CNL-20-089, Response to RAI Re Application for Tech. Specification Change Re Risk-Informed Justification for Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (BFN-TS-518)

Response to RAI Re Application for Tech. Specification Change Re Risk-Informed Justification for Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (BFN-TS-518)

ML20356A106
Person / Time
Site:	Browns Ferry
Issue date:	12/21/2020
From:	Polickoski J Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CNL-20-089, EPID L-2020-LLA-0058
Download: ML20356A106 (22)

Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-20-089 December 21, 2020 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Browns Ferry Nuclear Power Plant Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Dockets 50-259, 50-260, and 50-296

Subject:

Response to Request for Additional Information Regarding Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (BFN-TS-516)

(EPID L-2020-LLA-0058)

References:

1. TVA Letter to NRC, CNL-20-003, Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (BFN-TS-516), dated March 27, 2020 (ML20087P262)

2. NRC Electronic Mail to TVA, Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Request for Additional Information Regarding Request to Adopt TSTF-425 (EPID L-2020-LLA-0058), dated October 22, 2020 (ML20297A301)

In Reference 1, Tennessee Valley Authority (TVA) submitted a request for an amendment to the Technical Specifications (TS) for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3 to modify the BFN TS by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies." In Reference 2, the Nuclear Regulatory Commission (NRC) submitted a request for additional information (RAI) and requested a response by December 21, 2020. Enclosure 1 provides the TVA response to the RAI response.

As provided for in the Enclosure 1 RAI response, TVA proposes a new License Condition to ensure all BFN open Facts and Observations are closed prior to implementing any Surveillance Test Interval extensions. Enclosure 2 to this letter provides the existing BFN, Units 1, 2 and 3 Renewed Facility Operating Licenses marked-up to show the proposed changes respectively.

U.S. Nuclear Regulatory Commission CNL-20-089 Page 2 December 21, 2020 The enclosures to this letter do not change the no significant hazards consideration or the environmental considerations contained in Reference 1. Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and the enclosures to the Alabama State Department of Public Health.

If you should have any questions regarding this submittal, please contact Kimberly Hulvey, Senior Manager, Fleet Licensing, at 423-751-3275.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 18th day of December 2020.

Respectfully, James T. Polickoski Director, Nuclear Regulatory Affairs

Enclosures:

1. Response to Request for Additional Information Regarding Risk-Informed Justification For the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (EPID L-2020-LLA-0058)

2. BFN Units 1, 2, and 3 Renewed Facility Operating License Markup Pages cc (Enclosures):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Project Manager - Browns Ferry Nuclear Plant State Health Officer, Alabama State Department of Public Health

Enclosure 1 Response to Request for Additional Information Regarding Risk-Informed Justification For the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (EPID L-2020-LLA-0058)

The Nuclear Regulatory Commission (NRC) Request for Additional Information provided the following background discussion:

By letter dated March 27, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20087P262), the Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) regarding Browns Ferry Nuclear Plant, Units 1, 2, and 3 (Browns Ferry). The proposed amendments would modify the Browns Ferry Technical Specifications by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Industry (NEI) Topical Report 04-10, Revision 1, Risk-Informed Technical Specifications Initiative 5b - Risk-Informed Method for Control of Surveillance Frequencies (ADAMS Accession No. ML071360456). The licensee stated that the proposed changes are consistent with Nuclear Regulatory Commission (NRC)-approved

/Technical Specification Task Force (TSTF) Standard Technical Specification (STS) change TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b (ADAMS Accession No. ML090850642).

REGULATORY BASES AND GUIDANCE Regulatory Guide (RG) 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Section 2.2, provides regulatory guidance regarding peer reviews and the staff regulatory position on NEI 00-02, Probabilistic Risk Assessment (PRA) Peer Review Process Guidance (ADAMS Accession No. ML061510619), NEI 05-04, Process for Performing Follow-On [Internal Events]

PRA Peer Reviews Using the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard (ADAMS Accession No. ML083430462), and NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines (ADAMS Accession No. ML102230070).

The NEI guidance document NEI 04-10, Revision 1, provides guidance for relocating the surveillance frequencies from the Technical Specifications to a licensee-controlled program by providing an NRC-approved methodology for control of the surveillance frequencies. The guidance in NEI 04-10, Revision 1, is acceptable for referencing in licensing actions to the extent specified and under the limitations delineated in NEI 04-10, and the NRC safety evaluation providing the basis for NRC acceptance of NEI 04-10 (ADAMS Accession No. ML072570267).

REQUESTS FOR ADDITIONAL INFORMATION , Section 3 of the LAR states that the PRA models have been developed in accordance with the requirements of RG 1.200, Revision 2, subjected to peer review and the Facts and Observations (F&O) independent assessment process.

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The following requests for additional information (RAIs) are needed to enable the NRC staff to complete its review of the application:

The NRC RAI questions are provided in italics. The TVA responses are provided in non-italicized font.

APLB RAI 01: Peer Review History: Appendix X, Closure of Facts and Observations Section 3.4.2 of the LAR states, in part, that 95 Internals Events Finding F&Os identified in the peer review in May 2009 were subjected to an F&O Resolution Review FSPR [focused-scope peer review] in 2015, which followed the guidance from NEI 05-04. Section 3.5.1 of the LAR also states that [t]he focused scope peer review also assessed the closure of 68 F&Os resolved by TVA for the [Browns Ferry Nuclear Power Plant] BFN FPRA The staff is unable to determine that the 2015 focused scope peer reviews performed adequately assessed the closure actions incorporated into the Browns Ferry PRA models to ensure the associated SRs are met at capability category (CC) II for the applicable technical element in the ASME/ANS RA-Sa-2009 PRA Standard as endorsed by RG 1.200, Revision 2.

