View source for CNL-19-122, Seismic Probabilistic Risk Assessment Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights

{{Adams
| number = ML19351E391
| issue date = 12/17/2019
| title = Seismic Probabilistic Risk Assessment Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights
| author name = Polickoski J
| author affiliation = Tennessee Valley Authority
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000259, 05000260, 05000296
| license number = DPR-033, DPR-052, DPR-068
| contact person = 
| case reference number = CNL-19-122
| document type = Letter type:CNL, Response to Request for Additional Information (RAI)
| page count = 274
| project = 
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-19-122 December 17, 2019 10 CFR 50.4 10 CFR 50.54(f)
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296

==Subject:==
Seismic Probabilistic Risk Assessment for Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident

==References:==
: 1. NRC Letter, 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 March 12, 2012 (ML12053A340)
: 2. EPRI Report 1025287, Seismic Evaluation Guidance, Screening, Prioritization and Implementation Details [SPID] for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, dated November 2012 (ML12333A170)
: 3. TVA letter to NRC, Tennessee Valley Authoritys Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated March 31, 2014 (ML14098A478)
: 4. NRC letter to TVA, Browns Ferry Nuclear Plant, Units 1, 2 and 3 - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations Relating to Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident (TAC Nos. MF3764, MF3765 and MF3766), dated April 21, 2015 (ML15090A745)

U.S. Nuclear Regulatory Commission CNL-19-122 Page 2 December 17, 2019
: 5. NRC Letter, Final Determination of Licensee Seismic Probabilistic Risk Assessments Under the Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 Seismic of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated October 27, 2015 (ML15194A015)
On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued a Request for Information pursuant to Title 10 of the Code of Federal Regulations (CFR) Part 50.54(f)
(Reference 1) to all power reactor licensees. Enclosure 1 of the 50.54(f) letter requested addressees to reevaluate the seismic hazards at their respective sites using present-day NRC requirements and guidance, and to identify any actions taken or planned to address plant-specific vulnerabilities associated with the updated seismic hazards.
EPRI Report 1025287 (Reference 2) provides the guidance for screening, prioritization, and implementation details for the resolution of the Fukushima Near-Term Task Force (NTTF)
Recommendation 2.1: Seismic. The EPRI Screening, Prioritization and Implementation Details (SPID) guidance was used to compare the reevaluated seismic hazard to the design basis seismic hazard for Browns Ferry Nuclear Plant, Units 1, 2, and 3. As described in Reference 3, Enclosure 2, it was initially proposed by TVA that a seismic probabilistic risk assessment was not warranted because the reevaluated ground motion response spectra (GMRS) was bounded by a seismic response spectra developed for the Individual Plant Examination of External Events (IPEEE). It was understood, however, that the GMRS exceeded the design basis response spectrum in the 1 to 10 Hz screening range utilized by NRC.
Reference 4 is the NRC Staff Assessment for Browns Ferry Nuclear Plant, Units 1, 2, and 3, seismic hazard submittals which concluded that the reevaluated seismic hazards described in Reference 3, Enclosure 2, are suitable for other actions associated with NTTF Recommendation 2.1: Seismic. NRC also concluded that that a seismic risk evaluation was merited.
Following additional interaction with NRC, TVA chose to perform a seismic probabilistic risk assessment for Browns Ferry Nuclear Plant.
In the Reference 5 letter to multiple licensees, NRC indicated that a seismic probabilistic risk assessment was required for Browns Ferry Nuclear Plant, Units 1, 2, and 3, and should be submitted to NRC by December 31, 2019.
The Enclosure to this letter provides the Seismic Probabilistic Risk Assessment Summary Report for Browns Ferry Nuclear Plant, Units 1, 2, and 3, as requested in Reference 5. The Enclosure provides the information requested in Item (8)B of the 50.54(f) letter associated with NTTF Recommendation 2.1: Seismic.
The updated Browns Ferry Nuclear Plant internal events probabilistic risk assessment model was used as the basis for the Seismic Probabilistic Risk Assessment. The internal events probabilistic risk assessment finding-level peer review Facts and Observations were

U.S. Nuclear Regulatory Commission CNL-19-122 Page 3 December 17,2019 dispositioned as discussed in Appendix A of the Enclosure as part of the development of the Seismic Probabilistic Risk Assessment.
This letter contains no new regulatory commitments.
If you have any questions regarding this submittal, please contact Russell Thompson at (423) 751-2567.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 17th day of December 2019.
Respectfully, les T. Polickoski Director, Nuclear Regulatory Affairs

==Enclosure:==

Browns Ferry Nuclear Plant, Units 1, 2, and 3, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report cc (Enclosure):
NRR   Director - NRC Headquarters NRC   Regional Administrator - Region II NRC   Project Manager - Browns Ferry Nuclear Plant NRC   Senior Resident Inspector - Browns Ferry Nuclear Plant

ENCLOSURE Browns Ferry Nuclear Plant, Units 1, 2, and 3 Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report

Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3 Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic

==SUMMARY==
REPORT December 2019

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                         December 2019 Table of Contents Executive Summary ..................................................................................................... 5 1.0   Purpose and Objective ....................................................................................... 6 2.0   Information Provided in This Report ................................................................ 7 3.0   BFN Seismic Hazard and Plant Response...................................................... 11 3.1    Seismic Hazard Analysis ................................................................................................... 11 3.1.1 Seismic Hazard Analysis Methodology ...................................................................... 11 3.1.2 Seismic Hazard Analysis Technical Adequacy .......................................................... 22 3.1.3 Seismic Hazard Analysis Results and Insights .......................................................... 22 3.2    Comparison of NTTF 2.1 Seismic Hazard Submittal and PRA Supplemental Seismic Hazard Analysis .................................................................................................................. 27 3.3    Soil Failure and Fragility Analysis.................................................................................... 31 3.3.1 Soil Failure and Fragility Analysis Technical Adequacy ............................................. 33 4.0   Determination of Seismic Fragilities for the SPRA ........................................ 33 4.1    Seismic Equipment List ..................................................................................................... 33 4.1.1 SEL Development....................................................................................................... 33 4.1.2 Relay and Breaker Evaluation .................................................................................... 37 4.2    Walkdown Approach ........................................................................................................ 108 4.2.1 Significant Walkdown Results and Insights .............................................................. 109 4.2.2 Seismic Equipment List and Seismic Walkdowns Technical Adequacy .................. 109 4.3    Dynamic Analysis of Structures ..................................................................................... 110 4.3.1 Fixed-base Analysis ................................................................................................. 110 4.3.2 Soil-Structure Interaction (SSI) Analysis .................................................................. 110 4.3.3 Structure Response Models ..................................................................................... 112 4.3.4 Seismic Structure Response Analysis Technical Adequacy .................................... 115 4.4    SSC Fragility Analysis ..................................................................................................... 116 4.4.1 SSC Screening Approach......................................................................................... 116 4.4.2 SSC Fragility Analysis Methodology ........................................................................ 117 4.4.3 SSC Fragility Analysis Results and Insights............................................................. 119 4.4.4 SSC Fragility Analysis Technical Adequacy............................................................. 120 5.0   Plant Seismic Logic Model ............................................................................ 121 5.1    Development of the SPRA Plant Seismic Logic Model ................................................ 121 5.1.1 Seismic Initiating Event ............................................................................................ 121 5.1.2 Accident Sequences ................................................................................................. 122 5.1.3 Loss of Offsite Power ............................................................................................... 122 5.1.4 Very Small LOCA (VSLOCA) ................................................................................... 123 5.1.5 SLERF Analysis ........................................................................................................ 123 5.1.6 Summary of Resulting Correlated Component Groupings ....................................... 124 5.1.7 Summary of HRA methodology ................................................................................ 124 Page 2 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                            December 2019 5.1.8 Seismic-Fire .............................................................................................................. 125 5.1.9 Seismic-Flood ........................................................................................................... 125 5.2    SPRA Plant Seismic Logic Model Technical Adequacy ............................................... 126 5.3    Seismic Risk Quantification ............................................................................................ 126 5.3.1 SPRA Quantification Methodology ........................................................................... 126 5.3.2 SPRA Model and Quantification Assumptions ......................................................... 127 5.4    SCDF Results .................................................................................................................... 127 5.4.1 Overall SCDF ........................................................................................................... 127 5.4.2 SCDF as a Function of Hazard Interval.................................................................... 128 5.4.3 Fragility Group Importance for SCDF ....................................................................... 130 5.4.4 SCDF Component Importance (Non-Seismic Failures) ........................................... 137 5.4.5 Significant Human Failure Events ............................................................................ 142 5.4.6 Significant SCDF Accident Sequences .................................................................... 145 5.5    SLERF Results .................................................................................................................. 151 5.5.1 Overall SLERF .......................................................................................................... 151 5.5.2 SLERF as a Function of Hazard Interval .................................................................. 151 5.5.3 Fragility Group Importance for SLERF ..................................................................... 153 5.5.4 SLERF Component Importance (Non-Seismic Failures) ......................................... 161 5.5.5 Significant Human Failure Events ............................................................................ 164 5.5.6 Significant SLERF Accident Sequences .................................................................. 166 5.6    SPRA Quantification Uncertainty Analysis ................................................................... 170 5.6.1 Parameter Uncertainty.............................................................................................. 170 5.6.2 Model Uncertainty..................................................................................................... 178 5.6.3 Completeness Uncertainty ....................................................................................... 178 5.6.4 Truncation Study ...................................................................................................... 178 5.7    SPRA Quantification Sensitivity Analysis ..................................................................... 179 5.8    SPRA Logic Model and Quantification Technical Adequacy....................................... 182 6.0   Conclusions .................................................................................................... 182 7.0   References ...................................................................................................... 183 8.0   Acronyms and Abbreviations ........................................................................ 188 Appendix A ............................................................................................................... 193 A.1    Introduction ...................................................................................................................... 193 A.2    Peer Review of BFN SPRA .............................................................................................. 193 A.3    Revision of Model and Documentation .......................................................................... 203 A.4    Finding-Level F&O Independent Closure Review ......................................................... 203 A.5    Summary of SPRA Capability Relative to SPID Tables 6-4 through 6-5 ..................... 206 A.6    Summary of Technical Adequacy of the Seismic PRA ................................................ 230 A.7    Summary of Technical Adequacy of the BFN Internal Events PRA ............................ 230 A.8    Identification of Key Assumptions and Uncertainties.................................................. 262 Page 3 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                           December 2019 A.9    Identification of Plant Changes Not Reflected in the Seismic PRA ............................ 263 Appendix B ............................................................................................................... 264 Page 4 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 Executive Summary In response to the 10 CFR 50.54(f) letter issued by the Nuclear Regulatory Commission (NRC) on March 12, 2012, a Seismic Probabilistic Risk Assessment (SPRA) has been developed for Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3. The point estimate results of the BFN SPRA are summarized below:
Unit 1        Unit 2      Unit 3 Core Damage Frequency              6.30E-06       6.40E-06    7.13E-06 Large Early Release Frequency          3.00E-06       3.10E-06    3.31E-06 Sensitivity studies were performed to identify critical assumptions, test the sensitivity to quantification parameters and the seismic hazard, and identify potential areas to consider for the reduction of seismic risk. These sensitivity studies demonstrated that the model results were robust to the modeling and assumptions used. No seismic hazard vulnerabilities were identified, and no plant actions have been taken or are planned given the insights from the seismic risk assessment.
