ULNRC-06550, 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

ML20304A455
Person / Time
Site:	Callaway
Issue date:	10/30/2020
From:	Banker S Ameren Missouri, Union Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML20304A454	List: ML20304A456 ULNRC-06550, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors
References
ULNRC-06550
Download: ML20304A455 (52)

Text

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WAmeren Callaway Plant MISSOURI

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October 30, 2020 ULNRC-06550 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 10 CFR 50.90 10 CFR 50.69 Ladies and Gentlemen:

DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT NO. 1 UNION ELECTRIC CO.

RENEWED FACILITY OPERATING LICENSE NPF-30 APPLICATION TO ADOPT 10 CFR 50.69, "RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS" In accordance with the provisions of 10 CFR 50.69 and 10 CFR 50.90, Union Electric Company (dba Ameren Missouri) is requesting an amendment to the operating license (OL) for Callaway Plant, Unit No. 1.

The proposed amendment would modify the Callaway Plant licensing basis by the addition of a License Condition, i.e., License Condition 2.(C).(19), to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be oflow safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced.

This allows improved focus on equipment that has safety significance, resulting in improved plant safety.

Enclosure 1 to this letter provides the basis for the proposed change to the Callaway Plant, Unit No. 1, Operating License. The categorization process being implemented through this change is consistent with NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision O dated July 2005, which was endorsed by the U.S. Nuclear Regulatory Commission (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. Attachment 1 of Enclosure 1 provides a list of categorization prerequisites. Use of the categorization process on a plant system will only occur after the specified prerequisites are met.

Ameren Missouri intends to submit a separate license amendment request (LAR) for TSTF-505, "Risk-Informed Completion Time (RICT) Program," within the next three months using the same probabilistic risk assessment (PRA) models described in Enclosure 1. Ameren Missouri requests that the NRC coordinate their review of the PRA technical adequacy description in Section 3.2 and 3.3 of Enclosure 1 for both applications .

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ULNRC-06550 October 30, 2020 Page 2 of 4 This would reduce the number of Ameren Missouri and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action, as the details of the PRA models in each LAR are complete which will allow the NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

Additional documents are provided in Enclosures 2 through 4 of this letter. The aforementioned License Condition to be incorporated into the Callaway Plant OL is presented as a mark-up to the OL in Enclosure 2.

A "clean" copy of the affected OL page reflecting incorporation of the proposed License Condition is provided as Enclosure 3. Enclosure 4 documents a regulatory commitment being made for completion of activity in regard to the Fire PRA, as further described in Section 3 .3 of Enclosure 1.

Ameren Missouri requests approval of the proposed license amendment by October 30, 2021, with the amendment being implemented within 90 days.

In accordance with 10 CFR 50.91, a copy of this application, with its enclosures and attachments, is being provided to the designated Missouri Official.

If you should have any questions regarding this submittal, please contact Tom Elwood (Supervising Engineer, Regulatory Affairs/Licensing) at (314) 225-1905.

I declare under penalty of perjury that the foregoing is true and correct.

Sincerely,

~ Stephanie Banker Vice President, Engineering Executed on: __/_l)-+/_$_0...,../_J..D_M __

Enclosures:

1. Evaluation of the Proposed Change

2. Mark-up of Operating License (OL)

3. Clean Copy of Affected OL Pages (Reflecting Proposed Change)

4. Commitment

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ULNRC-06550 October 30, 2020 Page3 of 4 cc: Mr. Scott A. Morris Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Mahesh Chawla Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O8B1A Washington, DC 20555-0001

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ULNRC-06550 October 30, 2020 Page 4 of 4 Index and send hardcopy to QA File A160.0761 Hardcopy:

Certrec Corporation 6100 Western Place, Suite 1050 Fort Worth, TX 76107 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)

Electronic distribution for the following can be made via LER ULNRC Distribution:

F.M.Diya B. L. Cox F. J. Bianco S. P. Banker K. A. Mills S. G. Kovaleski R. C. Wink T. B. Elwood B. E. Ruhmann J. W. Hiller NSRB Secretary Performance Improvement Coordinator STARS Regulatory Affairs Mr. Jay Silberg (Pillsbury Winthrop Shaw Pittman LLP)

Ms. Katie Jo Wheeler (DNR)

Missouri Public Service Commission

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to ULNRC-06550 Page 1 of 32 Enclosure 1 Evaluation of the Proposed Change Table Of Contents 1

SUMMARY

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

2.3 DESCRIPTION

OF THE PROPOSED CHANGE.................................................5 3 TECHNICAL EVALUATION *********************************************************************************** 6 3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR S0.69(b)(2)(i)).7 3.1.1 Overall Categorization Process .......................................................... 7 3.1.2 Passive Categorization Process........................................................ 12 3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR S0.69(b)(2)(ii)) .......... 13 3.2.1 Internal Events and Internal Flooding ........................................... 13 3.2.2 Fire Hazards .....................................................................*.....................15 3.2.3 Seismic Hazards ..................................*.................................................15 3.2.4 High Winds Hazards ..........................................................................*.*15 3.2.5 Other External Hazards ......................................................................16 3.2.6 Low Power & Shutdown .....................................................................16 3.2.7 PRA Maintenance and Updates ***********************..************.************.***** 16 3.2.8 PRA Uncertainty Evaluations ******..**.***....****.*.*...**.***..********.*....**..***.. 17 3.2.9 Modeling of Flex....*...*.....*...........*..............**...****......**....*******....**....*.... 19 3.3 PRA REVIEW PROCESS RESULTS (10 CFR S0.69(b)(2)(iii)) ................. 22 3.4 RISK EVALUATIONS (10 CFR S0.69(b)(2)(iv)) ......................................... 25 3.5 FEEDBACK AND ADJUSTMENT PROCESS..................................................... 26 3.6 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA........................ 27 3.7 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS .................... 27

3.8 CONCLUSION

S...*.*.........*.*.*.********.**.***..********.*..**.*****.***.************************************** 29 4 ENVIRONMENTAL CONSIDERATION ************************************************************** 29 5 REFERENCES ******************************************************************************************************* 30 to ULNRC-06550 Page 2 of 32 LIST OF ATTACHMENTS : List of Categorization Prerequisites : Description of PRA Models Used in Categorization : Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items : External Hazards Screening : Progressive Screening Approach for Addressing External Hazards : Key Assumptions and Sources of Uncertainty : Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-1/11, CC-II/III, and CC-I/II/III

Enclosure 1 to ULNRC-06550 Page 3 of 32 1

SUMMARY

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

2 DETAILED DESCRIPTION 2.1 CURRENT REGULATORY REQUIREMENTS The 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 Structures, Systems and Components (SSCs) necessary to defend against the DBEs are defined as "safety-related," and these SSCs are the subject of many regulatory requirements, herein referred to as "special treatments," designed to ensure that they are of high quality and high reliability, and have the capability to perform during postulated design basis conditions. Treatment includes, but is not limited to, quality assurance, testing, inspection, condition monitoring, assessment, evaluation, and resolution of deviations.

The distinction between "treatment" and "special treatment" is the degree of NRC specification as to what must be implemented for particular SSCs or for particular conditions. Typically, the regulations establish the scope of SSCs that receive special treatment using one of three different terms: "safety-related," "important to safety," or "basic component." The terms "safety-related" and "basic component" are defined in

Enclosure 1 to ULNRC-06550 Page 4 of 32 the regulations, while "important to safety," used principally in the general design criteria of Appendix A to 10 CFR Part SO, is not explicitly defined.

2.2 REASON FOR PROPOSED CHANGE A probabilistic approach to regulation enhances and extends the traditional deterministic approach by allowing consideration of a broader set of potential challenges to safety, providing a logical means for prioritizing these challenges based on safety significance, and allowing consideration of a broader set of resources to defend against these challenges. In contrast to the deterministic approach, Probabilistic Risk Assessments (PRAs) address credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for common cause failures. The probabilistic approach to regulation is an extension and enhancement of traditional regulation by considering risk in a comprehensive manner.

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

The rule contains requirements on how a licensee categorizes SSCs using a risk-informed process, adjusts treatment requirements consistent with the relative significance of the SSC, and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four risk-informed safety class (RISC) categories.

The determination of safety significance is performed by an integrated decision-making process, as described by Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline" (Reference 1), which uses both risk insights and traditional engineering insights. The safety functions include the design basis functions, as well as functions credited for severe accidents (including external events). Special or alternative treatment for the SSCs is applied as necessary to maintain functionality and reliability, and is a function of the SSC categorization results and associated bases.

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

The rule does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility. Instead, the rule enables licensees to focus their resources on SSCs that make

Enclosure 1 to ULNRC-06550 Page 5 of 32 a significant contribution to plant safety. For SSCs that are categorized as high safety significant, existing treatment requirements are maintained or enhanced. Conversely, for SSCs that do not significantly contribute to plant safety on an individual basis, the rule allows an alternative risk-informed approach to treatment that provides a reasonable, though reduced, level of confidence that these SSCs will satisfy functional requirements.