Furthermore, the NRC notes that the table 1 of the LAR cites, Revision 3 of NEI 05-04 for several full scope and focused scope peer reviews performed, whereas the current version of RG 1.200, Revision 2 cites Revision 2 of NEI 05-04 listed as Reference 16. Provide clarification of which revision of NEI 05-04 was used.

To assess the adequacy for closure of the F&Os performed in 2015 for the IEPRA (includes internal floods) and fire PRA using the focused scope peer reviews, provide the following:

a. Discussion of the scope of the focused scope peer reviews performed in 2015: (1) and technical elements and supporting requirements (SRs) determined to be applicable, etc.), (2) any new methods incorporated into the PRA or identified as a result of the peer review.

b. Brief summary of the peer review teams conclusion(s) and comments on the focused scope peer reviews performed.

c. Describe how the resolution for the closure of each F&O was assessed to determine if it constituted a PRA upgrade or maintenance update, as defined in ASME/ANS RA-Sa-2009 and qualified by RG 1.200, Revision 2.

d. Explain how closure of the F&Os were 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.

e. Discuss whether the F&O closure review scopes included all finding-level F&Os, including those finding-level F&Os that are associated with Met SRs at CC-II. If not, identify and describe those findings that were excluded from the F&O closure review scope. For each identified finding-level F&O, describe the disposition and the impact of the F&O on the PRA as it pertains to this application.

OR CNL-20-089 E1-2 of 15

f. Perform a gap assessment between the 2015 focused scope peer reviews performed and the Appendix X to NEI 05-04/07-12/12-16, as accepted, with conditions by the NRC staff (ADAMS Accession Number ML17079A427). Provide the gaps identified, and a detailed summary of actions performed to address those gaps. The gap assessment should also include a detailed summary of the scope of the focused-scope peer reviews performed in 2015 for the IEPRA(includes internal floods) and fire PRA that discusses the technical elements and SRs determined to be applicable for the F&Os identified to be closed.

AND

g. Propose a mechanism that ensures for any F&Os identified not to be adequately assessed for closure at CC-II in response to the above portions of the RAI (a-e, or f) the F&Os are either resolved at CC-II for the applicable SR(s) using the ASME/ANS RA-Sa-2009 PRA Standard, endorsed by RG 1.200 or are assessed on a case-by-case basis consistent with NEI 04-10 for the individual STI (surveillance test interval) extension.

TVA Response to RAI APLB RAI 01:

In response to the RAI front matter, TVA confirms NEI 05-04 Revision 3 was used for the FSPR.

TVA has chosen Option a/b/c/d/e and g to respond to this RAI.

Response a Internal Events PRA F&O FSPR (2015) - The scope of the FSPR was a set of 65 SRs assessed as Not Met (or Category I), and associated Finding F&Os, from the 2009 Boiling Water Reactor Owners Group PRA Peer Review, and a set of additional Findings that were associated with SRs determined by the original peer review to be Capability Category II or better. There were no new methods identified as part of this peer review. The scope of the FSPR was for internal events only and did not assess any open Finding level Internal Flooding F&Os.

Fire PRA F&O FSPR (2015) - The scope of the FSPR, was to review the BFN Fire PRA against selected technical elements in Section 4 of the ASME/ANS PRA Standard. The detailed scope of the FSPR was based upon the results of the 2012 peer reviews. The scope of the review consisted of:

58 Supporting Requirements (SRs) identified by TVA for review. Two additional SRs were identified and addressed during the peer review, for a total of 60 SRs.

Review of TVA closure of 67 F&Os identified as Closed.

The review was focused on specific aspects of the FPRA that had changed. Most notably, the review focused on:

- Changes to PRA modeling of credit taken for the Alternate Shutdown Cooling (ASDC) mode of the Residual Heat Removal (RHR) System.

- Changes to PRA modeling associated with a planned modification to install a new Emergency High Pressure Makeup (EHPM) Pump and related support systems.

- Updated PRA methodologies/approaches.

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- Areas of focus identified by TVA in responses to RAIs from the NRC as part of the NFPA 805 transition process.

- There were no new methods identified as part of this peer review.

Response b Internal Events PRA F&O FSPR (2015) - The FSPR team used a consensus process for reaching conclusions regarding adequacy of closure of the 2009 finding level F&Os and their associated supporting requirements, and to judge if requirements were met. The FSPR determined that, of the 65 SRs previously determined to be Not Met or Not Cat II:

27 SRs are Met at Cat II or better; this includes 6 SRs originally graded as Cat I 4 SRs remain Met at Cat I 32 SRs remain Not Met 2 SRs that were originally noted as Met in the 2009 review remain as Met A total of 86 Finding level F&Os were evaluated. This number included 67 F&Os associated with the Not Met SRs from the 2009 Peer Review and 19 Findings associated with CC-II SRs.

37 Findings associated with Not Met SRs are Resolved and should no longer need to be considered issues for the BFN PRA 30 Findings associated with Not Met SRs are Not Resolved, and the reasons were provided in the report. Many of the shortcomings are related to documentation issues.

10 Findings associated with Met SRs are resolved and should no longer need to be considered issues for the BFN PRA 9 Findings associated with Met SRs are Not Resolved, and the reasons were provided in the report. Many of the shortcomings are related to documentation issues.

