GO2-21-133, License Amendment Request to Adopt 10 CFR 50.69, Risk Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

License Amendment Request to Adopt 10 CFR 50.69, Risk Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

ML21314A224
Person / Time
Site:	Columbia
Issue date:	11/09/2021
From:	Dittmer J Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GO2-21-133
Download: ML21314A224 (57)

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ENERGY J. Kent Dittmer Vice President, Engineering NORTHWEST P.O. Box 968, PE01 Richland, WA 99352-0968 Ph. 509.377.4248 l F. 509.377.2354 jkdittmer@energy-northwest.com November 9, 2021 10 CFR 50.90 GO2-21-133 10 CFR 50.69 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, "RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS

Dear Sir or Madam:

In accordance with the provisions of 10 CFR 50.69 and 10 CFR 50.90, Energy Northwest is requesting an amendment to the license of Columbia Generating Station (Columbia).

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

The enclosure to this letter provides the basis for the proposed change to the Columbia Operating License. The categorization process being implemented through this change is consistent with NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline," Revision 0, dated July 2005, which was endorsed by the Nuclear Regulatory Commission 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 the enclosure provides a list of categorization







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GO2-21-133 Page 2 of 3 prerequisites. Use of the categorization process on a plant system will only occur after these prerequisites are met.

Energy Northwest intends to submit a separate request titled "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b'" within the next two months using the same probabilistic risk assessment (PRA) models described in this license amendment request (LAR). Energy Northwest requests that the NRC coordinate their review of the PRA technical adequacy details in this LAR for both applications. This would reduce the number of Energy Northwest 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.

A proposed License Condition is described in Section 2.3 of the enclosure to this letter.

Energy Northwest requests approval of the proposed license amendment within one year of the submission date, with the amendment being implemented within 60 days.

In accordance with 10 CFR 50.91, Energy Northwest is notifying the State of Washington of this amendment request by transmitting a copy of this letter and enclosures to the designated State Official.

This letter and its enclosure contain no new commitments.

If there are any questions or if additional information is needed, please contact Mr. R.M. Garcia, Licensing Supervisor, at 509-377-8463.

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

Executed this ______  ! #

day of ___________ 2021.

Respectfully, J.Kent Dittmer Vice President, Engineering

Enclosure:

Evaluation of the Proposed Change







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GO2-21-133 Page 3 of 3 cc: NRC RIV Regional Administrator NRC NRR Project Manager NRC Senior Resident Inspector/988C CD Sonoda - BPA/1399 EFSECutc.wa.gov - EFSEC E Fordham - WDOH R Brice - WDOH L Albin - WDOH







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GO2-21-133 Enclosure Enclosure Evaluation of the Proposed Change TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 CURRENT REGULATORY REQUIREMENTS 2.2 REASON FOR PROPOSED CHANGE

2.3 DESCRIPTION

OF THE PROPOSED CHANGE

3.0 TECHNICAL EVALUATION

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

3.1.1 Overall Categorization Process 3.1.2 Passive Categorization Process 3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.59(b)(2)(ii))

3.2.1 Internal Events and Internal Flooding 3.2.2 Fire Hazards 3.2.3 Seismic Hazards 3.2.4 Other External Hazards 3.2.5 Low Power & Shutdown 3.2.6 Integral Assessment 3.2.7 PRA Maintenance and Updates 3.2.8 PRA Uncertainty Evaluations 3.2.9 Modeling of FLEX 3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii))

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

3.5 FEEDBACK AND ADJUSTMENT PROCESS

4.0 REGULATORY EVALUATION

4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA 4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS

4.3 CONCLUSION

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6.0 REFERENCES

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GO2-21-133 Enclosure LIST OF ATTACHMENTS Attachment 1: List of Categorization Prerequisites Attachment 2: Description of PRA Models Used in Categorization Attachment 3: Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items Attachment 4: External Hazards Screening Attachment 5: Progressive Screening Approach for Addressing External Hazards Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty Attachment 7: Marked-up Operating License Page ii







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GO2-21-133 Enclosure 1.0

SUMMARY

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

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

This deterministic approach establishes requirements for engineering margin and quality assurance in design, manufacture, and construction. In addition, it assumes that adverse conditions can exist (e.g., equipment failures and human errors) and establishes a specific set of design basis events (DBE). 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 (SSC) necessary to defend against the DBEs are defined as "safety-related," and these SSCs are the subject of many regulatory requirements, herein referred to as special treatments, designed to ensure that they are of high quality and high reliability and have the capability to perform during postulated design basis conditions. Treatment includes, but is not limited to, quality assurance, testing, inspection, condition monitoring, assessment, evaluation, and resolution of deviations.

The distinction between "treatment" and "special treatment" is the degree of NRC specification as to what must be implemented for particular SSCs or for particular conditions. Typically, the regulations establish the scope of SSCs that receive special treatment using one of three different terms: "safety-related," "important to safety," or "basic component." The terms "safety-related" and "basic component" are defined in the regulations, while "important to safety," used principally in the general design criteria (GDC) of Appendix A to 10 CFR Part 50, is not explicitly defined.

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GO2-21-133 Enclosure 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 (PRA) address credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for common cause failures. The probabilistic approach to regulation is an extension and enhancement of traditional regulation by considering risk in a comprehensive manner.

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

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

The determination of safety significance is performed by an integrated decision-making process, as described by NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0 (Reference 1), which uses both risk insights and traditional engineering insights. The safety functions include the design basis functions, as well as functions credited for severe accidents (including external events). Special or alternative treatment for the SSCs is applied as necessary to maintain functionality and reliability and is a function of the SSC categorization results and associated bases. Finally, periodic assessment activities are conducted to make adjustments to the categorization and/or treatment processes as needed so that SSCs continue to meet all applicable requirements.

The rule does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility. Instead, the rule enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as high safety significant, existing treatment requirements are maintained or enhanced. Conversely, for SSCs that do not significantly contribute to plant safety on an individual basis, the rule allows an alternative risk-informed approach to treatment that provides a Page 2 of 52







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GO2-21-133 Enclosure reasonable, though reduced, level of confidence that these SSCs will satisfy functional requirements.

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

2.3 DESCRIPTION

OF THE PROPOSED CHANGE Energy Northwest proposes the addition of the following condition to the renewed operating license of Columbia Generating Station to document the NRC's approval of the use 10 CFR 50.69.

Energy Northwest is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSC) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic risk; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 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 ASME/ANS PRA Standard RA-Sa-2009; as specified in License Amendment No. [XXX] dated

[DATE].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

3.0 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 Page 3 of 52







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GO2-21-133 Enclosure systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs.

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

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

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

The PRA models described within this license amendment request (LAR) are the same as those described within the Energy Northwest LAR to be submitted within the next two months, "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b.'"

The Internal Events and Seismic PRA models described within this LAR, and described within the TSTF-505 LAR, were previously reviewed for technical adequacy by the NRC. The Internal Events PRA model was used as the basis for developing the Seismic PRA model. Though routine maintenance updates have since been applied, the Seismic model was previously reviewed for technical adequacy as documented in the following letter:

Staff Review of Seismic Probabilistic Risk Assessment Associated with Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic (EPID No. L-2019-JLD-0009), dated April 28, 2020.

Energy Northwest requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the TSTF-505 LAR. This would reduce the number of Energy Northwest 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.

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

3.1.1 Overall Categorization Process Energy Northwest 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 Page 4 of 52







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GO2-21-133 Enclosure According to their Safety Significance (Reference 3). NEI 00-04, Section 1.5, states, Due to the varying levels of uncertainty and degrees of conservatism in the spectrum of risk contributors, the risk significance of SSCs is assessed separately from each of five risk perspectives and used to identify SSCs that are potentially safety-significant.

A separate evaluation is appropriate to avoid reliance on a combined result that may mask the results of individual risk contributors.

