-site power or surges in instrumentation output if grounding is not fully effective. The latter events often lead to reactor trips. Both events Lightning Y C4 are incorporated into the DNPS internal events model through the incorporation of generic and plant specific data.

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

Per UFSAR Section 9.2.5.2 (Reference [37]), the Kankakee River is the normal source of emergency cooling water for DNPS. In the event of a loss of this water source, there is a limited supply of water trapped, by design, in the intake canal, discharge canal, and cooling lake which would be used as a heat sink for C1 the long-term removal of decay heat from the reactors. Due to the topography Low Lake Level or of the circulating water canals and piping, approximately 9 million gallons of river River Stage Y C5 water are trapped within the canals, not including water in the cooling lake. ( C1)

UFSAR Section 9.2.5.3.1 discusses dam failure during normal operation. With offsite power available Units 2 and 3 could be safely shutdown using water from the UHS, on-site tanks, and circulating water piping providing 3.7 days of coping time. In addition, station procedure DOA -0010- 01, "DNPS Lock and Dam

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Screening Result

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

Failure," provides guidance for restoration of the UHS given a postulated dam failure, including restoration during loss of offsite power events. (C1, C5)

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

Per UFSAR Section 1.2.1 (Reference [37]), the design of components important to safety of the units and the station includes allowances for environmental phenomena (C1).

In addition, procedure OP-DR-108-111-1004, Cold Weather Strategy C1 (Reference [40]), contains guidance for inspections of various plant locations Low Winter and equipment during cold weather conditions (C5).

Temperature Y C4 C5 In addition, plant trips due to this hazard are covered in the definition of another event in the PRA model (e.g., transients, loss of condenser) (C4).

See also Ice Cover.

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

Meteorite or Per the IPEEE (Reference [36]), the frequency of a meteor or satellite strike is Satellite Impact Y PS4 judged to be so low as to make the risk impact from such events insignificant.

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Screening Result

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

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

IPEEE Section 5.2.3.3 discusses pipeline accidents (Reference [36]). This initiator falls under SEP topic 11-1.C, "Potential Hazards or Changes in Potential Hazards Due to Transportation, Institutional, Industrial, and Military Facilities."

NUREG-0823 (Reference [41]) reported that DNPS 2 "meets current criteria or was acceptable on another defined basis" for this SEP topic. The detailed NRC evaluation of this topic (Reference [42]) lists pipelines within five ( 5) miles of the plant. The IPEEE states that the NRC concluded that DNPS 2 (and DNPS 3 due to proximity to DNPS 2) is adequately protected from potential pipeline C1 ruptures. (C1, C3)

Pipeline Accident Y C3 An updated pipeline accident analysis discussion is provided in UFSAR Section PS4 2.2.2.3 (Reference [37]) for pipelines in the vicinity of the plant. The first seven pipelines listed in Table 2.2-3 pose the greatest potential hazard to the plant.

The other pipelines do not pose significant hazards to the plant because their diameters are smaller and they are more than two ( 2) miles from the plant.

Because of the proximity of the first seven pipelines to the DNPS safety-related structures, it was not possible to conclude that the peak overpressure would not exceed 1 psi (pressure below which no significant damage would be expected).

Therefore, the probability of exposure to pressure in excess of 1 psi was estimated for the five scenarios of concern discussed in the UFSAR Section

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Screening Result

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

2.2.2.3 using methodology described in Regulatory Guide 1.91 (Reference [43]). Four of the five scenarios were screened out due to all safety-related structures protected by distance (C3).

The fifth scenario involved an on-site natural gas line leak and explosion.

Analysis of gas release, dispersion, and accumulation showed the overall estimate of the frequency of an explosion that damages safety-related SSCs to be 2.72 x 10-8 per year. (PS4)

During the evaluation of this hazard, it was noted that a new natural gas pipeline is located within 5-miles from DNPS. This is a new pipeline, built by Alliance Pipeline, is not discussed in the current UFSAR (Reference [37]). An analysis was performed to evaluate the impact of a pipe rupture and gas leak on DNPS (Reference [44]).