Overall, 47 of the 2009 Peer Review Findings are determined to be Resolved (Closed), and 39 are Not Resolved (Open). There were an additional 9 Findings associated with Met SRs from the 2009 review that were not assessed during this review due to time constraints. The F&Os which the FSPR has assessed as Resolved can be considered to be closed, and no longer relevant for future risk-informed applications of the BFN PRA, and the associated SRs which the FSPR has assessed as Met (or Category II) can be considered as having that capability category.

Fire PRA F&O FSPR (2015) - Section 4 of the ASME/ANS combined PRA Standard contains a total of 183 SRs under 13 technical elements, and configuration control from Section 1.5. Of these 183 SRs, 24 were directly within the scope of the BFN Focused-Scope Peer Review. An additional 36 SRs back referenced Internal Events SRs were also within the scope. The determination of the Capability Category for each SR was based on consensus of the review team.

55 SRs are met at Capability Category (CC) II, or better 5 Are not applicable, back referenced from the Internal Events Peer Review. They are not applicable because grouping has not been utilized for the systems evaluated.

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Distribution of F&Os 46 F&Os were judged to be resolved and were closed 2 F&Os remained open 7 F&Os were not evaluated and remain open 2 new F&Os were generated Response c Internal Events PRA F&O FSPR (2015) - TVA treated all F&Os as upgrades, as defined in ASME/ANS RA-Sa-2009, by subjecting them to a FSPR. As described in the FSPR report, the methodology for the review is consistent with that defined in NEI 05-04 R3.

Information/documents were provided to the Peer Review team prior to the on-site review, including TVA resolution to each Finding level F&O, PRA quantification and others requested by the Team Leader. The F&O resolutions were assessed against the associated SRs and the technical adequacy was determined by assignment of a Capability Category. All F&Os determined to be resolved (closed) required a CC-II or better. Accordingly, the criteria for assessing the need for a model update has been met.

Fire PRA F&O FSPR (2015) - TVA treated all F&Os as upgrades, as defined in ASME/ANS RA-Sa-2009, by subjecting them to a FSPR. The FPRA FSPR was a tiered process in which the reviewer began with a relatively high level examination of the PRA technical element(s) against the supporting requirements, and progressed successively to additional levels of detail as necessary to ensure the robustness of the model until all of the requirements are adequately reviewed. The review involved a combination of a broad scope examination of the PRA element(s) based on what is found during the review. The SRs provided a structure, which in combination with the peer reviewers PRA experience provides the basis for examining the various PRA technical elements. The SRs helped to ensure completeness in the review, and included any clarifications provided in RG 1.200.

Response d Internal Events PRA F&O FSPR (2015) - A consensus process for reaching conclusion regarding adequacy of closure of the 2009 PRA Peer Review Finding level F&Os and consideration of the associated SRs was performed against the requirements of the 2009 ASME/ANS PRA Standard for SRs at Cat II. The team documented the basis for closure or the reason why an F&O remained open in the report. A lead reviewer was assigned for each F&O under consideration, discussion was held with the assigned reviewers. Then a consensus session was employed which consisted of all members of the Peer Review team.

Fire PRA F&O FSPR (2015) - The FSPR was performed using the process defined in NEI 07-12 against the selected technical elements of Section 4 of the 2009 ASME/ANS PRA Standard for SRs at Cat II. The industry guidance requires that in reaching their conclusions regarding the technical adequacy of the various elements and the FPRA as a whole, reviewers are expected to investigate the FPRA at different levels. The reviewers worked in small teams, and then presented their views to the entire Peer Review team, at which time a consensus process was used to determine the final capability category of each FPRA SR.

This process was applied to open F&Os as well.

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Response e Internal Events PRA F&O FSPR (2015) - A total of 86 Finding level F&Os from the 2009 review were evaluated. This included 67 findings associated with Not Met SRs from the 2009 review and 19 Findings associated with Met SRs from the 2009 review. Of these, the assessment of this review is that:

37 Findings associated with Not Met SRs are Resolved (Closed) and should no longer need to be considered issues for the BFN PRA.

30 Findings associated with Not Met SRs are Not Resolved (Open), for the reasons noted in the report. Many of these resolution gaps are related to PRA documentation.

Fire PRA F&O FSPR (2015) - A total of 48 Finding level F&Os were assessed, of which the determination of the status for each F&O was based on consensus of the review team.

Distribution of F&Os 46 F&Os were judged to be resolved and were closed 2 F&Os remained open 7 F&Os were not evaluated and remain open 2 new Finding level F&Os were generated Additional Information - The part of (e) that reads: For each identified finding-level F&O, describe the disposition and the impact of the F&O on the PRA as it pertains to this application relates to the F&Os that remain open. As provided in Enclosure 2, TVA proposes a License Condition to resolve all open F&Os using the Appendix X process prior to extending any surveillance test intervals.

Response g As provided for in Enclosure 2, TVA proposes the mechanism that ensures that all F&Os that were not closed during the 2015 FSPRs, is to close all remaining F&Os using the regulatory approved Appendix X to ASME/ANS RA-Sa-2009 and process prior to extending any surveillance test intervals.

