Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402

CNL-19-002

March 21, 2019

10 CFR 50.90

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

Sequoyah Nuclear Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328

Subject: Response to Request for Additional Information Regarding Application to Modify Sequoyah Nuclear Plant Units 1 and 2, Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors,” (SQN-TS-17-06) (EPID: L-2018-LLA-0066)

References: 1. TVA Letter to NRC, CNL-17-010 “Sequoyah Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, ‘Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors’,” dated March 18, 2018 (SQN-TS-17-06) (ML18075A365)

2. NRC Electronic Mail to TVA, “RAI - Sequoyah Nuclear Plant, Units 1 and 2, LAR to Adopt 10 CFR 50.69 Risk Informed SSC, (EPID: L 2018-LLA-0066),” dated January 15, 2019 (ML19015A419)

3. TVA Electronic Mail to NRC, “SQN 50.69 LAR RAI response extension,” dated February 27, 2019

In Reference 1, Tennessee Valley Authority (TVA) submitted a request for an amendment to Renewed Facility Operating License Nos. DPR-77 and DPR-79 for the Sequoyah Nuclear Plant (SQN) Units 1 and 2 respectively. The proposed license amendment request (LAR) would modify the SQN Renewed Facility Operating Licenses to allow for the implementation of the provisions of 10 CFR, Part 50.69, “Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors.”

U.S. Nuclear Regulatory Commission CNL-19-002 Page 2 March 21, 2019

In Reference 2, the NRC transmitted a request for additional information (RAI) and requested a response by March 1, 2019. In Reference 3, TVA transmitted a request for due date extension until March 22, 2019. NRC indicated that this revised due date was acceptable.

Enclosure 1 to this letter provides the TVA response to the RAI. As noted in Enclosure 1, the TVA response to RAl-03 requires a revision to the SON Units 1 and 2 Renewed Facility Operating Licenses that were provided in Reference 1. Enclosure 2 to this letter provides the existing SQN Unit 1 and Unit 2 Renewed Operating Licenses marked-up to show the proposed changes. Enclosure 3 to this letter provides the existing SQN Unit 1 and Unit 2 Renewed Operating Licenses re-typed pages to show the proposed changes. Enclosures 2 and 3 supersede those Operating License changes provided in Reference 1. Enclosure 4 provides a markup of Attachment 4 from Reference 1. Enclosure 5 provides a markup of Attachment 6 from Reference 1. Enclosure 6 provides retyped copies of Attachments 4 and 6. The revised Attachments 4 and 6 supersede those provided in Reference 1.

Consistent with the standards set forth in 10 CFR 50.92(c), TVA has determined that the additional information, as provided in this letter, does not affect the no significant hazards determination associated with the request provided in Reference 1.

There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Michael Anthony Brown at 423-751-3275.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 21st day of March 2019. ~~,iJtt-- Erin K. Henderson Director, Nuclear Regulatory Affairs

Enclosures cc (see Page 3) U.S. Nuclear Regulatory Commission CNL-19-002 Page 3 March 21, 2019

Enclosures:

1. Response to Request for Additional Information Regarding Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors” (SQN-TS-17-06) (EPID: L-2018-LLA-0066) 2. SQN Units 1 and 2 Renewed Operating Licenses Changes Markup 3. SQN Units 1 and 2 Renewed Operating Licenses Changes Retyped Copy 4. Markup of Original LAR, Attachment 4: External Hazards Screening Pages 5. Markup of Original LAR, Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty Pages 6. Attachments 4 and 6 Retyped Copy cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant NRC Project Manager – Sequoyah Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation Enclosure 1

Response to Request for Additional Information Regarding Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors” (SQN-TS-17-06) (EPID: L-2018-LLA-0066)

Nuclear Regulatory Commission (NRC) Introduction

The regulatory requirements and guidance which the U.S. Nuclear Regulatory Commission (NRC) staff is considering in its review of the application include the following:

RAI 01 – Appendix X, Close-out of Facts and Observations

Section 3.3 of the LAR states that [a] finding closure review was conducted on the identified PRA models on May 8 to May 10, 2017. Closed findings were reviewed and closed using the process documented in the NEI letter to the NRC, “Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&O) 1,” as accepted by the NRC on May 3, 20172. Provide the following information to clarify and confirm that the Independent Assessment for F&O(s) closure was performed consistent with Appendix X to NEI 05-04, NEI 07-12, and NEI 12-06 guidance, governing the process for “Close Out of Facts and Observations” as accepted, with conditions, by the NRC staff via letter dated May 3, 2017.

a. Confirm that the Independent Assessment team was provided with a written assessment and justification of whether the resolution of each F&O, within the scope of the Independent Assessment, constitutes a PRA upgrade or maintenance update, as defined in American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) RA-Sa-2009 and qualified by RG 1.200, Revision 2.

OR

Alternatively, perform a subsequent Independent Assessment for F&O(s) closure and/or addendum to the Independent Assessment report to address the inconsistency with Appendix X, as accepted, with conditions, by the NRC staff via letter dated May 3, 2017. Provide any F&Os or items remaining open as a result of this review. For each F&O and/or item that remains open, provide its associated disposition to demonstrate that it has no adverse impact on the 10 CFR 50.69 risk-informed application.

b. Appendix X guidance states in part, “[t]he relevant PRA documentation should be complete and have been incorporated into the PRA model and supporting documentation prior to closing the finding.” For closure of F&O(s) after the on-site review, Appendix X guidance explicitly states, “[t]he host utility may, in the time between the on-site review and the finalization of the independent assessment team report, demonstrate that the issue has been addressed, that a closed finding has

1 Anderson, V. K., Nuclear Energy Institute, letter to Stacey Rosenberg, U.S. Nuclear Regulatory Commission, “Final Revision of Appendix X to NEI 05-04/07-12/12-16, ‘Close-Out of Facts and Observations,’” dated February 21, 2017 (ADAMS Package Accession No. ML17086A431).

2 Giitter, J., and Ross-Lee, M. J., U.S. Nuclear Regulatory Commission, letter to Mr. Greg 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 (ADAMS Accession No. ML17079A427).

CNL-19-002 E1-1 of 29 Enclosure 1

been achieved, and that the documentation has been formally incorporated in the PRA Model of Record [MOR].”

i. Confirm that all model changes associated with the closure of all F&Os reviewed during the Independent Assessment performed in May 2017 were incorporated into the PRA and/or the supporting documentation at the time of the finalization of the Independent Assessment team report, consistent with Appendix X, as accepted, with conditions, by the NRC staff via letter dated May 3, 2017 (ADAMS Accession No. ML17079A427).

OR

ii. Perform a subsequent Independent Assessment for F&O closure and/or addendum to the Independent Assessment F&O closure report to address the identified inconsistency with Appendix X, as accepted, with conditions, by the NRC staff in letter dated May 3, 2017. Provide any F&Os that remain open as a result of this review. For each F&O and/or item that remains open, provide its associated disposition to demonstrate that it has no adverse impact on the 10 CFR 50.69 risk-informed application.

OR

iii. Alternatively, propose a mechanism that assures all the PRA model logic and all documentation changes reviewed by the Independent Assessment team for the closure of all F&Os in the final Independent Assessment report are incorporated into the MOR(s) prior to implementation of the 10 CFR 50.69 risk-informed categorization. Summarize why this deviation the May 3, 2017, Appendix X, whereby an intermediate PRA model and documentation is reviewed instead of the current MOR, is acceptable for use in the 10 CFR 50.69 risk-informed application.

c. Appendix X guidance states in part, “[i]n some cases, the Independent Assessment team may be assembled such that some reviewers are only needed for a limited number of finding reviews, and it may be possible to have these reviewers participate remotely. This remote participation should be supported with web and teleconference connection to the on-site review team, and the remote reviewers should participate in relevant consensus sessions.”

i. If remote (i.e. subsequent reviews) were performed following the Independent Assessment team’s onsite review, provide a brief summary describing the subsequent review performed. Include details for the NRC staff to confirm consistency with Appendix X (i.e., if the subsequent review and consensus session was remote using web conferencing, or face-to-face and the number of participants).

OR

ii. Alternatively, perform a subsequent Independent Assessment for F&O closure and/or addendum to the Independent Assessment report to address the identified inconsistency with Appendix X, as accepted, with conditions, by the NRC staff in letter dated May 3, 2017. Provide any F&Os that remain

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open as a result of this review. For each F&O and/or item that remains open, provide its associated disposition to demonstrate it has no adverse impact on the 10 CFR 50.69 risk-informed application.

d. Appendix X guidance states in part, the team will review the Supporting Requirement (SR) to ensure that the aspects of the underlying SR that were previously not met, or met at [Capability Category] CC I, are now met, or met at CC II.

i. Explain how closure of all F&Os was assessed to ensure that the capabilities of the PRA elements, or portions of the PRA within the elements, associated with the closed F&Os now meet ASME/ANS RA-Sa-2009 SRs at CC II.

ii. For any F&Os associated with SR IFQU-A6, and SR HR-I1 that were determined to be closed during the Independent Assessment performed in May 2017, include detailed justification for why the supporting SR was considered to be met at CC II.

OR

iii. Alternatively, perform a subsequent Independent Assessment for F&O closure and/or addendum to the Independent Assessment report to address the inconsistency with Appendix X, as accepted, with conditions, by the NRC staff in letter dated May 3, 2017. Provide any F&Os that remain open as a result of this review. For each F&O and/or item that remains open, provide its associated disposition to demonstrate that it has no adverse impact on the 10 CFR 50.69 risk-informed application.

TVA Response to RAI-01a

The Independent Assessment Team was not provided with a written assessment of whether the F&O resolutions constituted an upgrade or a maintenance update as part of the F&O Closure process defined by Appendix X to NEI 05-04/07-12/12-16. Therefore, TVA’s response to part b of this response is provided below.

TVA Response to RAI-01b

TVA performed a subsequent review of the F&O resolutions in December 2018. The results of this review are described below.

The F&O Closure states:

“TVA provided the assessment team descriptions of how the F&Os were resolved. However, the information provided did not include a self-assessment of whether each resolution constituted a PRA upgrade or use of a new PRA method, as required by NEI 05-04 Appendix X Section X.1.3. However, the absence of this update/upgrade self assessment did not negatively impact the ability of the assessment team in performance of their review, the conclusions reached for each F&O, or the team’s assessment of whether the resolution constituted a PRA Upgrade.”

CNL-19-002 E1-3 of 29 Enclosure 1

TVA Response to RAI-01b.iii

RAI-01b allows the option to perform items b.i, or b.ii, or b.iii. TVA is responding to option b.iii as follows.

TVA did not incorporate the F&O resolutions into the MOR. However, as documented in the F&O Closure Report, the changes initiated by the F&O resolutions were confirmed by the Independent Assessment Team to have been incorporated into the living model and associated documentation. Therefore, TVA is proposing a license condition that requires the MOR to be updated with the F&O resolutions from the closed F&Os prior to system categorization (see Attachment 1 to this enclosure). Enclosure 2 contains the markup of the proposed license condition and Enclosure 3 contains the retyped proposed license condition.

TVA Response to RAI-01c

RAI-01c allows the option to perform items c.i, or c.ii. TVA is responding to option c.ii as follows.

As noted in the response to RAI-01b, TVA performed a subsequent review of the F&Os in December 2018. The same team that performed the initial review of the F&Os, also performed the subsequent review, reviewing the same Probabilistic Risk Analysis (PRA) models and notebooks that were presented during the 2017 review. This review was performed remotely, with web-based consensus sessions used. A specific evaluation was also provided for each closed F&O to document whether the review team considered the F&O resolution a “PRA Maintenance Update” or a “PRA upgrade.” No F&Os remain open.

TVA Response to RAI-01d

RAI-01d allows the option to perform items d.i, and d.ii, or perform item d.iii. TVA is responding to option d.iii as follows.