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

2.3 DESCRIPTION

OF THE PROPOSED CHANGE Ameren Missouri proposes the addition of the following license condition to the renewed operating license of Callaway Plant, Unit No. 1, to document the NRC's approval of the use 10 CFR 50.69.

Ameren Missouri 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 SSCs using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, high winds, and seismic risk; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on a screening of other external hazards updated using the external hazard screening significance process identified in American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard RA-Sa-2009; as specified in License Amendment No. [XXX] dated [Date].

The above License Condition is proposed to be incorporated as License Condition 2.(C).19 of the Callaway Operating License (OL). As noted in the cover letter of this LAR, a mark-up of the affected OL page is provided in Enclosure 2. Additionally, a "clean" copy of the affected OL page reflecting incorporation of proposed License Condition 2.(C).(19) is provided as Enclosure 3.

Prior NRC approval, under 10 CFR 50.90, is required for implementation of the categorization process specified above, including approval of the proposed license condition. Upon approval of the above-described 10 CFR 50.69 process for Callaway, any future change to the categorization process described above (e.g., the change from a seismic margins approach to a seismic probabilistic risk assessment approach) would require NRC approval under 10 CFR 50.69.

to ULNRC-06550 Page 6 of 32 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)(l)(i).

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

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

It should be noted that Ameren Missouri intends to submit a separate license amendment request (LAR) for adoption of Technical Specification Task Force (TSTF)

TS Traveler TSTF-505, Revision 2, "Provide Risk- Informed Extended Completion Times

- RITSTF Initiative 4b," within the next three months using the same PRA models described in this Enclosure. Ameren Missouri requests that the NRC coordinate their review of the PRA technical adequacy description in Section 3.2 and 3.3 of this enclosure for both applications. This would reduce the number of Ameren Missouri and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action, as the details of the PRA models in each LAR are complete, which will allow the NRC staff to independently review and approve each LAR on its own merits without regard to the results from review of the other.

to ULNRC-06550 Page 7 of 32 3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(b)(2)(i))

3.1.1 Overall Categorization Process Ameren Missouri will implement the risk categorization process in accordance with NEI 00-04, Revision 0, as endorsed by Regulatory Guide (RG) 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance" (Reference 2). NEI 00-04 Section 1.5 states, "Due to the varying levels of uncertainty and degrees of conservatism in the spectrum of risk contributors, the risk significance of SSCs is assessed separately from each of five risk perspectives and used to identify SSCs that are potentially safety-significant."

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

1. PRA-Based Evaluations (e.g., the internal events, internal flooding, fire, seismic, and high winds PRAs)

2. Non-PRA Approaches (e.g., other external events screening, and shutdown assessment)

3. Seven Qualitative Criteria in Section 9.2 of NEI 00-04

4. The Defense-in-Depth Assessment

5. The Passive Categorization Methodology Categorization of SSCs will be completed per the NEI 00-04 process, as endorsed by RG 1.201, which includes the determination of safety significance through the various elements identified above. The results of these elements are used as inputs to arrive at a preliminary component categorization (i.e., High Safety Significant [HSS] or Low Safety Significant) that is presented to the Integrated Decision-Making Panel (IDP).

to ULNRC-06550 Page 8 of 32 Note: the term "preliminary HSS or LSS" is synonymous with the NEI 00-04 term "candidate HSS or LSS." A component or function is preliminarily categorized as HSS if any element of the process results in a preliminary HSS determination in accordance with Table 3-1 below. The safety significance determination of each element identified above is independent of the others, and therefore, the sequence of the elements does not impact the resulting preliminary categorization of each component or function. Consistent with NEI 00-04, the categorization of a component or function will only be "preliminary" until it has been confirmed by the IDP. Once the IDP confirms that the categorization process was followed appropriately, the final RISC category can be assigned.

The IDP may direct and approve detailed categorization of components in accordance with NEI 00-04 Section 10.2. The IDP may always elect to change a preliminary LSS component or function to HSS; however, the ability to change component categorization from preliminary HSS to LSS is limited. This ability is only available to the IDP for select process steps as described in NEI 00-04 and endorsed by RG 1.201.

Table 3-1 summarizes these IDP limitations described in NEI 00-04. The steps of the process are performed at either the function level, component level, or both, as also summarized in Table 3-1. A component is assigned its final RISC category upon approval by the IDP.

Table 3-1: Categorization Evaluation Summary IDP Categorization Drives Evaluation Change Element Step-NEI 00-04 Associated Level HSSto Section Functions LSS Internal Events (including Internal Not Yes Flooding) Base Allowed Case - Section 5.1 Internal Fire, Seismic, and High Risk (PRA Allowable No Winds Events Base Component Modeled) Case PRA Sensitivity Allowable No Studies Integral PRA Not Assessment - Yes Allowed Section 5.6

Enclosure 1 to ULNRC-06550 Page 9 of 32 IDP Categorization Drives Evaluation Change Element Step-NEI 00-04 Associated Level HSSto Section Functions LSS Other External Not Component No Risk (Non- Hazards Allowed modeled)

Shutdown - Function/ Not No Section 5.5 Component Allowed Core Damage - Function/ Not Yes Defense- Section 6.1 Component Allowed in-Depth Containment - Not Component Yes Section 6.2 Allowed Qualitative Considerations -

Criteria Section 9.2 Function Allowable1 N/A Segment/ Not Passive Passive - Section 4 No Component Allowed Table Notes:

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

The seven considerations are addressed preliminarily by the 50.69 categorization team for at least the system functions that are not found to be HSS due to any other categorization step. Each of the seven considerations requires a supporting justification for confirming (true response) or not confirming (false response) the consideration. If the 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 50.69 team (i.e. all considerations for all functions are reviewed). The IDP may confirm the preliminary function risk and associatedjustification or may direct that it be changed based upon their expert knowledge. Because the Qualitative Criteria are the direct responsibility of the IDP, changes may be made from preliminary HSS to LSS or from preliminary LSS to HSS at the discretion of the IDP. If the IDP determines any of the seven considerations cannot be confirmed (false response) for a function, then the final categorization of that function is HSS.

The mapping of components to system functions is used in some categorization process steps to facilitate preliminary categorization of components. Specifically, functions with mapped components that are determined to be HSS by the PRA-based assessment (i.e., internal events PRA or integral PRA assessment) or defense-in-depth evaluation will be initially treated as HSS. However, NEI 00-04 Section 10.2 allows to ULNRC-06550 Page 10 of 32 detailed categorization which can result in some components mapped to HSS functions being treated as LSS, and Section 4.0 discusses additional functions that may be identified (e.g., fill and drain) to group and consider potentially LSS components that may have been initially associated with an HSS function but which do not support the critical attributes of that HSS function. Note that certain steps of the categorization process are performed at a component level (e.g., Passive, Non-PRA-modeled hazards, as shown on 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 to HSS based on Table 3-1 above, or it may remain LSS.

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

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

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

• The decision criteria for the IDP for categorizing SSCs as safety significant or low safety-significant pursuant to § 50.69(f)(l) will be documented in Ameren Missouri procedures. Decisions of the IDP will be arrived at by consensus. Differing opinions will be documented and resolved, if possible. However, a simple majority of the panel is sufficient for final decisions regarding HSS and LSS.

• Passive characterization will be performed using the processes described in Section 3.1.2 of this enclosure. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

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

to ULNRC-06550 Page 11 of 32

• NEI 00-04 Section 7 requires assigning the safety significance of a function to be preliminary HSS if it is supported by an SSC determined to be HSS from the PRA-based assessment in-Section 5, but it does not require this for SSCs determined to be HSS from non-PRA-based, deterministic assessments in Section 5.

This requirement is further clarified in the Vogtle SER (Reference 8) which states,

" ... if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth assessment (Section 6), the associated system function(s) would be identified as HSS."

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

• With regard to the criteria that considers whether the active function is called out or relied upon in the plant Emergency/Abnormal Operating Procedures, Ameren Missouri 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 hazard is described below:

• Internal Event Risks: Internal events including internal flooding PRA model, Update 8.

• Fire Risks: Fire PRA model, Update 8.

• Seismic Risks: Seismic PRA model, Update 8.

• High Winds Risks: High Winds PRA model, Update 8.

• Other External Risks (e.g., external floods): PRA-OEH-ANALYSIS, "Other External Hazards: Screening Assessment Notebook," Update 8. The results of non-PRA evaluations are based on the screening of other external hazards updated using the external hazard screening significance process identified in American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard RA-Sa-2009, as formally endorsed in RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2 (Reference 12).

• 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, "Guidelines for Industry Actions to Assess Shutdown Management" (Reference 7), which provides guidance for assessing and enhancing safety during shutdown operations.