The process to categorize each system will be consistent with the guidance in NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, as endorsed by RG 1.201.

RG 1.201 states that the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence and that all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1)(iv). However, neither RG 1.201 nor NEI 00-04 prescribe a particular sequence or order for each of the elements to be completed.

Therefore, the order in which each of the elements of the categorization process (listed below) is completed is flexible and as long as they are all 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, and seismic).

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 LSS) that is presented to the Integrated Decision-Making Panel (IDP). Note: the term preliminary HSS or LSS is synonymous with the NEI 00-04 term candidate HSS or LSS. A component or function is preliminarily categorized as HSS if any element of the process results in a preliminary HSS determination in accordance with Table 3-1 below. The safety significance determination of each element identified above is independent of 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 Page 5 of 52







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GO2-21-133 Enclosure categorization process was followed appropriately, the final Risk Informed Safety Class (RISC) category can be assigned.

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

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

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

modeled)

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

Function Allowable1 N/A Criteria Section 9.2 Passive - Not Passive Segment/Component No Section 4 Allowed Page 6 of 52







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GO2-21-133 Enclosure Notes:

1 The assessments of the qualitative considerations are agreed upon by the IDP in accordance with NEI 00-04, Section 9.2. In some cases, a 50.69 categorization team may provide preliminary assessments of the seven considerations for the IDPs consideration, however, the final assessments of the seven considerations are the direct responsibility of the IDP.

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

Conversely, if all the seven considerations are confirmed, then the function is presented to the IDP as preliminary LSS.

The System Categorization Document, including the justifications provided for the qualitative considerations, is reviewed by the IDP. The IDP is responsible for reviewing the preliminary assessment to the same level of detail as the 50.69 team (i.e., all considerations for all functions are reviewed). The IDP may confirm the preliminary function risk and associated justification or may direct that it be changed based upon their expert knowledge. Because the Qualitative Criteria are the direct responsibility of the IDP, changes may be made from preliminary HSS to LSS or from preliminary LSS to HSS at the discretion of the IDP. If the IDP determines any of the seven considerations cannot be confirmed (false response) for a function, then the final categorization of that function is HSS.

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

- see Table 3-1). These components from the component level assessments will remain HSS (IDP cannot override) regardless of the significance of the functions to which they are mapped. Therefore, if a HSS component is mapped to a LSS function, that component will remain HSS. If an LSS component is mapped to an HSS function, that component may be driven HSS based on Table 3-1 above, or may remain LSS.

If an SSC supports two functions, one being LSS and the other being HSS, the SSC will receive a categorization of HSS. If an SSC supports only an LSS function but is considered HSS from a component level evaluation, the SSC will be considered HSS.

The only exception applies to those components that support an HSS function but do not have a credible means to fail the HSS function. These components may be considered LSS in accordance with the guidance in Section 10.2 of NEI 00-04.

NEI 00-04, Sections 4 and 7.1, will be followed for SSCs that support an interfacing system. Those SSCs will typically remain uncategorized until all interfacing systems are Page 7 of 52







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GO2-21-133 Enclosure categorized. In some cases, impacts that an interfacing component could have on an interfacing system can be fully determined and the interface component can be categorized (and alternative treatment implemented) without categorizing the entire interfacing system. In this event, an assessment of interface component risk associated with uncategorized systems will be limited to cases where the following two conditions are met: 1) the interface component failure cannot prevent performance of interface system functions, and 2) the risk is limited to passive failures assessed as LSS following the passive categorization process for the applicable pressure boundary segments.

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

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

x 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.

x The decision criteria for the IDP for categorizing SSCs as safety significant or low safety significant pursuant to § 50.69(f)(1) will be documented in Energy Northwest 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.

x 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.

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

x NEI 00-04, Section 7, requires assigning the safety significance of functions to be preliminary HSS if it is supported by an SSC determined to be HSS from the PRA-based assessment in Section 5, but does not require this for SSCs determined to be HSS from non-PRA-based, deterministic assessments in Section 5. This requirement is further clarified in the Vogtle SER (Reference 4) which states if any SSC is identified as HSS from either the integrated PRA component safety Page 8 of 52







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GO2-21-133 Enclosure 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.

x 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.

x With regard to the criteria that considers whether the active function is called out or relied upon in the plant Emergency/Abnormal Operating Procedures, Energy Northwest 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:

x Internal Event Risks: Internal events, including internal flooding, PRA model Revision 8.0.1, dated February 2019. The Internal Events PRA model described within this LAR is the same model as the one reviewed during the Staff Review of Seismic Probabilistic Risk Assessment Associated with Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1:

Seismic (EPID No. L-2019-JLD-0009), dated April 28, 2020 (Reference 2), as the Internal Events PRA model was used as the basis for development of the Seismic PRA model. The NRC staff also concluded in November 2016 that the Internal Events PRA model was of adequate quality to support the evaluation of changes under the Surveillance Frequency Control Program (Reference 33).

x Fire Risks: Fire PRA model Revision 8.2, dated September 2021. The Fire PRA model has not been previously reviewed by NRC.

x Seismic Risks: Seismic PRA model Revision 8.2, dated June 2020. The Seismic PRA model described within this LAR was developed using the Internal Events PRA Model. The Seismic PRA model was reviewed by NRC in the following application: Staff Review of Seismic Probabilistic Risk Assessment Associated With Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic (EPID No. L-2019-JLD-0009), dated April 28, 2020 (Reference 2). The NRC staff concluded that the technical adequacy of Columbias Seismic PRA model was sufficient to support regulatory decision-making associated with Phase 2 of the 50.54(f) letter.

x Other External Risks (e.g., tornados, external floods): Use of the IPEEE screening process described in Reference 5, as supplemented in January 2001 (Reference 6),

and as re-evaluated in May 2017 (References 7, 8). The other external hazards were determined to be insignificant contributors to plant risk.

x Low Power and Shutdown Risks: Qualitative defense-in-depth (DID) shutdown model for shutdown configuration risk management (CRM) based on the framework Page 9 of 52







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GO2-21-133 Enclosure for DID provided in NUMARC 91-06, Guidance for Industry Actions to Assess Shutdown Management (Reference 9), which provides guidance for assessing and enhancing safety during shutdown operations.

A change to the categorization process that is outside the bounds specified above will not be used without prior NRC approval.

The SSC categorization process documentation will include the following elements:

1. Program procedures used in the categorization

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

3. Mapping of components to support function(s)

4. PRA model results, including sensitivity studies

5. Hazards analyses, as applicable

6. Passive categorization results and bases

7. Categorization results including all associated bases and RISC classifications

8. Component critical attributes for HSS SSCs

9. Results of periodic reviews and SSC performance evaluations

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

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

The RI-RRA methodology is a risk-informed safety classification and treatment program for repair/replacement activities (RI-RRA methodology) for pressure retaining items and their associated supports. In this method, the component failure is assumed with a probability of 1.0 and only the consequence evaluation is performed. It additionally applies deterministic considerations (e.g., DID, safety margins) in determining safety significance. Component supports, if categorized, are assigned the same safety significance as the highest passively ranked component within the bounds of the Page 10 of 52







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GO2-21-133 Enclosure associated analytical pipe stress model. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

The use of this method was previously approved by the NRC in the final Safety Evaluation for Vogtle dated December 17, 2014 (Reference 4). The RI-RRA method as approved for use at Vogtle for 10 CFR 50.69 does not have any plant specific aspects and is generic. It relies on the conditional core damage and large early release probabilities associated with postulated ruptures. Safety significance is generally measured by the frequency and the consequence of the event. However, this RI-RRA process categorizes components solely based on consequence, which measures the safety significance of the passive component given that it ruptures. This approach is conservative compared to including the rupture frequency in the categorization as this approach will not allow the categorization of SSCs to be affected by any changes in frequency due to changes in treatment. The passive categorization process is intended to apply the same risk-informed process accepted by the NRC in the ANO2-R&R-004 for the passive categorization of Class 2, 3, and non-class components. This is the same passive SSC scope the NRC has conditionally endorsed in ASME Code Cases N-660 and N-662 as published in RG 1.147, Revision 15. Both code cases employ a similar risk-informed safety classification of SSCs in order to change the repair/replacement requirements of the affected LSS components. All ASME Code Class 1 SSCs with a pressure retaining function, as well as supports, will be assigned high safety-significant, HSS, for passive categorization which will result in HSS for its risk-informed safety classification and cannot be changed by the IDP. Therefore, this methodology and scope for passive categorization is acceptable and appropriate for use at Columbia 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, there are no PRA upgrades that have not been peer reviewed, and there are no open )DFWVDQG2EVHUYDWLRQV F&O . The PRA models described within this LAR are the same as those described within the Energy Northwest LAR to be submitted within the next two months, "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b.'"