The purpose of the subsequent review, as described in the response to RAI-01c, was to document for each closed F&O, that the F&O resolution met the CC II requirements of the ASME/ANS PRA Standard’s SRs that were referenced in the F&O. A specific evaluation was also provided for each closed F&O to document whether the review team considered the F&O resolution a “PRA Maintenance Update” or a “PRA upgrade.” No F&Os remain open.

RAI 02 – SSCs Categorization Based on Other External Hazards

Sections 50.69(c)(1)(ii) of 10 CFR require that the licensee determine SSC functional importance using an integrated, systematic process for addressing initiating events (internal and external), SSCs, and plant operating modes, including those not modeled in the plant- specific PRA.

Section 3.2.4 of the LAR states that [a]ll other external hazards were screened from applicability to Sequoyah Units 1 and 2 per a plant-specific evaluation in accordance with GL 88-20 and updated to use the criteria in the ASME/ANS PRA Standard RA-Sa-2009. This statement appears to indicate that TVA proposes to treat all SSCs as low-safety- significant (LSS) with respect to other external events risk. The LAR also states that “[a]s

CNL-19-002 E1-4 of 29 Enclosure 1 part of the categorization assessment of other external hazard risk, an evaluation is performed to determine if there are components being categorized that participate in screened scenarios and whose failure would result in an unscreened scenario,” and that “[c]onsistent with the flow chart in Figure 5-6 of Section 5.4 of NEI 00-04, these components would be considered [high-safety-significant] HSS”. Attachments 4 and 5 of the LAR provide a summary of the other external hazards screening results, but do not appear to address any considerations related to applying Figure 5-6 of NEI 00-04 to those hazards.

a. Identify the external hazards that will be evaluated according to the flow chart in NEI 00-04, Section 5.4, Figure 5-6. Provide detailed justification for screening external hazards (i.e., external flood, high winds, and tornados) using the criteria in Part 6 of ASME/ANS RA-Sa-2009. As applicable, the justification should include consideration of uncertainties in the determination of demonstrably conservative mean values as discussed in Section 6.2-3 of the ASME/ANS RA-Sa-2009 PRA Standard.

i. For screening criterion PS1 Attachment 5 of the LAR, provide justification for concluding that the external flooding, high winds and tornados hazard(s) cannot cause a core damage accident.

ii. For the external flooding hazard, provide detailed justification for concluding that screening criterion PS4 applies, i.e. external flooding hazard CDF is less than 1E-06/reactor-year.

iii. For the high winds and tornados hazard, provide detailed justification for concluding that the screening criterion PS3 applies, i.e. the mean frequency is less than 1×10-5 per reactor-year and the mean conditional core damage probability is less than 0.1.

b. Figure 5-6 of NEI 00-04 illustrates that if an SSC is included in a screened scenario(s), then for that SSC to be considered a candidate LSS, the licensee has to demonstrate that upon removal of the component, the screened scenario(s) would not become unscreened.

i. Identify and justify what type of SSCs, if any, are credited in the screening of the external hazard(s), including both passive, active, and temporary features.

ii. If there are any SSCs credited for screening of the external hazard(s), then explain and justify how the guidance in Figure 5-6 of NEI 00-04 will be applied for each of the external hazard(s). c. If the external hazards (i.e., external flood, high winds and tornados) cannot be screened out in item (a), discuss, using quantitative or qualitative assessments, how the risk from those hazards will be considered in the 10 CFR 50.69 categorization process. The discussion should include consideration of and, as applicable, the basis for the following factors:

• The frequency of the external hazard(s), • The impact of the external hazard(s) on plant SSCs and plant’s operation including the ability to respond to the external hazard initiating event, • The operating experience associated with reliability of the external hazard(s) protection measures (e.g., flood seals), and

CNL-19-002 E1-5 of 29 Enclosure 1

• The reliability of operator actions.

TVA Response to RAI-02a

NEI 00-04, Figure 5-6 “Other External Hazards” provides the NRC approved process to be used to determine Structures, Systems, and Components (SSC) safety significance for other external hazards (excluding internal fires and seismic hazards). TVA is following the NEI 00-04, Section 5.4 for assessment of other external hazards. Therefore, TVA is subjecting the external hazards (excluding internal fires and seismic hazards) to the process described by the flow chart in NEI 00-04, Figure 5-6. As part of the categorization assessment of “other external hazard” risk, an evaluation is performed to determine if there are components being categorized that participate in screened scenarios and whose failure would result in an unscreened scenario. Those components would be classified as high safety significant (HSS).

TVA Response to RAI-02a.i

In Attachment 4 of the License Amendment Request (LAR), TVA screened the external flooding hazard using criterion PS1 and PS4. Extreme winds and tornado hazards were screened using criterion PS1 and PS3. Based on further evaluation, TVA has determined that the more appropriate screening criteria should be C5 for external flooding and PS2 for extreme winds and tornados as described in the response to RAI-02a.iii. Enclosures 5 and 6 provide the re-typed and final versions, respectively, for the revision to Attachment 4 that was provided in the LAR reflecting the revised screening criteria.

TVA Response to RAI-02a.ii

ASME/ANS PRA 2009 Standard, Screening Table 6-2-3(b), Supporting Requirement EXT-B1, defines criterion C5 as:

“The event is slow in developing, and it can be demonstrated that there is sufficient time to eliminate the source of the threat or to provide an adequate response.”

With respect to external flooding, the Sequoyah Nuclear Plant (SQN) is designed such that there is sufficient warning time, given large rainfall or seismically-induced upstream dam failure, to shut the plant down and implement emergency procedures. Plant shutdown is based on a flood warning scheme divided into two stages (i.e., Stage I and Stage II). Stage I is a minimum of ten hours and Stage II is a minimum of 17 hours. During Stage I, preparation steps are taken for flood mitigation. If conditions persist, Stage II is entered whereby the operator moves to initiate plant shutdown. Therefore, the minimum time calculated that flooding could exceed plant grade is 27 hours. As noted in Section 2.4.10 of the SQN UFSAR:

“Any rainfall flood exceeding plant grade will be predicted at least 27 hours in advance by TVA's Reservoir Operations. Warning of seismic failure of key upstream dams will be available at the plant at least 27 hours before a resulting flood surge would reach plant grade. Hence, there is adequate time to prepare the plant for any flood.”

Therefore, the timing available to exceed plant grade represents a slow moving event that meets the criteria for C5 that there is sufficient time to provide an adequate response.

CNL-19-002 E1-6 of 29 Enclosure 1

In accordance with the referenced letter, TVA committed to provide the NRC with a revised SQN warning time analysis. Therefore, as noted in Attachment 1 to this enclosure, TVA will re-confirm that there is sufficient time to eliminate the source of the threat or to provide an adequate response in accordance with criterion C5, prior to 50.69 categorization.

Reference

TVA letter to NRC, CNL-16-176, “Mitigating Strategies Assessment for Flooding for Sequoyah Nuclear Plant, Units 1 and 2 - Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident,” dated December 23, 2016 (ML16363A382)

TVA Response to RAI-02a.iii

As noted in the response to RAI-02a.i, TVA is applying criterion PS2 for the extreme winds and tornado hazard. Section 6-2.3 of ASME/ANS Ra-Sa-2009, criterion (a) states an event can be screened out if it meets the criteria in the NRC’s 1975 SRP or a later revision. Because SQN was designed prior to the 1975 SRP, the approach taken in the Individual Plant Evaluation of External Events (IPEEE) was to review the design bases and compare them to the SRP requirements (screening criteria). Any changes that were made to the plant subsequent to the design analysis were reviewed to verify compliance with SRP criteria. For Other External Events, it was found that no vulnerabilities exist outside the screening thresholds of the SRP. Therefore, the SQN design meets the 1975 SRP criteria for extreme winds and tornadoes and the requirements for screening the hazards.

Enclosures 4 and 6 provide the markup and final versions, respectively, for the revision to Attachment 4 that was provided in the LAR to reflect the revised screening criteria.

TVA Response to RAI-02b.i

NEI 00-04, Figure 5-6, addresses the approach for risk-informed safety classification of SSCs for Other External Hazards (excluding seismic and internal fires). The SQN IPEEE concluded all Other External Hazards (excluding seismic and internal fires) were screened from further consideration in the examination of these hazards. There was no identification of SSCs that supported the screening of these hazards.

TVA Response to RAI-02b.ii

NEI 00-04, Figure 5-6, addresses the approach for risk-informed safety classification of SSCs for Other External Hazards (excluding seismic and internal fires). While there was no identification of SSCs that supported the screening of these hazards, TVA is providing this additional information with respect to the Other External Hazards review performed during the categorization process. As part of the categorization process for a given system, the SSCs listed for that system are assessed for safety significance, including whether that SSC is important for protecting against an Other External Hazard. That review includes a review of procedures and other documentation, as applicable, for Other External Hazards. The SSCs for a system are subjected to categorization and not limited to those types of components listed in RAI-02(b)(i) with respect to Other External Hazards.

CNL-19-002 E1-7 of 29 Enclosure 1

TVA Response to RAI-02c

As noted in the response to RAI-02a, the external hazards (i.e., external flood, extreme winds, and tornados) were screened out; therefore, no further response is required.

RAI 03 – License Condition for 10 CFR 50.69

The NEI 00-04 guidance allows licensees to implement different approaches, depending on the scope of their PRA (e.g., the approach if a seismic margins analyses is relied upon is different and more limiting than the approach if a seismic PRA is used). Regulatory Guide 1.201 states that “as part of the NRC's review and approval of a licensee's or applicant's application requesting to implement §50.69, the NRC staff intends to impose a license condition that will explicitly address the scope of the PRA and non-PRA approaches used in the licensee's categorization approach.”

The March 16, 2018 LAR proposed the following License Condition:

TVA is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the license amendment dated [XXXX].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach and or change from fire SSEL to a fire probabilistic risk assessment approach).

TVA shall complete the items listed in Attachment 1, List of Categorization Prerequisites, of TVA letter dated [XXXX], prior to implementation.

The proposed license condition does not explicitly address the PRA and non-PRA approaches that were used. Please provide a licensee condition that explicitly addresses the approaches, e.g.

TVA is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) model to evaluate risk associated with internal events, including internal flooding; the alternative method approved by the NRC staff in the safety evaluation using the results of the [fire safe shutdown equipment list for internal fire as supplemented in letter dated [MONTH, DAY YEAR] (ADAMS Accession No. ML XXXXXXX); the NUMARC 96-01 shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and 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 Unit [x] License Amendment [Number].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins

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approach to a seismic probabilistic risk assessment approach, change from alternative method for internal fire to a fire probabilistic risk assessment approach).

TVA shall complete the items listed in Attachment 1, List of Categorization Prerequisites, of TVA letter dated [XXXX], prior to implementation.

Note: The license condition may need to be expanded to address any implementation items identified in response to the RAIs.

TVA Response to RAI-03

In accordance with RAI-03, the proposed changes to the SQN Units 1 and 2 Renewed Facility Operating Licenses provided in the referenced letter will be modified as follows (see Enclosures 2 and 3 to this letter):

Adoption of 10 CFR 50.69, “Risk-informed categorization and treatment of structures, systems and components for nuclear power plants”

(1) TVA is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) model to evaluate risk associated with internal events, including internal flooding; using the fire safe shutdown equipment list in the SQN Fire Protection Report referenced in the Updated Final Safety Analysis Report to evaluate internal fire events; the NUMARC 96-01 shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the criteria in the endorsed ASME/ANS RA-Sa-2009 PRA Standard for other external hazard screening significance; as specified in Unit [x] License Amendment [Number].

(2) Prior to implementation of the provisions of 10 CFR 50.69, TVA shall complete the items below;

a. Items listed in Enclosure 1, Attachment 1, “SQN 10 CFR 50.69 PRA Implementation Items,” in TVA letter CNL-19-002, “Response to Request for Additional Information Regarding Application to Modify Sequoyah Nuclear Plant Units 1 and 2, Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors,” (SQN-TS-17-06) (EPID: L-2018-LLA-0066),” dated March 21, 2019.