Page 5 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                            December 2019 1.0 Purpose and Objective Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter requests that licensees and holders of construction permits under 10 CFR Part 50 reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance.
A comparison between the reevaluated seismic hazard and the design basis for Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3 has been performed, in accordance with the guidance in Electric Power Research Institute (EPRI) 1025287, Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, [2] and previously submitted to the NRC [3]. That comparison concluded that the Ground Motion Response Spectra (GMRS), which was developed based on the reevaluated seismic hazard, exceeds the design basis seismic response spectrum in the 1 to 10 Hz range, and a seismic risk assessment is required. An SPRA has been developed to perform the seismic risk assessment for BFN in response to the 50.54(f) letter, specifically item (8) in Enclosure 1 of the 50.54(f) letter.
This report describes the SPRA developed for BFN and provides the information requested in item (8)(B) of Enclosure 1 of the 50.54(f) letter and in Section 6.8 of the SPID [2]. The SPRA model has been peer reviewed (as described in Appendix A) and found to be of appropriate scope and technical capability for use in assessing the seismic risk for BFN, identifying which structures, systems, and components (SSCs) are important to seismic risk, and describing plant-specific seismic issues and associated actions planned or taken in response to the 50.54(f) letter.
This report provides summary information regarding the SPRA as outlined in Section 2.
The level of detail provided in the report is intended to enable the NRC to understand the inputs and methods used, the evaluations performed, and the decisions made as a result of the insights gained from the BFN SPRA.
Page 6 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 2.0 Information Provided in This Report The following information is requested in the 50.54(f) letter [1], Enclosure 1, Requested Information Section, paragraph (8)B, for plants performing an SPRA.
(1)   The list of the significant contributors to Seismic CDF (SCDF) for each seismic acceleration bin, including importance measures (e.g., Fussell-Vesely)
(2)   A summary of the methodologies used to estimate the SCDF and SLERF, including the following:
: i. Methodologies used to quantify the seismic fragilities of SSCs, together with key assumptions ii. SSC fragility values with reference to the method of seismic qualification, the dominant failure mode(s), and the source of information iii. Seismic fragility parameters iv. Important findings from plant walkdowns and any corrective actions taken
: v. Process used in the seismic plant response analysis and quantification, including the specific adaptations made in the internal events PRA model to produce the SPRA model and their motivation vi. Assumptions about containment performance (3)   Description of the process used to ensure that the SPRA is technically adequate, including the dates and findings of any peer reviews (4)   Identified plant-specific vulnerabilities and actions that are planned or taken Note that 50.54(f) letter Enclosure 1 paragraphs 1 through 6, regarding the seismic hazard evaluation reporting, also apply, but have been satisfied through the previously submitted BFN Seismic Hazard Submittal [3]. Further, 50.54(f) letter Enclosure 1 paragraph 9 requesting information on the Spent Fuel Pool has been satisfied [4,5].
Table 2.0-1 provides a cross-reference between the 50.54(f) reporting items noted above and the location in this report where the corresponding information is discussed.
The SPID [2] defines the principal parts of a SPRA, and the BFN SPRA has been developed and documented in accordance with the SPID. The main elements of the SPRA performed for BFN in response to the 50.54(f) Seismic letter correspond to those described in Section 6.1.1 of the SPID, i.e.:
        -   Seismic hazard analysis
        -   Seismic structure response and SSC fragility analysis
        -   Systems/accident sequence (seismic plant response) analysis
        -   Risk quantification Page 7 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                           December 2019 Table 2.0-2 provides a cross-reference between the reporting items noted in Section 6.8 of the SPID, other than those already listed in Table 2.0-1, and provides the location in this report where the corresponding information is discussed.
The BFN SPRA and associated documentation has been peer reviewed [6] against the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS)
PRA Standard [7, 8] in accordance with the process defined in Nuclear Energy Institute (NEI) 12-13 [9] as documented in the BFN SPRA Peer Review Report. The BFN SPRA, complete SPRA documentation, and details of the peer review are available for NRC review. Reference 7 is the ASME Code Case, which provides Part 5, Requirements for Seismic Events At-Power PRA. Reference 8 is the 2013 Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications and is included because the Code Case has supporting requirements (SR) that back reference the SRs from the 2013 Addenda. Throughout this document, these two documents will be collectively referred to as the PRA Standard.
Subsequent to the peer review, an independent assessment was performed of the closure of Finding-Level Facts and Observations (F&O) of record from the peer review.
The assessment was performed via NEI 12-13 Appendix X guidance, which has been accepted by the NRC [10]. The details of the Finding-Level F&O independent assessment are available for NRC review.
This submittal provides a summary of the SPRA development, results and insights, the peer review process and results, and the independent assessment, sufficient to meet the 50.54(f) information request in a manner intended to enable the NRC to understand and determine the validity of key input data and calculation models used, and to assess the sensitivity of the results to key aspects of the analysis.
The content of this report is organized as follows:
       -   Section 3 provides information related to the BFN seismic hazard analysis.
       -   Section 4 provides information related to the determination of seismic fragilities for BFN SSCs included in the seismic plant response.
       -   Section 5 provides information regarding the plant seismic response model (seismic accident sequence model) and the quantification of results.
       -   Section 6 summarizes the results and conclusions of the SPRA, including identified plant seismic issues and actions taken or planned.
       -   Section 7 provides references.
       -   Section 8 provides a list of acronyms and abbreviations used.
       -   Appendix A provides an assessment of SPRA Technical Adequacy for Response to NTTF 2.1 Seismic 50.54(f) letter, including a summary of BFN SPRA peer review and independent assessment as well as a discussion of the open findings related to the BFN Internal Events PRA (IEPRA).
Page 8 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                               December 2019
      -     Appendix B provides a response for each of the generic observations associated with the staffs review of SPRA reports provided in response to the March 12, 2012, 50.54(f) letter associated with reevaluated seismic hazards.
Table 2.0-1 Cross-Reference for 50.54(f) Enclosure 1 SPRA Reporting 50.54(f) Letter Description                              Location in this Report Reporting Item 1           List of the significant contributors to    The significant contributors are provided in SCDF for each seismic acceleration         Section 5.
bin, including importance measures 2           Summary of the methodologies               A summary of the methodologies utilized to used to estimate the SCDF and              estimate SCDF and SLERF are provided in SLERF                                      Sections 3, 4, and 5.
2i          Methodologies used to quantify the         Seismic methodologies are provided in seismic fragilities of SSCs, together      Section 4.
with key assumptions 2ii         SSC fragility values with reference        Tables 5.4-4, 5.4-5, 5.4-6, 5.5-4, 5.5-5 and to the method of seismic                   5.5-6 provides fragilities (Am, median qualification, the dominant failure        acceleration capacity, and beta, uncertainty mode(s), and the source of                 in capacity), failure mode information, and information                                method of determining fragilities for the top risk-significant SSCs based on Fussell-Vesely (F-V).
2iii         Seismic fragility parameters               Tables 5.4-4, 5.4-5, 5.4-6, 5.5-4, 5.5-5 and 5.5-6 provide fragilities (Am and beta),
failure mode information, and method of determining fragilities for the top risk-significant SSCs based on F-V.