3.2.1 Internal Events and Internal Flooding The Columbia categorization process for the internal events and internal flooding hazard will use the peer reviewed plant-specific PRA model. The Energy Northwest risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant. Attachment 2 of this enclosure identifies the applicable Internal Events PRA model that includes internal flooding. The NRC staff also concluded in November 2016 that the Internal Events PRA model was of adequate Page 11 of 52







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GO2-21-133 Enclosure quality to support the evaluation of changes under the Surveillance Frequency Control Program (Reference 33).

3.2.2 Fire Hazards The Columbia categorization process for fire hazards will use a peer reviewed plant-specific Fire PRA model. The internal Fire PRA model was developed consistent with NUREG/CR-6850 and only utilizes methods previously accepted by the NRC. At the time of development, RG 1.200 Revision 3 was the most current revision; no new methods were used during the development of the Fire PRA. The Energy Northwest risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant. Attachment 2 of this enclosure identifies the applicable Fire PRA model.

3.2.3 Seismic Hazards The Columbia categorization process for seismic hazards will use a peer reviewed plant-specific Seismic PRA model. The Seismic PRA for Columbia was originally developed in response to Recommendation 2.1 of the Near-Term Task Force (NTTF)

Review of Insights from the Fukushima Daiichi Accident (Reference 27). The Seismic PRA hazard methodology and analysis for the associated ground motion response spectra were submitted to the NRC (Reference 28) and found to be technically acceptable for application to the Columbia Seismic PRA (Reference 29). The Columbia Seismic PRA was reviewed by the NRC and found to be technically adequate for its purpose relative to NTTF 2.1 (Reference 2). The development of the Seismic PRA model that will be used for the categorization process conforms to all applicable Capability Category II supporting requirements of the ASME/ANS PRA Standard Code Case. The Seismic PRA model was developed following the guidance in EPRI 3002000709 (Reference 31) and EPRI 3002008093 (Reference 32).

The Energy Northwest risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant. 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 Other External Hazards All external hazards were screened from applicability to Columbia, with the exception of internal flooding, internal fire, and seismic activity, which are addressed by station PRA models. The screening was performed in a plant-specific evaluation in accordance with Generic Letter 88-20, Supplement 4 (Reference 11), and using the criteria in ASME PRA Standard RA-Sa-2009 (Reference 12), NUREG/CR-2300 (Reference 13), and NUREG-1407 (Reference 14). Attachment 4 to this enclosure provides a summary of the external hazards screening results. Attachment 5 provides a summary of the progressive screening approach for external hazards.

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GO2-21-133 Enclosure In the event that a change, such as a station modification, industry operating experience, or PRA model error or limitation, identifies that unavailability of an SSC results in an unscreened hazard, the applicable SSC will be considered HSS following the guidance in NEI 00-04, Figure 5-6. Similarly, if a passive SSC credited for an external hazard, such as external flooding (e.g., catch basin, drywell, roof drainage system) or tornado protection (e.g., permanent barrier structure), would result in an unscreened hazard if it is unavailable, the applicable SSC will be considered HSS following the guidance in NEI 00-04, Figure 5-6.

3.2.5 Low Power & Shutdown Consistent with NEI 00-04, the Columbia 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 of NEI 00-04 will be considered preliminary HSS.

3.2.6 Integral Assessment The Internal Events, Seismic, and Fire PRA models will remain as separate models when calculating importance measures for the 10 CFR 50.69 categorization process.

As such, integration of importance measures across all hazards will be performed manually using the methodology specified in NEI 00-04, Section 5.6.

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GO2-21-133 Enclosure 3.2.7 PRA Maintenance and Updates The Energy Northwest risk management process ensures that the applicable PRA model(s) used in this application continues to reflect the as-built and as-operated Columbia station. The process delineates the responsibilities and guidelines for updating the PRA models. The process will include criteria for both regularly scheduled and interim PRA model updates. The process will include 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 planned changes or errors have a significant impact on the PRA model, which is defined as greater than +/-25% core damage frequency (CDF) or large early release frequency (LERF) for the modeled hazard, or significant impacts to 50.69 basic event importance measures, then an unscheduled update will be performed to update the affected PRA model(s). If the changes in results are not easily measurable, a qualitative assessment will be performed to consider the potential change in basic event importance measures for each application. The qualitative assessment will utilize the experience and judgment of the PRA engineer to determine if there are any issues that are individually negligible but could collectively impact the 50.69 program. If there is a significant impact on the PRA model, the SSC categorization will be re-evaluated.

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

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

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

Consistent with the NEI 00-04 guidance, Energy Northwest will perform both an initial sensitivity study and a cumulative sensitivity study. The initial sensitivity study applies to the system that is being categorized. In the cumulative sensitivity study, the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components Page 14 of 52







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GO2-21-133 Enclosure 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 Internal Events, Seismic and Fire PRA models and documentation were reviewed for plant-specific and generic modeling assumptions and related sources of uncertainty.

The process to evaluate uncertainties is defined in NUREG-1855 (Reference 15),

EPRI TR-1026511 (Reference 16), and EPRI TR-1016737 (Reference 17). Each PRA model includes an evaluation of the potential sources of uncertainty for the base case models using the approach that is consistent with the ASME/ANS requirements (ASME/ANS RA-Sa-2009, Code Case for ASME/ANS RA-Sb-2013) for identification and characterization of uncertainties and assumptions.

Each PRA element notebook was reviewed for assumptions and sources of uncertainties. A total of 42 items were identified as potentially impacting the 50.69 categorization process. Each identified uncertainty was evaluated for its potential to significantly impact the importance evaluations for performing SSC categorization. The SSC categorization for 10 CFR 50.69 involves the use of Fussell-Vesely (F-V) and Risk Achievement Worth (RAW) importance measures based on CDF and LERF. If a source of uncertainty could be shown to satisfy the following considerations using a qualitative evaluation, then the uncertainty was determined to have a negligible impact on SSC categorization evaluations.

x Criterion #1: Uncertainties that are qualitatively shown to have a very small impact on total risk and would be expected to have a negligible impact on RAW and F-V (particularly uncertainties that pertain to parts of the model that would not impact components that are in the 10 CFR 50.69 Program, such as changes to non-support system initiating event frequencies, human error probabilities not related to 50.69-eligible equipment, etc.).

x Criterion #2: Uncertainties that are represented through conservative PRA modeling that would be expected to have a negligible impact on importance calculations.

x Criterion #3: Uncertainties that are identified as using current industry-accepted approaches and data are not considered as key sources of uncertainty. This is consistent with the ASME/ANS PRA Standard definition of a source of modeling uncertainty which states, a source is related to an issue in which there is no consensus approach or model and where the choice of approach or model is known to have an effect on the PRA model.

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GO2-21-133 Enclosure x Criterion #4: Uncertainties examined by sensitivity studies to confirm that the impact on baseline CDF and LERF are negligibly small are not considered as key sources of uncertainty for the 10 CFR 50.69 Program.