(3) Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach, change from alternative method for internal fire to a fire probabilistic risk assessment approach).

CNL-19-002 E1-9 of 29 Enclosure 1

Reference

TVA Letter to NRC, CNL-17-010 “Sequoyah Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors,” dated March 18, 2018 (SQN-TS-17-06) (ML18075A365)

RAI 04 – Key Assumptions and Sources of Uncertainties

Paragraphs 50.69(c)(1)(i) and (ii) of 10 CFR requires a licensee’s PRA be of sufficient quality and level of detail to support the SSC categorization process, and requires that all aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience. The guidance in NEI 00-04 specifies sensitivity studies to be conducted for each PRA model to address uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask the SSC(s) importance.

In Section 4.1 of the LAR, Sequoyah identifies RG 1.174, Revision 2, as an applicable regulatory requirement/criteria. RG 1.174 has been updated to Revision 3, dated January 2018 (ADAMS Accession No. ML17317A256). Regulatory Guide 1.174, Revision 3, cites NUREG-1855, Revision 1, as related guidance. In Section B of RG 1.174, Revision 3, the guidance acknowledges specific revisions of NUREG-1855 to include changes associated with expanding the discussion of uncertainties. Section 3.2.7 of the LAR states in part, [t]he detailed process of identifying, characterizing and qualitative screening of model uncertainties is found in Section 5.3 of NUREG-1855, [“Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making,” March 2009 (Revision 0) (ADAMS Accession No. ML090970525)] and Section 3.1.1 of EPRI Technical Report (TR)-1016737. Attachment 6 of the LAR provides four key assumptions and sources of uncertainties applicable to the internal events (includes flooding) PRA (IEPRA) model.

NUREG-1855 has been updated to Revision 1 as of March 2017 (ADAMS Accession No. ML17062A466). The NRC staff notes that NUREG-1855, Revision 1, provides guidance in stages A through E for how to treat uncertainties associated with PRA models in risk- informed decision-making. Revision 1 of NUREG-1855 cites EPRI TR-1026511, “Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty.” Considering these observations provide the following: a. A detailed summary of the process used to identify the key assumptions and sources of uncertainty presented in Attachment 6 of the LAR. The discussion should include:

i. How the process is consistent with NUREG-1855, Revision 1, or other NRC-accepted methods (e.g., NUREG-1855, Revision 0). If deviating from the current guidance provided in NUREG-1855, Revision 1, provide a basis to justify the appropriateness of any deviations for use in the 10 CFR 50.69 categorization process (e.g., exclusion/consideration of EPRI TR-1026511).

ii. A brief description of how the key assumptions and sources of uncertainties provided in Attachment 6 of the LAR were identified from the initial comprehensive list of PRA model(s) (i.e., base model) source of uncertainties

CNL-19-002 E1-10 of 29 Enclosure 1

and assumptions, including those associated with plant-specific features, modeling choices, and generic industry concerns. This can include an identification of the sources of plant-specific and applicable generic modeling uncertainties identified in the uncertainty analyses for the base internal events and internal flooding PRA. Include a disposition for each of the assumptions and/or uncertainties addressing their impact on the 10 CFR 50.69 risk-informed categorization process. For any source of uncertainty or assumption judged not to be key to the application, provide discussion for why it is not pertinent to the application and therefore does not need to be further addressed (i.e., sensitivity does not need to be performed). b. If the process used to identify, characterize, and assess the key assumption(s) and the treatment for the sources of uncertainty provided in Attachment 6 of the LAR cannot be justified for use in the 50.69 categorization process, provide the results of an updated assessment of the key assumptions, sources of uncertainty, and treatment of the sources of uncertainty performed in accordance with NUREG-1855, Revision 1, and NEI 00-04, Revision 0. If sensitivity studies are proposed to be performed for the treatment of a specific key assumption or source of uncertainty, include a detailed description of the proposed sensitivity studies and justify how it is bounding to address the specific key assumption and/or source of uncertainty.

TVA Response to RAI-04a.i

Substep E-1.1 of NUREG 1855, Revision 1 recommends using the detailed guidance and a generic list of sources of model uncertainty and related assumptions in EPRI 1016737 for the internal event hazards group, including Large Early Release Frequency (LERF), and using the examples of sources of model uncertainty for the internal fires, seismic, Low Power Shutdown and Level 2 hazard groups in EPRI 1026511. For SQN, this process was performed by reviewing PRA documentation for generic issues identified in Table A-1 of EPRI 1016737, as well as identifying plant-specific assumptions and uncertainties, and is therefore consistent with substep E-1.1 of NUREG-1855, Revision 1.

Substep E-1.2 of NUREG 1855, Revision 1 involves identifying those sources of model uncertainty and related assumptions in the base PRA that are relevant to an application. Those that are irrelevant can be screened from further discussion. However, since this application uses the internal events and internal flood model for both Core Damage Frequency (CDF) and LERF, all model uncertainties and related assumptions identified for these models are considered relevant. Therefore, SQN is consistent with substep E-1.2 of NUREG-1855, Revision 1.

Substep E-1.3 of NUREG 1855, Revision 1 involves characterizing the identified sources of model uncertainty and related assumptions. This characterization involves understanding how the identified sources of model uncertainty and related assumptions can affect the PRA. For the SQN uncertainty analysis, this was performed for all identified uncertainties and assumptions. Therefore, SQN is consistent with substep E-1.3 of NUREG-1855, Revision 1.

Substep E-1.4 is a qualitative screening process that involves identifying and validating whether consensus models have been used in the PRA to evaluate identified model uncertainties. As stated in NUREG 1855, Revision 1, the use of a consensus model eliminates the need to explore an alternative hypothesis. For the SQN uncertainty analysis,

CNL-19-002 E1-11 of 29 Enclosure 1 some uncertainties and assumptions were screened based on their use of a consensus method.

Once the relevant uncertainties and assumptions are identified, Steps E-1 and E-2 of NUREG 1855, Revision 1 provides guidance for identifying those sources of model uncertainty and related assumptions that are key to this application. The input to this step is the list of the relevant sources of model uncertainty identified in Step E-1. These sources of model uncertainty and related assumptions are then quantitatively assessed to identify those with the potential to impact the results of the PRA such that the application’s acceptance guidelines are challenged. This assessment is made by performing sensitivity analyses to determine the importance of the source of model uncertainty or related assumption to the acceptance criteria or guidelines. This assessment is consistent with substep E-1.4 of NUREG-1855, Revision 1.

For those uncertainties and related assumptions that are key to the application (i.e., it cannot be quantitatively shown that they do not have the potential to impact the acceptance criteria), Stage F (section 8) of NUREG 1855, Revision 1, provides guidance on justifying the strategy used to address the key uncertainties that contribute to risk metric calculations that challenge application-specific acceptance guidelines. This portion of the NUREG was not addressed in the original SQN uncertainty analysis.

TVA Response to RAI 04a.ii

In the referenced email, NRC revised the wording to the RAI as follows (revisions to the RAI are shown in bold italics):

“Include a disposition for each assumption and/or source of uncertainty determined to be key for the application, addressing its impact on the 10 CFR 50.69 risk- informed categorization process, and describing how it will be addressed for the application (e.g. use of a sensitivity study). If it is determined that a key source of uncertainty or assumption is not pertinent to the application, provide justification why it does not need to be further addressed (i.e., sensitivity does not need to be performed).”

The development of the PRA input to the SQN LAR to adopt 10 CFR 50.69 was performed in 2017. The applicable regulatory requirement/criteria at that time was Regulatory Guide (RG) 1.174 Revision 2. RG 1.174, Revision 3 was issued in January 2018, which endorsed NUREG-1855, Revision 1. SQN’s uncertainty and sensitivity analysis was performed in accordance with NUREG-1855, Revision 0. In NUREG-1855, Revision 1, Step E-1, identification of potential model uncertainties and related assumptions and determining their significance is performed. In Step E-2, the analyst takes the E-1 information and identifies key sources of model uncertainty and related assumptions.

These sources of model uncertainty and related assumptions are then quantitatively assessed to identify those with the potential to impact the results of the PRA such that the application’s acceptance guidelines are challenged. This assessment is made by performing sensitivity analyses to determine the importance of the source of model uncertainty or related assumption to the acceptance criteria or guidelines. In the SQN uncertainty analysis, for any uncertainties that were not previously screened qualitatively, they were analyzed quantitatively in order to determine the impact on the PRA results. Uncertainties and assumptions that were identified as having an impact on the PRA results were then assessed to determine those applicable to the 10 CFR 50.69 application. That

CNL-19-002 E1-12 of 29 Enclosure 1 subset was then reviewed further to identify those that could potentially have an impact on the risk metrics affecting the decision criteria for safety significance, which are referred to as “key.” The result of this assessment are the key assumption/sources of uncertainty (Attachment 6 of the LAR), with the exception of State of Knowledge Correlation (SOKC) uncertainty as discussed further in this RAI response.

For those uncertainties and related assumptions that are key to the application (i.e., it cannot be quantitatively shown that they do not have the potential to impact the acceptance criteria), Stage F (section 8) of NUREG 1855, Revision 1, provides guidance on justifying the strategy used to address the key uncertainties that contribute to risk metric calculations that challenge application-specific acceptance guidelines. This portion of the NUREG was not addressed in the original SQN uncertainty analysis.

As noted in the response to RAI-05a for internal flooding scenarios, a change (decrease) in the probability of detection (POD) that a flaw will be detected by inspection results in the integrity management factors increasing. This could potentially result in an increase in the CDF and the LERF, which in turn could result in increased risk importance measures (e.g., Fussell-Vesely (F-V) and Risk Achievement Worth (RAW)) of the affected equipment. Table 5-2 of NEI 00-04 states, “Any applicable sensitivity studies identified in the characterization of PRA adequacy.” Therefore, during categorization, this key uncertainty could be addressed by identifying the areas of the plant where the components are located, then adjusting the POD for the internal flooding sources affecting that area, then quantifying the model to determine the impact on the affected SSCs. In accordance with NEI 00-04, the results of the sensitivity study are given to the Integrated Decision-making Panel (IDP) for consideration in the final risk characterization for components initially classified as LSS that may be reclassified to HSS.

As noted in the response to RAI-05b.ii, one way to assess the effect of a different maintenance induced flood frequency would be to recalculate the frequency. NEI 00-04 Table 5-2, states “Any applicable sensitivity studies identified in the characterization of PRA adequacy.” Therefore this uncertainty could be assessed by adjusting the maintenance induced flood frequencies. Once the maintenance-induced flood frequency is calculated, it is then added to the human-induced flood frequency and the passive pipe rupture frequency in order to obtain the appropriate internal flooding initiator frequency. The impact on the PRA metrics is determined by re-quantifying the model. By changing the maintenance-induced flood frequency, certain flooding initiator frequencies could potentially increase resulting in a higher core damage frequency caused by those flooding scenarios. This may cause the calculated RAW values for the equipment credited to mitigate core damage, or release from this scenario to potentially increase. In accordance with NEI 00-04, the results of the sensitivity study are given to the IDP for consideration in the final risk characterization for components initially classified as LSS that may be reclassified to HSS.

As noted in the response to RAI-05c, the SQN demand data includes successful demands from post-maintenance testing (PMT). NEI 00-04, Table 5-2, states “Any applicable sensitivity studies identified in the characterization of PRA adequacy.” Therefore, during categorization, the impact of PMTs will be determined by a sensitivity study that will be used to account for the additional PMT test data. The sensitivity study would be to provide a PMT successful demand contribution that is conservative. By changing the number of PMTs used in the type code data, the calculated CDF and LERF could potentially increase. This may cause the calculated RAW values for the equipment credited to mitigate core damage

CNL-19-002 E1-13 of 29 Enclosure 1 and/or release to change. In accordance with NEI 00-04, the results of the sensitivity study are given to the IDP for consideration in the final risk characterization for components initially classified as LSS that may be reclassified to HSS.