Enclosure 1 to ULNRC-06550 Page 12 of 32 A change to the categorization process that is outside the bounds specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach) will not be used without prior NRC approval. The SSC categorization process documentation will include the following elements:

1. Program procedures used in the categorization

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

3. Mapping of components to support function(s)

4. PRA model results, including sensitivity studies

5. Hazards analyses, as applicable

6. Passive categorization results and bases

7. Categorization results including all associated bases and RISC classifications

8. Component critical attributes for HSS SSCs

9. Results of periodic reviews and SSC performance evaluations

10. IDP meeting minutes and qualification/training records for the IDP members 3.1.2 Passive Categorization Process For the purposes of 10 CFR 50.69 categorization, passive components are those components that have a pressure retaining function. Passive components and the passive function of active components will be evaluated using the Arkansas Nuclear One (ANO) Risk-Informed Repair/Replacement Activities (RI-RRA) methodology contained in Reference 9 (ML090930246) consistent with the related Safety Evaluation (SE) issued by the Office of Nuclear Reactor Regulation.

The RI-RRA methodology is a risk-informed safety classification and treatment program for repair/replacement activities (RI-RRA methodology) for pressure retaining items and their associated supports. In this method, the component failure is assumed with a probability of 1.0 and only the consequence evaluation is performed. It additionally applies deterministic considerations (e.g., DID, safety margins) in determining safety significance. Consistent with the Reference 9 ANO-2 Safety Evaluation, pipe supports were not required to be in the scope of the evaluation, but may be included in the scope at the licensee's discretion. 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.

Enclosure 1 to ULNRC-06550 Page 13 of 32 The RI-RRA method was also approved to be used for a 10 CFR 50.69 application as documented in the NRC's final Safety Evaluation for Vogtle, dated December 17, 2014 (Reference 8). The RI-RRA method as approved for use at Vogtle for 10 CFR 50.69 does not have any plant specific aspects and is generic. It relies on the conditional core damage and large early release probabilities associated with postulated ruptures.

Safety significance is generally measured by the frequency and the consequence of the event. However, this RI-RRA process categorizes components solely based on consequence, which measures the safety significance of the passive component given that it ruptures. This approach is conservative compared to including the rupture frequency in the categorization, as this approach will not allow the categorization of SSCs to be affected by any changes in frequency due to changes in treatment.

The passive categorization process is intended to apply the same risk-informed process accepted by the NRC in the Reference 9 ANO2-R&R-004 document for the passive categorization of Class 2, 3, and non-class components. This is the same passive SSC scope the NRC has conditionally endorsed in ASME Code Cases N-660 and N-662 as published in Regulatory Guide 1.147, Revision 15. Both code cases employ a similar risk-informed safety classification of SSCs in order to change the repair/replacement requirements of the affected LSS components. All ASME Code Class 1 SSCs with a pressure retaining function, as well as supports, will be assigned high safety-significant, HSS, for passive categorization, which will result in HSS for its risk-informed safety classification and cannot be changed by the IDP. Therefore, it is proposed that this methodology and scope for passive categorization is acceptable and appropriate for use at Callaway Plant, Unit No. 1, 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, except as noted in Section 3.3 under the Fire PRA.

3.2.1 Internal Events and Internal Flooding The Callaway Plant, Unit No. 1, categorization process for the internal events and flooding hazard will use the peer reviewed plant-specific PRA model. The Ameren Missouri risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for the Callaway Plant, Unit No. 1.

Related to the technical adequacy of the internal events model, the Internal Events discussion under Section 3.3 describes implementation of the methodology provided in PWROG-18027-NP (Reference 32) for assessing the loss of room cooling in PRA modeling. Following, but unrelated to, implementation of the method provided in PWROG-18027-NP into the Callaway PRA, this method was chosen by the PWROG and NEI to pilot the Newly Developed Methods (NDM) peer review process established in NEI 17-07 (Reference 33). The NEI 17-07 process was successfully completed with

Enclosure 1 to ULNRC-06550 Page 14 of 32 all applicable NDM attributes met at capability category (CC) CC 1/11 and no open peer review Findings against the method in PWROG-18027-NP. While the NEI 17-07 process was completed successfully, it is recognized that this process is not an approved process until RG 1.200 Revision 3 is issued, which is expected near the end of 2020. Despite the Callaway assessment, and acknowledgement by the PWROG, that the method provided in PWROG-18027-NP does not necessarily meet the definition of a NDM, it is recognized that the NRC staff may decide to independently review the method in PWROG-18027-NP for technical adequacy.

To support this decision, the following summary information on the noted method is provided.

PRA modeling practices generally tend to reference equipment operating ambient temperature limits (room temperatures, for example) as the criteria for determining whether or not room cooling is required to maintain equipment operability. If the room temperature is expected to exceed the equipment qualification temperature following loss of room cooling within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, (which is the mission time typically used in a PRA model), it is assumed that the equipment will fail. This can require extensive modeling of room cooling (ventilation and air conditioning) and often introduces additional conservatisms into the model. The equipment qualification (EQ) temperatures are often intended for continual operation of equipment for the qualified life of the component, not for single use as a limiting high temperature. Limited use of equipment at temperatures above the qualification temperature will not necessarily fail the component, but only shorten its qualified life. The methodology provides screening criteria to remove modeling the failure of HVAC as a failure mode of front-line equipment needed for accident mitigation. Implementation of this method requires additional plant/component-specific input such as room heatup calculations as well as identifying limiting (temperature sensitive) subcomponents. The standard Arrhenius model is used in this method and demonstrates considerable margin in most PRA equipment; it thus would not consider most PRA equipment vulnerable to failure by room heatup until reaching the screening temperature provided in the method.

As with existing approaches, this method compiles industry data and EQ information for common components and characterizes the components as thermally robust or potentially sensitive, relative to the specified screening temperature. The method identifies the component types and provides criteria for components that require additional analysis prior to applying the screening temperature. The PWROG-18027-NP report contains the technical basis for the acceptability of the method and is available for NRC audit. of this enclosure identifies the applicable internal events (including internal flooding) PRA models.

to ULNRC-06550 Page 15 of 32 3.2.2 Fire Hazards The Callaway Plant, Unit No. 1, categorization process for fire hazards will use a peer reviewed plant-specific fire PRA model. The internal fire PRA model wa-s developed consistent with NUREG/CR-6850 and only utilizes methods previously accepted by the .

NRC. Callaway Plant, Unit No. 1 was approved to implement NFPA-805 in January 2014, and since that time, there have been numerous updates to the approved methods through the issuance of fire PRA frequently asked questions and new or revised guidance documents. New or revised guidance is specifically addressed through the PRA maintenance and update process. The Ameren Missouri risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for the Callaway Plant, Unit No. 1.

It should also be noted that, as part of transition to NFPA-805, there were several committed modifications and implementation items as documented in NFPA-805 LAR Attachment S, "Plant Modifications and Items to be Completed during Implementation," which described the Callaway plant modifications necessary to implement the NFPA 805 licensing basis. All NFPA-805 LAR Attachment S items have been implemented; therefore, there are no NFPA-805 open items impacting this application. of this enclosure identifies the applicable fire PRA model.

3.2.3 Seismic Hazards The Callaway Plant, Unit No. 1, categorization process for seismic hazards will use a peer reviewed plant-specific seismic PRA model. The Ameren Missouri risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for the Callaway Plant, Unit No. 1. Industry standard methods were utilized in the development of the seismic hazards for the seismic PRA.

Updates to seismic hazard curves will be reflected in the PRA used for the categorization in accordance with the PRA model maintenance process. Attachment 2 of this enclosure identifies the applicable seismic PRA model.

3.2.4 High Winds Hazards The Callaway Plant, Unit No. 1, categorization process for the high winds hazard will use a peer-reviewed plant-specific high winds PRA model. The Ameren Missouri risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for the Callaway Plant, Unit No. 1. Attachment 2 of this enclosure identifies the applicable high winds PRA model.

Enclosure 1 to ULNRC-06550 Page 16 of 32 3.2.5 Other External Hazards All other external hazards were screened for applicability to Callaway Plant, Unit No. 1, per a plant-specific evaluation using the external hazard screening significance process identified in American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard RA-Sa-2009. Attachment 4 of this enclosure provides a summary of the other external hazards screening results. Attachment 5 provides a summary of the progressive screening approach for external hazards.

All remaining hazards not explicitly modeled were screened from applicability and considered insignificant for every SSC. Therefore, they will not be considered during the categorization process.

3.2.6 Low Power & Shutdown Consistent with NEI 00-04, the Callaway Plant, Unit No. 1, categorization process will use the shutdown safety management plan described in NUMARC 91-06 for evaluation of safety significance related to low power and shutdown conditions. The overall process for addressing shutdown risk is illustrated in Figure 5-7 of NEI 00-04.