The overall process for application of Criteria #1 and #2 is clarified as described below:

1. Determine the impacts of the uncertainty on the systems, procedures, and scenarios involved, such as having a negligible impact on system or plant response when considering, for example, expected system flow rates, heat removal capacities, and maximum expected flow diversions.

2. Determine the possible impacts on specific initiating events and component failure modes of concern if the candidate uncertainty is applicable based on expected system or plant operation through a review of the all-hazards PRA results to identify those that would have negligible impacts on CDF and LERF. This review was performed by PRA engineers, considering such factors as the frequencies and conditional core damage probabilities for initiating events as well as F-V, RAW and Risk Reduction Worth importance measures for specific failures or appropriate surrogate events. The evaluations specifically consider the impact on the importance calculations that are used to categorize SSCs.

There were 26 key assumptions and uncertainties identified from the internal events PRA model. After review of each uncertainty was completed, there was one key assumption or uncertainty identified that could impact the 10 CFR 50.69 categorization process and will be addressed with a sensitivity.

For the Fire PRA model, there were 11 key assumptions and uncertainties identified.

After review of each uncertainty was completed, there were two key assumptions or uncertainties identified that could impact the 10 CFR 50.69 categorization process.

Both of the identified items (non-credited equipment and cable routing) were related to assumed failure of events in the Fire PRA. Therefore, a single sensitivity will be used to address both items.

Lastly, there were five key assumptions and uncertainties identified from the Seismic PRA model. After review of each uncertainty was completed, there were no key assumptions or uncertainties identified that could impact the 10 CFR 50.69 categorization process.

The key assumptions and uncertainties for the 10 CFR 50.69 application are discussed in more detail in Attachment 6. Attachment 6 also provides discussion on the uncertainties associated with human error probabilities and FLEX, although both of these areas were determined not to be key assumptions or uncertainties with respect to the 10 CFR 50.69 application.

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GO2-21-133 Enclosure The sensitivity studies described in Attachment 6 will be run as part of the system categorization process and the results will be provided to the IDP as an additional input to their categorization decision.

3.2.9 Modeling of FLEX FLEX strategies are credited in the Columbia Internal Events, Fire, and Seismic PRA models. The FLEX strategies that are credited in the PRA models are:

x Powering the battery chargers for continued operation of the battery chargers.

x Providing an alternate source of low pressure reactor pressure vessel (RPV) injection should other sources of high and low pressure injection fail.

x Venting containment using the permanently installed Hardened Containment Vent System (HCVS).

The first two FLEX strategies listed above, powering the battery chargers and providing an alternate source of low pressure RPV injection, are modeled only for Station Blackout (SBO) and Extended Loss of Alternating Current Power (ELAP) sequences with Reactor Core Isolation Cooling (RCIC) initially available and operating for four hours. FLEX strategies are not credited for sequences with no RPV injection at time zero due to insufficient time to perform the alignment.

Diesel Generator 4 (DG-4) or Diesel Generator 5 (DG-5) is used to power the battery chargers. DG-4 is permanently staged in the area where electrical connections exist to connect DG-4 to the 480V electrical bus which is used to power the battery chargers.

DG-5 is considered portable FLEX equipment stored onsite and it must be transported to the area where the electrical hookup is performed. Should a loss of all onsite and offsite power occur, the operating crew would utilize DG-4 (preferred) or DG-5 to power the battery chargers. Clear procedural guidance is provided to the operating crew for the startup and electrical connection of the FLEX generators.

One of the FLEX diesel powered pumps, FLEX-P-1 or B.5.b pumper truck, is used to provide an alternate source of low pressure RPV injection. These are redundant pumps which are used as an alternate source of low pressure injection to the RPV when no other source of low pressure injection is available. The FLEX diesel pumps are portable FLEX equipment stored on site and must be transported to the required area, near Service Water pond A or B. Hoses are run from the FLEX pump to the Service Water pond for suction, and from the discharge of the FLEX pump into the reactor building and connected into one of the Residual Heat Removal divisional loops near valves RHR-V-63A/B/C. Clear procedural guidance is provided to the operating crew for the transport, connection, and start of the FLEX diesel powered pumps.

Venting containment using the HCVS uses permanently installed components. The credited equipment is similar to other permanently installed plant components with Page 17 of 52







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GO2-21-133 Enclosure sufficient plant-specific or generic industry data. The operator actions are similar to other operator actions evaluated using approaches consistent with the endorsed ASME/ANS RA-Sa-2009 PRA standard.

Internal Events PRA Model and Human Error Probabilities (HEP)

The FLEX strategies were incorporated into the Internal Events PRA model (Revision 8.0) as part of the PRA maintenance process. The incorporation of FLEX strategies into the Internal Events PRA model does not constitute a PRA model upgrade because modeling inclusion of FLEX has been performed in a manner that:

x Is consistent with other modeling aspects used in the PRA model and no new methodology was used.

x Does not result in a change in scope or capability that impacts the significant accident sequences or the significant accident progression.

Therefore, a focused scope peer review was not performed.

Two separate FLEX actions associated with the operation of DG-4 and DG-5 are modeled in the PRA for alignment of each diesel generator.

EACHUMN-DG4-XTIE= 4.72E-01 and EACHUMN-DG5-XTIE= 8.25E-01 However, if the operators fail to successfully align DG-4, the PRA explicitly models that operators will also fail to connect DG-5.

For RPV low pressure injection using the FLEX diesel powered pumps (FLEX-P-1 or B.5.b pumper truck), a single operator action is used to capture the action for starting either pump.

RPVHUMNFLEX-P1= 1.12E-01 These credited FLEX operator actions are evaluated using approaches consistent with the endorsed ASME/ANS RA-Sa-2009 PRA standard and are documented in the Human Reliability Analysis (HRA) notebook.

Seismic PRA Model The Internal Events PRA model (Revision 8.0) was used to develop the Seismic PRA model. The Seismic PRA model, which includes the FLEX strategies described above, was peer reviewed against the requirements of the Code Case for PRA Standard ASME/ANS RA-Sb-2013 (see LAR Section 3.3).

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GO2-21-133 Enclosure Fire PRA Model The Internal Events PRA model (Revision 8.0) was used to develop the Fire PRA model. However, the Fire PRA model only credits DG-4 to power the battery chargers as part of the FLEX strategies to mitigate fire scenarios. DG-5 and the FLEX diesel powered pumps (FLEX-P-1 and B.5.b pumper truck) are not credited in the Fire PRA model, as they are portable equipment. The Fire PRA model, which includes FLEX diesel generator DG-4, was peer reviewed against the requirements of PRA Standard ASME/ANS-RA-Sa-2009 (see LAR Section 3.3).

FLEX Equipment Data The failure to start and failure to run data for the FLEX equipment was developed using the generic values in NUREG/CR-6928 (Reference 26) for diesel generators and diesel pumps.

Unavailability of the diesel generators and diesel pumps was modeled in the PRA.

Sensitivity Analysis A sensitivity analysis was performed to measure the risk increase associated with completely removing credit for DG-4, DG-5, FLEX-P-1, and the B.5.b pumper truck from the Internal Events, Fire, and Seismic PRA models. The results of the sensitivity analysis are provided in the following table:

Delta Internal Events CDF 1.00E-08 Internal Events LERF 0.00 Fire CDF 4.20E-07 Fire LERF 0.00 Seismic CDF 0.00 Seismic LERF 0.00 Total Delta CDF 4.30E-07 Total Delta LERF 0.00 3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii))

The PRA models described in Section 3.2 have been assessed against RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, (References 18, 19).

The Internal Events and Seismic PRA models were subject to full scope peer reviews, focused scope peer reviews, and closure peer reviews in accordance with RG 1.200, Revision 2 (Reference 18). The Fire PRA model was subject to a full scope peer review and closure review in accordance with RG 1.200, Revision 3 (Reference 19).