As noted in the response to RAI-05d, the key uncertainty described in Attachment 6 of the LAR is no longer applicable because the SOKC has been removed from the SQN MOR. However, in review of RAI-05d, TVA identified that addressing SOKC for the Interfacing System Loss of Cooling Accident (ISLOCA) analysis is required to meet CC II for SR QU-A3, which states, “ESTIMATE the mean CDF accounting for the state-of-knowledge correlation between event probabilities when significant.” As stated in the response to RAI-05d, TVA has revised Attachment 2 to this enclosure to include SOKC in the PRA model prior to categorization. An example of how SOKC sensitivity can be addressed is with a multiplier method that increases the SOKC contribution to risk.

Reference

NRC Electronic Mail to TVA, “FW: Clarification Call Follow Up for RAI 04,” dated March 8, 2019.

TVA Response to RAI 04b

The following figure from NUREG-1855, Revision 0 describes the process used to identify, characterize, and assess the key assumption(s) and the treatment for the sources of uncertainty provided in Attachment 6 of the LAR, which the SQN analysis followed.

Sources of Model Uncertainty and Related Assumptions Rekwant to Application

Define and Justify Sensitivity Cases Individual Source of Model Uncertainty Logical Combinations

Acceptance Perform Screening Sensitivity Analyses Guidelines Associated with Conservative Application Realistic

Sources of Model Uncertainty and Related No Sources of Model Assumptions Challenge Uncertainty and Related Acceptance Criteria? Assumptions NOT Key to Application Yes

Sources of Model Unce rtainty and Related Assumptions Key to Application

Figure 4

CNL-19-002 E1-14 of 29 Enclosure 1

As a comparison, the process described in NUREG-1855, Revision 1 is shown in the following figure.

Step E-2

Sources of model uncertainty and related Acceptance guidelines associated assumptions relevant to the application with application ~ , I ~ r ..- I Perform screeninQ and sensitivity I

Do the sources of model uncertainty and re lated assumptions result in an initial challenge to or exceedance of the accepta nce guidelines? ' YES ' NO ' Sources of model uncertainty Sources of model uncertainty and related assumptions are and related assumptions are treated as key to the application NOT key to the application I I + Stage F Li censee Application Development Process

Figure 5

The difference between the two figures is the addition of Stage F “Licensee Application Development Process.” In Stage F, a strategy is developed to address any key uncertainties and to propose any compensatory measures or performance monitoring programs, as appropriate. In accordance with 10 CFR 50.69(e), Section 12 of NEI 00-04, and Section 3.1.1 “Overall Categorization Process” of the LAR, TVA will perform periodic reviews of SSC performance. This review is in accordance with Stage F of NUREG-1855, Revision 1 and is an appropriate method for dealing with uncertainties and related assumptions that challenge or exceed the acceptance guidelines. Section 8.5 of NUREG-1855, Revision 1, states:

“Performance monitoring can be used to demonstrate that; “following a change to the design of the plant or operational practices, there has been no degradation in specified aspects of plant performance that are expected to be affected by the change. This monitoring is an effective strategy when no predictive model has been developed for plant performance in response to a change.”

As noted in the response to RAI-04a.ii, the four key sources of uncertainty listed in Attachment 6 of the LAR are discussed. The over-conservatisms introduced by the indicated sensitivity studies bound the impact of the uncertainty on the risk metrics and is

CNL-19-002 E1-15 of 29 Enclosure 1 used as an input for the IDP in determining the safety significance of SSCs initially given an LSS classification.

RAI 05 – Dispositions of Key Assumptions and Sources of Uncertainties

Paragraph 50.69(c)(1)(i) of 10 CFR requires the licensee to consider the results and insights from the PRA during categorization. The guidance in NEI 00-04 specifies sensitivity studies to be conducted for each PRA model. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask importance of components. NEI 00-04 guidance states that additional “applicable sensitivity studies” from characterization of PRA adequacy should be considered.

NRC staff observes that even small impacts to each hazard PRA model’s CDF and LERF (both increases and decreases) could potentially increase the risk importance values for certain SSCs above the NEI 00-04 Section 5 threshold criteria for determining safety significance. Additionally, conservatisms can mask the importance measures of other SSCs. The dispositions presented in Attachment 6 of the LAR for key assumptions and modeling uncertainties did not provide sufficient information for the NRC staff to conclude that the uncertainties do not impact the 10 CFR 50.69 categorization results. Considering these observations, address the following: a. For the disposition of the key assumption that a leaking pipe will be detected by visual inspection with a 0.9 probability, the licensee stated that internal flooding is a significant contributor to plant risk and that the prescribed sensitivity studies in Section 5 of NEI 00-04 will be followed.

Describe which of the sensitivity studies outlined in Section 5 of NEI 00-04 is directly applicable for this key assumption. Include in the description, justification that addresses: (1) how the sensitivity study bounds the source of uncertainty being addressed and (2) how the potential to mask/skew the importance measures of other SSCs is considered.

b. Attachment 6 of the LAR identifies passive pipe break failures; human-induced flooding and maintenance induced flooding as a key source of uncertainty. The disposition provides statement of how the impact can be treated (e.g., random sample via Monte Carlo method), but it does not provide an explicit disposition for the application.

i. Provide justification for why the uncertainty associated with passive pipe break failures, human-induced flooding and maintenance induced flooding has no adverse impact (mask/skew the importance measures of other SSCs) on the 10 CFR 50.69 risk-informed categorization process, and therefore does not need to be addressed,

OR

ii. Describe how the uncertainty associated with passive pipe break failures, human-induced flooding and maintenance induced flooding will be addressed during implementation of the 10 CFR 50.69 categorization process, consistent with NEI 00-04 (e.g., sensitivity study). Include justification to address: (1) how the sensitivity study bounds the source of uncertainty under

CNL-19-002 E1-16 of 29 Enclosure 1

consideration and (2) how the potential to mask/skew the importance of other SSCs is considered.

c. Attachment 6 of the LAR identified a key assumption/uncertainty that equipment type code data include successful post-maintenance testing (PMT) and that this can result in an under-estimation of the failure probabilities.

i. Provide discussion for what data type codes are impacted by PMT demands for this identified assumption/uncertainty.

ii. Provide justification for why this uncertainty has no adverse impact (mask/skew the importance measures of other SSCs) on the 10 CFR 50.69 risk-informed categorization, and therefore does not need to be addressed;

OR

iii. Alternatively, describe how this uncertainty will be addressed during implementation of the 10 CFR 50.69 categorization process consistent with NEI 00-04 (e.g., sensitivity study). Include in the description, justification to address (1) how the sensitivity study bounds the source of uncertainty being addressed and (2) how the potential to mask or skew the importance of other SSCs is considered. d. Attachment 6 of the LAR identifies the State of Knowledge Correlation (SOKC) as a source of uncertainty. The discussion for this source of uncertainty states in part “[t]hose events that are used are considered correlated, which implies that the same distribution applies to all sampled events when using a Monte Carlo approach.” The disposition of this source of uncertainty states that the multiplier method used for interfacing system loss of coolant accidents (ISLOCA) is used to address correlation.

i. Describe the method used to address the SOKC uncertainty. Include justification to confirm that the method applied is consistent with Appendix -6-A of NUREG-1855, Revision 1.

OR

ii. If an alternative method was applied to address SOKC that deviates from NUREG-1855, Revision 1; provide justification for why the deviation is acceptable for use in the 10 CFR 50.69 categorization process. Refer to RAI 04(a).i if applicable.

TVA Response to RAI-05a

In accordance with NEI 00-04, the applicable sensitivity study allowed by Table 5-2 states, “Any applicable sensitivity studies identified in the characterization of PRA adequacy.” This would govern the sensitivity study that would be performed during categorization for the key assumption that a leaking pipe will be detected by visual inspection with a 0.9 probability.

CNL-19-002 E1-17 of 29 Enclosure 1

Additionally, the following provides an example of the sensitivity study:

In the Internal Flooding Analysis, EPRI Report 3002000079, Revision 3, “Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments,” is used as it provides a method to reduce the probability of pipe failure by applying a Markov statistical model to represent the probability of observing a leak in a system prior to having a rupture of the system.

The EPRI report lists two different types of detection; the first is a Non-Destructive Evaluation (NDE) and the second is periodic visual inspections for piping. An Integrity Management Factor (IMF) can be applied which is presented in tables throughout Section 3 of the EPRI report. These factors represent curves for different Equivalent Break Sizes (EBS), for visual inspections, NDE testing, and a combination of the two. These factors are used as a multiplier (<1.0) on the pipe break frequencies; therefore, reducing the frequency of the initiating event.

The POD is a probability that a flaw will be detected when a segment of piping is inspected. This parameter is related to the reliability of NDE inspection and is a conditional probability given that the location being inspected has a flaw that meets the criteria for repair according to the ASME Section XI code. The assumption made in the PRA model is that the inspection process (visual or NDE) has a 90% probability of detection for flaws, as defined by the ASME code.

To assess the effect of a different detection probability, each of the IMFs would have to be recalculated and applied to the corresponding flooding scenarios. The flooding analysis would then be re-quantified to observe the impact the various IMFs would have on the results.

By decreasing the POD for the inspections, the IMF will increase, potentially resulting in a higher CDF or release caused by the flooding scenario. This may increase the calculated RAW values for the equipment credited to mitigate core damage (or release) from this scenario. Per NEI 00-04, the sensitivity study reports are provided to the IDP for consideration to determine if any candidate LSS components should be upgraded to HSS.

TVA Response to RAI-05b

Attachment 6 of the LAR identifies passive pipe break failures, human-induced flooding, and maintenance induced flooding as a key source of uncertainty. Passive pipe breaks and human induced flooding are addressed in the response to RAI-05b.i, while maintenance induced flooding is addressed in the response to RAI-05b.ii.

TVA Response to RAI-05b.i

As noted in Attachment 6 of the LAR, the Internal Flooding analysis calculation uses a summation of three different frequencies; passive pipe break failures, human-induced flooding, and maintenance induced flooding each having their own inherent uncertainties. The internal flooding calculation inputs are variable (such as the repair interval for a component) and were assumed to occur under a static timeframe to bound the analysis but still achieve reasonable results.

CNL-19-002 E1-18 of 29 Enclosure 1

The uncertainty associated with the passive pipe break failures are treated the same as other events within the Computer Aided Fault Tree Analysis (CAFTA) fault tree model and are assessed through the UNCERT tool. The uncertainty associated with human-induced flooding events is analyzed through the Human Reliability Analysis (HRA) Calculator program and is treated the same as the other HRA actions within the model.

TVA Response to RAI-05b.ii

In accordance with NEI 00-04, the applicable sensitivity study allowed by Table 5-2 states, “Any applicable sensitivity studies identified in the characterization of PRA adequacy.” This would govern the sensitivity study that would be performed for this assumption during categorization. Additionally, the following provides an example of the sensitivity study:

Regarding the maintenance-induced flooding events:

The maintenance induced flood is determined using equation (1 e ) f −λt activity Where: − ∗ λ = the generic frequency for an MOV internal rupture (NUREG/CR-6928) t = the mission time of the repair (7 days) f = the frequency of the maintenance activity in years

activity The internal rupture of a motor-operated valve (MOV) probability is taken from NUREG/CR-6928, “Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants,” dated February 2017 (ML070650650), thus, the uncertainty associated with that value is inherent within the generic number. This is a conservative value as there have been no incidents at SQN with an MOV large internal rupture. Updating the MOV internal rupture frequency with plant specific data would reduce this probability. The mission time of the repair interval is seven days. Should the maintenance activity last more than seven days, the maintenance-induced frequency would increase by 8E-08 per year for each day over the seven-day timeframe. The frequency of the maintenance activity is analyzed based on the documented frequency within the procedure. Any associated conditions that could result in additional maintenance are not assessed, as those are unknown throughout the year until the condition is reached and the activity performed. If additional maintenance activities were included to account for plant conditions that initiate the maintenance activity, the maintenance-induced frequency would increase.