2iv          Important findings from plant              Section 4.2 addresses walkdowns and walkdowns and any corrective               walkdown insights.
actions taken 2v           Process used in the seismic plant          Section 5 provides the processes used in response analysis and                      the seismic plant response.
quantification, including specific adaptations made in the IEPRA model to produce the SPRA model and their motivation 2vi          Assumptions about containment              Sections 4.3 and 5.5 address containment performance                                and related SSC performance.
3           Description of the process used to         Appendix A describes the assessment of ensure that the SPRA is technically        SPRA technical adequacy for the 50.54(f) adequate, including the dates and          submittal and results of the SPRA peer findings of any peer reviews               review and subsequent independent assessment.
4           Identified plant-specific                  Section 6 addresses the plant-specific vulnerabilities and actions that are       vulnerabilities. No vulnerabilities were planned or taken                           identified, and no actions are planned as a result of the SPRA.
Page 9 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                         December 2019 Table 2.0-2 Cross-Reference for Additional SPID Section 6.8 SPRA Reporting SPID Section 6.8 Item Description                            Location in this Report A report should be submitted to the NRC summarizing the SPRA        Entirety of the report addresses this.
inputs, methods, and results.
The level of detail needed in the submittal should be sufficient to Entirety of the report addresses this.
enable NRC to understand and determine the validity of all input    The key methods of analysis and data and calculation models used.                                   referenced codes and standards are identified in the report.
The level of detail needed in the submittal should be sufficient to Entirety of the report addresses this.
assess the sensitivity of the results to all key aspects of the     Results sensitivities are discussed in analysis.                                                           Section 5.7 (SPRA Quantification Sensitivity Analysis).
The level of detail needed in the submittal should be sufficient to Entirety of the report addresses this.
make necessary regulatory decisions as a part of NTTF Phase 2 activities.
It is not necessary to submit all the SPRA documentation for such   Entirety of report addresses this.
an NRC review. Relevant documentation should be cited in the        This report summarizes important submittal and be available for NRC review in easily retrievable     information from the SPRA, with form.                                                               detailed information in lower-tier documentation.
Documentation criteria for an SPRA are identified throughout the    This is an expectation relative to PRA Standard. Utilities are expected to retain that documentation   documentation of the SPRA that the consistent with the PRA Standard.                                   utility retains to support application of the SPRA to risk-informed plant decision-making.
Note (1): The items listed here do not include those designated in SPID Section 6.8 as guidance.
Page 10 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 3.0 BFN Seismic Hazard and Plant Response 3.1   Seismic Hazard Analysis This section discusses the seismic hazard methodology, presents the final seismic hazard results used in the SPRA, and discusses important assumptions and important sources of uncertainty.
The seismic hazard analysis determines the annual frequency of exceedance (AFE) for selected ground motion parameters. The analysis involves use of earthquake source models, ground motion attenuation models, characterization of the site response (e.g.,
soil column), and accounts for the uncertainties and randomness of these parameters to arrive at the site seismic hazard. Detailed information regarding the BFN site hazard was provided to the NRC in the seismic hazard information submitted to the NRC in response to the NTTF 2.1 Seismic information request [3]. As further discussed below, a supplemental seismic hazard analysis has been performed for BFN [11].
3.1.1 Seismic Hazard Analysis Methodology A supplemental seismic hazard analysis [11] was performed for the BFN SPRA in lieu of the NTTF 2.1 Submittal [3] since the site analysis develops the additional elements required for the SPRA such as updated seismic source catalog, Foundation Input Response Spectra (FIRS), hazard-consistent strain-compatible properties, and vertical ground motions. In addition, a site-specific surface geophysics investigation was performed to support the development of the BFN soil profiles.
The GMRS at BFN is defined at the foundation control point corresponding to the base of the Reactor Building (RB).
The following four FIRS were developed for the structures listed in Table 3.1-1 and are summarized below:
* GMRS/FIRS1 - equivalent to the GMRS. GMRS/FIRS1 is located at a control point corresponding to an outcrop spectra at Elevation (EL) 515 ft, re: Mean Sea Level (MSL) at the top of the rock (Fort Payne Formation), at the base of the RB.
The GMRS/FIRS1 soil profile includes 50 ft of compacted earth fill at the top with the zero depth corresponding to EL 565 ft, re: MSL. The soil profile for GMRS/FIRS1 is designated as GMRS/FIRS1/FIRS4 soil profile (since FIRS4 uses the same profile but corresponds to a control point defined at the ground surface as discussed below). The three branches used to develop the GMRS/FIRS1/FIRS4 soil profile adopted the same single branch below EL 515 ft, re: MSL, as discussed in the development of the FIRS3 soil profile below.
* FIRS2 - corresponds to the outcrop spectra at EL 556 ft, re: MSL, located at a control point at the base of the Diesel Generating Building (DGB). The FIRS2 soil profile includes 12 ft of compacted earth fill from EL 565 ft to EL 553 ft, re: MSL, followed by 38 ft of crushed rock fill down to EL 515 ft, re: MSL, below which the rock subsurface (Fort Payne Formation and deeper units) profile ties in. The soil profile for FIRS2 is designated as FIRS2. The three branches used to develop the FIRS2 soil profile adopted the same single branch below EL 515 ft, re: MSL, as discussed in the development of the FIRS3 soil profile below.
Page 11 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019
* FIRS3 - corresponds to the outcrop spectra at EL 515 ft, re: MSL, located at a control point at the base of the Intake Pumping Station (IPS). The FIRS3 soil profile is identical to FIRS1, FIRS2, and FIRS4 when the top 50 ft of compacted earth fill/crushed rock fill are removed. Since this profile only includes the rock (Fort Payne Formation and deeper units), and based on the geophysics interpretations completed as part of the supplemental seismic hazard analysis
[11], a single branch was used to model the epistemic uncertainty for the shear-wave velocity profile. A direct application of the SPID [2] guidelines relative to implementation of epistemic uncertainty by developing multiple branches would not have been appropriate for rock layers below EL 515 ft, re: MSL, due to the knowledge obtained from surface geophysics campaign completed during the supplemental seismic hazard analysis, which was collected around the site perimeter and confirmed the uniformity of those rock layers. While the geophysics campaign did not provide any basis for implementing epistemic uncertainty explicitly in our modeling, and a single branch was used, it is recognized that the uncertainty attributed to aleatory variability (0.25 natural log sigma) in principle does include a minor portion attributed to epistemic uncertainty. However, due to the interpretation of the underlying geology and the lack of any data collected through the geophysics program to support modeling epistemic uncertainty explicitly, a single value was used for modeling purposes in the form of the aleatory variability, with the recognition that a very minor component of that variability is in reality a representation of epistemic uncertainty, which would have an insignificant impact on the quantification of the total uncertainties. The soil profile for FIRS3 is designated as FIRS3.
* FIRS4 - corresponds to a surface-founded FIRS located at control point EL 565 ft, re: MSL, at the base of the Yard Equipment. The FIRS4 soil profile is identical to FIRS1 with the exception that the FIRS is located at the ground surface at EL 565 ft, re: MSL, as opposed to FIRS1, at which the FIRS was located at 50 ft depth (EL 515 ft, re: MSL). The soil profile for FIRS4 is the GMRS/FIRS1/FIRS4 soil profile defined above. The three branches used to develop the GMRS/FIRS1/FIRS4 soil profile adopted the same single branch below EL 515 ft, re: MSL, as discussed in the development of the FIRS3 soil profile above.
To perform the site response analyses for BFN, the random vibration theory approach was employed. This process is consistent with existing NRC guidance and the SPID.
The guidance contained in Appendix B of the SPID on incorporating epistemic uncertainty in shear-wave velocities, non-linear dynamic properties and source spectra was followed for BFN in addition to development of High Frequency (HF) and Low Frequency (LF) controlling earthquakes (control motions) per recommendations in NRC Regulatory Guide (RG) 1.208 [13] for mean annual frequency of exceedance (MAFE) corresponding to 10-2, 10-3, 10-4, 10-5, 5x10-6, and 10-6 at reference rock.
Idealized shear-wave velocity profiles were developed incorporating the existing geotechnical data, onshore geophysics survey, and the derived geologic profile at depth derived for the BFN NTTF 2.1 Seismic Hazard submittal [3], along with the general guidelines included in the SPID to account for the soil profiles epistemic uncertainty and aleatory variability. The idealized shear-wave velocities developed for each of the three Page 12 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                         December 2019 base-case soil profiles for GMRS/FIRS1/FIRS4, FIRS2, and FIRS3 are presented in Figures 3.1-1 to 3.1-6.