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

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

3.2.7 PRA Maintenance and Updates The Ameren Missouri risk management process ensures that the applicable PRA models used in this application continue to reflect the as-built and as-operated plant for the Callaway Plant, Unit No. 1. 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, Ameren Missouri will implement a process that addresses the requirements in NEI 00-04, Section 11, "Program Documentation and Change Control." The process to ULNRC-06550 Page 17 of 32 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.

3.2.8 PRA Uncertainty Evaluations Uncertainty evaluations associated with any applicable baseline PRA models used in this application were evaluated during the assessment of PRA technical adequacy and confirmed through the peer review process as discussed in Section 3.3 of this enclosure.

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

In the overall risk sensitivity studies, Callaway Plant, Unit No. 1 will utilize a factor of 3 to increase the unavailability or unreliability of LSS components, consistent with that approved for Vogtle (Reference 9). Consistent with the NEI 00-04 guidance, Callaway Plant, Unit No. 1 will perform both an initial sensitivity study and a cumulative sensitivity study. 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 importance of the SSC(s).

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 performing qualitative screening of model uncertainties is found in Section 5.3 of NUREG-1855 (Revision 1) and in Section 3.1.1 of EPRI TR-1016737 (Reference 14). The process described in these references was mostly developed to evaluate the uncertainties associated with the internal events PRA model; however, the approach can be applied to other types of hazard groups.

The list of assumptions and sources of uncertainty were reviewed to identify those that would be significant for the evaluation of this application. If the Callaway Plant, Unit No. 1 PRA model used a non-conservative treatment or methods that are not commonly

Enclosure 1 to ULNRC-06550 Page 18 of 32 accepted, the underlying assumption or source of uncertainty was reviewed to determine its impact on this application. Only those assumptions or sources of uncertainty that could significantly impact the risk calculations were considered key for this application. To identify these assumptions and sources of uncertainty, both plant-specific and generic sources of uncertainty (as identified in EPRI TR-1016737) were considered. All PRA notebooks were reviewed, and sources of uncertainty were compiled and characterized. The identification and characterization of the sources of uncertainty was performed consistent with the requirements of the ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009). This evaluation meets the intent of steps C-1 and E-1 of NUREG-1855, Revision 1.

To assess the impact of sources of uncertainties on 10 CFR 50.69 system categorizations, a review of the base case sources of uncertainty for the Callaway Plant, Unit No. 1 PRA models was performed. Each identified uncertainty was evaluated with respect to its potential to significantly impact the risk ranking evaluations that will be performed to support the categorization effort. This evaluation meets the intent of the screening portion for steps C-2 and E-2 of NUREG-1855, Revision 1.

Callaway Plant, Unit No. 1 PRA model specific assumptions and sources of uncertainty for this application were identified and dispositioned. Attachment 6, "Key Assumptions and Sources of Uncertainty," (provided with this enclosure) documents the conclusion of this review, which found that no additional sensitivity analyses are required to address the Callaway Plant, Unit No. 1 model specific assumptions or sources of uncertainty.

At the time of this submittal, all open Finding-level facts and observations (F&Os) on the Callaway Plant, Unit No. 1 Internal Events, Internal Flood, Fire, High Winds and Seismic PRA models have been closed using the process documented in Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13 as accepted by the NRC (per Reference 18). The results of these independent assessments have been documented and are available for NRC audit; therefore, no additional sensitivities are required to address open Finding F&Os against the Callaway Plant, Unit No. 1 PRA models.

As discussed under the Fire PRA description in section 3.3, the resolution of Suggestion F&O FSS-B1-03 was determined to be an upgrade. This upgrade has not yet been reviewed by an independent assessment; if it remained unreviewed this may represent an uncertainty in the FPRA results as they relate to this application. However, Ameren Missouri will conduct a focused scope peer review on this upgrade and close any new Finding F&Os, using an endorsed closure process, prior to using the Fire PRA to support implementation of any alternative treatments under the 50.69 application. Therefore, no additional sensitivities are required to address the current upgrade in the Callaway Plant, Unit No. 1 Fire PRA model. This new commitment is documented in Enclosure 4 to this application.

Enclosure 1 to ULNRC-06550 Page 19 of 32 3.2.9 Modeling of Flex 3.2.9.1 Background The NRC has been issuing a "generic" Request for Additional Information (RAI) regarding crediting of FLEX equipment in PRA models. (See References 34 and 35 for more detail.) The Limerick RAI (Reference 34) is summarized below.

The NRC memorandum dated May 30, 2017, "Assessment of the Nuclear Energy Institute 16-06,

'Crediting Mitigating Strategies in Risk-Informed Decision Making, ' Guidance for Risk-Informed Changes to Plants Licensing Basis" (ADAMS Accession No. ML17031A269), provides the NRC's staff assessment of the challenges of incorporating diverse and flexible (FLEX) coping strategies and equipment into a PRA model in support of risk-informed decision-making in accordance with the guidance of RG 1.200, Revision 2 (ADAMS Accession No. ML090410014). Docketed information does not indicate if [PLANT NAME] has credited FLEX equipment or actions in the

[PRA MODEL]. As such, please address the following:

a. Discuss whether [UTILITY] has credited FLEX equipment or mitigating actions into the

[PLANT NAME] [PRA MODEL]. If not incorporated or their inclusion is not expected to impact the PRA results used in the RfCT program, no additional response is requested.

b. If FLEX equipment or operator actions have been credited in the PRA, address the following, separately for the internal events (including internal flooding), and other PRAs.

i. Summarize the supplemental equipment and compensatory actions, including FLEX strategies that have been quantitatively credited for each of the PRA models used to support this application. Include discussion of whether the credited FLEX equipment is portable or permanently installed equipment.

ii. Discuss whether the credited equipment (regardless of whether it is portable or permanently-installed) are like other plant equipment (i.e. SSCs with sufficient plant-specific or generic industry data) and whether the credited operator actions are similar to other operator actions evaluated using approaches consistent with the endorsed ASMEIANS RA-Sa-2009 PRA Standard.

iii. If any credited FLEX equipment is dissimilar to other plant equipment credited in the PRA (i.e., SSCs with sufficient plant-specific or generic industry data), discuss the data and failure probabilities used to support the modeling and provide the rationale for using the chosen data. Discuss whether the uncertainties associated with the parameter values are in accordance with the ASMEIANS PRA Standard as endorsed by RG 1.200, Revision 2.

iv. If any operator actions related to FLEX equipment are evaluated using approaches that are not consistent with the endorsed ASMEIANS RA-Sa-2009 PRA Standard (e.g. using surrogates), discuss the methodology used to assess human error probabilities for these operator actions. The discussion should include:

1. A summary of how the impact of the plant-specific human error probabilities and associated scenario-specific pe,formance shaping factors listed in (a)-(j) of supporting requirement HR-G3 of the ASMEIANS RA-Sa-2009 PRA Standard were evaluated.

2. Whether maintenance procedures for the portable equipment were reviewed for possible pre-initiator human failures that renders the equipment to ULNRC-06550 Page 20 of 32 unavailable during an event, and if the probabilities of the pre-initiator human failure events were assessed as described in HLR-HR-D of the ASME/ANS RA-Sa-2009 PRA Standard.

3./f the procedures governing the initiation or entry into mitigating strategies are ambiguous, vague, or not explicit, a discussion detailing the technical bases for probability offailure to initiate mitigating strategies.

c. The ASMEIANS RA-Sa-2009 PRA Standard defines PRA upgrade as the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences. Section 1-5 of Part 1 of ASMEIANS RA-Sa-2009 PRA Standard states that upgrades of a PRA shall receive a peer review in accordance with the requirements specified in the peer review section of each respective part of this Standard. Provide an evaluation of the model changes associated with incorporating mitigating strategies, which demonstrates that none of the following criteria is satisfied: (1) use of new methodology, (2) change in scope that impacts the significant accident sequences or the significant accident progression sequences, and (3) change in capability that impacts the significant accident sequences or the significant accident progression sequences.

d. Section 2.3.4 of NE! 06-09, Revision 0-A, states that PRA modeling uncertainties shall be considered in application of the PRA base model results to the RfCT program. The NRC SE for NE! 06-09, Revision 0, states that this consideration is consistent with Section 2.3.5 of RG 1.177, Revision 1. NE! 06-09, Revision 0-A, further states that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties which could potentially impact the results of a RfCT calculation.

NRC staff notes that the impact of model uncertainty could vary based on the proposed RICTs. NE! 06-09, Revision 0-A, also states that the insights from the sensitivity studies should be used to develop appropriate compensatory RMAs including highlighting risk significant operator actions, confirming availability and operability of important standby equipment, and assessing the presence of severe or unusual environmental conditions.