In order to assess the effect of a different maintenance induced flood frequency, the frequency would have to be recalculated. Once the maintenance-induced flood frequency is calculated, it is then added to the human-induced flood frequency and the passive pipe rupture frequency in order to obtain the appropriate internal flooding initiator frequency. The flooding analysis would then have to be re-quantified to ascertain the associated impact.

By changing the maintenance-induced flood frequency, certain flooding initiator frequencies could potentially increase resulting in a higher core damage frequency caused by those flooding scenarios. This may cause the calculated RAW values for the equipment credited to mitigate core damage, or release from this scenario, to potentially increase. In accordance with NEI 00-04, the results of the sensitivity study are given to the IDP for consideration in the final risk characterization for components initially classified as LSS that may be reclassified to HSS.

CNL-19-002 E1-19 of 29 Enclosure 1

TVA Response to RAI-05c

RAI-05c allows the option to perform items c.i, c.ii, or c.iii. TVA is responding to option c.iii as follows.

In accordance with NEI 00-04, the applicable sensitivity study allowed by Table 5-2 states, “Any applicable sensitivity studies identified in the characterization of PRA adequacy.” This would govern the sensitivity study that would be performed for this assumption during categorization. Additionally, the following provides an example of the sensitivity study.

Attachment 6 of the LAR states that the SQN PRA data includes some successful PMT demands, and that sensitivity studies have been performed that show the inclusion of PMT data has a small impact on CDF. When plant specific data was gathered for the fails-to-start type codes, the PMTs were tracked for a subset of the overall data window. A number of PMT demands were identified and removed from the number of successful demands for each type code. As data was updated, PMTs performed subsequent to the initial data window have not been removed. The inclusion of additional data that includes PMT demands results in potentially lower failure probabilities for these type codes.

In support of the categorization process, in order to assess the impact from the additional PMTs, a sensitivity study will need to be performed. The sensitivity approach that would be used to account for the additional PMT test data would be to provide a PMT successful demand contribution that is conservative. Previously, a seven-year data window was reviewed which resulted in a worse case PMT demand contribution for any given type code of three percent of the total number of demands.

By changing the number of PMTs used in the type code data, the calculated CDF and LERF could potentially increase. This may cause the calculated RAW values for the equipment credited to mitigate core damage and/or release to change. Per NEI 00-04, sensitivity information is provided to the IDP, which considers this information in the determination of safety significance.

TVA Response to RAI-05d

In Attachment 6 of the LAR, treatment of the SOKC was described. Further review of the SQN MOR indicates that SOKC is not applied. This error in the LAR has been entered into the TVA Corrective Action Program. WCAP 17154-P suggests application of SOKC for evaluation of ISLOCA. In accordance with ASME/ANS RA-Sa-2009, inclusion of SOKC is required to meet Capable Category II for Supporting Requirement QU-A3. As noted in Attachment 1 to this enclosure, TVA will incorporate SOKC into the MOR consistent with NUREG-1855, Revision 1, prior to using the PRA model to support categorization of SSCs under 10 CFR 50.69. Additionally, SOKC remains on the list of key uncertainties listed in Attachment 6, but for not being included in the PRA model.

Based on the above discussion for RAI-05d.i, no alternate method was employed as mentioned in RAI-05d.ii.

CNL-19-002 E1-20 of 29 Enclosure 1

RAI 06 – Qualitative Function Categorization

NEI 00-04, Section 9.2.2, "Review of Safety Related Low Safety-Significant Functions/SSCs," states "in making their assessment, the IDP should consider the impact of loss of the function/SSC against the remaining capability to perform the basic safety functions." This section also provides seven questions that should be considered for making the final determination of the safety-significance for each system function/SSC. It is unclear from the LAR how the IDP will collectively assess these seven specific questions. LAR Table 3-1 contains the entry “Allowable” at the intersection of the “IDP change HSS to LSS” column and “Qualitative Criteria” row.

a. Clarify how the IDP will collectively assess the seven specific questions to identify a function/SSC as LSS as opposed to HSS. For example, a function/SSC is considered HSS when the answer to any one question is false. b. If the criteria provided in part (a) considers more than one question is false for the IDP to assign a category of HSS to an SSC, provide justification to support rationale for why this is appropriate to use in the 10 CFR 50.69 risk-informed application.

TVA Response to RAI-06a

The assessment of the qualitative considerations (seven statements of consideration) is agreed upon by the IDP in accordance with NEI 00-04 Section 9.2. As noted in Attachment 1 to this enclosure, TVA procedures governing the IDP will require that if any one of the seven statements for consideration has a ‘FALSE’ response, the function risk will be assigned a classification of HSS. If all seven responses are ‘TRUE’, a function risk of LSS will be assigned; however, each ‘TRUE’ response requires a supporting justification for confirming the basis for the decision.

Additionally, TVA is providing the following information to the following statement RAI-06, which states: “LAR Table 3-1 contains the entry ‘Allowable’ at the intersection of the ‘IDP change HSS to LSS’ column and ’Qualitative Criteria’ row”.

The element column of Table 3-1 has the elements for the categorization steps performed by the Categorization Team (CT). In the Categorization Step column, the Qualitative Criteria row lists Section 9.2 from NEI 00-04. NEI 00-04 Section 9.2 provides responsibilities of the IDP, including the Qualitative Criteria, and the seven statements of consideration found in Section 9.2.2. As noted in Attachment 1 to this enclosure, TVA procedures will require the CT to consider the seven statements of consideration in addition to the IDP. However, in accordance with NEI 00-04 Section 9.2, the IDP is allowed to change the safety function classification made by the CT, in accordance with Table 3.1. As stated in this response, if all responses to the seven statements of consideration are ‘TRUE’, a function risk of LSS will be assigned; however, each ‘TRUE’ response requires a supporting justification for confirming the basis for the decision.

CNL-19-002 E1-21 of 29 Enclosure 1

Table 3-1: IDP Changes from Preliminary HSS to LSS IDP Dri ves Categori zation Step - NEI Evaluation Element Change Associated 00 -04 Section Level HSS to LSS Functions Internal Events Base Case Not Allowed Yes - Section 5.1 Fire, Seismic an d Other Risk (PRA External Events Base Allowable No Component Modeled) Case PRA Sensitivitv Studies Allowable No Integral PRA Assessment Not Allowed Yes - Section 5.6

Fire, Seismic and Other Risk (Non- Component Not Allowed No Extern al Hazards - modeled) Shutdown - Section 5.5 Function/Component Not Allowed No Core Damage - Section Defense- Function/Component Not Allowed Yes 6.1 in-Depth Containment - Section 6.2 Component Not Allowed Yes Qual itative Considerations - Section Function Al lowable N/A Criteria 9.2 Passive Passive - Section 4 Segment/Component Not Allowed No

TVA Response to RAI-06b

The criteria presented in the response to RAI-6a states that a function will be assigned an HSS classification for one or more ‘FALSE’ responses to the seven statements of consideration.

RAI 07 – Alternate Non-PRA Method for Fire to Categorize SSCs

Sections 50.69(c)(1)(ii) of 10 CFR requires that the licensee determine the SSC’s functional importance using an integrated, systematic process for addressing initiating events (internal and external), SSCs, and plant operating modes, including those not modeled in the plant- specific PRA.

Section 3.2.2 of the LAR states in part, “[t]he SQN categorization process will use the Fire SSEL for evaluation of safety significance related to fire hazards.” It further states that this approach addresses conditions defined by 10 CFR 50, Appendix R, NRC Branch Technical Position CMEB 9.5-1, regulatory exemptions, and fire-induced Multiple Spurious Operations to identify equipment. The LAR states that the alternate approach proposed is considered to be a conservative method, compared to FIVE or fire PRA, based on industry assessments.

Section 3.3 of NEI 00-04, Revision 0 provides limited guidance for determining the technical adequacy attributes required for these types of analyses for this specific application. RG 1.201, Revision 0 states in part, “as part of the plant-specific application requesting to implement §50.69, the licensee or applicant will provide the bases supporting the technical adequacy of its…non-PRA-type analyses for this application.”

Address the following regarding the proposed alternate approach:

a. Provide justification that the Fire SSEL method is technically adequate relative to the acceptable methods in NEI 00-04. Include in the justification, (1) the industry assessments referenced in the LAR (2) a summary of the industry evaluations and results that support the conclusion that Sequoyah’s proposed approach to use the

CNL-19-002 E1-22 of 29 Enclosure 1

fire SSEL is conservative, and (3) discussion for how additional SSCs will be assigned HSS in comparison to using an acceptable method (e.g., additional HSS SSCs would not be identified by a FIVE or fire PRA analysis).

b. The first paragraph of 3.2.2 states that, “[t]he Fire Safe Shutdown paths identify the safety functions and associated sets of equipment credited to achieve and maintain safe shutdown under postulated fire conditions” and that, “[t]he Fire SSEL identifies the credited equipment.” In review of Figure 3.1 of the LAR, it appears there are other SSCs, not on the fire SSEL, necessary for safe shutdown. According to Figure 3.1, if an SSC is not already on the SSEL, the next step in the process is to question whether the SSC is relied upon to maintain safe shutdown for a fire. An affirmative response to this question would categorize the SSC as candidate HSS.

i. Provide clarification along with a rationale for the additional equipment that will be identified as HSS for a fire event that is not on the SSEL.

ii. Confirm that all the SSCs identified as candidate HSS per Figure 3.1 of the LAR will remain HSS at the end of the categorization and cannot be re- categorized by the IDP.

c. Clarify whether fire detection and suppression (and fire dampers) equipment is included in Sequoyah’s SSEL. If not included, summarize how the risk-significance of this equipment will be evaluated to determine whether the equipment is HSS or LSS.

d. Fire protection actions can be credited if they are feasible, but PRA actions generally are not credited unless they are proceduralized and have a failure probability assigned. Some feasible actions have a high failure probability. Provide discussion to justify how the probability of failure for operator actions (which could be high) is considered in the analysis for determining SSCs identified on the SSEL.

TVA Response to RAI-07a

The proposed approach for identifying HSS SSCs for Internal Fire Hazards, by use of the Safe Shutdown Equipment List (SSEL), is similar to the NEI 00-04 acceptable method for Seismic Hazards in that the measure of safety significance identifies all system functions and associated SSCs that are involved in the safe-shutdown success paths as HSS. The justification is provided below.

At an NRC public meeting held on September 6, 2017 (ML17228A732), NEI and industry stakeholders met with NRC to describe a proposed approach for identifying HSS SSCs in the 10 CFR 50.69 application for Internal Fire Hazards.

The industry 10 CFR 50.69 Coordinating Committee performed a study involving several plants to compare the number of HSS SSCs identified by each of three approaches, 1) Fire Probabilistic Risk Assessment (FPRA), 2) Fire Induced Vulnerability Evaluation (FIVE), and 3) SSEL, each are more conservative in approach resulting in more HSS SSCs. The reason the SSEL approach is more conservative than using FIVE results, is that FIVE uses a successive screening methodology and the SSEL does not. The industry assessments referenced by the LAR are contained in ML17249A072.

CNL-19-002 E1-23 of 29 Enclosure 1

The following information supports the conclusion that SQN’s proposed approach to use the fire SSEL is conservative.

As shown in the graph below, a summary of the industry evaluations performed as part of the study concluded that the proposed SSEL approach is conservative by introducing significantly more SSCs assigned a HSS classification than use of a FPRA or FIVE. Additionally, the SSEL approach included all the SSCs identified by the FPRA and the FIVE approach.