Table 3.1-1: Category I Structures and Geotechnical Foundation Material Category I Structure                    Geotechnical. Foundation                 Applicable Elevation Material Reactor Building                        Fort Payne Formation at EL 515 ft,       515 ft with 50 ft of compacted earth fill on top (to EL 565 ft)
Diesel Generator Building               3 ft of compacted earth fill above 38    556 ft ft of crushed rock fill below EL 556 ft, followed by Fort Payne Formation at EL 515 ft, with 9 ft of compacted earth fill (to EL 565 ft) above Intake Pumping Station                  Fort Payne Formation at EL 515 ft        515 ft Ground Surface (Yard Equipment)         Fort Payne Formation at EL 515 ft,       565 ft with 50 ft of compacted earth fill on top (to EL 565 ft)
Page 13 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                     December 2019 Figure 3.1-1: Idealized Shear-wave Velocity (VS) Profiles (GMRS/FIRS1/FIRS4)
Page 14 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                       December 2019 Figure 3.1-2: Idealized Shear-wave Velocity (VS) Profiles in Top 200 ft (GMRS/FIRS1/FIRS4)
Page 15 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                        December 2019 Figure 3.1-3: Idealized Shear-wave Velocity (VS) Profile (FIRS2)
Page 16 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                       December 2019 Figure 3.1-4: Idealized Shear-wave Velocity (VS) Profile in Top 200 ft (FIRS2)
Page 17 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                        December 2019 Figure 3.1-5: Idealized Shear-wave Velocity (VS) Profile (FIRS3)
Page 18 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                       December 2019 Figure 3.1-6: Idealized Shear-wave Velocity (VS) Profile in Top 200 ft (FIRS3)
Page 19 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                            December 2019 To accommodate the full range in expected dynamic material behavior for the firm rock profiles, linear and nonlinear soil dynamic models were included, with equal weights given to each approach. Shear modulus reduction and hysteretic damping curves were used for the various soil layers for the four FIRS. The base-case profiles were randomized to account for aleatory variability in shear-wave velocities and dynamic material properties; sixty randomized profiles were generated.
The results of the site response analyses consist of amplification factors that describe the amplification (or de-amplification) of hard reference rock motion as a function of frequency and input reference rock amplitude. The amplification factors are represented in terms of a median amplification value and an associated standard deviation (sigma) for each oscillator frequency and input rock amplitude. Consistent with the SPID, a minimum median amplification value of 0.5 was employed in the present analysis.
The site amplification factors (SAFs) and logarithmic standard deviations are inputs to develop the full set of site-specific hazard curves that accommodate the randomness and uncertainty in the local dynamic material properties. Sample amplification factors are presented in Figure 3.1-7.
The seismic hazard calculations use a minimum earthquake moment magnitude of 5.0 since the cumulative absolute velocity filter is not used. Soil seismic hazard curves are calculated for frequencies of 0.5, 1, 2.5, 5, 10, and 25 Hz and peak ground acceleration (PGA) (100 Hz). Horizontal uniform hazard response spectrum (UHRS) are calculated for AFEs of 10-2, 10-3, 10-4, 10-5, 5x10-6, and 10-6.
The GMRS and FIRS were developed in accordance with NRC RG 1.208 [13]. Sixty randomizations were generated for the site response for each epistemic branch in the soil logic tree, compared to a minimum of 30 recommended in the SPID. The site response analyses were completed using the HF and LF control motions. Site-specific horizontal hazard curves for each of the FIRS site conditions were used and were developed using Approach 3 of NUREG/CR-6728 [14].
Vertical spectra are developed using vertical-over-horizontal (V/H) scaling relations. The idealized V/H ratios are used to derive the vertical design response spectra from their horizontal equivalents.
For GMRS/FIRS1 and FIRS3, these FIRS are very close to Central and Eastern United States (CEUS) hard reference rock with the average time-weighted shear-wave velocity in the top 100 ft below the ground surface being 8,462 ft/s. As a result, the CEUS V/H scaling relation in NUREG/CR-6728 [14] was used with no modifications. For FIRS2 and FIRS4, the average time-weighted shear-wave velocity in the top 100 ft below those control points was equal to 2,304 ft/s and 1,613 ft/s, respectively. For these two FIRS, in the absence of CEUS V/H scaling relations appropriate for the FIRS, a logic tree was adopted to incorporate epistemic uncertainty by weighting alternative models consistent the methodology described in EPRI 3002004396 [15].
The reference earthquake ground motion to which the fragilities are referenced is represented by the horizontal UHRS at AFE corresponding to 1E-05 at the RB foundation control point. The PGA hazard curve is the ground motion parameter used for the SPRA.
Page 20 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                     December 2019 Figure 3.1-7: GMRS/FIRS1 Site Amplification Factor and Logarithmic Sigmas (100 Hz, 25 Hz, and 10 Hz)
Page 21 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 3.1.2 Seismic Hazard Analysis Technical Adequacy The BFN SPRA hazard methodology and analysis was subjected to an independent peer review against the pertinent requirements in the PRA Standard. After completion of the subsequent independent assessment, the full set of SRs was met. The seismic hazard analysis was determined to be acceptable for use in the SPRA.
The peer review assessment and subsequent disposition of peer review findings through an independent assessment are further described in Appendix A and references [6] and [16].
3.1.3 Seismic Hazard Analysis Results and Insights Table 3.1-2 and Figure 3.1-8 present the mean and fractile exceedance frequencies for the control point corresponding to GMRS/FIRS1 at 100 Hz. Table 3.1-3 provides the final seismic hazard results used as input to the BFN SPRA, in terms of exceedance frequencies as a function of PGA level at the GMRS/FIRS1 control point.
Table 3.1-2: BFN GMRS/FIRS1 Mean and Fractile Exceedance Frequencies at PGA (100 Hz)
Amplitude                                          Fractile Hazard Curves Mean (g)                         0.05       0.16             0.5          0.84         0.95 0.0001        1.641E-01    9.997E-02  1.267E-01        1.630E-01    2.010E-01    2.320E-01 0.00025        1.300E-01    7.065E-02  1.023E-01        1.297E-01    1.600E-01    1.822E-01 0.0005        1.022E-01    4.698E-02  7.965E-02        1.015E-01    1.292E-01    1.469E-01 0.00075        8.487E-02    3.410E-02  6.495E-02        8.369E-02    1.087E-01    1.288E-01 0.001        7.232E-02    2.650E-02  5.419E-02        7.054E-02    9.440E-02    1.153E-01 0.0015        5.513E-02    1.785E-02  3.959E-02        5.333E-02    7.300E-02    9.608E-02 0.002        4.392E-02    1.325E-02  2.962E-02        4.191E-02    5.787E-02    8.388E-02 0.003        3.044E-02    8.529E-03  1.819E-02        2.887E-02    3.966E-02    6.835E-02 0.005        1.804E-02    4.685E-03  8.959E-03        1.656E-02    2.336E-02    4.933E-02 0.0075        1.159E-02    2.730E-03  4.808E-03        1.010E-02    1.582E-02    3.640E-02 0.01         8.428E-03    1.827E-03  3.128E-03        6.615E-03    1.189E-02    2.841E-02 0.015        5.345E-03    9.230E-04  1.688E-03        3.951E-03    8.150E-03    1.889E-02 0.02         3.821E-03    5.199E-04  1.031E-03        2.733E-03    5.848E-03    1.377E-02 0.03         2.301E-03    2.411E-04  4.392E-04        1.405E-03    3.885E-03    8.889E-03 0.05         1.122E-03    9.118E-05  1.456E-04        5.010E-04    2.032E-03    4.701E-03 0.075        5.903E-04    4.492E-05  6.967E-05        2.106E-04    9.901E-04    2.629E-03 0.1         3.603E-04    2.868E-05  4.532E-05        1.193E-04    5.485E-04    1.567E-03 0.15         1.680E-04    1.547E-05  2.465E-05        5.804E-05    2.190E-04    6.621E-04 0.2         9.208E-05    9.688E-06  1.512E-05        3.579E-05    1.143E-04    3.132E-04 0.3         3.654E-05    4.358E-06  7.123E-06        1.652E-05    4.991E-05    1.041E-04 0.5         1.062E-05    1.211E-06  2.332E-06        5.939E-06    1.666E-05    2.786E-05 0.75         3.838E-06    3.463E-07  7.886E-07        2.183E-06    6.492E-06    1.099E-05 1          1.799E-06    1.179E-07  3.179E-07        9.992E-07    3.099E-06    5.613E-06 1.5         5.706E-07    1.591E-08  6.850E-08        2.785E-07    9.617E-07    2.022E-06 Page 22 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                 December 2019 Table 3.1-2: BFN GMRS/FIRS1 Mean and Fractile Exceedance Frequencies at PGA (100 Hz)
Amplitude                                          Fractile Hazard Curves Mean (g)                          0.05        0.16             0.5          0.84         0.95 2          2.350E-07    1.645E-09   1.693E-08        1.010E-07    3.933E-07    9.215E-07 3          5.961E-08    4.555E-17   5.849E-10        1.905E-08    9.632E-08    2.781E-07 5          8.475E-09    5.263E-29   1.611E-14        1.532E-09    1.254E-08    4.847E-08 7.5         1.488E-09    2.202E-29   1.492E-21        1.377E-10    1.858E-09    9.810E-09 10          3.922E-10    2.200E-29   1.069E-24        2.207E-11    4.271E-10    2.622E-09 Figure 3.1-8: PGA (100 Hz) GMRS/FIRS1 Soil Profile Fractile Hazard Curves for BFN 3.1.3.1 Uncertainties in the Seismic Hazard Result from Input Parameters and Models The epistemic and aleatory uncertainties in components of the model, including seismic source characterization and ground motion models, were incorporated using logic trees.
Sensitivity analyses were also performed to assess the input parameters. Sensitivity analyses were performed on the ground motion models and several of the seismic source characterization, including alternatives for magnitude completeness, alternate earthquake recurrence rates, and maximum magnitude alternatives. Based on the sensitivity analyses performed, the epistemic uncertainty in the ground motion models dominates the contribution to the total epistemic uncertainty for the BFN site.