The analysis concluded that the radius of a jet fire will not reach the site, which is approximately 3,300 feet from the site of the closest potential location of a pipeline break. The blast overpressure at the nearest plant structure is below 1 psi and would therefore have no effect on plant safety systems. (C3)

Other potential threats such as the flammable area of the vapor cloud have been shown to have acceptable consequences. The Control Room methane concentration following the postulated pipeline break is not a concern because

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Screening Result

External Hazard Screened? Screening (Y/N) Criterion Comment (Note a) the outdoor concentration is well below the PAC-1 limit1. Because the outdoor methane concentration is well below the PAC-1 limit, no protective actions by the Control Room operators are required.

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

See Section 3.2.4.1 for the discussion of External Flooding.

The risk impact of the Local Intense Precipitation flood mechanism was Precipitation, screened by crediting ten normally closed exterior doors. The ten doors shown Intense Y C1 in Table 3-1 will be categorized as High Safety Significant during the10 CFR 50.69 categorization process.

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

IPEEE Section 5.2.3.4, (Reference [36]), discusses results of a survey for Release of C1 potentially toxic chemicals stored or transported onsite or within a five ( 5) mile Chemicals in Y radius that refers to the then current version (~1997 timeframe) of the UFSAR.

Onsite Storage PS2 The UFSAR includes the conclusion that the estimated probabilities of control room uninhabitability due to release of ammonia or ethylene oxide (the toxic

1 The PAC-1 limit (65,000 ppm) is the airborne concentration above which it is predicted that the general population, including susceptible individuals, when exposed for more than one hour, could experience notable discomfort, irritation, or certain asymptomatic, non-sensory effects.

However, these effects are not disabling and are transient and reversible upon cessation of exposure.

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Screening Result

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

gases identified as concerns) are an order of magnitude below the Standard Review Plan Section 2.2.3 criterion for realistic estimates. (PS2)

The current UFSAR (Reference [37]) discusses onsite toxic chemicals and Table 2.2-7 lists potentially toxic chemicals stored within the DNPS site boundary. UFSAR Section 6.4.4.2.2 states that the onsite chemicals listed in Table 2.2-7 were analyzed and screened based on a 2015 study (Reference [45]) to determine if the release of any of those chemicals could pose a threat to control room habitability. The screening shows that none of the chemicals stored onsite pose a threat. (C1)

See also Toxic Gas.

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

Per the FHRR, (Reference [46]), DNPS reported that channel migration is not considered to be a potential contributor to flooding at DNPS. A review of C1 historical data and site information indicates that the Kankakee River and Des River Diversion Y C3 Plaines River has not exhibited a tendency to meander towards the DNPS ( C3, C5). Channel diversion impacts at DNPS are not anticipated to occur as a result C5 of landslide because of the low landslide potential in the area as well as the relatively flat surrounding floodplain areas. ( C1)

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Screening Result

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

In the NRC staffs assessment report of the FHRR (Reference [47]), the NRC confirmed that the reevaluated hazard from channel migrations or diversions is not bounded by the current design-basis flood hazard. However, the staff also confirmed the licensee's conclusion that the flood hazard from channel migrations or diversions alone would not inundate the site and that this hazard did not need to be included within the scope of the Integrated Assessment.

See also External Flooding.

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

Per the IPEEE (Reference [36]), the mid-western location of DNPS prevents sandstorms. More common wind-borne dirt can occur but poses no significant Sandstorm Y C3 risk given the robust structures and protective features of the plant.

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

Per UFSAR Section 2.4.5 (Reference [37]), flooding due to seiches is not applicable to DNPS.

Seiche Y C3 The NRC accepted this in its review of the DNPS Flood Hazard Reevaluation Report (Reference [47]) confirming that this flood-causing mechanism was considered not plausible since DNPS is an inland location and does not connect directly with any bodies of water capable of producing seiche.

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Screening Result

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

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

Seismic Activity N/A N/A The DNPS Seismic PRA model addresses risk from seismic events.

This hazard is slow to develop and can be identified via monitoring and managed via normal plant processes ( C5).

C4 Snow Y Potential flooding impacts are accounted for under external flooding screening C5 (C4).