APLB RAI 02 Disposition for Open F&Os Tables 11 and 13 of the LAR, Open Internal Events With Internal Flooding PRA Open F&Os, and Fire PRA Open F&Os, respectively, provide the licensees disposition for each of the PRA F&Os. Address the following:

a. For several of the findings (i.e., 4-18, 4-25, 6-30, and IFQU A6-01) TVA concludes for impact on the surveillance frequency control program (SFCP) that STI

[surveillance test interval] changes are not affected by human error probabilities (HEPs) as the calculation determines the change in risk due to changes in reliability, and goes on to further state that minimal or no impact on the PRA results is expected. A PRA model uses the as- built, as-operated plant, which includes operational practices that can influence the results of the PRA. For the SFCP, the results of the PRA assess changes in the core damage frequency (CDF) and large early release frequency (LERF) to determine the extension for an STI. For the above CNL-20-089 E1-6 of 15

F&Os (i.e., 4-18, 4-25, 6-30, and IFQU A6-01), it is unclear to the NRC staff why the operator responses and HEPs will not impact potential STI changes.

i. Provide sufficient justification for these F&Os that supports how it was determined that future STI extensions are not adversely impacted (e.g., results of a sensitivity) or identify how NEI 04-10 (generic sensitivities) addresses future STI extensions within the established process for each of the F&Os. If any F&Os are determined to need to be addressed on a case-by-case basis for future STI evaluations, include that in the justification.

b. The finding (F&O 1-33) pertaining to supporting requirement LE-F2 observes that TVA did not review LERF contributors for reasonableness. TVA stated that its documentation provides a listing of addressed phenomena and failures postulated to lead to LERF and explained that the Browns Ferry model provides mapping to these postulated events in the QU notebook and includes a comparison of absolute frequency to similar designs. TVA further states that this is a documentation issue and that a reasonableness check of results ensures the actual results obtained align with expected results and, therefore, no impact is expected on the STI change evaluations, it is unclear to the NRC staff how TVA assured that conservatisms have not skewed the results (level of plant-specificity is appropriate for significant contributors, etc.). Either:

i. Provide sufficient justification that confirms the expectation that the F&O will have no impact on the STI change evaluations.

OR ii. Perform and document a review of the reasonableness of the contributors to LERF consistent with SR LE-F2 at capability category (CC)-II of the ASME/ANS RA-Sa- 2009 PRA standard and provide a summary of those results to the NRC staff.

c. For several of the findings (i.e., IFSN-A10-01, IFSN-A10-02, IFEV-A1-01, and IFQU A9- 01) TVA concludes for impact on the SFCP that [the] issues impact potential initiating events, but the scenario response would be characterized by scenarios already modeled [and] STI changes are not affected by flood initiating events It is unclear to the NRC staff how the potential impact of the initiating events has been appropriately characterized to ensure that the scenario responses are adequate to ensure that future STI changes are not affected.

i. For the finding-level F&Os (i.e., IFSN-A10-01, IFSN-A10-02, IFEV-A1-01, and IFQU A9-01), provide sufficient justification that determines that future STI changes are not adversely impacted (e.g., results of a sensitivity) or identify how NEI 04-10 (generic sensitivities) addresses future STI changes within the established process for each of the F&Os. If any F&Os are determined to need to be addressed on a case-by-case basis for future STI evaluations, include in the justification.

d. For several of the findings related to the fire PRA (FPRA) (i.e., 2-38, 2-39, 2-50, AS-A5, 4-21, 4-12 for SR AS-A5), and 9-4,) the analysis does not reflect the as-built, as- operated plant (e.g., existing Browns Ferry procedures cannot be updated CNL-20-089 E1-7 of 15

until the NFPA 805 modifications and completion of the post-transition safe shutdown procedures have been completed). For impact on SFCP the licensee concludes the STI changes are not affected by either HEPs or spurious operations events, etc. Furthermore, in Section 7.1 of the LAR the licensee provides a table that includes the FPRA results for Units 1, 2, and 3. It is unclear to the NRC staff why the operator responses, HEPs, spurious operations events, etc. will not impact future STI evaluations. Also, it is not clear to the NRC staff if the results for the FPRA provided in Table 16 of the LAR reflect the current as-built, as-operated plant or the future state after the NFPA-805 transition has been completed.

i. Provide sufficient justification for these F&Os that supports how it was determined that future STI extensions are not adversely impacted (e.g., results of a sensitivity) or identify how NEI 04-10 addresses future STI extensions (within the established process) for each of the F&Os.

AND ii. Clarify, if the FPRA results provided in Table 16 of the LAR represent the current as- built, as-operated Browns Ferry plant or the future state after the NFPA-805 transition has been completed. If the FPRA results represent the future state of the plant, propose a mechanism that ensures that any open F&Os will be addressed as applicable for the STI at the time the evaluation ,

PRA model used will represent the as-built, as-operated plant).

TVA Response to APLB RAI 02

a. i. As provided for in Enclosure 2, TVA has proposed a License Condition to ensure all open F&Os are closed prior to any surveillance test interval extensions are implemented.

This will justify that these F&Os will not adversely affect any STI extensions.

b. i. As provided for in Enclosure 2, TVA has proposed a License Condition to ensure all open F&Os are closed prior to any surveillance test interval extensions are implemented.

This will justify the expectation that F&O 1-33 will have no impact on the STI change evaluations.

c. i. As provided for in Enclosure 2, TVA has proposed a License Condition to ensure all open F&Os are closed prior to any surveillance test interval extensions are implemented.

This provides justification that F&O IFSN-A10-01, IFSN-A10-02, IFEV-A1-01, and IFQU A9-01 will not adversely affect future STI extensions.

d. i. As provided for in Enclosure 2, TVA has proposed a License Condition to ensure all open F&Os are closed prior to any surveillance test interval extensions are implemented.