The Central and Eastern United States Seismic Source Characterization (CEUS-SSC) concluded its data gathering efforts in 2008. As a result, a literature search of published Page 23 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                            December 2019 and unpublished data was completed to identify any data that may have an impact on the CEUS-SSC, or any other site-specific modifications based on new information. An updated CEUS-SSC seismicity catalog was developed for the whole CEUS-SSC Study Region for the period of January 1, 2009 through April 30, 2016 for the region encompassed by the 250-mile (400-km) radius around the BFN site. The final seismicity catalog used for the BFN Probabilistic Seismic Hazard Analysis (PSHA) is the combination of the original CEUS-SSC seismicity catalog (1568 through 2008) and the updated BFN site regional catalog (January 1, 2009 through April 30, 2016). After the review and study of new information, it was concluded that the CEUS-SSC recurrence parameters did not require an update.
The PSHA performed incorporated the entire CEUS-SSC logic tree published in NUREG-2115 with its revisions published in 2015 [12]. The only simplification performed to the entire CEUS-SSC was related to using point sources for the background sources. No seismic sources were screened out of the analyses. The use of point sources for modeling the background sources is supported by the sensitivities presented in NUREG-2115.
3.1.3.2 Horizontal and Vertical GMRS This section provides the control point horizontal and vertical GMRS.
The GMRS at the control point is provided in Table 3.1-3 and plotted in Figure 3.1-9.
The insights are summarized in Section 3.1.1 and further described in detail in the BFN PSHA report [11].
3.1.3.2.1 Vertical GMRS Vertical ground motions were developed by applying V/H ratios to the horizontal GMRS and FIRS. For GMRS/FIRS1 and FIRS3, these FIRS are very close to the CEUS hard reference rock with the average time-weighted shear-wave velocity in the top 100 ft below the ground surface being 8,462 ft/s. As a result, the CEUS V/H scaling relation in NUREG/CR-6728 [14] was used with no modifications. In addition to the V/H scaling relations developed above for development of the vertical GMRS and FIRS, V/H scaling relations were also developed at annual frequency of exceedance of 10-5. As stated above, for GMRS/FIRS1 and FIRS3, these FIRS are very close to CEUS hard rock with an average time-weighted shear wave velocity in the top 100 ft below the ground surface of 8,462 ft/s. As a result, the CEUS Vertical over Horizontal (V/H) scaling relation in NUREG/CR-6728 (USNRC, 2001) was used with no changes (corresponding to PGA between 0.2g and 0.5g for the GMRS level where the PGA was equal to 0.254g). At annual frequency of exceedance of 10-5, the same V/H scaling relation could be used for GMRS/FIRS1 and FIRS3 in spite of the 10-5 hazard level PGA being 0.5155g.
For FIRS2 and FIRS4, the average time-weighted shear-wave velocity in the top 100 ft below those control points was equal to 2,304 ft/s and 1,613 ft/s, respectively. For these two FIRS, in the absence of CEUS V/H scaling relations appropriate for the FIRS, a logic tree was adopted to incorporate epistemic uncertainty by weighting alternative Page 24 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                          December 2019 models consistent the methodology described in EPRI 3002004396 [15]. The development of the V/H ratios is documented in the BFN PSHA report.
Table 3.1-3 summarizes the horizontal and vertical response spectra at the control point. Figure 3.1-9 provides a plot of the vertical and horizontal GMRS as well as V/H ratios.
Page 25 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 Table 3.1-3 Smoothed Horizontal and Vertical GMRS/FIRS1/FIRS3 and V/H Ratios Horizontal                 Vertical Frequency GMRS/FIRS1/FIRS3           GMRS/FIRS1/FIRS3       V/H Ratio (Hz)
(g)                       (g) 0.1                  1.64E-02                  1.23E-02             0.750 0.125                  2.06E-02                  1.54E-02             0.750 0.15                   2.49E-02                  1.87E-02             0.750 0.2                  3.38E-02                  2.53E-02             0.750 0.3                  5.30E-02                  3.97E-02             0.750 0.4                  7.28E-02                  5.46E-02             0.750 0.5                  8.94E-02                  6.71E-02             0.750 0.6                  9.98E-02                  7.49E-02             0.750 0.7                  1.06E-01                  7.98E-02             0.750 0.8                  1.11E-01                  8.36E-02             0.750 0.9                  1.17E-01                  8.74E-02             0.750 1                   1.26E-01                  9.41E-02             0.750 1.25                   1.55E-01                  1.16E-01             0.750 1.5                  1.85E-01                  1.38E-01             0.750 2                   2.36E-01                  1.77E-01             0.750 2.5                  2.63E-01                  1.97E-01             0.750 3                   2.82E-01                  2.12E-01             0.750 4                   3.21E-01                  2.40E-01             0.750 5                   3.54E-01                  2.65E-01             0.750 6                   3.89E-01                  2.92E-01             0.750 7                   4.22E-01                  3.17E-01             0.750 8                   4.53E-01                  3.40E-01             0.750 9                   4.80E-01                  3.60E-01             0.750 10                   5.01E-01                  3.75E-01             0.750 12.5                   5.28E-01                  4.07E-01             0.771 15                   5.34E-01                  4.21E-01             0.788 20                   5.08E-01                  4.20E-01             0.826 25                   4.67E-01                  4.11E-01             0.880 30                   4.27E-01                  4.00E-01             0.937 35                   3.91E-01                  3.83E-01             0.981 40                   3.64E-01                  3.80E-01             1.042 45                   3.50E-01                  3.86E-01             1.102 50                   3.43E-01                  3.86E-01             1.124 60                   3.29E-01                  3.75E-01             1.137 70                   3.01E-01                  3.40E-01             1.128 80                   2.75E-01                  2.99E-01             1.090 90                   2.60E-01                  2.70E-01             1.038 100                   2.54E-01                  2.54E-01             1.000 Page 26 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Figure 3.1-9: Horizontal and Vertical GMRS/FIRS1/FIRS3 and V/H Ratio 3.2 Comparison of NTTF 2.1 Seismic Hazard Submittal and PRA Supplemental Seismic Hazard Analysis The BFN SPRA used the supplemental seismic hazard analysis documented in the BFN PSHA report [11]. Table 3.1-3 and Figure 3.1-9 provide the vertical and horizontal GMRS.
A site-specific surface geophysics program encompassing BFN was completed to better define the shear-wave velocities of the BFN units. The existing geotechnical information available at the project site was used to characterize the depth of the various units, e.g.,
Fort Payne formation, and the shear-wave velocities from the geophysics were then assigned to their corresponding units, since the geophysics surveys profiles were acquired at the perimeter of the BFN site.
Figures 3.2-1 to 3.2-3 compare the NTTF 2.1 Seismic Hazard submittal [3], assessed by the NRC staff, with the SPRA Supplemental Seismic Hazard Analysis.
Figures 3.2-1 and 3.2-2 show the idealized site profiles developed. The key difference between the base profile developed in the current study and the NTTF 2.1 Submittal study is that the profile developed in the current study is stiffer (faster) except for the top 25 ft, where the profile developed is slightly softer (slower) compared to the NTTF 2.1 Submittal. Due to the uniformity observed from the onshore geophysics program Page 27 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 encompassing BFN, a single branch was used in the SPRA Supplemental Seismic Hazard Analysis. Another difference included the addition of a softer layer at approximate EL 435 ft, re: MSL with a significant impedance reversal. Sensitivity analyses were performed and documented in the BFN PSHA report, where this softer layer was removed, and it showed an insignificant impact on the final PSHA results.
Figure 3.2-3 compares the NTTF 2.1 Submittal GMRS and the current GMRS/FIRS1/FIRS3. Overall, the shapes of the spectra are comparable. At the lower frequencies, there are insignificant differences. At higher frequencies, GMRS/FIRS1/FIRS3 is lower than or equal to the NTTF 2.1 Seismic Hazard Submittal.
The small differences between the two spectra could be attributed to a number of factors. Excluding factors like the use of different software or variations in the randomization algorithms for the base-case profiles, the total uncertainty (combined epistemic uncertainty and aleatory variability) are lower in the SPRA Supplemental Seismic Hazard Analysis compared to the NTTF 2.1 Submittal. This is due to the site-specific measurements that were completed at BFN, which confirmed the relative uniformity of the site. Another difference includes the slightly softer (slower) subsurface conditions in the top 25 ft on average. Reduction of the total sigmas would be expected to reduce the mean hazard directly even if the profiles were identical. The slightly softer (slower) subsurface conditions in the top 25 ft would be expected to slightly shift the spectrum towards the lower frequency, which is what is observed in Figure 3.2-3, where the spectral peak is at a slightly lower frequency (approximately 15 Hz as opposed to 25 Hz).
Page 28 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                   December 2019 Figure 3.2-1: Comparison of Base-Case Soil Profiles NTTF 2.1 Seismic Hazard Submittal and SPRA Supplemental Seismic Hazard Analysis Page 29 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                   December 2019 Figure 3.2-2: Comparison of Base-Case Soil Profiles NTTF 2.1 Seismic Hazard Submittal in Top 500 ft and SPRA Supplemental Seismic Hazard Analysis Page 30 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Figure 3.2-3: Comparison of Horizontal GMRS/FIRS1/FIRS3 NTTF 2.1 Seismic Hazard Submittal and Seismic PRA Supplemental Seismic Hazard Analysis 3.3   Soil Failure and Fragility Analysis The SPRA soil failure and fragility analysis is performed in the report CJC-BFN-C-001
[17]. Soil failure modes considered in the analysis include liquefaction, seismic-induced settlements, seismic-induced lateral deformation, slope stability, sliding of earth and building structures, and seismic bearing capacity. The evaluations performed and described in this report followed an overall graded approach for developing soil failure mode fragilities for inputs into the BFN SPRA model. The graded approach uses increasing levels of rigor for screening out or estimating soil fragilities depending upon the contribution to risk of a given soil failure mode.