In the graph below3, for SQN, the far right column would also include components identified in deviations/exceptions taken by the Fire Protection Program (see the response to RAI-07b.ii below), and Fire Protection System SSCs (including detection and suppression SSCs, and fire dampers).

rire Safe Shutdown Program

Internal EvenlS PRA

Use of the Fire SSEL provides a conservative alternative approach to 18 addressing fire for 50.69 categorization ~ I

Therefore, the SQN approach is conservative and inclusive with respect to identifying HSS SSCs that would be identified by a FPRA or FIVE approach.

The following information describes how additional SSCs will be assigned HSS in comparison to using an acceptable method.

During system categorization, the SSCs associated with the system are assessed for safety significance consistent with the NEI 00-04 process. Qualitatively, the additional SSCs that are assigned an HSS classification are those that are not on the Appendix R SSEL, but are credited by the Fire Protection Program to maintain safe shutdown. Additionally, under the proposed approach Fire Protection Equipment, including detection, suppression, and fire dampers, is assigned an HSS classification. This is addressed further in the response to RAI-07b.i.

TVA Response to RAI-07b.i:

As shown in Figure 3.1 of the LAR, beginning with the “Select Component” box, following the logic, the question is asked “Is the SSC on the Fire SSEL?” If the answer is ‘yes’, the logic proceeds downward and the SSC is “Candidate High Safety Significant.” If the answer is ‘no’, the logic follows to the diamond box on the right, “Is the SSC relied upon to maintain safe shutdown for Fire?” The reliance referred to in this diamond includes SSCs credited by

3 Slide 18 from ML17249A072

CNL-19-002 E1-24 of 29 Enclosure 1 the Fire Protection Program to mitigate multiple spurious operations (MSOs), SSCs credited for exemptions or deviations taken by the Fire Protection Program, and fire protection equipment SSCs (including fire dampers). Therefore, the Fire SSEL is a subset of the total number of SSCs assigned an HSS classification for internal fire hazards.

Select Component

Candidate Low Safety Significant

Candidate High Safe ty Significant

Identify Safety Significant Attributes of Component

Figure 3.1: Safety Significance Process for Structures, Systems, and Components Addressed in Fire Safe Shutdown Program

TVA Response to item RAI 7b.ii:

As shown below in Table 3-1 from the LAR, SSCs assigned as candidate HSS for non-modeled hazards (including fire) are not allowed to be re-categorized to LSS by the IDP; therefore, they remain HSS. Table 3-1 shows the Fire Hazards treatment as follows:

Table 3-1: IDP Changes from Preliminary HSS to LSS Drives Categorization Step - Evaluation IDP Change Element Associated NEI 00-04 Section Level HSS to LSS Function Risk (Non- Fire, Seismic and modeled) Other External Component Not Allowed No Hazards -

TVA Response to RAI-7c:

While not specifically included in the Internal Fire Protection Program SSEL, for the purposes of 10 CFR 50.69 categorization, the fire protection system SSCs are included within the scope of components assigned a HSS classification for Internal Fire Hazards, including fire detection equipment, suppression equipment, and fire dampers.

CNL-19-002 E1-25 of 29 Enclosure 1

TVA Response to RAI-07d:

The SSEL developed by the Fire Protection Program is based on credible methods (including operator actions) for safely shutting down the plant and maintaining safe-shutdown for a period of time. Therefore, failure probabilities are not included in the analysis. The proposed approach for identification of HSS SSCs for Internal Fire Hazards is a process similar to the NEI 00-04 identification of HSS SSCs for Seismic Hazards. Operator actions are not explicitly considered in the safety classification; therefore, there is no assignment of operator action failure probabilities. As noted in the response to RAI-07a, the study performed by the industry 50.69 Coordinating Committee shows that a FPRA or FIVE approach did not identify any SSCs beyond those identified by the proposed approach. Therefore, the identification of HSS SSCs for Internal Fire Hazards is not impacted by excluding operator action failure probabilities.

RAI 08 – Addition of FLEX to the PRA Model

The NRC memorandum dated May 30, 2017, “Assessment of the Nuclear Energy Institute 16-06, ‘Crediting Mitigating Strategies in Risk-Informed Decision Making,’ Guidance for Risk-Informed Changes to Plants Licensing Basis” (ADAMS Accession No. ML17031A269), provides the NRC’s staff assessment of identified challenges and strategies for incorporating FLEX equipment into a PRA model in support of risk-informed decision making in accordance with the guidance of RG 1.200. The LAR does not state whether or not the licensee has incorporated FLEX mitigating strategies and associated equipment into the PRA models at Sequoyah. For the NRC staff to assess the potential incorporation of FLEX equipment into the Sequoyah PRA model(s), provide the following:

a. State whether FLEX equipment and strategies have been credited in the PRA model(s). If not incorporated or their inclusion is not expected to impact the PRA results used in the categorization process no additional response is requested.

b. If the equipment or strategies have been credited, and their inclusion is expected to impact the PRA results used in the categorization process, provide the following information separately for each of the PRA model(s) (i.e., IEPRA (includes flooding), external hazards PRA(s)), and external hazards screening as appropriate:

i. A discussion detailing the extent of incorporation, i.e. summarize the supplemental equipment and compensatory actions, including FLEX strategies that have been quantitatively credited for each of the PRA models used to support this application.

ii. A discussion detailing the methodology used to assess the failure probabilities of any modeled equipment credited in the licensee’s mitigating strategies (i.e., FLEX). The discussion should include justification explaining the rational for parameter values, and whether the uncertainties associated with the parameter values are considered in accordance with the ASME/ANS PRA Standard as endorsed by RG 1.200.

CNL-19-002 E1-26 of 29 Enclosure 1

iii. A discussion detailing the methodology used to assess operator actions related to FLEX equipment and the licensee personnel that perform these actions. The discussion should include:

1. A summary of how the licensee evaluated the impact of the plant- specific human error probabilities and associated scenario-specific performance shaping factors listed in (a)-(j) of supporting requirement HR-G3 of the ASME/ANS RA-Sa-2009 PRA standard.

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

3. If the licensee’s procedures governing the initiation or entry into mitigating strategies are ambiguous, vague, or not explicit, a discussion detailing the technical bases for probability of failure to initiate mitigating strategies. c. The ASME/ANS RA-Sa-2009 PRA standard defines PRA upgrade as the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences. Section 1-5 of Part 1 of ASME/ANS RA-Sa-2009 PRA Standard states that upgrades of a PRA shall receive a peer review in accordance with the requirements specified in the peer review section of each respective part of this Standard.

i. Provide an evaluation of the model changes associated with incorporating mitigating strategies, which demonstrates that none of the following criteria is satisfied: (1) use of new methodology, (2) change in scope that impacts the significant accident sequences or the significant accident progression sequences, (3) change in capability that impacts the significant accident sequences or the significant accident progression sequences;

OR

ii. Propose a mechanism to ensure that a focused-scope peer review is performed on the model changes associated with incorporating mitigating strategies, and associated F&Os are resolved to Capability Category II prior to implementation of the 10 CFR 50.69 categorization program. An example would be a table of listed implementation items referenced in a license condition.

TVA Response to RAI-08:

The SQN PRA does not credit FLEX equipment or FLEX strategies.

CNL-19-002 E1-27 of 29 Enclosure 1

Attachment 1 SQN 10 CFR 50.69 PRA Implementation Items

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

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

• Qualitative assessment of system functions. System functions are qualitatively categorized as preliminary HSS or LSS based on the seven criteria in Section 9 of NEI 00-04 (see Section 3.1.1 of the original LAR (CNL-17-010)). Any component supporting an HSS function is categorized as preliminary HSS. Components supporting an LSS function are categorized as preliminary LSS.

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

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

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

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

• Internal Event Risks: Internal Events including internal flooding PRA model Revision 3, dated August 5, 2014 or later updated model as described in section 3.2.6 of this enclosure will be used. This model was accepted by NRC in NRC Letter to TVA, “Sequoyah Nuclear Plant, Units 1 and 2 - Issuance of Amendments for the Conversion to the Improved Technical Specifications with Beyond Scope Issues (TAC Nos. MF3128 and MF3129),” dated September 30, 2015 (ML15238B460).

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

• Documentation requirements per Section 3.1.1 of enclosure 1 of the LAR (CNL-17-010)(ML 18075A365).

• TVA procedures governing the IDP will require that if any one of the seven statements for consideration has a ‘FALSE’ response the function risk will be assigned a classification of HSS.

• TVA procedures will require the Categorization Team to consider the seven statements of consideration in addition to the IDP.

CNL-19-002 E1-28 of 29 Enclosure 1

In addition to the procedure changes above, TVA will also preform the following actions.

• As documented in the F&O Closure Report, all changes initiated by the F&O resolutions were confirmed by the Integrated Assessment Team to have been incorporated into the living model and associated documentation. TVA shall update the Model of Record (MOR) with this information prior to system categorization.

• TVA shall re-introduce the State of Knowledge Correlation (SOKC) into the MOR prior to using the PRA model to support categorization of SSCs under 10 CFR 50.69.

• With respect to the external flooding hazards, TVA shall re-confirm that there is sufficient time to eliminate the source of the threat or to provide an adequate response in accordance with screening criterion C5, prior to 50.69 categorization.

CNL-19-002 E1-29 of 29 Enclosure 2 SQN Units 1 and 2 Renewed Operating License Changes Markup - 14a -

(d) For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance subject to the modified acceptance criteria is due at the end of the first Surveillance interval that began on the date the Surveillance was last performed prior to the implementation of this amendment.

D. Exemptions from certain requirements of Appendices G and J to 10 CFR Part 50 are described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 1. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. The exemptions are, therefore, hereby granted. The granting of these exemptions are authorized with the issuance of the License for Fuel Loading and Low Power Testing, dated February 29, 1980. The facility will operate, to the extent authorized herein, Act, and the regulations of the Commission.

E. Physical Protection

(1) The licensee shall fully implement and maintain in effect all provisions of the Commission- approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Sequoyah Nuclear Plant Security Plan, Training And Qualification Plan, And Safeguards Contingency Plan" submitted by letter dated May 8, 2006.

(2) The licensee shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The licensee CSP was approved by License Amendment No. 329, as amended by changes approved by License Amendment Nos. 333 and 337.

Amendment 334, 337 XXX Renewed License No. DPR 77 October 3, 2016 - 13a -

relocation of the requirements to the specified documents, as described in Table R, Relocated Specifications and Removed Detail Changes, attached to the NRC staff’s Safety Evaluation, which is enclosed in this amendment.

2. Schedule for New and Revised Surveillance Requirements (SRs) The schedule for performing SRs that are new or revised in License Amendment 327 shall be as follows:

(a) For SRs that are new in this amendment, the first performance is due at the end of the first Surveillance interval, which begins on the date of implementation of this amendment.

(b) For SRs that existed prior to this amendment, whose intervals of performance are being reduced, the first reduced Surveillance interval begins upon completion of the first Surveillance performed after implementation of this amendment.

(c) For SRs that existed prior to this amendment, whose intervals of performance are being extended, the first extended Surveillance interval begins upon completion of the last Surveillance performed prior to implementation of this amendment.

(d) For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance subject to the modified acceptance criteria is due at the end of the first Surveillance interval that began on the date the Surveillance was last performed prior to the implementation of this amendment.

Amendment No. 327 XXX Renewed License No. DPR 79 September 30, 2015 Enclosure 3 SQN Units 1 and 2 Renewed Operating License Change Retyped Copy - 14a -

(d) For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance subject to the modified acceptance criteria is due at the end of the first Surveillance interval that began on the date the Surveillance was last performed prior to the

(33) Adoption of 10 CFR 50.69, “Risk-Informed categorization and treatment of structures, systems and components for nuclear power plants”

(1) TVA is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) model to evaluate risk associated with internal events, including internal flooding; using the fire safe shutdown equipment list in the SQN Fire Protection Report referenced in the Updated Final Safety Analysis Report to evaluate internal fire events; the NUMARC 96-01 shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the criteria in the endorsed ASME/ ANS RA-Sa-2009 PRA Standard for other external hazard screening significance; as specified in Unit 1 License Amendment [Number].