This provides justification that F&Os 2-38, 2-39, 2-50, AS-A5, 4-21, 4-12, and 9-4 will not adversely affect future STI extensions.

and CNL-20-089 E1-8 of 15

ii. The BFN NFPA 805 LAR made the following commitments (as rendered enforceable under License Conditions 2.C.(13), 2.C.(14), and 2.C.(7) for BFN Units 1, 2, and 3, respectively):

NFPA 805 LAR, Attachment S, Table S-3, Item 32 - Update the Fire PRA model, as necessary, after all modifications are complete (returned to operation) and in their as-built configuration. The update will include a verification of the validity of the reported change in risk on as-built conditions after the modifications are completed.

If this verification determines that the risk metrics have changed such that the RG 1.174 acceptance guidelines are not met, the Nuclear Regulatory Commission (NRC) will be notified and additional analytical efforts, and/or procedure changes, and/or plant modifications assure the RG 1.174 risk acceptance criteria are met.

NFPA 805 LAR, Attachment S, Table S-3, Item 33 - Update the fire HRA (Human Reliability Analysis) upon completion of all procedure updates, all modifications and all training. The update will include a verification of the validity of the reported change in risk on as-built conditions after the procedure updates, modifications, and training are completed. If this verification determines that the risk metrics have changed such that the RG 1.174 acceptance guidelines are not met, the Nuclear Regulatory Commission (NRC) will be notified and additional analytical efforts, and/or procedure changes, and/or plant modifications will be made to assure the RG 1.174 risk acceptance criteria are met.

The commitments made in NFPA 805 LAR, Attachment S, Table S-3, Items 32 and 33 have been completed. Therefore, all plant modifications and procedure updates credited in the BFN NFPA 805 LAR have been incorporated into the current Fire PRA model, which represents the current as-built, as-operated plant. The BFN Fire PRA results included in the SFCP LAR were generated using this updated model and thus represent the results for the current as-built, as-operated plant.

APLB/C RAI 03: Key Assumptions and Sources of Uncertainties NEI 04-10, Revision 1, Step 5 discusses how Regulatory Guide RG 1.200, Revision 2, provides attributes of importance for risk determinations relative to external events, seismic, internal fires, and shutdown. This RG specifically addresses the need to evaluate important assumptions that relate to key modeling uncertainties and the need to evaluate parameter uncertainties and demonstrate that calculated risk metrics (e.g., CDF and LERF) represent mean values.

Sections 3.4.4, 3.5.3, and 3.6.3 of Attachment 2 to the LAR describes the approach for the licensee used for the identification of Internal Events (includes internal floods), fire, and seismic key assumptions and sources of uncertainty. For seismic, the licensee states that the seismic PRA (SPRA) was built off of the full-power IEPRA; therefore, any assumptions that are key in the full-power IEPRA are also key assumptions for the SPRA. However, the licensee did not provide a complete list identifying the key assumptions and sources of uncertainty, and how impacts for this application were assessed.

a. Provide a brief description of how the key assumptions and sources of uncertainties for the IEPRA (includes internal floods), FPRA, and SPRA were identified from the initial comprehensive list of PRA model(s) (i.e., base model) source of uncertainties and assumptions, including those associated with plant-specific features, modeling choices, and generic industry concerns. Include a disposition for each of the key assumptions and/or key sources of uncertainties, addressing their impact on the risk-CNL-20-089 E1-9 of 15

informed application. For any key source of uncertainty or key 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. Identify appropriate sensitivity cases that will be used to support the disposition for this application or use a qualitative discussion to justify that the identified key assumption would not affect this application.

TVA Response to APLB/C RAI 03 Overview To identify the key assumptions and uncertainties associated with the PRA models supporting this application, the generic issues identified in Table A.1 of EPRI 1016737 and EPRI 1026511 were reviewed, as well as the BFN PRA documentation for plant-specific assumptions and uncertainties. The BFN PRA documentation reviewed included the notebooks supporting the internal events PRA, internal floods PRA, FPRA, and SPRA. BFN does not have a low power and shutdown (LPSD) PRA, therefore this type of hazard was not considered further.

Each key assumption and source of uncertainty was evaluated to determine whether it was key for the application, using an approach based on screening criteria. A disposition was generated, addressing the impact on the risk metrics of interest for the risk-informed application.

The risk metrics of interest for TSTF-425 are core damage frequency (CDF) and large early release frequency (LERF).

The screening criteria that were used to determine whether an assumption or source of uncertainty that is key to one or more of the PRA models was key to the application are as follows. If the PRA model key assumption or source of uncertainty satisfied any of these criteria, it was considered to be not key to the application. If the PRA model key assumption or source of uncertainty does not satisfy any of these criteria, it was considered to be key to the application.

1. The assumption or source of uncertainty is addressed by implementing a consensus model as defined in NUREG-1855 Revision 1. EPRI 1013491 elaborates on the definition of a consensus model to include those areas of the PRA where extensive historical precedent is available to establish a model that has been accepted and yields PRA results that are considered reasonable and realistic. Thus, assumptions and sources of uncertainty for which there is extensive historical precedent, and which produce results that are reasonable and realistic, are screened from further consideration.

2. The assumption or source of uncertainty has no impact or insignificant impact on the PRA results and therefore no impact or insignificant impact on the calculated change in risk due to changing surveillance frequencies.

3. The assumption or source of uncertainty introduces a realistic conservative bias in the PRA model results. EPRI 1013491 uses the term realistic conservatisms and notes that judiciously applied realistic conservatism can provide a PRA that avoids many of the traps associated with the use of excess conservatism. This criterion, which allows screening of sources of conservative bias, is intended to be less restrictive than the previous criterion, which does not distinguish between conservative and CNL-20-089 E1-10 of 15

nonconservative bias. Thus, using this criterion, assumptions that introduce realistic (slight) conservatisms can be screened from further consideration.