Seismic-induced soil failure primarily results in lateral and/or vertical displacements and increased lateral pressure on building walls for embedded structures founded on rock at the BFN site. CJC-BFN-C-001 provides estimates of vertical and lateral deformations due to soil failures for earthquakes ranging from AFE of 1E-4 to 1E-7.
Page 31 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                          December 2019 The Category I structures at BFN that may be susceptible to damage as a result of ground motions due to earthquakes were identified and either screened out, evaluated using scaling of results of existing analyses, or analyzed to develop estimates of deformation and behavior, as shown in Table 3.3-1. These analyses were based on geotechnical data available at the site using contemporary methodologies to estimate slope stability, vertical settlement, and lateral deformation, as appropriate.
The effect of soil deformation on buried piping strain levels are developed in report CJC-BFN-C-001 [17]. Piping fragility is determined based on the strain levels induced in the embedded piping due to localized and distributed soil deformations. The localized deformation case resulted in the largest strain demands in the piping. Strain levels are computed for ground motion levels corresponding to AFE 1E-4 to 1E-7.
The ground motion levels and associated SAFs for the analysis are taken from the BFN PSHA report [11].
Table 3.3-1: BFN Soil Failure and Fragility Analysis Geotechnical Structure                       Foundation                     Evaluation Material Reactor Building and Steel Containment         Shale/limestone                 Screened out1 Vessel                           bedrock Reinforced Concrete Chimney                    Bedrock                   Screened out Intake Pumping Station (IPS), including              Bedrock                   Screened out1 the Residual Heat Removal Service Water (RHRSW) Intake Structure Gate Structure No. 2 and 3             Shale/limestone                 Screened out bedrock Diesel Generator Building (DGB)           Compacted earth/                 Screened out crushed rock fill Standby Gas Treatment Building                 Compacted                   Screened out earth backfill Off-gas Treatment Building                   Bedrock                   Screened out Equipment Access Lock                    Compacted              Vertical Settlement and earth backfill            Lateral Deformation Vacuum Pipe Building                   Compacted              Vertical Settlement and earth                Lateral Deformation backfill Condensate Storage Tanks (CST)                  Compacted              Vertical Settlement and earth backfill            Lateral Deformation Retaining Wall Between DGB &                  Compacted                   Screened out Radwaste Building                   earth backfill Intake Channel Slope                   In-situ soils               Screened out Earth Berm south of Reactor Building             Compacted              Vertical Settlement and backfill and             Lateral Deformation in-situ soils North Bank of Cool Water Channel                In-situ soils          Vertical Settlement and Lateral Deformation Cool Water Discharge Dike                  In-situ soils               Screened out and Hydro fill 1Lateral soil pressures calculated Page 32 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                            December 2019 3.3.1 Soil Failure and Fragility Analysis Technical Adequacy The BFN Soil Failure and Fragility Analysis methodology and analysis was subjected to an independent peer review against the pertinent requirements in the PRA Standard.
After completion of the subsequent independent assessment, the full set of SRs was met. The seismic hazard analysis was determined to be acceptable for use in the SPRA.
The peer review assessment, and subsequent disposition of peer review findings through an independent assessment [6] are further described in Appendix A [16].
4.0 Determination of Seismic Fragilities for the SPRA This section provides a summary of the process for identifying and developing fragilities for SSCs that participate in the plant response to a seismic event for the BFN SPRA.
The subsections provide brief summaries of these elements.
4.1   Seismic Equipment List For the BFN SPRA, a seismic equipment list (SEL) was developed to include SSCs that are important to achieving safe shutdown following a seismic event and to mitigating radioactivity release if core damage occurs, and that are included in the SPRA model.
The methodology used to develop the SEL is consistent with the guidance provided in EPRI 3002000709 [18].
4.1.1 SEL Development The comprehensive SEL was developed by starting with the list of components modeled in the BFN IEPRA, including internal flooding. That list was then augmented by reviewing equipment contained in the BFN individual plant examination of external events (IPEEE), fire safe shutdown equipment lists (SSELs), and the NTTF 2.3 seismic walkdown equipment list. Diverse and flexible coping strategies (FLEX) systems included in the model were added to the SEL. Table 4.1-1 includes a list of systems considered in the SEL development. In addition, a separate effort was conducted by the Human Reliability Analysis (HRA) analyst to identify instrumentation needed by operators to support actions modeled in the IEPRA. Components typically not modeled in IEPRAs, such as cable trays; conduits; motor control centers (MCCs); electrical cabinets and panels; heating, ventilating, and air conditioning (HVAC) ducting; and piping, were identified and included in the SEL. The SEL was also updated after the seismic walkdowns to incorporate additional items such as block walls. The SEL includes structures, buildings, and substructures that either contain safety-related equipment or whose failure could impact safety functions or cause a reactor trip. The SEL includes nuclear steam supply system (NSSS) components and components required for containment integrity.
The resulting SEL includes a total of about 6,800 component entries for all three units combined (including common components). The final SEL was documented in the BFN SPRA Seismic Equipment List [19].
Page 33 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                       December 2019 Table 4.1-1: List of Systems Considered in the SEL Development System                                                 Mitigation                        In 2.3                    Containment System                                 In IE PRA     In IPEEE           In SEL Number                                                 Potential                         SWEL                       Isolation 001                   Main Steam                        Yes         Yes          Yes    Yes        Yes               Yes 002                   Condensate                         No         Yes           No     No        Yes               No 003                Feedwater System                      No         Yes          Yes    Yes        Yes               Yes 004           Hydrogen Water Chemistry                   No         No            No     No        No                No 005                 Extraction Steam                     No         No            No     No        No                No 006              Heater Drains & Vents                   No         No            No     No        No                No 008     Turbine Drains & Miscellaneous Piping            No         No            No     No        No                No 010            Rx Vessel Vents & Drains                  No         No           Yes    Yes        Yes               No 012                  Auxiliary Boiler                    No         No            No     No        No                No 018                      Fuel Oil                       Yes         Yes          Yes     No        Yes               No 020                  Lubricating Oil                     No         No            No     No        No                No 023  Residual Heat Removal Service Water System         Yes         Yes          Yes    Yes        Yes               No 024                Raw Cooling Water                     No         Yes           No     No        Yes               No 025                Raw Service Water                     No         No            No     No        No                No 026          High Pressure Fire Protection               No         Yes           No     No        Yes               No 027          Condenser Circulating Water                 No         Yes           No     No        Yes               No 030                Normal Ventilation                   Yes         Yes          Yes    Yes        Yes               No 031                 CREV/Ventilation                     No         No            No     No        No                No 031                      Chillers                       Yes         Yes           No    Yes        Yes               No 032                    Control Air                      Yes         Yes          Yes    Yes        Yes               No 033                    Service Air                      Yes         Yes           No     No        Yes               No 034                 Vacuum Priming                       No         No            No     No        No                No 035          Generator Hydrogen Cooling                  No         No            No     No        No                No 036       Auxiliary Boiler FW SEC Treatment              No         No            No     No        No                No 037                Gland Seal Water                      No         No            No     No        No                No 040                 Station Drainage                     No         No            No     No        No                No 043                  Chemistry Heat                      No         No           Yes     No        Yes               No 044                   Building Heat                      No         Yes           No     No        Yes               No 046                Feedwater Control                    Yes         Yes           No     No        No                No 047                   EHC Control                       Yes         Yes           No     No        Yes               No 049                   Breathing Air                      No         No            No     No        No                No 050               Sodium Hypochlorite                    No         No            No     No        No                No 051             Raw Water Chlorination                   No         No            No     No        No                No 052                Seismic Monitoring                    No         No            No     No        No                No 053           Demineralizer Backwash Air                 No         No            No     No        No                No 055                   Annunciators                       No         No            No     No        No                No 056             Temperature Monitoring                   No         No            No     No        No                No 063              Standby Liquid Control                 Yes         Yes          Yes    Yes        Yes               Yes 064A               Primary Containment                   Yes         Yes          Yes    Yes        Yes               Yes 064B                Containment Purge                     No         No           Yes    Yes        Yes               Yes 064C             Secondary Containment                   Yes         Yes          Yes    Yes        Yes               Yes 064D     Primary Containment Isolation System            Yes         Yes          Yes    Yes        Yes               Yes Page 34 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                     December 2019 Table 4.1-1: List of Systems Considered in the SEL Development System                                                Mitigation                        In 2.3                    Containment