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GO2-21-133 Enclosure RG 1.200, Revision 2, endorsed PRA Standard ASME/ANS RA-Sa-2009 (Reference

12) for preparation of a technically acceptable PRA that can be used to implement 10 CFR 50.69. RG 1.200, Revision 3, was subsequently issued in December 2020.

Changes implemented by Revision 3 included:

x Endorsement of NEI 17-07 Revision 2, "Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard" x Endorsement, with staff exceptions and clarifications, of requirements in ASME/ANS RA-S Case 1, "Case for ASME/ANS RA-Sb-2013 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment of Nuclear Power Plant Applications" x Endorsement of the requirements for peer review of Newly Developed Methods (NDM), the process for determining whether a change to a PRA is classified as PRA maintenance or a PRA upgrade, and the definitions related to NDMs, PRA maintenance, and PRA upgrade from PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review" x Enhancement of guidance related to key assumptions and sources of uncertainty x A glossary of key terms x A list of hazards to be considered in the development and use of PRA.

Consequently, the use of RG 1.200, Revision 2, or Revision 3, are acceptable for implementation of risk-informed categorization of structures, systems and components (Reference 20).

Internal Events and Internal Flooding PRA Model The Full Power Internal Events, including internal flood, PRA model was subject to a full scope peer review conducted in August 2009 using the NEI 05-04 process, the ASME PRA Standard (ASME/ANS RA-Sa-2009), and RG 1.200 Revision 2 (Reference 18).

A finding level F&O independent assessment peer review was performed in March 2018. The purpose of the F&O independent assessment was to review close-out of the F&Os from the August 2009 peer review using Appendix X of NEI 05-04. All F&Os were documented as successfully closed. The March 2018 review included a focused scope peer review of a PRA model upgrade. The upgrade was related to re-evaluating the methodology from the Accident Sequence Evaluation Program (ASEP) method to the Human Cognitive Reliability/Operator Reliability Experiments (HCR/ORE) method.

An F&O independent assessment peer review was performed remotely in May and June 2018. The purpose of the independent assessment was to review close-out of the Page 20 of 52







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GO2-21-133 Enclosure F&Os from the March 2018 focused scope peer review. All F&Os were documented as successfully closed.

Seismic PRA Model The Seismic PRA model was peer reviewed in December 2018 using the NEI 12-13 process, the ASME PRA Standard (ASME/ANS RA-Sb-2013, Code Case 1), and RG 1.200, Revision 2. The peer review was a full-scope review of the Columbia at-power Seismic PRA model against all technical elements in ASME/ANS RA-Sb-2013, Code Case 1. The Code Case is an approved alternative to meet the requirements of Part 5 in ASME/ANS RA-Sb-2013 (Reference 25).

An F&O independent assessment peer review was performed in July 2019 to review close-out of F&Os from the December 2018 peer review. All F&Os were documented as successfully closed, except F&O 20-10, which was partially closed. The July 2019 review included a focused scope peer review of an upgrade to the Seismic PRA model.

The upgrade was related to recalculating fragilities using a scaling approach. No F&Os were generated from the focused scope peer review.

A follow up F&O independent assessment peer review was conducted remotely in November 2019 to review close out of F&O 20-10. F&O 20-10 was documented as successfully closed.

A focused scope peer review was conducted in November 2019 to assess the technical adequacy of a Seismic PRA model upgrade. The PRA model upgrade was related to a secondary containment effectiveness model of a reactor water clean-up line break to support a Seismic PRA LERF reduction.

An F&O independent assessment peer review was conducted in March 2020 to review close-out of F&Os from the November 2019 focused scope peer review. All F&Os were documented as successfully closed.

Fire PRA Model The Fire PRA model was subject to a full scope peer review in February 2021. The peer review was conducted in accordance with the NEI 17-07 process, PRA Standard ASME/ANS-RA-Sa-2009, and RG 1.200, Revision 3.

An F&O independent assessment peer review was conducted in July 2021 to review the close-out of finding level F&Os from the February 2021 peer review. All findings were documented as successfully closed.

Closure Reviews Internal Events and Seismic PRA peer review findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-Page 21 of 52







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GO2-21-133 Enclosure out of Facts and Observations (F&O) (Reference 21) as accepted by the NRC in the letter dated May 3, 2017 (Reference 22). Fire PRA model peer review findings were reviewed and closed using the independent assessment guidance in NEI 17-07, Revision 2 (Reference 30), as endorsed by RG 1.200, Revision 3. The independent assessment teams assessing F&O closure assessed each F&O to conclude if the F&O constituted a PRA upgrade or maintenance update.

There are no open items from the Columbia RG 1.200 self-assessments or open findings from the peer reviews of the PRA models. The results of the reviews have been documented and are available for NRC audit.

The self-assessment and peer reviews demonstrate that the Internal Events, Fire and Seismic PRA models are of sufficient quality and level of detail to support the categorization process and have been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC as required by 10 CFR 50.69(c)(1)(i).

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

The Columbia 10 CFR 50.69 categorization process will implement the guidance in NEI 00-04. The overall risk evaluation process described in the NEI guidance addresses both known degradation mechanisms and common cause interactions and meets the requirements of §50.69(b)(2)(iv). Sensitivity studies described in NEI 00-04, Section 8, will be used to confirm that the categorization process results in acceptably small increases to CDF and LERF. The failure rates for equipment and initiating event frequencies used in the PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, and human errors). Subsequent performance monitoring and PRA updates required by the rule will 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 occurring at least once every two refueling outages will evaluate new insights resulting from available risk information (i.e., PRA model or other analysis used in the categorization) changes, design changes, operational changes, and SSC performance. If it is determined that these changes have affected the risk information or other elements of the categorization process such that the categorization Page 22 of 52







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GO2-21-133 Enclosure results are more than minimally affected, then the risk information and the categorization process will be updated. This review will include:

x A review of plant modifications since the last review that could impact the SSC categorization x A review of plant specific operating experience that could impact the SSC categorization x A review of the impact of the updated risk information on the categorization process results x A review of the importance measures used for screening in the categorization process x 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.

4.0 REGULATORY EVALUATION

4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA The following NRC requirements and guidance documents are applicable to the proposed change.

x 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".

x NRC RG 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, May 2006.

x RG 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.

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

x RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 3, December 2020.

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GO2-21-133 Enclosure The proposed change is consistent with the applicable regulations and regulatory guidance.

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS Energy Northwest 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.

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of 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?

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GO2-21-133 Enclosure Response: No.

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

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

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

Response: No.

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

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

Based on the above, Energy Northwest concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c),

and, accordingly, a finding of "no significant hazards consideration" is justified.

4.3 CONCLUSION

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

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GO2-21-133 Enclosure

5.0 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.

6.0 REFERENCES

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

2. NRC letter to B.J. Sawatzke, Energy Northwest, Columbia Generating Station -

Staff Review of Seismic Probabilistic Risk Assessment Associated With Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic (EPID No. L-2019-JLD-0009), dated April 28, 2020 (ML20076A547)

3. NRC 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

4. NRC letter to C.R. Pierce, Southern Nuclear Operating Company, Inc., Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments Re: Use of 10 CFR 50.69 (TAC Nos. ME9472 and ME9473), dated December 17, 2014 (ML14237A034)

5. J.V. Parrish, Energy Northwest, to NRC, WNP-2, Operating License NPF-21, Initial Submittal of Individual Plant Examination for External Events (IPEEE) (TAC No. 74489), dated June 26, 1995

6. D.W. Coleman, Energy Northwest, to NRC, Columbia Generating Station, Operating License NPF-21, Initial Submittal of Individual Plant Examination for External Events (IPEEE), Status of Improvements, dated January 24, 2001 (ML010300150)

7. Columbia Generating Station, IPEEE-2-RE-001, Re-examination of External Events in the IPEEE, Revision 0, dated May 22, 2017 Page 26 of 52