Uncertainty exists in modeling FLEX equipment and actions related to assumptions regarding the failure probabilities for FLEX equipment used in the model, the corresponding operator actions, and pre-initiator failure probabilities. Therefore, FLEX modeling assumptions can be key assumptions and sources of uncertainty for RICTs proposed in this application. In light of these observations:

i. Describe the sensitivity studies that will be used to identify the RICTs proposed in this application for which FLEX equipment and.lor operator actions are key assumptions and sources of uncertainty (e.g., use of generic industry data for non-safety related equipment). Explain andjustify the approach (e.g., any multipliers for failure probabilities) used to perform the sensitivity studies.

ii. Described how the results of the sensitivity studies which identify FLEX equipment and/or operator actions as key assumptions and sources of uncertainty will be used to identify RMAs prior the implementation of the RICT program, consistent with the guidance in Section 2.3.4 of NE! 06-09, Revision 0-A.

iii. Demonstrate the approaches described in items (i) and (ii) above using an example sensitivity study for the nominal configuration of a proposed RfCT where the FLEX equipment and/or operator actions are identified as key assumptions and sources of uncertainty.

Enclosure 1 to ULNRC-06550 Page 21 of 32 Section 3.2.9.2 (below) provides responses, as applicable, to the above questions regarding modeling of FLEX equipment in the Callaway Plant, Unit 1 PRA models. The responses are provided in a consolidated form instead of individual responses to each question.

3.2.9.2 Discussion FLEX strategies are credited in the Callaway internal events (IE) and seismic PRA (S-PRA) models. Specifically, three strategies are modeled:

1. FLEX SG Makeup AFW Pump - FLEX pumps are used to supply water to the steam generators when all other sources fail. This strategy is intended as a backup to the TDAFW pump and is not expected to be available for SG makeup until the Phase 2 FLEX staffing is available at the 6-hour point in the scenario. This strategy involves portable equipment.

2. 480 VAC Portable Backup Generators - FLEX diesel generators are used to supply power to battery chargers. Specifically, the backup generators are used to supply power to the battery chargers for 125 VDC buses NK0l, NK02, and NK04. This strategy involves portable equipment.

3. HCST Credit - Realignment of the turbine-driven AFW pump (TDAFP) recirculation from the condensate storage tank (CST) to the hardened condensate storage tank (HCST),

and isolation of MDAFP recirculation. This strategy involves permanently installed equipment.

The first two strategies discussed above are included in the S-PRA model. Each strategy is modeled as a single basic event (i.e., component failures and human errors aren't explicitly modeled) and assigned a total failure probability of 0.99. The failure probability of 0.99 is conservative and ensures minor, if not negligible, credit is taken for FLEX. The value is used as a place holder for failure of the strategy until industry guidance is available to explicitly model component failures and human errors.

Sensitivity studies are performed in the S-PRA model to understand the impact of crediting FLEX. More specifically, since the FLEX strategies are assumed to have a failure probability of 0.99, sensitivity studies are performed to understand any benefit of reducing the failure probability of the strategy. Assuming a failure probability of 1.0 is expected to have negligible difference on the S-PRA model results as compared to the current treatment.

Additionally, it should be noted that these portable equipment FLEX strategies are incorporated into the baseline PRA model used to support the S-PRA model. This was done in preparation for a one-top all hazards model. When quantifying hazards other than seismic, these strategies are assumed failed (i.e., set to TRUE).

Regarding the third strategy discussed above, in addition to the normal supply to each auxiliary feedwater pump from the CST, the hardened condensate storage tank (HCST) serves as an additional back-up source of water. Swap-over to the HCST would provide an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of AFW supply. For the safety-related AFW pumps, swap over to the HCST is automatic and is provided by a single air-operated, fail-open valve which is operated on low suction to ULNRC-06550 Page 22 of 32 pressure. The HCST may also be aligned to the non-safety AFW pump, but this alignment requires a local manual operator action and is not currently credited in the PRA. The HCST swap over to the safety-related AFW pumps is credited in the IE PRA model (and therefore used in all other hazard models).

The incorporation of the FLEX strategies into the PRA model does not constitute a PRA model upgrade. Modeling inclusion of FLEX has been performed in a manner that:

• Is consistent with other modeling aspects used in the PRA model

• Is commensurate with the supporting requirements of the ASME/ ANS PRA Standard

• Does not add any additional scope to the PRA

• Does not and any new capability of the PRA

• Does not significantly impact significant accident sequences or accident sequence progression 3.3 PRA REVIEW PROCESS RESULTS (10 CFR S0.69(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, including the clarifications included therein, and consistent with NRC RIS 2007-06. All F&O closure reviews were performed in accordance with the process documented in Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13 as accepted by the NRC (Reference 18), including the NRC clarifications and specific documentation expectations, as well as the requirements published in the ASME/ANS PRA Standard (RA-Sa-2009).

INTERNAL EVENTS AND INTERNAL FLOOD PRA MODEL The information provided in this section demonstrates that the Callaway Plant, Unit 1 internal events PRA model (including internal flooding) meets the expectations for PRA scope and technical adequacy as presented in ASME/ANS RA-Sa-2009 and RG 1.200, Revision 2.

The Internal Events/ Internal Flooding PRA was peer reviewed in April 2019. This peer review was a full-scope review of the technical elements of the internal events and internal flooding at-power PRA as documented in PWROG-19012-P (Reference 22).

An Independent Assessment of F&Os was conducted in November 2019 and documented in PWROG-19034-P (Reference 24). The scope of the assessment included all Facts and Observations (F&Os) generated in the April 2019 peer review.

All F&Os except for one were closed. The remaining F&O was related to implementation of the methodology provided in PWROG-18027-NP (Reference 32) for assessing the loss of room cooling in PRA modeling. Following, but unrelated to, incorporation of the method provided in PWROG-18027-NP into the Callaway PRA, this method was chosen by the PWROG and NEI to pilot the Newly Developed Methods to ULNRC-06550 Page 23 of 32 (NDM) peer review process established in NEI 17-07 (Reference 33). Despite the Callaway assessment, and acknowledgement by the PWROG, that the method provided in PWROG-18027-NP did not necessarily meet the definition of a NDM, Callaway decided to suspend resolution of the associated F&O until the NDM peer review and closure of any F&Os were completed using the process established in NEI 17-07. Also during the November 2019 independent assessment, two F&O resolutions were determined to be upgrades to the Internal Events/ Internal Flooding PAA. Thus, a focused-scope peer review was required. Based on this focused scope peer review, one new Internal Events F&O was generated.

During February and March 2020 a new peer review, following the guidance in NEI 17-07 Revision 2, was conducted on the method provided in PWROG-18027-NP and documented in PWROG-19020-NP (Reference 31 ). Based on the results of this review all applicable NDM attributes are met at CC 1/11 and there are no open peer review Findings against the method in PWROG-18027-NP.

In June 2020, an independent assessment of F&O resolution and a focused scope peer review, completing the review of PWROG-18027-NP implementation, were conducted on the Callaway Plant internal events and fire PAA models. The focused scope peer review determined that all of the SRs that were examined, including the SR associated with the F&O related to implementation of the method in PWROG-18027-NP, satisfy CC II, or higher, requirements as documented in AMN#PES00031-REPT-001 (Reference 29). The independent assessment of F&Os included an assessment of all remaining open F&O Findings. The results of this review are documented in AMN#PES00031-REPT-002 (Reference 30).

There are no open peer review Findings for the Internal Events/ Internal Flooding PAA models.

HIGH WINDS PRA MODEL The High Winds PAA was peer reviewed in April 2019 and documented in PWROG-19022-P (Reference 23). The scope of this work was to review the Callaway External Hazards Screening Assessment and High Winds PAA against the technical elements in Sections 6 and 7 of the ASME/ANS RA-Sa-2009 Standard, and in RG 1.200, Revision 2.

An Independent Assessment of F&O resolution was conducted in November 2019 and documented in PWROG-19034-P (Reference 24). The scope of the assessment included all F&Os generated in the April 2019 peer review. All F&Os were closed.

There are no open peer review Findings for the Other External Hazards Screening or the High Winds PAA model.

Enclosure 1 to ULNRC-06550 Page 24 of 32 SEISMIC PRA MODEL The Seismic PRA was peer reviewed in June 2018 and documented in PWROG-18044-P (Reference 25). This peer review was conducted against the requirements of the Code Case for ASME/ANS RA-Sb-2013, as amended by the Nuclear Regulatory Commission on March 12, 2018. The Code Case is an approved alternative to Part 5 of ASME/ANS RA-Sb-2013 Addendum B, the American Society of Mechanical Engineers (ASME) / American Nuclear Society (ANS) Probabilistic Risk Assessment (PRA)

Standard.

An Independent Assessment of F&Os was conducted in March 2019. The scope of the assessment included all but two of the F&Os generated in the June 2018 peer review.

All in-scope F&Os were closed as documented in PWROG-19011-P (Reference 26).

Also in the March 2019 review documented in PWROG-19011-P, three SRs were the subject of a focused-scope peer review based on the closures of associated F&Os being assessed as upgrades. As a result of that peer review, the three SRs were determined to be met at CC II.

Subsequently, another Independent Assessment of F&Os was conducted in June 2020 and documented in AMN#PES00031-REPT-002 (Reference 30). The scope of the assessment included all remaining F&Os generated in the June 2018 peer review. All F&Os were closed.