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

Per UFSAR Section 2.5.4 (Reference [37]), examination of cores from borings at the site and excavation for the construction of Units 1 and 2 show that all footings for major structures have a foundation of sound rock which eliminates Soil Shrink-Swell the potential problems of soil consolidation and differential settlement. The load-Consolidation Y C1 bearing capacity of the rock formation foundation is significantly in excess of that necessary for the support of the plant structures.

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

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Screening Result

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

Per UFSAR Section 2.4.5 (Reference [37]), flooding due to surges is not applicable to DNPS.

The NRC accepted this assessment in its review of the DNPS Flood Hazard Reevaluation Report (Reference [47]) confirming that this flood-causing Storm Surge Y C3 mechanism was considered not plausible since DNPS is an inland location and does not connect directly with any bodies of water capable of producing storm surge.

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

The UFSAR (Reference [37]) discusses onsite toxic chemicals and Table 2.2-7 lists potentially toxic chemicals stored within the DNPS site boundary. UFSAR Section 6.4.4.2.2 states that the onsite chemicals listed in Table 2.2-7 were analyzed and screened based on a 2015 study (Reference [45]) to determine if the release of any of those chemicals could pose a threat to control room Toxic Gas Y C1 habitability. The screening shows that none of the chemicals stored onsite pose a threat.

Per UFSAR Section 6.4.4.2.2, the offsite chemicals evaluated included the chemicals listed in Regulatory Guide 1.78 (Reference [48]) along with those chemicals listed in the original Bechtel Control Room Habitability Study (Reference [49]). Each of the chemicals included in the analysis was evaluated

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Screening Result

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

based on toxic, physical, and chemical properties. Some were eliminated based on Regulatory Guide 1.78 (Table C-2) criteria (Reference [48]). The remaining chemicals were analyzed assuming a fresh air intake of 2000 ft3/min to the air handling system and no isolation.

Per UFSAR Section 6.4.4.2.3, operators are protected against those remaining chemicals by placing the control room HVAC system in the isolation/recirculation mode. This isolation mode provides for 100% recirculated air with no outside makeup. Operator action to isolate the control room is required (within 2 minutes after detection of odor) for chemicals whose control room concentrations would otherwise exceed the toxicity limits after that time. The chemicals requiring operator action are pure formaldehyde and pure hydrogen fluoridehydrochloric acid, chlorine, and 1,3-butadiene.

Since the isolation/recirculation mode of the control room HVAC system is credited for screening the Toxic Gas hazard, components fulfilling the isolation/recirculation mode function of the control room HVAC system will be considered high safety significant should the system be categorized under 10 CFR 50.69.

See also Industrial or Military Facility Accident and Release of Chemicals from Onsite Storage

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

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Screening Result

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

Per the IPEEE Section 5.2.3 (Reference [36]), rail, barge, aircraft, and pipeline transportation accidents are insignificant hazards to the site. (C1, C3)

Highway, Railway, and Waterway Transportation is discussed in UFSAR Sections, 2.2.3.1.2, 2.2.3.1.3, 2.2.3.1.4; respectively (Reference [37]).

For highways, the worst event postulated according to Regulatory Guide 1.91 (Reference [43]) is the explosion of a truck carrying 50,000 pounds of TNT on the nearest road. It was found that the corresponding safe standoff distance for which blast overpressure does not exceed 1.0 psi is 1660 feet. Since the C1 closest road is more than 1660 feet from the Class 1 structures of the plant, the Transportation transport of explosive materials on nearby roads does not present a hazard to Accident Y C3 plant safety. (C3)

For railways, an accident postulated in Regulatory Guide 1.91 is the simultaneous explosion of three boxcar-loads of TNT (396,000 pounds) on the nearest railroad. It was found that the corresponding safe standoff distance for which blast overpressure will not exceed 1.0 psi is 0.63 miles. Since the nearest railroad is more than 0.63 miles from the Class 1 structures of the plant, the transport of explosive materials on nearby railroads does not present a hazard to plant safety. (C3)

For waterways, a review of the materials passing by the site area indicates that the worst event would be the explosion of an empty petroleum barge (one

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Screening Result

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

containing the vapors of the previous cargo) on the river 0.5 miles from the plant. For such an explosion to occur, it is assumed that the empty tank contains an adequate vapor-air mixture and that a proper detonating stimulus is applied to this mixture. Under these assumptions, the corresponding distance at which the blast overpressure attenuates to 1.0 psi is approximately 600 feet, as determined using Regulatory Guide 1.91. Since the closest Class 1 structure (crib house) is more than 600 feet from the Illinois River, empty fuel barges do not present a hazard to plant safety. ( C3)

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

Per UFSAR Section 2.4.6 (Reference [37]), flooding due to tsunamis is not applicable to DNPS.