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GO2-21-133 Enclosure

8. A.L. Javorik, Energy Northwest, to NRC, Columbia Generating Station, Docket No.

50-397, Response to Request for Additional Information Related to License Amendment Request - Revise Technical Specification 5.5.12 for Permanent Extension of Type A Test and Type C Leak Rate Test Frequencies, dated January 2, 2018 (reference response to RAI 4) (ML18002A501)

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

10. NRC letter to Vice President, Operations, Arkansas Nuclear One, Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative AN02-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems (TAC No. MD5250), dated April 22, 2009 (ML090930246)

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

12. The American Society of Mechanical Engineers, ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, dated February 2009

13. NUREG/CR-2300, PRA Procedures Guide, dated January 1983

14. NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, dated June 1991

15. NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, Revision 1, dated March 2017

16. Electric Power Research Institute (EPRI), Technical Report (TR) 1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, dated December 2012

17. Electric Power Research Institute (EPRI), Technical Report (TR) 1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, dated December 2008

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

19. Regulatory Guide 1.200, Acceptability of Probabilistic Risk Assessment Results for Page 27 of 52







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GO2-21-133 Enclosure Risk-Informed Activities, Revision 3, dated December 2020

20. NRC letter to D.P. Rhoades, Exelon Generation Company, LLC, LaSalle County Station, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 249 and 235 Related to Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (EPID L-2020-LLA-0017), dated May 27, 2021 (ML21082A422)

21. 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), dated February 21, 2017 (ML17086A431)

22. NRC letter to G. Krueger, Nuclear Energy Institute, U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os), dated May 3, 2017 (ML17079A427)

23. 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, dated January 2018

24. ASME/ANS RA-S Case 1, Case for ASME/ANS RA-Sb-2013 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, dated November 2017

25. NRC letter to R.C. Grantom and R.J. Budnitz, U.S. Nuclear Regulatory Commission Acceptance of ASME/ANS RA-S Case 1, dated March 12, 2018

26. NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants, dated February 2007

27. J.K. Dittmer, Energy Northwest, to NRC, Columbia Generating Station Docket No. 50-397 Seismic Probabilistic Risk Assessment Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights from the Fukushima Dai-ichi Accident (MF3726, MF3727), dated September 26, 2019

28. D.A Swank, Energy Northwest, to NRC, Columbia Generating Station Docket No. 50-397 Seismic Hazard and Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights from the Fukushima Dai-ichi Accident, dated March 12, 2015

29. NRC letter to M.E. Reddemann, Columbia Generating Station - Staff Assessment of Information provided Under Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Page 28 of 52







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GO2-21-133 Enclosure Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (CAC No. MF5274), dated November 4, 2016

30. NEI 17-07, "Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard, Revision 2, dated August 2019

31. EPRI 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013

32. EPRI 3002008093, An Approach to Human Reliability Analysis for External Events with a Focus on Seismic, December 2016

33. NRC letter to M.E. Reddemann, Energy Northwest, Columbia Generating Station -

Issuance of Amendment Re: Adoption of Technical Specification Task Force Traveler TSTF-425, Revision 3 (CAC No. MF6042), dated November 3, 2016 (ML16253A025)

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GO2-21-133 Enclosure Attachment 1: List of Categorization Prerequisites Energy Northwest will establish procedures prior to the use of the categorization process on a plant system. The procedures will contain the elements/steps listed below.

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

x 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.

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

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

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

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

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

x Documentation requirements per Section 3.1.1 of the enclosure.

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GO2-21-133 Enclosure Attachment 2: Description of PRA Models Used in Categorization Baseline Baseline Comments Model CDF LERF Internal Events PRA Model, NRC reviewed F&Os during including Internal Flooding, Staff Review of Seismic Revision 8.0.1 PRA associated with the re-dated evaluated seismic hazard February 2019 for implementation of NTTF, 2.36E-06 1.60E-07 Recommendation 2.1:

Peer reviewed against Seismic.

RG 1.200, Rev. 2 (see LAR Section 3.3)

Seismic PRA Model, NRC reviewed model during Revision 8.2 Staff Review of Seismic dated PRA associated with the re-June 2020 evaluated seismic hazard 1.73.E-05 5.16E-06 for implementation of NTTF, Peer reviewed against Recommendation 2.1:

RG 1.200, Rev. 2 Seismic.

(see LAR Section 3.3)

Fire PRA Model, Revision 8.2 dated September 2021 4.06E-05 3.34E-06 Peer reviewed against RG 1.200, Rev. 3 (see LAR Section 3.3)

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GO2-21-133 Enclosure Attachment 3: Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items This attachment is not applicable. There are no open peer review findings or open self-assessment items.

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GO2-21-133 Enclosure Attachment 4: External Hazards Screening Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

The IPEEE external hazard evaluation was performed in 1995.

An updated evaluation regarding aircraft impact was performed as part of the Re-Examination of Aircraft Impact Y PS4 External Events Evaluation in the IPEEE in 2017. Section R.1 of the evaluation demonstrated that aircraft impact and skid frequencies remain less than 1E-07/yr.

Plant site is not located near large mountains where snow avalanches are prevalent. Columbia Avalanche Y C3 Generating Station (CGS) is located in the middle of a plain with minimal variations in height across topography.

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

Actions committed to and completed by CGS in response to C1 Generic Letter 89-13 provide Biological Event Y ongoing control of biological C5 hazards. These controls are described in the Service Water Reliability Program. In addition, station actions taken in response to INPO SOER 07-2 provide an additional layer of biological hazard management.

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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

Not applicable to the site because Coastal Erosion Y C3 of location.

Drought is a slowly developing hazard. The plant location (riverine site with seven dams upstream and four dams downstream) precludes the impact on the plant due to this Drought Y C5 hazard. The river surface water level near CGS is primarily controlled by regulation of the 35 million acre-feet capacity of upstream reservoir projects.

The external flooding hazard at the site was recently updated as a result of the post-Fukushima 50.54(f) Request for Information and the flood hazard re-evaluation report (FHRR) which was submitted to NRC for review on October 6, 2016.

The results indicate that flooding from rivers and streams (precipitation based) and dam External Flooding failure, and the combined effect for and Intense Y C1 dam failure with coincident wind-Precipitation wave activity, do not pose a challenge to the plant below the mean sea level (MSL) critical elevation at CGS. Natural topography maintains natural drainage away from the site.

To provide adequate surface drainage during severe precipitation conditions such as heavy rainfall and fast snowmelts (local intense precipitation events),

a system of catch basins and dry Page 34 of 52







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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a) wells is constructed with inlet elevations lower than the finished floor elevation of the nearest building(s). The roofs of safety-related buildings are designed to handle local intense precipitation events with adequate drainage.

The design basis tornado maximum wind speed is 200 mph. The atmospheric pressure at the center of the tornado is 0.9 psi below ambient. The 0.9 psi external pressure drop is assumed to occur at a rate of 0.3 psi/sec.

Per Table 6-1 of NUREG/CR-4461, Revision 2, the 1E-7 probability tornado wind speed is 210 mph, based on the F-scale, and C1 167 mph, based on the more recent Extreme Wind or Y EF-scale. CGS calculated the Tornado PS4 probability of a design basis tornado striking CGS to be 9.6E-7.

Sections 3.3 and 3.5.2 of the CGS FSAR describe the capability of SSCs to withstand wind and tornado loadings to include missile protection for SSCs required to bring the plant to a safe shutdown condition. All the SSCs are protected from external missiles by passive permanent barrier structures or redundant systems.

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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

The principal effects of such events would be to cause a loss of offsite power (LOOP) and are addressed Fog Y C1 in the weather-related LOOP initiating events in the Internal Events PRA model for CGS.

The most significant consequence of a forest or range fire is a LOOP which is evaluated in the Internal Events PRA model.