4. For an assumption under consideration, there is no reasonable alternative assumption or reasonable modeling refinement that would change the risk profile of the plant. Here, reasonable alternative assumption is taken with the meaning given in Regulatory Guide 1.200, Revision 2.

5. For an assumption under consideration, there is no reasonable alternative assumption or reasonable modeling refinement to address the uncertainty that is at least as sound as the assumption under consideration.

6. The potential conservatism could result in an overstated delta risk, which could influence a decision made based on the TSTF application. However, the undesired consequence, if any, of overstated risk would be to forego decreases in surveillance frequencies that might otherwise be allowed. Foregoing reductions in surveillance frequencies preserves the existing safety basis of the plant based on the existing technical specifications.

Because retaining the potential conservatism is in that sense fail-safe, the assumption or source of uncertainty is determined to be not key for the application and is not investigated further.

For each BFN identified assumption or source of uncertainty, a qualitative discussion was sufficient to demonstrate that it met one or more of the above screening criteria and therefore was not key. In some cases, the qualitative evaluation was complemented by insights from PRA results to help screen the assumption or source uncertainty, based on insignificant impact on CDF and LERF.

As a result of this review, it was determined that there are no assumptions or sources of uncertainty that qualify as key to the application.

No sensitivity cases were required, due to the fact that there are no key sources of uncertainty relevant to the application.

Evaluation of Potential Key Assumptions and Sources of Uncertainty The following assumptions from the BFN PRA notebooks were identified as potential key sources of uncertainty due to a degree of conservatism that has not been demonstrated to be slight in accordance with Criterion 3. However, in accordance with Criterion 6, the uncertainties are determined not to be key uncertainties for the TSTF-425 application.

Assumptions in the system notebooks for control rod drive hydraulics, core spray, heating ventilation and air conditioning, and residual heat removal potentially introduce a conservative bias in the PRA model results because they assume that the reactor building quadrant room HVAC failure would fail the system although there is a probability that the system would not fail. Likewise, and assumption in the system notebook for the emergency diesel generators potentially introduces a conservative bias in the PRA model results because it assumes that HVAC failure would fail the emergency diesel generators although there is a probability that they would not fail.

In accordance with an assumption in the primary containment isolation system notebook, the suppression chamber drains were modeled in the containment isolation analysis CNL-20-089 E1-11 of 15

because they are normally open valves. The assumption potentially introduces a conservative bias in the LERF model results because it does not account for the potential mitigation of radioactive release by components downstream of the valves.

Assumptions in the raw cooling water system notebook assume that plugging of the RCW intakes affects all three plant units in a similar manner. The intake plugging initiator is a multiunit initiator that scrams all three units due to common cause plugging off all RCW intakes. The plugging and mitigation probabilities of the intake plugging model are conservative screening values.

An assumption in the reactor protection system (RPS) notebook applies a conservative upper bound on the RPS system for failure to SCRAM. A consensus model for CCF is used to estimate the CCF probabilities. However, the estimation of the failure probability of the RPS is subject to high uncertainty as no actual failures have occurred. The estimate is most likely a conservative upper bound on RPS performance during that period, given previous estimates of RPS unavailabilities.

Assumptions in the RPS notebook indicate that no credit is taken for the operator initiating a manual scram, due to the inability to ensure that adequate time exists to initiate the scram prior to exceeding peak reactor pressure vessel (RPV) design pressure. If the RPS fails to scram the reactor, the operator will attempt a manual scram upon recognition that the automatic scram failed. No credit for manual scrams is taken because adequate time cannot be assured to exist for all conditions. This introduces conservatism in the PRA results.

In accordance with an assumption in the recirculation pump trip notebook, the end of cycle portion of the recirculation pump trip system (EOC-RPT) is not credited in the model. During a load reject condition or following a turbine trip, the end-of-cycle (EOC) portion of the RPT system assists the RPS in shutting down the reactor. The EOC trip provides additional negative reactivity at normal end-of-cycle conditions when the control rods may all be fully withdrawn and thermal neutron flux may have shifted upwards in the core, thus delaying the effect of negative reactivity from a control rod scram. Not crediting the Unit 1 EOC-RPT introduces a conservatism in the PRA model because it assumes a redundant trip mechanism available for Unit 1 fails (it is disabled in Units 2 and 3).

In accordance with an assumption in the residual heat removal notebook, no credit is taken for mitigation of Anticipated Transient Without Scram by RPT when vessel level is lowered to top of active fuel because detailed analysis is not available to show it will be successful for BFN.

In accordance with an assumption in the standby liquid control notebook, failure to isolate the reactor water cleanup (RWCU) system may not result in immediate failure of the standby liquid control (SLC) system; however, this isolation failure is conservatively assumed to result in SLC function failure. The assumption may introduce a conservative bias in the PRA model results because it is not certain that the operation of the RWCU system would dilute the injected sodium pentaborate solution enough to fail the SLC function.

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In accordance with an assumption in the SLC notebook, SLC is not modeled as a successful RPV injection source. The assumption may introduce a conservative bias in the PRA model results because a potential RPV injection source is not credited.

In accordance with an assumption in the LERF notebook, two vacuum breaker lines failed open constitutes vapor suppression failure (i.e., bypass). Extensive MAAP calculations and MELCOR calculations for typical Boiling Water Reactor Mark I containments have indicated that even with two vacuum breakers failed open, the containment pressurization at RPV breach will not result in containment pressures greater than the ultimate capability. Therefore, the assumptions being used in the BFN PRA regarding vapor suppression failure appear to be conservative.