The NRC accepted this in its review of the DNPS Flood Hazard Reevaluation Tsunami Y C3 Report (Reference [47]) confirming that this flood-causing mechanism was considered not plausible since DNPS is an inland location and does not connect directly with any bodies of water capable of producing a tsunami.

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

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Screening Result

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

For main turbine missiles, per NRC RG 1.115 (Reference [50]), the NRC-preferred option for unfavorably oriented main turbines such as at DNPS is to limit P1, the annual probability of "low trajectory" turbine missile generation resulting in the ejection of turbine disk (or internal structure) fragments through the turbine casing, to 1E-5.

A 2019 evaluation (performed consistent with RG 1.115 approaches) was performed to calculate the P1 value for various cases of test intervals of turbine overspeed protection (OPS) components for the DNPS main turbine (Reference [51]).

Turbine-Generated The value of P1 for DNPS is the sum of the annual probabilities of the two RG Missiles Y PS3 1.115 missile generation scenarios: 1) probability of generating a turbine missile due to brittle failure of a rotor disk at design speed and 2) probability of generating a turbine missile due to ductile failure of the rotor.

Based on the monthly main turbine surveillance tests corresponding to the "Case 3 Test Interval" from Table 2-1 of Reference [51], the total P1 probability is 8.01E-6.

Per Table 1 of RG 1.115, the likelihood of main turbine missile impact on essential equipment is 1E-7/yr for a plant with an unfavorably oriented main turbine demonstrating a P1 value of less than 1E-5/yr. As such, the estimated CDF from a DNPS postulated main turbine missile is less than 1E-7/yr.

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Screening Result

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

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

Per the IPEEE (Reference [36]), the hazard is not applicable to the site because of location in the Midwest.

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

Refer to Section 3.2.4.1.

Waves Y C4 Based on this review, the Waves hazard is included in the Precipitation and Combined Effects Flooding hazards. Therefore, waves can be considered to be negligible and screened.

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

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Figure A4-1: External Flooding Hazard Curve for Combined Effects Flooding

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Attachment 5: Progressive Screening Approach for Addressing External Hazards

Event Analysis Criterion Source

C1. Event damage potential is < events NUREG/CR-2300 and for which plant is designed. ASME/ANS Standard RA-Sa-2009 C2. Event has lower mean frequency and NUREG/CR-2300 and no worse consequences than other ASME/ANS Standard events analyzed. RA-Sa-2009 Initial Preliminary C3. Event cannot occur close enough to NUREG/CR-2300 and Screening the plant to affect it. ASME/ANS Standard RA-Sa-2009 C4. Event is included in the definition of NUREG/CR-2300 and another event. ASME/ANS Standard RA-Sa-2009 C5. Event develops slowly, allowing ASME/ANS Standard adequate time to eliminate or mitigate the RA-Sa-2009 threat.

PS1. Design basis hazard cannot cause ASME/ANS Standard a core damage accident. RA-Sa-2009 PS2. Design basis for the event meets NUREG-1407 and the criteria in the NRC 1975 Standard ASME/ANS Standard Progressive Review Plan (SRP). RA-Sa-2009 Screening PS3. Design basis event mean frequency NUREG-1407 as modified in is < 1E-5/y and the mean conditional ASME/ANS Standard core damage probability is < 0.1. RA-Sa-2009 NUREG-1407 and PS4. Bounding mean CDF is < 1E-6/y. ASME/ANS Standard RA-Sa-2009 Screening not successful. PRA needs to NUREG-1407 and Detailed PRA meet requirements in the ASME/ANS ASME/ANS Standard PRA Standard. RA-Sa-2009

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Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty

The process defined in NUREG 1855 Rev. 1 (Reference [52]) and Electric Power Research Institute (EPRI) Technical Reports 1016737 (Reference [53]) and 1026511 (Reference [54]) was used to evaluate uncertainties in this application. These include:

• Identification of plant-specific Internal Events/Internal Flooding PRA model uncertainty sources, as well as generic sources per EPRI 1016737.