There are no major wooded areas close enough to the site to pose a significant fire hazard. Areas adjacent to CGS, major buildings, C1 and auxiliary facilities are Forest or Range Fire Y maintained to prevent weed growth C4 by landscaping, ground cover, and weed control spraying to help reduce the likelihood of brush fires.

FSAR Section 3.1.2.2.10 confirms that the control room ventilation is established by recirculating air through HEPA filters in the event of excessive smoke in the air near the normal control room ventilation intake.

The principal effects of such events would be to cause a LOOP and are C1 addressed in the weather-related Frost Y LOOP initiating events in the C4 Internal Events PRA model for CGS.

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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

The principal effects of such events would be to cause a LOOP and are C1 addressed in the weather-related Hail Y LOOP initiating events in the C4 Internal Events PRA model for CGS.

High summer temperatures are of negligible impact on the site. This C1 phenomenon provides a large High Summer Y amount of time for preparation Temperature C5 (weather forecast) with time for implementation of appropriate mitigation actions.

High tide or lake level are not C1 applicable to the site because of location. Impact of high river stage High Tide, Lake Y C3 is slow to develop with time for Level, or River Stage implementation of appropriate C4 mitigation actions.

See also External Flooding.

The Pacific Northwest location of Hurricane Y C3 the CGS site precludes the possibility of a hurricane.

There have been on average seven glaze days per year where glaze refers to a clear coating of ice containing some air pockets. Two C1 instances of severe traffic Ice Cover Y disruption have occurred, but there C4 has been no known damage to transmission lines due to ice.

Ice flooding will not affect the capability to shut down the reactor in a safe and orderly manner. The maximum water surface elevation Page 37 of 52







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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a) due to ice induced flooding is bounded by other flood causing mechanisms. Also, the daily fluctuating stage of the river at the intake location will discourage formation of sheet ice as well as ice jams. Ice flows, should they occur, will normally pass over the intake structure due to relatively high winter discharge in the river.

Therefore, the principal effects of ice cover would be to cause a LOOP and are addressed in the weather-related LOOP initiating events in the Internal Events PRA model for CGS.

There are no military facilities within the proximity to the plant site.

Industrial facilities within 5 miles of the plant were screened from further evaluation as they do not pose a challenge to the safe operation of the plant due to frequency of occurrence.

The Energy Northwest Hydrogen PS2 Storage and Supply Facility is Industrial or Military Y located 0.6 miles from the site. An Facility Accident PS4 analysis shows that an explosion and subsequent missile generation from a random tank rupture at normal pressure would not affect the plant due to the remote distance of the facility. A second analysis shows that an overpressurization event and subsequent rupture is a credible event; however, the total annual probability for any missiles Page 38 of 52







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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a) generated is less than 10-7 and the storage containers have relief valves for overpressurization.

Therefore, the hydrogen storage facility does not pose a challenge to the plant.

The CGS Internal Events PRA Internal Flooding N None model includes evaluation of risk from internal flooding events.

The CGS Fire PRA model includes Internal Fire N None evaluation of risk from internal fires.

The plant site is located on level Landslide Y C3 terrain and is not subject to landslides.

The principal effect of such events would be to cause a LOOP which is C1 addressed in the weather-related Lightning Y LOOP initiating events in the C4 Internal Events PRA model for CGS.

Low lake level is not applicable to the site because of location.

Impacts of low river stage are slow Low Lake Level or Y C5 to develop with time for River Stage implementation of appropriate mitigation actions (e.g., plant power reduction or shutdown).

Low winter temperatures are of C1 negligible impact on the site. This Low Winter Y phenomenon provides a large Temperature C5 amount of time for preparation (weather forecast) with time for implementation of appropriate Page 39 of 52







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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a) mitigation actions (e.g., plant power reduction or shutdown).

The possibility of plant damage from water damage from a meteorite and subsequent tsunami Meteorite or Satellite Y PS4 is not credible based on the site Impact location. The sum of meteor land impacts and air burst impacts frequencies is 4.7E-08/yr.

There are no challenges presented to CGS as a result of pipeline accidents. There are no pipelines in the vicinity of CGS as the two closest pipelines are natural gas Pipeline Accident Y C3 pipelines at distances of 12 and 24 miles from the site.

See "Industrial or Military Facility Accident" for discussion of the Energy Northwest Hydrogen Storage and Supply Facility.

In accordance with the Control Room Envelope Habitability Program, hazardous chemical evaluations have been performed for all of the chemicals stored onsite. The impact of releases of Release of C4 chemicals in onsite storage do not Chemicals in Onsite Y pose a risk to the site.

Storage PS1 Chemicals on the CGS site were analyzed in accordance with the guidance provided in RG 1.78, Revision 1. Most of the chemicals screened out due to being small quantities in small containers, whereas the rest of the chemicals Page 40 of 52







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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a) were analyzed assuming the maximum control room intake (1300 cfm in two train pressurization mode) is unfiltered.

The results showed that all the chemicals were well below the toxicity limits. Therefore, none of the chemicals pose a threat to control room operators at CGS.

See also Transportation Accidents.

Columbia River flow is controlled by the operation of upstream reservoirs by the U.S. Army Corps of Engineers (USACE). There is no historical or topographical evidence indicating that flow in the Columbia River can be diverted from its River Diversion Y C3 present course. The river is wide and well defined, and there are no deeply incised gorges upstream that could cause a landslide that would cut off river flow. Therefore, it is very unlikely that the Columbia River would be diverted from its present course by natural causes.

C1 Sand or dust storm is bounded by a Sand or Dust Storm Y postulated volcanic ash event.

C4 See also Volcanic Activity.

C1 Flooding due to seiches is not Seiche Y relevant for CGS per FSAR C3 Section 2.4.11.2.

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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

The CGS Seismic PRA model Seismic Activity N None includes evaluation of risk from seismic activity.

The principal effect of snow events would be to cause a LOOP which is addressed in the weather-related LOOP initiating events in the Internal Events PRA model for CGS.

FSAR Section 2.3.1.2.2 states the following: ANSI value of 20 lb/ft2 was used as the design load for all CGS structures, which corresponds to a depth of 3.2 feet. The largest 24-hour snowfall was 10.2 inches in February 1993 and a record depth C1 of approximately 12 inches lasted Snow Y four days in December 1964.

C4 These depths would correspond to snow loads of 5.3 and 6.24 lb/ft2, respectively.

Based upon the above information, the design of the CGS structures is able to withstand any postulated snow load. The effects of the snowfall (as opposed to the snow load following the precipitation) is addressed by the weather-centered LOOP initiator in the Internal Events PRA model; no other treatment is necessary for risk-informed regulatory applications.

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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

The existing loose to medium dense sand was excavated down to the underlying very dense Ringold gravel beneath all Seismic C1 Category I structures and replaced Soil Shrink-Swell Y with a denser state by compaction.

Consolidation C5 Therefore, the potential for this hazard is low at the site. The plant design considers this hazard and the hazard is slow to develop and can be mitigated.

Given the inland location and no connections with any water bodies considered for meteorological Storm Surge Y C3 events associated with a storm surge, flooding due to a storm surge is not plausible at CGS.

In accordance with the Control Room Envelope Habitability Program, hazardous chemical evaluations have been performed for all of the chemicals stored onsite and for those chemicals that exist within a 5-mile radius from the Toxic Gas Y C4 plant or transported using roads around the plant.

The hazards associated with toxic gas are screened elsewhere in this table (i.e., Release of Chemicals in Onsite Storage, Industrial or Military Facility Accident, and Transportation Accident).

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GO2-21-133 Enclosure Water Transportation: The Columbia River is not used for barge traffic in the vicinity of the plant site as the river channel is too shallow and the current is too swift.

Rail Transportation: The Department of Energy (DOE)

Hanford Site railroad system connects with commercial rail systems in Richland and Kennewick, Washington. Railroad operations that pass through the CGS property are restricted to only those trains that have been authorized by Energy Northwest Security. The rail line is physically C3 blocked at the two points where the plant vehicle barrier crosses the C4 tracks (FSAR Section 2.2.1). DOE Transportation Y shipments of large quantities of Accident PS2 hazardous materials within the exclusion area of the plant site PS4 during initial licensing of CGS are no longer made (FSAR Section 2.2.2.2).