System                                In IE PRA     In IPEEE           In SEL Number                                                Potential                         SWEL                       Isolation 065                       SGT                          Yes         Yes           No     No        Yes               No 066                      Off-Gas                        No         No            No     No        Yes               No 067      Emergency Equipment Cooling Water             Yes         Yes          Yes    Yes        Yes               No 068                   Recirculation                     No         Yes           No     No        Yes               Yes 069             Reactor Water Clean Up                 Yes         Yes          Yes    Yes        Yes               Yes 070     Reactor Building Closed Cooling Water           No         Yes          Yes    Yes        Yes               Yes 071          Reactor Core Isolation Cooling            Yes         Yes          Yes    Yes        Yes               Yes 072          Auxiliary Decay Heat Removal               No         No            No     No        No                No 073         High Pressure Coolant Injection            Yes         Yes          Yes    Yes        Yes               Yes 074              Residual Heat Removal                 Yes         Yes          Yes    Yes        Yes               Yes 075                Core Spray System                   Yes         Yes          Yes    Yes        Yes               Yes 076            Containment Air Monitoring               No         No           Yes    Yes        Yes               Yes 077                     Rad waste                       No         Yes          Yes    Yes        Yes               Yes 078                 Fuel Pool Cooling                   No         No            No    Yes        Yes               No 079                    Refuel Tools                     No         No            No     No        No                No 080  Primary Containment/Temperature Monitoring         No         No            No     No        No                No 082          Emergency Diesel Generators               Yes         Yes          Yes    Yes        Yes               No 084                       CAD                          Yes         Yes          Yes    Yes        Yes               Yes 085                 Control Rod Drive                  Yes         Yes          Yes    Yes        Yes               Yes 086                  D/G Starting Air                  Yes         Yes          Yes    Yes        Yes               No 090               Radiation Monitoring                  No         No            No     No        No                No 092                Neutron Monitoring                   No         No           Yes    Yes        Yes               Yes 094                    TIP System                       No         No            No     No        No                Yes 096        Reactor Recirculation Flow Control           No         No            No     No        Yes               No 099            Reactor Protection System               Yes         Yes          Yes    Yes        Yes               No 111            Cranes (Reactor & Turbine)               No         No            No     No        Yes               No 202                 4-kV Unit Boards                    No         Yes           No     No        Yes               No 203              4-kV Common Boards                     No         Yes           No     No        Yes               No 204           4-kV Unit Start Board & Bus               No         Yes           No     No        Yes               No 205         4-kV Cooling Tower Switch Gear              No         No            No     No        No                No 206             4-kV Bio-Thermal Board                  No         No            No     No        No                No 210                4-kV Bus Tie Board                  Yes         Yes           No     No        Yes               No 211        4-kV Shutdown Board and Buses               Yes         Yes          Yes    Yes        Yes               Yes 215              480-V Common Board                    Yes         Yes           No     No        Yes               No 219             480-V Diesel Aux Board                 Yes         Yes           No    Yes        Yes               No 225                480-V Unit Boards                    No         Yes           No     No        Yes               No 231             480-V Shutdown Boards                  Yes         Yes          Yes    Yes        Yes               Yes 232           480-V Cooling Tower Boards                No         No            No     No        No                No 233             480-V Biothermal Boards                 No         No            No     No        No                No 236                Main Transformers                   Yes         Yes           No     No        Yes               No 237       480-V Service Building Main Board             No         No            No     No        No                No 238       480-V Transformer Yard Distribution           No         Yes           No     No        Yes               No 239              480-V Lighting Boards                  No         No            No     No        No                No 240            480-V Water Supply Board                 No         Yes           No     No        Yes               No Page 35 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                       December 2019 Table 4.1-1: List of Systems Considered in the SEL Development System                                                    Mitigation                        In 2.3                    Containment System                                 In IE PRA     In IPEEE           In SEL Number                                                    Potential                         SWEL                       Isolation 241                   161-kV Switchyard                     No         Yes           No     No        No                No 242                     Main Generator                      No         Yes           No     No        Yes               No 243           Unit Station Service Transformer              No         Yes           No     No        Yes               No 244                    Communications                       No         No            No     No        No                No 245       Common Station Service Transformer                No         Yes           No     No        Yes               No 246             Cooling Tower Transformer                   No         No            No     No        No                No 248                    250V DC System                      Yes         Yes          Yes    Yes        Yes               Yes 249               Plant Preferred 120V AC                  Yes         Yes           No     No        No                No 250            Plant Non-Preferred 120V AC                  No         Yes           No     No        No                No 251                48-V DC Power System                     No         No            No     No        No                No 252                Unit Preferred 120V AC                  Yes         Yes           No     No        Yes               Yes 253              120-V AC Inst & Ctrl PWR                  Yes         Yes          Yes    Yes        Yes               No 254                  Diesel 125-V DC Sys                   Yes         Yes          Yes    Yes        Yes               Yes 255                      Data Logger                        No         No            No     No        No                No 256                        Inverters                       Yes         Yes          Yes    Yes        Yes               No 258                  Operation Recorder                     No         No            No     No        No                No 259             480-V Load Shedding Logic                   No         No            No     No        No                No 260                        Security                         No         No            No     No        No                No 261                   Process Computer                      No         No            No     No        No                No 262         Generator Bus Duct Cooling System               No         No            No     No        No                No 265     480-V Reactor Building Ventilation Boards           No         No           Yes    Yes        Yes               No 266            480-V Control Bay Vent Board                 No         Yes           No     No        Yes               No 268             480-V Reactor MOV Boards                   Yes         Yes           No    Yes        Yes               No 269         480-V Turbine Building MOV Board                No         Yes           No     No        Yes               No 270          480-V Condensate Demineralizer                 No         No            No     No        No                No 271               480-V Aux Boiler Boards                   No         No            No     No        No                No 272              480-V Water & Oil Storage                  No         No            No     No        No                No 273               480-V Radwaste Boards                     No         No            No     No        No                No 274             480-V Service Building Vent                 No         No            No     No        No                No 275          480-V Office Building Vent Board               No         No            No     No        No                No 276                 480-V Power Cabinets                    No         No            No     No        No                No 277               Gatehouse Panel Board                     No         No            No     No        No                No 278                     Dist. Cabinets                      No         Yes           No     No        Yes               No 280                     Battery Boards                     Yes         Yes          Yes    Yes        Yes               No 281             250-V Reactor MOV Boards                   Yes         Yes          Yes    Yes        Yes               No 282            250-V DC Distribution Boards                Yes         Yes           No     No        Yes               No 283                24-V DC Power System                    Yes         Yes          Yes    Yes        Yes               No 284                  480-V Power Outlets                    No         No            No     No        No                No 285          Computer Universal Power Supply                No         No            No     No        Yes               No 360                          FLEX                          Yes         No            No     No        Yes               No Various               Miscellaneous Panels                   Yes         Yes          Yes    Yes        Yes               No Various             Miscellaneous Structures                 Yes         No            No     No        Yes               Yes Various Additional potential SIFF fire & flooding sources     No         Yes          Yes    Yes        Yes               Yes Page 36 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 4.1.2 Relay and Breaker Evaluation During a seismic event, vibratory ground motion can cause relays and breakers to chatter. The chattering of relays potentially can result in spurious signals to equipment.
The chattering of breakers potentially can result in equipment either losing power or starting when it is not desired. Relay/breaker chatter can be acceptable (does not impact the associated equipment), self-correcting, or recovered by operator action. An extensive relay/breaker chatter evaluation was performed for all three BFN units [20] in accordance with SPID [2], Section 6.4.2, and PRA Standard, Section 5-2.2. The evaluation resulted in many relay/breaker chatter scenarios screened out from further evaluation based on no impact to component function. The 407 relays that were not screened out are listed in Table 4.1-2.