• Consideration of parameter and completeness uncertainties.

• Identification of plant-specific Internal Fire PRA model sources, along with generic sources per Appendix B of EPRI 1026511.

• Identification of seismic PRA model plant-specific sources, and generic sources per Appendix C of EPRI 1026511.

• Consideration of generic Level 2 model sources per Appendix E of EPRI 1026511, as applicable to Large Early Release Frequency (LERF).

The DNPS Internal Events/Internal Flooding and Fire and Seismic PRA models and documentation were reviewed for generic and plant-specific modeling assumptions and related sources of uncertainty. The applicable lists of EPRI-identified generic sources of uncertainty per EPRI 1016737 and EPRI 1026511 were also reviewed.

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

[22]) requirements for identification and characterization of uncertainties and assumptions. The evaluations identify those sources of uncertainty that are important to the PRA results and may be important to PRA applications. The approach meets the intent of steps C-1 and E-1 of NUREG-1855, Revision 1 (Reference [52]).

The results of the base PRA evaluations were reviewed to determine which potential uncertainties could impact applications, including the 50.69 P rogram. This evaluation meets the intent of the screening portion of steps C-2 and E-2 of NUREG-1855, Revision 1.

A detailed evaluation of assumptions and sources of uncertainties for this application is provided in Reference [55].

In addition, for the10 CFR 50.69 Program, the guidance in NEI 00 04 (Reference [1]) specifies sensitivity studies to be conducted for each PRA model to address key sources of uncertainty.

The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask the SSC(s) importance. Regulatory Guide 1.174, Revision 3 [56] cites NUREG-1855, Revision 1, as related guidance. In Section B of RG 1.174, Revision 3, the guidance acknowledges specific revisions of NUREG 1855 to include changes associated with expanding the discussion of uncertainties.

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Note: As part of the required10 CFR 50.69 PRA categorization sensitivity cases directed by NEI 00 04, Internal Events / Internal Flood and Fire and Seismic PRA models, human failure events (HFEs) and common cause failure (CCF) basic events are increased to their 95th percentile and also decreased to their 5th percentile values. In addition, maintenance unavailability terms are set to 0.0. For the Fire PRA model, a sensitivity case is required to allow no credit for manual suppression. The seismic PRA model sensitivity requires use of correlated fragilities for all SSCs in an area. These results are capable of driving a component and respective functions to HSS. The uncertainty of the modeling of HFEs and CCF basic events in the PRA, and the probabilities associated with those events (human error probabilities (HEPs) and CCF probabilities) are accounted for in the10 CFR 50.69 Application.

Based on the above evaluations, key assumptions and sources of uncertainty which could affect this application were identified and dispositioned with sensitivity analyses using the DNPS Internal Events/Internal Flooding and FPRA, and seismic PRA models as documented in Reference [55]. The results are summarized in Tab le A6 below.

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Table A6: Key Assumptions/Sources of Uncertainty, Their Impact, and Disposition Sources of Assumption/ 10 CFR 50.69 Impact Model Sensitivity and Disposition Uncertainty (10 CFR 50.69)

Internal Events and Internal Flood (IE/IF) Model

Core Cooling Success Following A sensitivity analysis was performed that Containment Failure or Venting increased the conditional probability by a Through Non-Hard Pipe Vent Paths factor of 10 (to 6.0E -1) that a large drywell FW / Condensate / SBCS and CRD are failure would result in loss of the feedwater, Following containment failure, credited for success after containment condensate, SBCS, and CRD injection injection from CRD and failure, but an additional basic event capabilities.

FW/Condensate could still be (1CNPVDWRUPT--R -- large DW maintained, but if a large containment containment failure causes loss of injection) The results demonstrate that FPIE PRA and failure occurs, injection paths may be is included that represents the likelihood that FPRA CDF and LERF results are sensitive disrupted leading to loss of these the containment failure size and location to the failure probability associated with external sources. This failure disrupts the capability of failure of all of the selected injection probability is based on a detailed FW/Condensate/SBCS and CRD to inject. systems following a large drywell failure.

structural analysis of the Mark I However, a factor of 10 increase in the containment design and large scale conditional failure probability, applied to all ultimate failure testing of steel of the selected injection systems, is not containments. considered credible.