There are no open peer review Findings for the Seismic PRA model.

FIRE PRA MODEL The Fire PRA was prepared using the methodology defined in NUREG/CR-6850, "Fire PRA Methodology for Nuclear Power Facilities," to support a transition to National Fire Protection Association (NFPA) Standard 805, "Performance Based Standard for Fire Protection for Light ":later Reactor Electric Generating Plants."

The Fire PRA was peer reviewed to ASME/ANS RA-Sa-2009 and RG 1.200 Revision 2 in October 2009. The review is documented in LTR-RAM-11-10-019 (Reference 27).

An Independent Assessment of F&Os was conducted in June 2019 and documented in AMN#PES00021-REPT-001 (Reference 28).

In June 2020, an independent assessment of F&Os and a focused scope peer review were held for the Callaway Plant internal events and Fire PRA models. The focused scope peer review generated additional Fire PRA related F&Os as documented in AMN#PES00031-REPT-001 (Reference 29). The independent assessment of F&Os included an assessment of all remaining open F&O Findings. As documented in AMN#PES00031-REPT-002 (Reference 30), all Finding F&Os were closed, including the Fire PRA Findings identified in the Focused Scope peer review.

Enclosure 1 to ULNRC-06550 Page 25 of 32 There are no open peer review Findings for the Fire PRA model.

The discussion above, along with the referenced reports, demonstrates that the PRA is of sufficient quality and level of detail to support the categorization process, and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC as required 10 CFR 50.69(c)(l)(i).

While all Finding F&Os against the Fire PRA are closed, the resolution of Suggestion F&O FSS-81-03 was determined to be an upgrade during the F&O closure review documented in ANM#PES00021-RPT-001 (Reference 28). Being outside the scope of the closure review, the upgrade was not reviewed at the time of identification and has not yet been reviewed by an independent peer review.

FSS-81-03 relates to documentation associated with the original use of screening values for quantifying human error probabilities during Main Control Room (MCR) abandonment scenarios. To resolve FSS-81-03 and also close NFPA-805 LAR Table S-3 Implementation item 13-805-001, the Fire PRA was updated to perform the Human Reliability Analysis (HRA) evaluations for MCR abandonment using detailed assessments within HRA Calculator rather than the screening values discussed in this F&O. The HRA evaluations were performed using the NUREG-1921 methodology and quantified using the HRA Calculator methods used for other existing human failure events (HFEs) in the Fire PRA. This change was determined by a recent independent peer review team to be an upgrade.

PRA scope and technical adequacy guidance requires upgrades to be independently peer reviewed and Finding F&Os to be closed, or dispositioned for a given application.

Therefore, Ameren Missouri will conduct a focused scope peer review on this upgrade and close any new Finding F&Os, using an endorsed closure process, prior to using the Fire PRA to support implementation of any alternative treatments under the 50.69 application. This new commitment is presented in Enclosure 4 to this application.

Completion of this commitment will ensure the Fire PRA is of sufficient quality and level of detail to support the categorization process, and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC as required 10 CFR 50.69(c)(l)(i).

3.4 RISK EVALUATIONS {10 CFR S0.69{b){2){iv))

The Callaway Plant, Unit No. 1, 10 CFR 50.69 categorization process will implement the guidance in NEI 00-04. The overall risk evaluation process described in the Nuclear Energy Institute 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 and large early release

Enclosure 1 to ULNRC-06550 Page 26 of 32 frequency. The failure rates for equipment and initiating event frequencies used in the PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, and human errors).

Subsequent performance monitoring and PRA updates required by the rule will continue to capture this data and provide timely insights into the need to account for any important new degradation mechanisms.

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

Scheduled periodic reviews, at a frequency of at least once every two refueling cycles, will evaluate new insights resulting from available risk information (i.e., PRA model or other analysis used in the categorization) changes, design changes, operational changes, and SSC performance. If it is determined that these changes have affected the risk information or other elements of the categorization process such that the categorization results are more than minimally affected, then the risk information and the categorization process will be updated. This review will include:

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

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

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

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

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

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

Enclosure 1 to ULNRC-06550 Page 27 of 32 3.6 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA The following NRC requirements and guidance documents are applicable to the proposed change.

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

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

Revision 1, May 2006.

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

Revision 2, April 2015.

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

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

3.7 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS Ameren Missouri proposes to modify the licensing basis to allow for the voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance, resulting in improved plant safety.

Ameren Missouri has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

to ULNRC-06550 Page 28 of 32

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations.

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

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations.

The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed.

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment

Enclosure 1 to ULNRC-06550 Page 29 of 32 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, Ameren Missouri 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.

3.8 CONCLUSION

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

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

to ULNRC-06550 Page 30 of 32 5 REFERENCES

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

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

Revision 1, May 2006

3. Internal Events (including Internal Flooding) PRA Model of Record 8.00, Update 8

4. Fire PRA Model of Record 8.00, Update 8

5. Seismic PRA Model of Record 8.00, Update 8

6. High Winds PRA Model of Record 8.00, Update 8

7. NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management," December 1991

8. NRC letter to Southern Nuclear Operating Company, "Issuance of Amendments Re:

Use of 10 CFR 50.69 (TAC Nos. ME9472 and ME9473)", dated December 17, 2014 (ADAMS Accession No. ML14237A034)

9. 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,"' April 22, 2009 (TAC NO. MD5250) (ADAMS Accession No. ML090930246)

10. Callaway Energy Center Report PRA-OEH-ANALYSIS, "Other External Hazards:

Screening Assessment Notebook," Update 8

11. 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

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

13. NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," Revision 1, March 2017

14. EPRI TR-1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," December 2008 to ULNRC-06550 Page 31 of 32

15. EPRI TR-1026511, "Practical Guidance on the use of Probabilistic Risk Assessment in Risk Informed Applications with a Focus on the Treatment of Uncertainty,"

December 2012

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

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

18. NRC Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os)," May 3, 2017, (Accession No. ML17079A427)

19. NRC Letter to Mr. Oliver Martinez, "U.S. Nuclear Regulatory Commission (NRC)

Comments on 'Addenda to a Current ABS: ASME RA-SB - 20XX, 'Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications"'.

20. NEI 12-13, "External Hazards PRA Peer Review Process Guidelines," Revision O, Nuclear Energy Institute, August 2012

21. NRC Letter to Mr. Biff Bradley (NEI), "U.S. Nuclear Regulatory Commission Comments on Nuclear Energy Institute 12-13, 'External Hazards PRA Peer Review Process Guidelines,' Dated August 2012," November 16, 2012, (Accession No. ML12321A280)

22. PWROG-19012-P, "Peer Review of the Callaway Internal Events and Internal Flood Probabilistic Risk Assessment," April 2019

23. PWROG-19022-P, "Peer Review of the Callaway External Hazard Screening Assessment and High Winds Probabilistic Risk Assessment," April 2019

24. PWROG-19034-P, "Independent Assessment of Facts & Observations Closure and Focused Scope Peer Review of the Callaway Probabilistic Risk Assessments,"

November 2019

25. PWROG-18044-P, "Peer Review of the Callaway Seismic Probabilistic Risk Assessment," June 2018 to ULNRC-06550 Page 32 of 32

26. PWROG-19011-P, "Independent Assessment of Facts & Observations Closure and Focused Scope Peer Review of the Callaway Seismic Probabilistic Risk Assessment," March 2019

27. LTR-RAM-II-10-019, "Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements From Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for The Callaway Nuclear Plant Fire Probabilistic Risk Assessment," October 2009

28. AMN#PES00021-REPT-001, "Callaway Energy Center Fire Probabilistic Risk Assessment Peer Review F&Os Closure," June 2019

29. AMN#PES00031-REPT-001, "Callaway Energy Center Probabilistic Risk Assessment Focused Scope Peer Review," July 2020

30. AMN#PES00031-REPT-002, "Callaway Energy Center Probabilistic Risk Assessment Peer Review F&Os Closure," July 2020

31. PWROG-19020-NP Revision 1, "Newly Developed Method Peer Review Pilot -

General Screening Criteria for Loss of Room Cooling in PRA Modeling," April 2020

32. PWROG-18027-NP Revision 0, "Loss of Room Cooling in PRA Modeling," April 2020.

33. NEI 17-07, Revision 2, "Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard," July 2019 (ADAMS Accession No. ML19228A242)

34. ML19192A031, "2019/08/10 NRR E-mail Capture - Limerick-Request for Additional Information: Risk-Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Completion Times - RITSTF Initiative 4B" (EPID L-2018-LLA-0567)

35. ML20192A144, "2020/07/07 NRR E-mail Capture - RE: Request for Additional Information RE: Prairie Island License Amendment Request to Adopt TSTF-505 (EPID: L-2019-LLA-0283) to ULNRC-06550 Page 1 of 2 Attachment 1: List of Categorization Prerequisites The PRA model to be used for categorization credits no modifications to achieve an overall core damage frequency (CDF) and large early release frequency (LERF) consistent with U.S. Nuclear Regulatory Commission Regulatory Guide 1.174 risk limits that were not adequately addressed in the PRAs proposed for application regarding the Callaway Plant, Unit No. 1, 10 CFR 50.69 Program.