(2) Prior to implementation of the provisions of 10CFR 50.69, TVA shall complete the items below;

a. Items listed in Enclosure 1, Attachment 1, "SQN 10 CFR 50.69 PRA Implementation Items," in TVA letter CNL-19-002, “Response to Request for Additional Information Regarding Application to Modify Sequoyah Nuclear Plant Units 1 and 2, Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, (SQN-TS-17-06)(EPID: L-2018-LLA-0066),” dated March 21, 2019.

(3) Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach, change from alternative method for internal fire to a fire probabilistic risk assessment approach).

Amendment Renewed License No. DPR 77 - 14b -

D. Exemptions from certain requirements of Appendices G and J to 10 CFR Part 50 are described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 1. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. The exemptions are, therefore, hereby granted. The granting of these exemptions are authorized with the issuance of the License for Fuel Loading and Low Power Testing, dated February 29, 1980. The facility will operate, to the extent authorized herein, Act, and the regulations of the Commission.

E. Physical Protection

(1) The licensee shall fully implement and maintain in effect all provisions of the Commission- approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Sequoyah Nuclear Plant Security Plan, Training And Qualification Plan, And Safeguards Contingency Plan" submitted by letter dated May 8, 2006. (2) The licensee shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The licensee CSP was approved by License Amendment No. 329, as amended by changes approved by License Amendment Nos. 333 and 337.

Amendment Renewed License No. DPR 77 - 13a - relocation of the requirements to the specified documents, as described in Table R, Relocated Specifications and Removed Detail Changes, attached to the NRC staff’s Safety Evaluation, which is enclosed in this amendment.

2. Schedule for New and Revised Surveillance Requirements (SRs) The schedule for performing SRs that are new or revised in License Amendment 327 shall be as follows:

(a) For SRs that are new in this amendment, the first performance is due at the end of the first Surveillance interval, which begins on the date of implementation of this amendment.

(b) For SRs that existed prior to this amendment, whose intervals of performance are being reduced, the first reduced Surveillance interval begins upon completion of the first Surveillance performed after implementation of this amendment.

(c) For SRs that existed prior to this amendment, whose intervals of performance are being extended, the first extended Surveillance interval begins upon completion of the last Surveillance performed prior to implementation of this amendment.

(d) For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance subject to the modified acceptance criteria is due at the end of the first Surveillance interval that began on the date the Surveillance was last performed prior to the implementation of this amendment.

(26) Adoption of 10 CFR 50.69, “Risk-Informed categorization and treatment of structures, systems and components for nuclear power plants”

(1) TVA is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) model to evaluate risk associated with internal events, including internal flooding; using the fire safe shutdown equipment list in the SQN Fire Protection Report referenced in the Updated Final Safety Analysis Report to evaluate internal fire events; the NUMARC 96-01 shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the criteria in the endorsed ASME/ANS RA-Sa-2009 PRA Standard for other external hazard screening significance; as specified in Unit 2 License Amendment [Number].

Amendment No. Renewed License No. DPR 79 - 13b -

(2) Prior to implementation of the provisions of 10CFR 50.69, TVA shall complete the items below;

a. Items listed in Enclosure 1, Attachment 1, "SQN 10 CFR 50.69 PRA Implementation Items," in TVA letter CNL-19-002, “Response to Request for Additional Information Regarding Application to Modify Sequoyah Nuclear Plant Units 1 and 2, Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, (SQN-TS-17-06)(EPID: L-2018-LLA-0066),” dated March 21, 2019.

(3) Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach, change from alternative method for internal fire to a fire probabilistic risk assessment approach).

Amendment No. Renewed License No. DPR 79 Enclosure 4 Markup of Original LAR, Attachment 4: External Hazards Screening Pages Attachment 4: External Hazards Screening

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

Plant design eliminates drought as a concern. In addition, this event is slowly Y C5 developing. TVA’s management of the Drought Tennessee River level is such that drought will not jeopardize the ultimate heat sink’s capability.

Flood protection plans, designed to minimize impact of floods above plant grade on safety- related facilities are in place, and procedures for predicting rainfall floods, arrangements to warn of upstream dam failure floods, and lead times available and types of action to be taken to meet safety requirements for sources are thoroughly analyzed and compensatory measures are well planned.

Some wind driven waves are likely when the probable maximum flood (PMF) crests at SQN. Analysis shows that the probability PS1 that this would occur on the specific day of External Y the PMF is on the order of 1E-10 or lower. Flooding PS4 Considering the simultaneous events over a 40-yr period is on the order of 1E-06. Therefore, wind driven waves are not a concern to plant safety.

Snowmelt is not a factor in generating maximum floods at the plant site due to the plant location, which is within a temperate zone.

SQN conforms to regulatory position of RG 1.59, “Design Basis Floods for Nuclear Power Plants.”

Enclosure 4, CNL-19-002 Page 1 of 2 Attachment 4: External Hazards Screening

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

The SQN plant Category I structures are designed for a 95-mph wind (including a 1.1 gust factor) 30 feet above grade with a 100- year period of recurrence. The basis was determined from American Society of Civil Engineers (ASCE) Paper 3269, “Wind Forces on Structures.” PS1 Extreme Wind or Y The SQN plant is designed for tornados Tornado PS3 having a maximum rotational velocity of 300 mph, and up to a translational speed of 60 mph.

SQN has been designed to resist tornado wind and missile effects equivalent to the Design Basis Tornado and meets the intent of RG 1.76 and 1.117.

Fog and mist may increase the frequency of accidents involving aircraft, ships, or Fog Y C4 vehicles. This weather condition is included implicitly in the accident rate for these Transportation Accidents

The ground has been cleared for two- Forest or Range Y C3 thousand feet around plant buildings. There Fire are no wooded areas close enough to present a hazard from forest fires.

Implicitly included in weather-related LOOP, Frost Y C4 such as hail and snow.

Enclosure 4, CNL-19-002 Page 2 of 2

Enclosure 5 Markup of Original LAR, Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty Pages Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty

Assumption/Uncertainty Discussion Disposition

which, all introduce some level of uncertainty into the calculation. The large internal rupture of an MOV is assumed in NUREG/CR-6928 to be a factor of 0.02 less than that of a small internal leak on an MOV as there has been no actual large internal rupture events in the industry. The mission time is also assumed based on a seven day repair interval, this number could potentially be greater than that if the component is not covered by an Technical Specification or, more likely, less than the assumed seven day repair time. The final area of uncertainty is the frequency of the activity. Most of the procedures reviewed have frequencies as well as conditions. These conditions could cause the actual maintenance activity to occur more often than the frequency noted in the procedure. This was addressed by performance of sensitivity studies that indicate inclusion of the number of Equipment type code data Inclusion of successful PMT demands can assumed PMT demands has a small impact on CDF. If includes successful post- result in an under-estimation of the failure the number of successful PMT starts is significantly maintenance testing (PMT). probability of a component type. overestimated it can have a more profound affect on CDF and LERF. The SOKC causes point estimates to differ To account for the potential correlation of various type from the mean values, which are higher. codes, factors are employed that multiply the value of SOKC is a statistical effect that appears when the cutset to ensure the resultant frequency is in line State of Knowledge Correlation a pool of data is used to characterize the with the mean. These factors were calculated using the (SOKC). uncertainty distribution for all events within multiplier method, where an additional basic event is the data pool. Those events that are used are appended to the cutsets with correlated basic events. considered correlated, which implies that the Since the normal cutset frequency calculation is

Enclosure 5, CNL-19-002 Page 1 of 2 Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty

Assumption/Uncertainty Discussion Disposition

same distribution applies to all sampled performed via the multiplication method, this eases use events when using a Monte Carlo approach. of these factors. The factors were based on use of the multiplier method discussed in WCAP 17154-P. This method requires the recovery rules to be written to add factors onto each applicable cutset. The factors are calculated based on the uncertainty distribution type and the associated alpha and beta shaping factors. Both the Beta distribution and the Gamma distribution were calculated for the SOKC.

Enclosure 5, CNL-19-002 Page 2 of 2 Enclosure 6 Attachments 4 and 6 Retyped Copy Attachment 4: External Hazards Screening

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

The Dallas Bay Sky Park is located beyond five miles of the plant. Also, Federal Airway V333 passes directly over the site. The PS2 Chattanooga Airport is located approximately Y Aircraft Impact 14.5 miles from the plant site. These are the PS4 only facilities of potential significance to the safe operation of the plant, and based on evaluations these activities will pose no hazard.

SQN is located in the Tennessee Valley on the Tennessee River. Avalanches are not a Y C3 Avalanche viable external initiator because of climate and topography, as SQN is located within a temperate zone.

Sudden influxes are not applicable to the plant design. Slowly developing growth can be detected and mitigated by surveillance. Control of organic fouling is provided by use Y C5 of strainers in the supply headers and Biological Event biocide treatments. Asiatic clams are controlled by a combination of straining and biocide treatments. Microbiologically induced corrosion (MIC) is controlled by injecting biocide.

SQN is located in the Tennessee Valley; Y C3 Coastal Erosion therefore, coastal erosion is not a concern as a credible hazard.

Enclosure 6, CNL-19-002 Page 1 of 14 Attachment 4: External Hazards Screening

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

Plant design eliminates drought as a concern. In addition, this event is slowly Y C5 developing. TVA’s management of the Drought Tennessee River level is such that drought will not jeopardize the ultimate heat sink’s capability.

Flood protection plans, designed to minimize impact of floods above plant grade on safety- related facilities are in place, and procedures for predicting rainfall floods, arrangements to warn of upstream dam failure floods, and lead times available and types of action to be taken to meet safety requirements for sources are thoroughly analyzed and compensatory measures are well planned.

With respect to external flooding, SQN is designed such that there is sufficient warning time, given large rainfall or seismically- induced upstream dam failure, to shut the External Y C5 plant down and implement emergency Flooding procedures. Plant shutdown is based on a flood warning scheme divided into two stages (i.e., Stage I and Stage II). Stage I is a minimum of ten hours and Stage II is a minimum of 17 hours. During Stage I preparation steps are taken for flood mitigation. If conditions persist, Stage II is entered whereby the operator moves to initiate plant shutdown. Therefore, the minimum time calculated that flooding could exceed plant grade is 27 hours. As noted in Section 2.4.10 of the SQN UFSAR, “Any rainfall flood exceeding plant grade will be predicted at least 27 hours in advance by TVA's Reservoir Operations. Warning of seismic failure of key upstream dams will be

Enclosure 6, CNL-19-002 Page 2 of 14 Attachment 4: External Hazards Screening

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

available at the plant at least 27 hours before a resulting flood surge would reach plant grade. Hence, there is adequate time to prepare the plant for any flood.” Therefore, the timing available to exceed plant grade represents a slow moving event that meets the criteria for C5 that there is sufficient time to provide an adequate response.

SQN conforms to regulatory position of RG 1.59, “Design Basis Floods for Nuclear Power Plants.”

The SQN plant Category I structures are designed for a 95-mph wind (including a 1.1 gust factor) 30 feet above grade with a 100- year period of recurrence. The basis was determined from American Society of Civil Engineers (ASCE) Paper 3269, “Wind Forces on Structures.”

The SQN plant is designed for tornados having a maximum rotational velocity of 300 mph, and up to a translational speed of 60 mph. Extreme Wind or Y PS2 Tornado SQN has been designed to resist tornado wind and missile effects equivalent to the Design Basis Tornado and meets the intent of RG 1.76 and 1.117.

SQN was designed prior to the 1975 SRP, the approach taken in the Individual Plant Evaluation of External Events (IPEEE) was to review the design bases and compare them to the SRP requirements (screening criteria). Any changes that were made to the plant subsequent to the design analysis were reviewed to verify compliance with SRP

Enclosure 6, CNL-19-002 Page 3 of 14 Attachment 4: External Hazards Screening

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

criteria. For Other External Events, it was found that no vulnerabilities exist outside the screening thresholds of the SRP. As such, the SQN design meets the intent of the 1975 SRP for high winds and tornadoes and the requirements for screening the hazard.