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Sources of Assumption/ 10 CFR 50.69 Impact Model Sensitivity and Disposition Uncertainty (10 CFR 50.69)

The requirement for room cooling for various HPCI mission times, with and without failure of the gland seal condenser, is identified as a candidate source of model HPCI Room Cooling uncertainty.

HPCI room cooling is supplied by A sensitivity analysis was performed that DGCW1. It requires manual startup HVAC dependencies for HPCI are not increased the failure probability for the action. No room cooling is required included for early operation but are included HPCI gland seal hotwell pump failing to for HPCI mission time as long as there (fan only) for extended operation beyond 24 start by about a factor of 100 (to 1.0E-1).

is no gland seal condenser failure. hours.

For gland seal failures, the HPCI The results demonstrate that FPIE PRA system is assigned failure directly. CDF and LERF have little sensitivity to the failure probability associated with failure of HPCI room cooling. However, a factor of about 100 increase in the pumps failure probability is conservative and not consistent with observed behavior.

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Sources of Assumption/ 10 CFR 50.69 Impact Model Sensitivity and Disposition Uncertainty (10 CFR 50.69)

Reliability values from vendor studies For this sensitivity analysis, the failure demonstrating that the system performance probability for the digital feedwater Digital Feedwater Control Failure would result in less than 0.1 transients per controller failing to control feedwater such Probabilities year are used for the key components of the that the RPV overfills and floods the steam system. line was increased by a factor of 100.

There are model uncertainties associated with modeling digital Basic events representing the reliability The results demonstrate that FPIE PRA systems, such as those related to values for the auto level controller, the field CDF and LERF are not sensitive to the determining the failure modes of these buses, false signal from the redundant failure probability associated with failure of systems and components. reactivity control system, and false signal the digital feedwater controller to stop from the Level 8 trip system are included in feedwater prior to vessel overfill.

the system logic model.

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Sources of Assumption/ 10 CFR 50.69 Impact Model Sensitivity and Disposition Uncertainty (10 CFR 50.69)

The recovery probability assigned for recovering instrument air is considered It is assumed the containment vent valves reasonable and is supported by data.

cannot be opened by local manipulation of However, use of this value could lead to a Instrument Air (IA) System Recovery the valves or their air operators. They slightly optimistic assessment of (Containment Vent Valve Dependency require instrument air to provide the force to containment vent success.

on Air) open the containment vent AOVs. This requires instrument air availability to fulfill For this sensitivity analysis, the probability The containment vent valves do not the containment vent function. For some of failing to recover balance of plant have accumulator backups to provide low probability sequences, instrument air is systems (including instrument air) after IA a method of successful venting given not available due to system failures. failure due to random causes was a loss of IA scenario. Currently, the However, the model uses data from NSAC-increased to 1.0.

model credits IA recovery at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 161 (Reference [57]) to provide a basis to to restore the hard pipe vent path. support the restoration of instrument air or The results demonstrate that FPIE PRA its support systems. CDF and LERF are relatively insensitive to the failure probability associated with failure to recover instrument air (balance of plant, as well as specific to support of venting).

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Sources of Assumption/ 10 CFR 50.69 Impact Model Sensitivity and Disposition Uncertainty (10 CFR 50.69)

The HEPs and equipment reliabilities used for FLEX may be underestimated given the current state of knowledge about FLEX.