Ameren Missouri will establish a procedure(s) prior to the use of the categorization process on a plant system. The procedure(s) will contain the elements/steps listed below.

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

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

• Component safety significance assessment. The safety significance of active components is assessed through a combination of probabilistic risk assessment (PRA) and non-PRA methods, covering all hazards. The 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, are upgraded to HSS.

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

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

to ULNRC-06550 Page 2 of 2

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

• Documentation requirements per Section 3.1.1 of Enclosure 1.

to ULNRC-06550 Page 1 of 2 Attachment 2: Description of PRA Models Used in Categorization Unit Model Baseline Baseline Comments CDF LERF For Peer Review Report and Internal Events Associated F&O Closure (Excluding 1 5.79E-06 9.24E-08 Information, see References Internal 1, 3, 8, 9 and 10 on page 2 Flooding) PRA of this attachment.

For Peer Review Report and Associated F&O Closure Internal 1 7.07E-06 1.64E-08 Information, see References Flooding PRA 1 and 3 on page 2 of this attachment.

For Peer Review Report and Associated F&O Closure 1 Fire PRA 7.40E-06 2.79E-08 Information, see References 6, 7, 8 and 9 on page 2 of this attachment.

For Peer Review Report and Associated F&O Closure 1 Seismic PRA 5.59E-05 2.82E-06 Information, see References 4, 5 and 9 on page 2 of this attachment.

For Peer Review Report and Associated F&O Closure High Winds 1 5.60E-06 2.6E-07 Information, see References PRA 2 and 3 on page 2 of this attachment.

Total 8.18E-05 3.22E-06 Aggregate Risk to ULNRC-06550 Page 2 of 2 REFERENCES

1. PWROG-19012-P, "Peer Review of the Callaway Internal Events and Internal Flood Probabilistic Risk Assessment"

2. PWROG-19022-P, "Peer Review of the Callaway External Hazard Screening Assessment and High Winds Probabilistic Risk Assessment"

3. PWROG-19034-P, "Independent Assessment of Facts & Observations Closure and Focused Scope Peer Review of the Callaway Probabilistic Risk Assessments"

4. PWROG-18044-P, "Peer Review of the Callaway Seismic Probabilistic Risk Assessment"

5. PWROG-19011-P, "Independent Assessment of Facts & Observations Closure and Focused Scope Peer Review of the Callaway Seismic Probabilistic Risk Assessment"

6. LTR-RAM-11-10-019, "Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements From Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications For The Callaway Nuclear Plant Fire Probabilistic Risk Assessment"

7. AMN#PES00021-REPT-001, "Callaway Energy Center Fire Probabilistic Risk Assessment Peer Review F&Os Closure"

8. AMN#PES00031-REPT-001, "Callaway Energy Center Probabilistic Risk Assessment Focused Scope Peer Review"

9. AMN#PES00031-REPT-002, "Callaway Energy Center Probabilistic Risk Assessment Peer Review F&Os Closure"

10. PWROG-19020-NP Revision 1, "Newly Developed Method Peer Review Pilot -

General Screening Criteria for Loss of Room Cooling in PRA Modeling" to ULNRC-06550 Page 1 of 1 Attachment 3: Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items There are no open Peer Review Findings for the Callaway Internal Events, Internal Flood, Fire, Seismic, and High Winds PRA models or for the screening of Other External Hazards.

to ULNRC-06550 Page 1 of 5 Attachment 4: External Hazards Screening Screening Result External Screening Hazard Screened?

Criterion Comment (Y/N)

(Note 1)

Airports, military installations, and flight corridors have been considered. There are three low-altitude airways and three high-Aircraft Impact y PS4 altitude airways that pass near the plant. A bounding analysis of aircraft impact associated with these airways results in a CDF

< lE-6/yr.

y Not applicable to the site because of climate Avalanche C3 and topography.

Service Water and Essential Service Water systems include traveling water screens and Biological backwash strainers to prevent intake of Event y cs foreign matter. These design features would provide sufficient time to detect and mitigate the hazard.

The site is not in proximity to any ocean or Coastal y C3 large body of water; therefore, costal erosion Erosion is not an applicable hazard.

y Plant design eliminates drought as a concern.

Drought Cl, CS In addition, this event is slowly developing.

External flooding is incorporated into High Tide, Lake Level, or River Stage, Seiche, External y Cl,C3 Tsunami, Storm Surge, and Waves. Local Flooding Intense Precipitation (LIP) was analyzed and shown to be within the plant design basis.

Extreme Wind Callaway Plant, Unit No. 1 developed a High N None or Tornado Winds PRA that addresses this hazard.

y Negligible impact on the plant. Implicitly Fog C4 included in air, land, and water transportation.

to ULNRC-06550 Page 2 of 5 Screening Result External Screening Hazard Screened?

Criterion Comment (Y/N)

(Note 1)

Limited occurrence and bounded by other events, primarily loss of offsite power (LOOP),

Forest or y for which the plant is designed. Not Cl, C3, C4 Range Fire applicable to site due to limited vegetation and inability of hazard to propagate into Protected Area.

Limited occurrence and bounded by other Frost y Cl, C4 events for which the plant is designed. Frost impacts covered under ice and snow hazards.

Limited occurrence and bounded by other events for which the plant is designed.

y Consequences of this hazard are bounded by Hail C4 a loss of offsite power (LOOP), which is included in the Internal Events PRA weather-induced LOOP frequency.

Plant is designed for this hazard. Associated High Summer y plant trips have not occurred and are not Cl, CS Temperature expected. In addition, this event is slowly developing.

The site is not in proximity to any oceans or High Tide, y lakes, and is not susceptible to flooding by Lake Level, or C3 rivers. This hazard is not applicable due to River Stage the site location.

y Covered under Extreme Wind or Tornado, Hurricane C4 Intense Precipitation, and Storm Surges.

y Plant is designed for freezing temperatures, Ice Cover Cl which are infrequent and short in duration.

The only industrial or military facility within 5 miles of the site is Mertens Quarry, located 4.5 miles northwest of the site. The amount Industrial or of explosive material required to potentially Military Facility y C3 damage the plant is significantly greater than Accident the amount that would ever be stored at a quarry. Therefore, industrial or military facility accidents cannot occur close enough to the plant to affect it.

to ULNRC-06550 Page 3 of 5 Screening Result External Screening Hazard Screened?

Criterion Comment (Y/N)

{Note 1)

Callaway Plant, Unit No. 1 developed an Internal N None Internal Flooding PRA that addresses this Flooding hazard.

Callaway Plant, Unit No. 1 developed an Internal Fire N None Internal Fire PRA that addresses this hazard.

The site is located on a plateau, and there is y no significantly higher ground within 5 miles.

Landslide C3 The topography of the site precludes this hazard.

Lightning strikes causing loss of offsite power or turbine trip are contributors to the initiating y event frequencies for these events. However, Lightning Cl other causes are also included. The impacts are no greater than already modeled in the internal events PRA.

Low Lake y The plant is designed for such events and Level or River Cl,CS their impacts are slow to develop. '

Stage Extended freezing temperatures are rare the Low Winter y Cl plant is designed for such events I and th~ir Temperature

• impacts are slow to develop.

The frequency of meteorites greater than Meteorite or y 100 lb striking the plant is around lE-8/y and Satellite PS4 corresponding satellite impacts is around Impact 2E-9/y.

There are no pipelines or tank farms within 5 Pipeline y miles of the site. Pipelines are not close C3 Accident enough to significantly impact plant structures.

The plant is operated and designed for such Release of y events. No control room habitability problems Chemicals in Cl from the potential release of hazardous Onsite Storage chemicals have been identified.

to ULNRC-06550 Page 4 of 5 Screening Result External Screening Hazard Screened?

Criterion Comment (Y/N)

(Note 1)

The Missouri River is strictly managed and highly regulated. It is extremely improbable that naturally occurring or man-made River Diversion y Cl, CS diversions would be allowed to continue unchecked or uncontrolled. Even so, the plant is designed for such events, and their impacts are slow to develop.

There are no recorded instances of sandstorms affecting Callaway county or any Sand or Dust y neighboring counties. The plant is designed Cl, C3 Storm for such events. Also, a procedure instructs operators to replace filters before they become inoperable.

Site is not located near any bodies of water y for which seiche flooding would apply. Onsite Seiche C3 reservoirs and spray ponds are designed for seiches.

Seismic Callaway Plant, Unit No. 1 developed a N None Activity Seismic PRA that addresses this hazard.

The event damage potential is less than other events for which the plant is designed.