Fog and mist may increase the frequency of accidents involving aircraft, ships, or Fog Y C4 vehicles. This weather condition is included implicitly in the accident rate for these Transportation Accidents

The ground has been cleared for two- Forest or Range Y C3 thousand feet around plant buildings. There Fire are no wooded areas close enough to present a hazard from forest fires.

Implicitly included in weather-related LOOP, Frost Y C4 such as hail and snow.

Hail events could cause complete or partial losses of offsite power (LOOP) (Reference: NUREG/CR-5042 Supplement 2). The C1 potential effects of a LOOP and station Y blackout are addressed by the Internal Hail C4 Events PRA and the frequency of hail events is subsumed by the LOOP initiating event frequency for weather related events. Therefore, this event is bounded by other events for which the plant is designed.

Enclosure 6, CNL-19-002 Page 4 of 14 Attachment 4: External Hazards Screening

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

SQN meets General Design Criteria 22 and 23 that require the design of protection systems assures that the effects of natural phenomena and other conditions do not result in a loss of the protection function.

C1 Furthermore, the design of protection High Summer Y systems fail in the safe state or into a state Temperature C4 determined to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air) or postulated adverse environments (e.g., extreme heat, cold) are experienced.

SQN conforms to regulatory position of RG 1.59, “Design Basis Floods for Nuclear High Tide, Lake Y C4 Power Plants.” Level, or River Stage See External Flooding for additional information.

Due to the location of the SQN plant in the Tennessee Valley, a hurricane is not a C3 credible event. However, the downgraded Y Hurricane storm is expected to cause other hazards C4 that are covered under Extreme Wind or Tornado, and Intense Precipitation (i.e. External Flooding).

Because of SQN’s location in a temperate Y C3 climate, significant amounts of ice do not Ice Cover form on Tennessee Valley rivers and lakes.

Enclosure 6, CNL-19-002 Page 5 of 14 Attachment 4: External Hazards Screening

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

There are no industrial or military facilities within five miles of the SQN plant that would potentially pose a hazard to the safe Industrial or Y C3 operation of the plant. Facilities of interest Military Facility beyond five miles include the Volunteer Army Accident Ammunition plant that is the only industrial or military facility of potential significance to the safe operation of the plant.

N N/A Internal Flooding SQN has an Internal Flooding Model

N N/A Internal Fire Addressed in Section 3.2.2

Y C3 Not applicable to SQN because of Landslide topography.

Lightning strikes causing loss of offsite power or turbine trip are contributors to the C1 initiating event frequencies for these events. Y Lightning However, other causes are also included in C4 the loss of offsite power initiator. Impacts from lightning are no greater than already modeled in the internal events PRA.

This event is slowly developing and well Low Lake Level Y C5 controlled based on TVA’s management of or River Stage the Tennessee River.

Enclosure 6, CNL-19-002 Page 6 of 14 Attachment 4: External Hazards Screening

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

Extended freezing temperatures are rare in C1 the Tennessee Valley where the SQN plant Low Winter Y is located. The plant is designed for such Temperature C5 events, and their impacts are slow to develop.

Extraterrestrial activity includes both natural and manmade objects that enter earth’s Meteorite or Y PS4 atmosphere from space. Because the Satellite Impact probability of a meteorite strike is very small. Therefore, it can be dismissed on the basis of its very low initiating event frequency.

There are no major natural gas pipelines within five miles of SQN. The nearest eight-inch diameter pipeline is located at Pipeline Y C3 more than 0.5 miles at its closest approach Accident to the site. Any postulated rupture of this pipe at that distance would not pose a hazard to the safety-related SSCs at SQN.

The main control room habitability during postulated hazardous chemical releases at or near the plant was evaluated by TVA. The evaluation conformed to RG 1.78 and concluded that the main control room Release of habitability is not jeopardized by accident Chemicals in Y C2 releases of the chemicals. SQN maintains a Onsite Storage list of hazardous materials, storage location, and quantities. There are no industrial or military facilities where large quantities of toxic chemicals could be stored with a 5-mile radius of the plant.

Enclosure 6, CNL-19-002 Page 7 of 14 Attachment 4: External Hazards Screening

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

River (channel) diversion is not a potential problem for SQN. There are no channel diversions upstream of the plant that would cause diverting or rerouting of the source of Y C3 plant cooling water, and none are anticipated River Diversion in the future. The flood plain is such that large floods do not produce major channel meanders or cutoffs. Analyses indicate the Tennessee River has essentially maintained its present alignment for over 35,000 years.

Sand or Dust Y C3 Due to plant location, SQN is not subject to Storm sand or dust storms.

SQN is located on the Tennessee River (Chickamauga Lake). No conceivable Y C3 Seiche hurricane or cyclonic-type winds could produce wave heights required to reach the plant grade.

NRC’s IPEEE Safety Evaluation for SQN dated February 21, 2001 (Reference 7) acknowledged that SQN was categorized as a full-scope plant (per NUREG-1407), having a 0.3g peak ground acceleration. SQN performed a full-scope seismic margin N N/A analysis using the methodology described in Seismic Activity EPRI NP-6041-SL, Revision 1, “A Methodology for Assessment of Nuclear Power Plant Seismic Margin.”

The NRC Staff concluded that the aspect of seismic hazards was adequately addressed for SQN.

Enclosure 6, CNL-19-002 Page 8 of 14 Attachment 4: External Hazards Screening

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

Roofs of structures at SQN are designed for snow loading greater than observed or expected for the location of the plant site, Y C3 Snow which is within a temperate zone.

Potential flooding impacts covered under external flooding.

The SQN Updated FSAR, Chapter 2.5 describes the characteristics of the C1 stratigraphy, rocks, foundation, soils, and Soil Shrink-Swell Y backfill. For Category I soil-supported Consolidation C5 structures, the allowable capacity for sustained loading is at least a factor of three less than the ultimate bearing. The potential for this hazard is low at the SQN site.

SQN is located in the Tennessee Valley on Y C3 Storm Surge the Tennessee River; therefore storm surge is not a viable external initiator for SQN.

The main control room habitability during postulated hazardous chemical releases at or near the plant was evaluated by TVA. The evaluation conformed to RG 1.78 and concluded that the main control room Y C2 habitability is not jeopardized by accident Toxic Gas releases of the chemicals. SQN maintains a list of hazardous materials, storage location, and quantities. There are no industrial or military facilities where large quantities of toxic chemicals could be stored with a 5-mile radius of the plant.

Enclosure 6, CNL-19-002 Page 9 of 14 Attachment 4: External Hazards Screening

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

An examination of the impact from potential accidents on the transportation routes

(i.e., railroad, aircraft, highways, and barge traffic) concluded that their contribution to plant risk is negligible.

PS2 Railroads -As noted in SER for the SQN IPEEE Section 2.3.3.2, (Reference 7) “The nearest mainline railroad is 5.5 miles away, which is greater than the Regulatory Guide 1.91 safe distance.” There is a spur that leads into the plant; however, it is limited to delivery of large components to the plant on an infrequent basis.

Transportation Y C4 Aircraft - see Aircraft Impact Accident C3 Highways - As noted in Reference 7, Section 2.3.3.2, “The nearest highway is two miles away, which is greater than the RG 1.91 safe distance.”

PS4 Barge Traffic - The potential for damage to the SQN plant from a barge explosion is negligible. Materials considered included TNT, gasoline, liquid natural gas and fertilizers.

Barge impacts on plant structures are conservatively calculated to be in the 1E-05/yr (upstream tow), and 1E-08/yr (Downstream Drifting) range.

Enclosure 6, CNL-19-002 Page 10 of 14 Attachment 4: External Hazards Screening

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

SQN is located in the Tennessee Valley, Y C3 which is not adjacent to any large bodies of Tsunami water. Therefore, a tsunami event is not applicable to SQN because of location.

The SQN plant arrangement is such that safety-related SSCs are essentially protected from low trajectory turbine missiles. The Turbine- Y PS2 probability of this event has been estimated Generated at less than 1E-07/yr. SQN has an Missiles inspection program of the low pressure turbine discs that is performed on a regular basis.

Y C3 SQN is located in the Tennessee Valley Volcanic Activity which is not prone to volcanic activity.

SQN is located in the Tennessee Valley and not adjacent to any large body of water that C3 can result in significant waves. Therefore Y large wave events are not applicable to the Waves C4 SQN site.

Note: Waves associated with External Flooding are covered under that hazard.

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

Enclosure 6, CNL-19-002 Page 11 of 14 Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty

Assumption/Uncertainty Discussion Disposition

The assumed 0.9 value is consistent with EPRI Report 3002000079 R3, Pipe Rupture Internal Flooding - The assumption Frequencies for Internal Flooding PRA. Areas Uncertainty evaluations associated with the risk that a leaking pipe will be detected with potentially HSS SSCs could be sensitive categorization process are addressed using the by visual inspection is given a 0.9 to higher flooding probabilities. Since internal processes discussed in NEI 00-04 Section 8 and in the probability. Similarly, detection flooding is a significant contributor to plant prescribed sensitivity studies discussed in Section probabilities for Non-Destructive risk, and the detection probability is 5. The TVA process will follow those requirements to Testing (NDT) uses probabilities significant to reducing the impact of a flooding address these and similar assumptions. up to 0.9. scenario, detection is a major factor in the analysis. Passive pipe break failures rates have been given an uncertainty parameter. The impact of these uncertainties can be treated by the use of a random sampling Monte Carlo process.

Human induced flooding is analyzed by use of the HRA The Internal Flooding analysis Calculator program which creates an assumed calculation uses a summation of uncertainty term for any HRA action. Since the human three different frequencies, Each of these flooding events has its own induced flooding events is a combination of both pre- passive pipe break failures, inherent uncertainties. initiating event and post initiating event, each portion human-induced flooding, and has an independent uncertainty term. The HRA maintenance induced flooding. Calculator program also arbitrarily assigns an uncertainty term to HRA actions based on the calculated probabilities.

Maintenance induced flooding events has three main inputs to the calculation of this frequency, failure rate of

Enclosure 6, CNL-19-002 Page 12 of 14 Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty

Assumption/Uncertainty Discussion Disposition

an MOV, mission time, and frequency of the activity, of which, all introduce some level of uncertainty into the calculation. The large internal rupture of an MOV is assumed in NUREG/CR-6928 to be a factor of 0.02 less than that of a small internal leak on an MOV as there has been no actual large internal rupture events in the industry. The mission time is also assumed based on a seven day repair interval, this number could potentially be greater than that if the component is not covered by an Technical Specification or, more likely, less than the assumed seven day repair time. The final area of uncertainty is the frequency of the activity. Most of the procedures reviewed have frequencies as well as conditions. These conditions could cause the actual maintenance activity to occur more often than the frequency noted in the procedure. This was addressed by performance of sensitivity studies that indicate inclusion of the number of Equipment type code data Inclusion of successful PMT demands can assumed PMT demands has a small impact on CDF. If includes successful post- result in an under-estimation of the failure the number of successful PMT starts is significantly maintenance testing (PMT). probability of a component type. overestimated it can have a more profound affect on CDF and LERF. Review of the SQN PRA model indicates that the SOKC was not included in the Interfacing TVA proposes a licensing condition to update the PRA State of Knowledge Correlation System Loss of Cooling Accident (ISLOCA) model with SOKC. (SOKC). analysis as required by the ASME/ANS RA- Sa-2009 PRA Standard to meet Capability Category II for Supporting Requirement (SR)

Enclosure 6, CNL-19-002 Page 13 of 14 Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty

Assumption/Uncertainty Discussion Disposition

QU-A3 which reads: “ESTIMATE the mean CDF accounting for the sate-of-knowledge correlation between event probabilities when significant.”

Enclosure 6, CNL-19-002 Page 14 of 14