FLEX and Hardened Containment The FLEX component failure rates can be For this sensitivity case, credit for FLEX Vent System (HCVS) Heps and represented by using the failure rates of was completely removed from the model Equipment Failure Rates like -components (e.g., emergency diesel through use of a FLAG event.

generator (EDG)) as surrogates (e.g., for the For the DNPS IE PRA model there FLEX diesel generators). The HEPs can be The results demonstrate that FPIE PRA were no industry-approved data represented using screening values (ranging CDF and LERF essentially have no sources for FLEX equipment from 1E-3 to 0.5). FLEX component failures sensitivity to the availability of FLEX. This reliability. The DNPS PRA models are estimated based on like-components result is driven by the fact that the use of FLEX component failures and human by increasing that like-components failure FLEX at the DNPS is constrained to failure events associated with failure rate by a factor of two. Human error situations declared as Extended Loss of AC to align the FLEX equipment. probabilities employ screening values. Power (ELAP) scenarios (including station blackout (SBO)). These scenarios are rare, and thus there are few sequences in the PRA for which FLEX is credited. Removing that credit therefore has very little impact on the quantified results.

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Sources of Assumption/ 10 CFR 50.69 Impact Model Sensitivity and Disposition Uncertainty (10 CFR 50.69)

The base model includes credit for the BlackStarTech (BST) portable equipment (which includes the capability to supply specific AC and DC power to select BlackStarTech (BST) portable carts Use of BST has the same constraints as components). Equipment failure equipment and human action reliability FLEX and HCVS (i.e., ELAP). The BST probabilities are based on similar component failure rates can be represented components (e.g., batteries), and human For the DNPS PRA model there were by using the failure rates of like-error probabilities are calculated using no industry-approved data sources for components (e.g., batteries and battery detailed human reliability analysis BlackStarTech (BST) equipment chargers) as surrogates for BST equipment. techniques.

reliability. The DNPS PRA models The HEPs can be represented using BST component failures and human standard HRA techniques (as employed for For this sensitivity case, credit for BST was failure events associated with failure other human failure events throughout the completely removed from the model to align the BST equipment. PRA). through use of a FLAG event. The results demonstrate that FPIE PRA CDF and LERF have no sensitivity to the availability of BST.

Use of BST has the same constraints as FLEX and HCVS (i.e., ELAP).

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Sources of Assumption/ 10 CFR 50.69 Impact Model Sensitivity and Disposition Uncertainty (10 CFR 50.69)

Fire Model

For this sensitivity case, equipment or cables assumed failed or credited by assumed routing in the base FPRA were assumed to always be available.

Fire PRA component selection The results demonstrate that FPRA results involves the selection of components The Fire PRA assumes that, at a minimum, are sensitive to changes when the to be treated in the analysis in the a plant trip occurs. This is consistent with equipment is assumed available. However, context of fire initiating events and accepted industry practice. The Fire PRA a review of the sensitivity analysis results mitigation. The potential sources of does not credit some equipment or systems identified the reasons for the decrease in uncertainty include those inherent in that are credited in the full power internal FPRA results were non-conservative given the internal events PRA model as that events PRA. Systems are included based on the contributing equipment is not expected model provides the foundation for the an iterative process to include equipment to be available.

FPRA. that may be significant to the fire risk.

The scope of credited equipment and cables and assumed cable routing is based on reviews of the applicable systems and the PRA model. Therefore, the scope credited equipment in the FPRA provides best estimate results.

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The use of 1E-6 is consistent with industry The Fire PRA includes conservative guidance for FPRA and is an accepted adjustments to the HFEs to account for practice.

The Human Error Probabilities (HEPs) adverse impacts of fire events. The Fire used in the FPRA were adjusted to PRA does not include credit for all operator A sensitivity case was performed for the consider the additional challenges that actions, including fire response actions. The base FPRA using a minimum joint HEP of may be present given a fire. The Fire PRA does not include credit for all 1E-5. The results demonstrate that using a HEPs included the consideration of instrument cues that may be available. A higher FPRA minimum joint HEP has a degradation or loss of necessary cues minimum joint HEP was applied for the HRA slight impact on FPRA results.

due to fire. Given the methodology dependency analysis. Applying a minimum used, the impact of any remaining joint HEP may skew the results by artificially Further, as directed by NEI 00- 04, the fire uncertainties is expected to be small. increasing the risk due to human failure model human error basic events are events. The HEPs are propagated in the increased to their 95th percentile and also parametric uncertainty evaluation based on decreased to their 5th percentile values as the uncertainty parameters from the HRAC. part of the required10 CFR 50.69 PRA categorization sensitivity cases.