Consequences of this hazard are bounded by Snow y Cl a loss of offsite power {LOOP), which is included in the Internal Events PRA weather-induced LOOP frequency. Potential flooding impacts covered under external flooding.

Soil y The potential for this hazard is low at the site; Shrink-Swell Cl the plant design considers this hazard.

Consolidation The site is not located near any large bodies y of water for which storm surge flooding would Storm Surge C3 apply. This hazard is not applicable to the site because of location.

Toxic gas covered under release of chemicals Toxic Gas y C4 in onsite storage, industrial or military facility accident, and transportation accident.

to ULNRC-06550 Page 5 of 5 Screening Result External Screening Hazard Screened?

Criterion Comment

{Y/N)

{Note 1)

Road and highway, railroad, and water transport cannot occur close enough to the plant to affect it. Water transport collisions with intake structure are of lesser potential Transportation y Cl, C3, C4, damage than the events for which the plant Accident cs was designed, and would provide sufficient time to respond. Aviation and pipeline accidents covered under those specific categories.

The site is not located near any large bodies y of water for which tsunami flooding would Tsunami C3 apply. This hazard is not applicable to the site because of location.

The event is of equal or lesser damage potential than the events for which the plant Turbine-y has been designed, and the core damage Generated Cl,PS4 frequency (calculated using a bounding or Missiles demonstrably conservative analysis) has a mean frequency <lE-6/yr.

The site is not located near any active Volcanic y C3 volcano. This hazard is not applicable to the Activity site because of location.

The plant design considers this hazard. Water Waves y Cl levels in the UHS retention pond after wave run-up do not reach the critical water level.

Note 1: See Attachment 5 for descriptions of the screening criteria.

A detailed description of the external events screening and evaluation process is presented in Callaway Plant, Unit No. 1, report PRA-OEH-ANALYSIS, "Other External Hazards: Screening Assessment Notebook," Update 8.

to ULNRC-06550 Page 1 of 2 : Progressive Screening Approach for Addressing External Hazards Event Criterion Source Comments Analysis Initial C1. Event damage potential NUREG/CR-2300 and Preliminary is < events for which plant is ASME/ANS Standard Screening designed. RA-Sa-2009 C2. Event has lower mean NUREG/CR-2300 and frequency and no worse ASME/ANS Standard consequences than other RA-Sa-2009 events analyzed.

C3. Eventcannotoccur NUREG/CR-2300 and close enough to the plant to ASME/ANS Standard affect it. RA-Sa-2009 C4. Event is included in the NUREG/CR-2300 and Not used to screen.

definition of another event. ASME/ANS Standard Used only to include RA-Sa-2009 within another event.

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

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

PS2. Design basis for the NUREG-1407 and Not utilized herein event meets the criteria in the ASME/ANS Standard because NRC 1975 Standard Review RA-Sa-2009 conformance to the Plan (SRP). SRP does not guarantee that the CDF is less than 1x10-5 per year.

to ULNRC-06550 Page 2 of 2 Event Criterion Source Comments Analysis PS3. Design basis event NUREG-1407 as mean frequency is < 1E-5/y modified in and the mean conditional ASME/ANS Standard core damage probability is RA-Sa-2009

< 0.1.

PS4. Bounding mean GDF is NUREG-1407 and

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

A detailed description of the external events screening and evaluation process is presented in Callaway Plant, Unit No. 1, report PRA-OEH-ANALYSIS, "Other External Hazards: Screening Assessment Notebook," Update 8.

Enclosure 1 to ULNRC-06550 Page 1 of 3 Attachment 1: Key Assumptions and Sources of

, Uncertainty The Callaway Plant, Unit No. 1 list of assumptions and sources of L1ncertainty associated with Internal Events, Fire, Seismic, and High Winds PRA models were reviewed to identify those that would be significant for the evaluation of 10 CFR 50.69 application.

If the Callaway Plant, Unit No. 1 PRA model used a non-conservative treatment or methods that are not commonly accepted, the underlying assumption or source of uncertainty was reviewed to determine its impact on this application. Only those assumptions or sources of uncertainty that could significantly impact the risk calculations were considered key for this application.

To identify these assumptions and sources of uncertainty both plant-specific and generic sources of uncertainty (as identified in EPRI TR-1016737) were considered. All PRA notebooks were reviewed, and sources of uncertainty were compiled and characterized in the Callaway Plant, Unit No. 1 PRA Uncertainty Analysis Notebook (Reference 6 of this attachment). The identification and characterization of the sources of uncertainty was performed consistent with the requirements of the ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009). This evaluation meets the intent of steps C-1 and E-1 of NUREG-1855, Revision 1.

To assess the impact of sources of uncertainties on 10 CFR 50.69 system categorizations, a review of the base case sources of uncertainty for the Callaway Plant, Unit No. 1 PRA models was performed. As documented in the Callaway Plant, Unit No.

1 PRA Uncertainty Analysis Notebook APP 6 (Reference 7), each identified uncertainty was evaluated with respect to its potential to significantly impact the risk ranking evaluations that will be performed to support the categorization effort. Only the sources of uncertainties and related assumptions with the potential to challenge the risk ranking evaluation guidelines are considered key. This evaluation meets the intent of the screening portion for steps C-2 and E-2 of NUREG-1855, Revision 1.

Regulatory Guide 1.174, Revision 3 (Reference 1) 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, which is addressed in this attachment.

The guidance in NEI 00-04 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 (CCF), and maintenance probabilities) do not mask the SSC(s) importance. As part of the required 50.69 PRA categorization sensitivity cases directed by NEI 00-04, human error probabilities (HEPs) and common cause basic events are increased to their 95th percentile and also decreased to their 5th percentile values for all models

Enclosure 1 to ULNRC-06550 Page 2 of 3 applicable to this application. These results are capable of driving a component and respective functions HSS, and therefore, the uncertainty of the PRA modeled HEPs and CCFs are accounted for in the 50.69 application. Furthermore, the sensitivity study where all LSS component failure rates are increased by a factor of three accounts not only for postulated increases due to the removal of special treatment requirements but also addresses uncertainty in component performance related to various sources of epistemic uncertainty.

Callaway Plant, Unit No. 1 PRA model specific assumptions and sources of uncertainty for this application were identified and dispositioned. The results of the evaluation of PRA model sources of uncertainty, as described above, are evaluated relative to the 50.69 application in PRA Uncertainty Analysis Notebook APP 6 (Reference 7). The evaluation found that no assumptions or sources of uncertainty challenged the risk ranking evaluation guidelines of the 50.69 application. Therefore, there are no additional sensitivity analyses required to address the Callaway Plant, Unit No. 1 model specific assumptions or sources of uncertainty.

In addition, at the time of this submittal, all open Finding level facts and observations (F&Os) on the Callaway Plant, Unit No. 1 Internal Events, Internal Flood, Fire, High Winds and Seismic PRA models have been closed using the process documented in Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13 as accepted by the NRC (Reference 8). The results of these independent assessments have been documented and are available for NRC audit; therefore, no additional sensitivities are required to address open Finding F&Os against the Callaway Plant, Unit No. 1 PRA models.

Furthermore, as discussed under the Fire PRA description in section 3.3, the resolution of Suggestion F&O FSS-B1-03 was determined to be an upgrade. This upgrade has not yet been reviewed by an independent assessment. If it remained unreviewed, this may represent an uncertainty in the Fire PRA results as they relate to this application. As this issue is related to human error probability calculations, the standard sensitivities contained in the NEI 00-04 methodology would adequately address associated uncertainties. However, Ameren Missouri will conduct a focused scope peer review on this upgrade and close any new Finding F&Os using an endorsed closure process, prior to using the Fire PRA to support implementation of any alternative treatments under the 50.69 application. Therefore, no additional sensitivities are required to address the current upgrade in the Callaway Plant, Unit No. 1 Fire PRA model.

to ULNRC-06550 Page 3 of 3 REFERENCES

1. 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 (ADAMS Accession No. ML17317A256)

2. NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Final Report," Revision 1, March 2017

3. EPRI 1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," December 2008

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

December 2012

5. EPRI 1013491, "Guideline for the Treatment of Uncertainty in Risk-Informed Applications," October 2006

6. PRA-IE-UNCERT, "Probabilistic Risk Assessment (PRA), Uncertainty Analysis Notebook," Revision 000

7. PRA-IE-UNCERT Appendix 6, "Disposition of Key Uncertainties: Risk Informed Engineering Programs (10CFR50.69)," Revision 000.

a. NRC Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os)," May 3, 2017, (Accession No. ML17079A427) to ULNRC-06550 Page 1 of 1 Attachment 7: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-1/11, CC-11/111, and CC-I/II/III The Callaway Plant, Unit No.1, probabilistic risk assessments (PRAs) to be applied in the 10 CFR 50.69 Program were all performed and peer reviewed based on Regulatory Guide 1.200, Revision 2, standard requirements.

Therefore, this comparison is not required for the PRAs to be applied in this program.