The generic sources of modeling uncertainty from EPRI Report 1016737 were evaluated for the baseline BFN internal events PRA. The following generic sources of uncertainty were identified as potential key sources of uncertainty with conservatism that is acceptable for the application.

The extent of the conservatism in each case has not been demonstrated. However, in accordance with Criterion 6, the potential conservatism is acceptable for the TSTF-425 application.

Topic 9: Room heat-up calculations. Loss of heating ventilation and air conditioning (HVAC) can result in room temperatures exceeding equipment qualification limits. This uncertainty is discussed in the bulleted list above.

Topic 12: Containment sump/strainer performance. There is not a consistent method for the treatment of suppression pool strainer performance. The low pressure coolant injection, containment spray, and suppression pool cooling modes of the residual heat removal (RHR) system share a common suction path from the suppression pool. This path begins with suction strainers inside the torus shell. There are four stainless-steel mesh suction strainers connected to four lines connecting the suppression pool to the 30 emergency core cooling system ring header. The screens are mounted above the bottom of the suppression pool to minimize plugging. Larger strainers have been installed on all units to address industry concerns about strainer plugging. The individual probability of debris-induced loss of NPSH to the RHR pumps has been addressed realistically by using a generic prior distribution for strainer plugging from industry data supplemented by a Bayesian update using plant specific data. Common cause failure (CCF) is addressed by using OR logic for the four strainers, implying complete CCF dependence. The assumption of complete CCF dependence is conservative because adequate suction flow does not require success of all four strainers.

Generic sources of modeling uncertainty from EPRI Report 1026511 were evaluated for the TSTF-425 application. EPRI Report 1026511 focuses on sources of uncertainty from (a) fire PRAs, (b) seismic PRAs, (c) low power and shutdown (LPSD) PRAs, and (d) Level 2 PRAs.

LPSD PRA is not relevant to the TSTF-425 application. This leaves sources of uncertainty from fire, seismic, and Level 2 PRAs. For the TSTF-425 application, the risk metric of interest after core damage is the large early release frequency (LERF). The evaluation determined that none of the generic sources of uncertainty from EPRI Report 1026511 have the potential to be key sources of uncertainty for the TSTF-425 application.

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Conclusion The evaluation of PRA assumptions and the generic sources of uncertainty determined that there are no key sources of uncertainty for the TSTF-425 application. Several conservative assumptions and generic sources of uncertainty that are addressed by conservative approaches were initially evaluated as having the potential to be key sources of uncertainty for the TSTF-425 application. However, further evaluation determined that the conservatisms are not considered key sources of uncertainty and are acceptable for the TSTF-425 application because the potential consequence of the conservatisms is to forego decreases in surveillance frequency that might otherwise be allowed in accordance with the TSTF-425 application.

Foregoing reductions in surveillance frequency preserves the existing safety basis of the plant based on the existing technical specifications. No sensitivity studies were required to disposition potential key uncertainties because the potential key uncertainties were conservatisms that were found not to be key to the TSTF-425 application.

APLC RAI 04: Considerations of Tornado Missiles NEI 04-10, Revision 1, states that external events risk impact may be considered quantitatively or qualitatively. The NRC staffs safety evaluation on NEI 04-10, Revision 1, states that a qualitative screening analysis may be used when the surveillance frequency impact on plant risk can be shown to be negligible or zero.

Section 4.1 of Attachment 2 to the LAR provides the considerations of extreme wind and tornado. The licensee screened extreme wind and tornado based on the current plant design.

However, the licensee did not provide any discussion on tornado missile impacts on the systems, structures, and components that are considered in this LAR. Therefore, it is unclear to the NRC staff whether and how the tornado missile impact has been considered for this application.

Explain how the risk from tornado generated missiles on systems, structures, and components, including, but not limited to, the non-conformances identified in the licensees LAR to use the Tornado Missile Risk Evaluator (TMRE) methodology (ADAMS Accession No. ML20127H904),

is considered for this application. The explanation should include either a description of the approach that will be followed for considering such risks and its consistency with the endorsed guidance in NEI 04-10, Revision 1, or justification for screening the risk from tornado generated missiles for this application.

TVA Response to APLC RAI 04 BFN is designed to protect against tornado missiles per the Updated Final Safety Analysis Report (UFSAR). The non-conformances identified during the internal response to Regulatory Issue Summary (RIS) 2015-006, Tornado Missile Protection, were shown to be of low significance through the LAR to incorporate the TMRE methodology evaluation into the BFN UFSAR (ML20127H904), which is currently under NRC review. The change in CDF and LERF both fall below the Regulatory Guide 1.174 guidelines and are therefore considered very small.

Table 14 of the TSTF-425 LAR provides a basis for why tornados were screened from the SFCP analysis. Due to the TMRE analysis showing a low significance for the non-conformances, the screening provided in Table 14 remains the same.

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APLC RAI 05: Considerations of External Flooding NEI 04-10, Revision 1, states that external events risk impact may be considered quantitatively or qualitatively. The NRC staffs safety evaluation on NEI 04-10, Revision 1, states that a qualitative screening analysis may be used when the surveillance frequency impact on plant risk can be shown to be negligible or zero.