Seismic Model

Change in SLERF due to Seismic The reasonable success in offsite The "early" component of the Large Early Impact on offsite Evacuation evacuation given a General Emergency Radionuclide Release risk metric is based declaration is inherent in the LERF on assumptions of evacuation times for The definition of the LERF risk metric definition. The LERF end state assignment surrounding areas given declaration of a includes the radionuclide release time is assigned to accident sequences based on General Emergency. Typical of most U.S.

frame referenced to the timing of the accident sequence progression timings and NPP SPRAs, the same definition of LERF declaration per plant procedure of a (i.e., same " Early" time frame hours and

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General Emergency. The LERF release magnitudes specific to each same "Large" magnitude range in fraction of timing assumes reasonable success accident sequence. core CsI released to environment) that is in the offsite evacuation and used in the FPIE PRA and FPRA PRA is sheltering. Seismic-induced failures used for the SPRA. This is to facilitate the of offsite infrastructure may occur for addition of LERF results from the various high magnitude seismic events that PRA models using a consistent definition.

would hamper the basic assumption of This approach, along with investigating the reasonable success of evacuation and topic in a sensitivity study for the base sheltering. SPRA analysis, is recommended in Reference [58] guidance.

As part of the DNPS SPRA base quantification, a sensitivity study was performed to estimate the impact on SLERF if the surrounding infrastructure is assumed significantly disrupted (road surfaces buckled, traffic lights inoperable, building facades fallen, etc.) for large earthquakes such that the evacuation timing assumptions inherent in the definition of early are questionable. This sensitivity is performed by assuming that all seismic events with magnitude >0.5g result in sufficient delay in the evacuation time such that they are modeled as leading directly to

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the SLERF endstate. This sensitivity approach is reflective of sensitivity studies used in other U.S. NPP SPRAs; detailed fragility calculations and revised evacuation studies for offsite infrastructure is not a realistic alternative for typical SPRAs. The sensitivity study showed that total SLERF would significantly increase (by 68%) for such a modeling assumption.

Use of a LERF definition that is consistent across all the PRA models is judged reasonable and typical of U.S. NPP PRAs.

As such, this topic is not carried forward as an SPRA model uncertainty for 50.69 risk sensitivity calculations.

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The DNPS RPV Internals are calculated to have a relatively low seismic capacity of Am

= 0.75g PGA due to the identified governing failure mode of the upper and lower clamps on the core shroud tie rods. Seismic failure of the RPV internals is modeled in the DNPS SPRA (typical of U.S. SPRAs) with a 100% likelihood of failing successful Further Evaluation of Failure to insertion of the control rods into the reactor Scram Fragility (SCRAM), resulting in an ATWS scenario.

The SPRA modeling assumptions are The DNPS SPRA failure to scram conservative for this issue (e.g., failure of fragility is dominated by seismic - the core shroud clamps is assumed to induced failure of the upper and lower Seismic-induced failure to scram fragilities result in instantaneous failure of welds due clamps of the reactor core shroud are incorporated into the reactor scram to rapid expansion of previously identified plant modification. Further function system fault tree logic. weld indications and sufficient failure of the investigation into the fragility shroud and core geometry that leads to calculations of the reactor internals failure to scram).

and failure to scram may be postulated to lower the ATWS risk As part of the DNPS SPRA base contribution in the DNPS SPRA. quantification, a sensitivity study was performed that increased the seismic capacity of the failure to scram fragility (a capacity of Am = 1.25g PGA was assumed). The sensitivity study showed that increasing the Am for the core shroud tie rod failure from 0.75g to 1.25g would decrease SLERF significantly (by 37.5%).

However, modification to the core shroud tie RS 056, DNPS Application to Adopt 10 CFR 50.69 Attachment 6 Page 74 of 74

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rods would require a significant design effort and significant work within the reactor vessel, as well as inherent plant risk associated with implementing the modifications. In addition, while the core shroud tie rod failure currently controls the fragility of the RPV internals, there are several other components with Am values at or below 1.0g.

It is uncertain as to the precision of the seismic fragility but the current modeling is judged reasonable for use in the base DNPS SPRA and for use in 50.69 risk calculations. As such, this topic is not carried forward as an SPRA model uncertainty for 50.69 risk sensitivity calculations.