Land Transportation: CGS is serviced by a paved access road connected to the DOE roadway system. State Highway 240 comes within 7 miles of CGS at its closest point.

Chemical hazards stored and transported in the vicinity of the plant were analyzed in FSAR Section 2.2.3. The analysis concluded that toxic chemicals transported or stored within the vicinity of the plant do not pose a threat to the plant.

Not applicable to the site because of location.

Tsunami Y C3 Page 44 of 52







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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

The probability of damage to safety-related systems by turbine missiles is acceptably low, due to:

(a) the protection provided by reinforced-concrete structural barriers, (b) the calculated PS2 probability of turbine missile Turbine-Generated Y generation, and (c) periodic testing Missiles PS4 and inspection of turbine overspeed protection systems with associated corrective action as required.

The highest overall damage probability for postulated turbine-generated missiles is less than or equal to 1x10-7 per year.

Due to the distance of CGS from the two major volcanoes (Mt. Adams, 165 km, and Mt. St. Helens, 220 km), only ash fall poses a hazard as mud flows, avalanches, pyroclastic rock flows, lava flows and shock waves are confined to a volcanos immediate area.

Enhancements to seismic Volcanic Activity Y C1 monitoring capability over the past decade, as well as ash fall monitoring by the Volcanic Ash Advisory Center (VAAC) over the past several decades, improved the timeliness and detail of information available to CGS in advance of potential ash fall. Therefore, there is adequate time to take mitigating actions (e.g., plant power reductions or shutdown).

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GO2-21-133 Enclosure Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

The CGS Ash Fall procedure includes six attachments that detail procedural steps and timing required to perform filter replacements. It includes direction to change FLEX component original equipment manufacturer (OEM) filters to ensure that in the event of a Station Blackout/Extended Loss of Alternating Current Power (SBO/ELAP) event the related FLEX components are maintained available. Therefore, there is adequate time to take mitigating actions (e.g., plant power reductions or shutdown).

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

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

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GO2-21-133 Enclosure Attachment 5: Progressive Screening Approach for Addressing External Hazards Event Analysis Criterion Source Comments C1. Event damage NUREG/CR-2300 and potential is less than ASME/ANS Standard events for which plant is RA-Sa-2009 designed.

C2. Event has lower NUREG/CR-2300 and mean frequency and no ASME/ANS Standard worse consequences RA-Sa-2009 than other events analyzed.

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

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

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

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

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

PS3. Design basis event NUREG-1407 as mean frequency is modified in

< 1E-05 per year and the ASME/ANS Standard mean conditional core RA-Sa-2009 damage probability is <

0.1.

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GO2-21-133 Enclosure Event Analysis Criterion Source Comments PS4. Bounding mean NUREG-1407 and CDF is < 1E-06 per year. ASME/ANS Standard RA-Sa-2009 Screening not NUREG-1407 and successful. PRA needs ASME/ANS Standard Detailed PRA to meet requirements in RA-Sa-2009 the ASME/ANS PRA Standard.

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GO2-21-133 Enclosure Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty Assumption/ Disposition Discussion Uncertainty The Internal Events PRA The failure of HPCS as a A sensitivity study will be model assumes that result of the containment performed as part of the failure of the failure is a conservative categorization process to containment, in the lower modeling assumption. This determine if any changes in one-third of the structure, assumption has the potential HSS/LSS determination would results in failure of High to affect the categorization for occur as a result of the Pressure Core Spray some components. uncertainty of HPCS failure (HPCS) due to following containment failure.

environmental effects.

Assumed failure of The uncertainty is the A sensitivity study will be components not credited potential benefit of including performed as part of the or assumed failures of the non-credited equipment categorization process to cables without routing in the scope of the FPRA. A determine if any changes in information in the FPRA. sensitivity has shown some HSS/LSS determination will benefit if these components occur as a result of are protected from fire uncertainties of assumed failed damage. components.

Human error probabilities Human Reliability Analysis This item does not represent a (HEP) represent a (HRA) uncertainties could key source of uncertainty in the potentially large have some impact on the Fire PRA model but sensitivity uncertainty for the Fire risk. Accepted industry to HEP variations will be PRA model given the methods (e.g., EPRI HRA assessed as part of the 50.69 importance of human Calculator) were used to process per Table 5-3 of actions in the base perform the Human Reliability NEI 00-04.

model. Since many of Analysis.

the HEP values were adjusted for fire, the joint dependency multipliers developed for the Internal Events PRA model also represent a potential for introducing a degree of conservatism.

No FLEX Equipment The primary purpose of the A sensitivity analysis was FLEX modifications is to performed in the CGS Fire provide emergency cooling PRA Uncertainty and and RPV injection over an Sensitivity Analysis (UNC)

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GO2-21-133 Enclosure Assumption/ Disposition Discussion Uncertainty extended period during an Notebook to measure the risk Extended Loss of AC Power increase associated with (ELAP). The only FLEX completely removing FLEX function credited in the Fire credit from the model. Based PRA is alignment of DG-4 for on the results of this sensitivity, ELAP sequences, which crediting the FLEX supplies the battery chargers modifications provides a for RCIC; other FLEX negligible risk benefit.

functions are modeled but not credited in the Fire PRA. This item does not represent a key source of uncertainty in the PRA model and will not be an issue for 50.69 calculations.

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GO2-21-133 Enclosure Attachment 7: Marked-up Operating License Page (one page follows)

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(34) The information in the FSAR supplement, submitted pursuant to 10 CFR 54.21(d), as supplemented by Commitment Nos. 1, 5, 13, 14, 17, 18, 23, 24, 26, 27, 28, 32, 36, 38, 40, 41, 42, 43, 48, 49, 50, 53, 55, 58, 59, 60, 61, 63, 64, 65, 66, 67, 68, 69, and 70 of Appendix A of NUREG-2123, "Safety Evaluation Report Related to the License Renewal of Columbia Generating Station" dated May 2012, is henceforth part of the FSAR which will be updated in accordance with 10 CFR 50.71(e). As such, the licensee may make changes to the programs and activities described in the UFSAR supplement and Commitment Nos. 1, 5, 13, 14, 17, 18, 23, 24, 26, 27, 28, 32, 36, 38, 40, 41, 42, 43, 48, 49, 50, 53, 55, 58, 59, 60, 61, 63, 64, 65, 66, 67, 68, 69, and 70 of Appendix A of NUREG-2123 provided the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(35) The licensee's FSAR supplement submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, and as supplemented by Commitment Nos. 1, 5, 13, 14, 17, 18, 23, 24, 26, 27, 28, 32, 36, 38, 40, 41, 42, 43, 48, 49, 50, 53, 55, 58, 59, 60, 61, 63, 64, 65, 66, 67, 68, 69, and 70 of Appendix A of NUREG-2123, describes certain future programs and activities to be completed before the period of extended operation. Energy Northwest shall complete these activities no later than June 20, 2023, and shall notify the NRC in writing when implementation of these activities is complete.

(36) To prevent lateral motion of the core plate, the licensee shall install core plate wedges around the periphery of the core plate within the shroud on or before December 20, 2021. Upon completion of the core plate wedge installation, the licensee shall submit a written report to the NRC staff summarizing the results of the installation. The licensee shall also submit a written report regarding any corrective action taken related to core plate rim hold-down bolts or core plate wedges and the results of extent of condition reviews on or before December 20, 2021.

(37) 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Energy Northwest is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSC) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic risk; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 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 ASME/ANS PRA Standard RA-Sa-2009; as specified in License Amendment No. [XXX] dated

[DATE].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

Renewed License No. NPF-21 Amendment No. 225