Section 4.1 of Attachment 2 to the LAR provides the considerations of external flooding, including the probable maximum flood (PMF). The licensee screened external flooding based on the current plant design. However, the licensee did not discuss other external flooding mechanism, such as the local intense precipitation (LIP) event. The licensees flood hazard re-evaluation report (FHRR; ADAMS Accession No. ML15072A130) and its corresponding NRC staff assessment (ADAMS Accession No. ML16196A088) identified LIP as an external flooding hazard that exceeds the plants design basis. Therefore, it is unclear to the NRC staff whether and how the risk from LIP has been considered for this application.

Explain how the risk from LIP, is considered for this application. The explanation should include either a description of the approach that will be followed for considering the risk and its consistency with the endorsed guidance in NEI 04-10, Revision 1, or justification for screening the risk for this application.

TVA Response to APLC RAI 05 TVA chooses to provide a justification for screening the risk of LIP, relative to this application.

The external flooding hazard at BFN was evaluated as a result of the post- Fukushima 10 CFR 50.54(f) Request for Information. The FHRR was submitted to NRC for review on March 12, 2015. The results indicate that flooding from all hazards, except LIP, are bounded by the current plant design basis and do not pose a challenge to the plant. The FHRR included commitments describing required actions to provide BFN protection against the re-evaluated LIP hazard. Flooding from LIP was subsequently evaluated in a Focused Evaluation (FE) which concluded that effective flood protection was provided for LIP, and that the actions required for the LIP protection strategy were determined to be feasible. The BFN FE was submitted to the NRC in a letter dated October 11, 2019 (ML19284F761). This FE was accepted in an NRC Staff Evaluation, dated May 6, 2020 (ML20112F485). This justifies screening the risk of LIP for this application.

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Enclosure 2 BFN Units 1, 2, and 3 Renewed Facility Operating License Markup Pages (4 Pages)

CNL-20-089 E2-1 of 1

INSERT 1 (XX) TVA shall close all open Facts and Observations (F&Os) listed in Tables 11 and 13 to Attachment 2 of TVA Letter CNL-20-003, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (BFN-TS-516), dated March 27, 2020, prior to implementing any Surveillance Test Interval extensions under the Surveillance Frequency Control Program. The F&O closures will be performed in accordance with the ASME/ANS RA-Sa-2009 PRA Standard, as endorsed by Regulatory Guide 1.200.

-6f-Insert 1 (23) Maximum Extended Load Line Limit Analysis Plus (MELLLA+) Implementation Prior to the first implementation of MELLLA+, TVA shall perform reload safety analyses using codes that have been corrected for the errors described in TVA letter CNL-19-125, dated December 19, 2019.

D. The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21(d), shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71(e)(4) following the issuance of this renewed operating license. Until that update is complete, TVA may make changes to the programs and activities described in the supplement without prior Commission approval, provided that TVA evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

E. The UFSAR supplement, as revised, describes certain future activities to be complete prior to the period of extended operation. TVA shall complete these activities no later than December 20, 2013, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

F. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of the most recent NRG-approved version of the Boiling Water Reactor Vessels and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) appropriate for the configuration of the specimens in the capsule. Any changes to the BWRVIP ISP capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules _placed in storage BFN-UNIT 1 Renewed License No. DPR-33 Amendment No. 310

-6f-Insert 1 (23) Maximum Extended Load Line Limit Analysis Plus (MELLLA+) Implementation Prior to the first implementation of MELLLA+, TVA shall perform reload safety analyses using codes that have been corrected for the errors described in TVA letter CNL-19-125, dated December 19, 2019.

D. The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21(d), shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71(e)(4) following the issuance of this renewed operating license. Until that update is complete, TVA may make changes to the programs and activities described in the supplement without prior Commission approval, provided that TVA evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

E. The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. TVA shall complete these activities no later than June 28, 2014, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

F. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of the most recent NRC-approved version of the Boiling Water Reactor Vessels and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) appropriate for the configuration of the specimens in the capsule. Any changes to the BWRVIP ISP capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage BFN-UNIT 2 Renewed License No. DPR-52 Amendment 333

-6f-(16) Radiological Consequences Analyses Using Alternative Source Terms TVA shall perform facility and licensing basis modifications to resolve the non-conforming/degraded condition associated with the Alternate Leakage Treatment pathway such that the current licensing basis dose calculations (approved in License Amendment Nos. 251/282 (Unit 1),

290/308 (Unit 2) and 249/267 (Unit 3)) would remain valid. These facility and licensing basis modifications shall be complete prior to initial power ascension above 3458.

(17) Prior to extending the frequency for the Integral Leakage Rate Testing described in TS 5.5.12, the licensee shall implement the modifications, that are modeled in the Fire PRA and described in Table S-2, Plant Modifications Committed, of Tennessee Valley Authority letter CNL-18-100, dated October 18, 2018; as supplemented by letter CNL-19-027, dated February 13, 2019.

(18) Maximum Extended Load Line Limit Analysis Plus (MELLLA+)

Special Consideration The licensee shall not operate the facility within the MELLLA+

operating domain more than a 10°F reduction in feedwater temperature below the design feedwater temperature.

( 19) Maximum Extended Load Line Limit Analysis Plus (MELLLA+)

Implementation Insert 1 Prior to the first implementation of MELLLA+, TVA shall perform reload safety analyses using codes that have been corrected for the errors described in TVA letter CNL-19-125, dated December 19, 2019.

I I D. The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21(d),

shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71(e)(4) following the issuance of this renewed operating license. Until that update is complete, TVA may make changes to the programs and activities described in the supplement without prior Commission approval, provided that TVA evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

E. The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. TVA shall complete these activities no later than July 2, 2016, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

BFN-UNIT 3 Renewed License No. DPR-68 Amendment 293