Page 37 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-1-FCV-071-0002, RCIC BFN-1-FCV-071-0002          STEAM LINE INBD               13A-K33            BFN-1-RLY-071-13A-K33   12HFA51A1F    SEIS_14-1R1-1 ISOLATION VLV BFN-1-FCV-071-0002, RCIC BFN-1-FCV-071-0002          STEAM LINE INBD               13A-K32            BFN-1-RLY-071-13A-K32  12HFA151A1F    SEIS_14-1R1-2 ISOLATION VLV BFN-1-FCV-071-0002, RCIC BFN-1-FCV-071-0002          STEAM LINE INBD               13A-K12            BFN-1-RLY-071-13A-K12  12HFA51A41F    SEIS_14-1R2-2 ISOLATION VLV BFN-1-FCV-071-0002, RCIC BFN-1-FCV-071-0002          STEAM LINE INBD               13A-K13            BFN-1-RLY-071-13A-K13  12HGA11A51F    SEIS_14-1R3-2 ISOLATION VLV BFN-1-FCV-071-0002, RCIC BFN-1-FCV-071-0002          STEAM LINE INBD               13A-K38            BFN-1-RLY-071-13A-K38  12HGA11A51F    SEIS_14-1R3-2 ISOLATION VLV BFN-1-FCV-071-0002, RCIC BFN-1-FCV-071-0002          STEAM LINE INBD               13A-K15            BFN-1-RLY-071-13A-K15  12HFA51A41F    SEIS_14-1R3-2 ISOLATION VLV BFN-2-FCV-071-0002, RCIC BFN-2-FCV-071-0002          STEAM LINE INBD               13A-K33            BFN-2-RLY-071-13A-K33   12HFA51A1F    SEIS_14-1R1-1 ISOLATION VLV BFN-2-FCV-071-0002, RCIC BFN-2-FCV-071-0002          STEAM LINE INBD               13A-K32            BFN-2-RLY-071-13A-K32   12HFA51A1F    SEIS_14-1R1-2 ISOLATION VLV BFN-2-FCV-071-0002, RCIC BFN-2-FCV-071-0002          STEAM LINE INBD               13A-K12            BFN-2-RLY-071-13A-K12  12HFA51A41F    SEIS_14-1R2-2 ISOLATION VLV Page 38 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-2-FCV-071-0002, RCIC BFN-2-FCV-071-0002          STEAM LINE INBD               13A-K13            BFN-2-RLY-071-13A-K13  12HGA11A51F    SEIS_14-1R3-2 ISOLATION VLV BFN-2-FCV-071-0002, RCIC BFN-2-FCV-071-0002          STEAM LINE INBD               13A-K38            BFN-2-RLY-071-13A-K38  12HGA11A51F    SEIS_14-1R3-2 ISOLATION VLV BFN-2-FCV-071-0002, RCIC BFN-2-FCV-071-0002          STEAM LINE INBD               13A-K15            BFN-2-RLY-071-13A-K15  12HFA51A41F     SEIS_15-1R1 ISOLATION VLV BFN-3-FCV-071-0002, RCIC BFN-3-FCV-071-0002          STEAM LINE INBD               13A-K33            BFN-3-RLY-071-13A-K33   12HFA51A1F    SEIS_14-1R1-1 ISOLATION VLV BFN-3-FCV-071-0002, RCIC BFN-3-FCV-071-0002          STEAM LINE INBD               13A-K32            BFN-3-RLY-071-13A-K32  12HFA151A1F    SEIS_14-1R1-2 ISOLATION VLV BFN-3-FCV-071-0002, RCIC BFN-3-FCV-071-0002          STEAM LINE INBD               13A-K12            BFN-3-RLY-071-13A-K12  12HFA51A41F    SEIS_14-1R2-2 ISOLATION VLV BFN-3-FCV-071-0002, RCIC BFN-3-FCV-071-0002          STEAM LINE INBD               13A-K13            BFN-3-RLY-071-13A-K13  12HGA11A51F    SEIS_14-1R3-2 ISOLATION VLV BFN-3-FCV-071-0002, RCIC BFN-3-FCV-071-0002          STEAM LINE INBD               13A-K38            BFN-3-RLY-071-13A-K38  12HGA11A51F    SEIS_14-1R3-2 ISOLATION VLV BFN-3-FCV-071-0002, RCIC BFN-3-FCV-071-0002          STEAM LINE INBD               13A-K15            BFN-3-RLY-071-13A-K15  12HFA51A41F     SEIS_15-1R1 ISOLATION VLV Page 39 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-1-FCV-071-0003, RCIC BFN-1-FCV-071-0003        STM LINE OUTBD ISOL             13A-K16            BFN-1-RLY-071-13A-K16  12HFA51A41F    SEIS_14-1R2-2 FLOW CONT VLV BFN-1-FCV-071-0003, RCIC BFN-1-FCV-071-0003        STM LINE OUTBD ISOL             13A-K32            BFN-1-RLY-071-13A-K32  12HFA151A1F    SEIS_14-1R1-2 FLOW CONT VLV BFN-1-FCV-071-0003, RCIC BFN-1-FCV-071-0003        STM LINE OUTBD ISOL             13A-K12            BFN-1-RLY-071-13A-K12  12HFA51A41F    SEIS_14-1R2-2 FLOW CONT VLV BFN-1-FCV-071-0003, RCIC BFN-1-FCV-071-0003        STM LINE OUTBD ISOL             13A-K13            BFN-1-RLY-071-13A-K13  12HGA11A51F    SEIS_14-1R3-2 FLOW CONT VLV BFN-1-FCV-071-0003, RCIC BFN-1-FCV-071-0003        STM LINE OUTBD ISOL             13A-K38            BFN-1-RLY-071-13A-K38  12HGA11A51F    SEIS_14-1R3-2 FLOW CONT VLV BFN-1-FCV-071-0003, RCIC BFN-1-FCV-071-0003        STM LINE OUTBD ISOL             13A-K15            BFN-1-RLY-071-13A-K15  12HFA51A41F    SEIS_14-1R3-2 FLOW CONT VLV BFN-1-FCV-071-0003, RCIC BFN-1-FCV-071-0003        STM LINE OUTBD ISOL             13A-K33            BFN-1-RLY-071-13A-K33   12HFA51A1F    SEIS_14-1R1-1 FLOW CONT VLV BFN-2-FCV-071-0003, RCIC BFN-2-FCV-071-0003         STEAM LINE OUTBD               13A-K16            BFN-2-RLY-071-13A-K16  12HFA51A41F    SEIS_14-1R2-2 ISOLATION VLV BFN-2-FCV-071-0003, RCIC BFN-2-FCV-071-0003         STEAM LINE OUTBD               13A-K32            BFN-2-RLY-071-13A-K32  12HFA151A1F    SEIS_14-1R1-2 ISOLATION VLV Page 40 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-2-FCV-071-0003, RCIC BFN-2-FCV-071-0003         STEAM LINE OUTBD               13A-K12            BFN-2-RLY-071-13A-K12  12HFA51A41F    SEIS_14-1R2-2 ISOLATION VLV BFN-2-FCV-071-0003, RCIC BFN-2-FCV-071-0003         STEAM LINE OUTBD               13A-K13            BFN-2-RLY-071-13A-K13  12HGA11A51F    SEIS_14-1R3-2 ISOLATION VLV BFN-2-FCV-071-0003, RCIC BFN-2-FCV-071-0003         STEAM LINE OUTBD               13A-K38            BFN-2-RLY-071-13A-K38  12HGA11A51F    SEIS_14-1R3-2 ISOLATION VLV BFN-2-FCV-071-0003, RCIC BFN-2-FCV-071-0003         STEAM LINE OUTBD               13A-K15            BFN-2-RLY-071-13A-K15  12HFA51A41F     SEIS_15-1R1 ISOLATION VLV BFN-2-FCV-071-0003, RCIC BFN-2-FCV-071-0003         STEAM LINE OUTBD               13A-K33            BFN-2-RLY-071-13A-K33   12HFA51A1F    SEIS_14-1R1-1 ISOLATION VLV BFN-3-FCV-071-0003, RCIC BFN-3-FCV-071-0003         STEAM LINE OUTBD               13A-K16            BFN-3-RLY-071-13A-K16  12HFA51A41F    SEIS_14-1R2-2 ISOLATION VLV BFN-3-FCV-071-0003, RCIC BFN-3-FCV-071-0003         STEAM LINE OUTBD               13A-K32            BFN-3-RLY-071-13A-K32  12HFA151A1F    SEIS_14-1R1-2 ISOLATION VLV BFN-3-FCV-071-0003, RCIC BFN-3-FCV-071-0003         STEAM LINE OUTBD               13A-K12            BFN-3-RLY-071-13A-K12  12HFA51A41F    SEIS_14-1R2-2 ISOLATION VLV BFN-3-FCV-071-0003, RCIC BFN-3-FCV-071-0003         STEAM LINE OUTBD               13A-K13            BFN-3-RLY-071-13A-K13  12HGA11A51F    SEIS_14-1R3-2 ISOLATION VLV Page 41 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-3-FCV-071-0003, RCIC BFN-3-FCV-071-0003          STEAM LINE OUTBD              13A-K38            BFN-3-RLY-071-13A-K38  12HGA11A51F    SEIS_14-1R3-2 ISOLATION VLV BFN-3-FCV-071-0003, RCIC BFN-3-FCV-071-0003          STEAM LINE OUTBD              13A-K15            BFN-3-RLY-071-13A-K15  12HFA51A41F     SEIS_15-1R1 ISOLATION VLV BFN-3-FCV-071-0003, RCIC BFN-3-FCV-071-0003          STEAM LINE OUTBD              13A-K33            BFN-3-RLY-071-13A-K33   12HFA51A1F    SEIS_14-1R1-1 ISOLATION VLV BFN-1-XX-071-0009, RCIC TURB STOP VLV TRIP BFN-1-XX-071-0009                                         13A-K8            BFN-1-RLY-071-13A-K8   12HFA51A41F     SEIS_15-1R1 SOLENOID (SPECIAL POWER SUPPLY)
BFN-1-XX-071-0009, RCIC TURB STOP VLV TRIP BFN-1-XX-071-0009                                        13A-K15            BFN-1-RLY-071-13A-K15  12HFA51A41F    SEIS_14-1R3-2 SOLENOID (SPECIAL POWER SUPPLY)
BFN-1-XX-071-0009, RCIC TURB STOP VLV TRIP BFN-1-XX-071-0009                                        13A-K39            BFN-1-RLY-071-13A-K39  12HFA51A41F     SEIS_15-1R1 SOLENOID (SPECIAL POWER SUPPLY)
BFN-1-XX-071-0009, RCIC TURB STOP VLV TRIP BFN-1-XX-071-0009                                         13A-K6            BFN-1-RLY-071-13A-K6   12HGA11A51F     SEIS_15-1R2 SOLENOID (SPECIAL POWER SUPPLY)
BFN-1-XX-071-0009, RCIC TURB STOP VLV TRIP BFN-1-XX-071-0009                                         13A-K7            BFN-1-RLY-071-13A-K7   12HGA11A51F     SEIS_15-1R2 SOLENOID (SPECIAL POWER SUPPLY)
Page 42 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                                 Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                  Disposition BFN-1-XX-071-0009, RCIC TURB STOP VLV TRIP BFN-1-XX-071-0009                                        1-71-9C             BFN-1-RLY-071-009C    Agastat GPFNR     SEIS_15-1R2 SOLENOID (SPECIAL POWER SUPPLY)
BFN-1-XX-071-0009, RCIC TURB STOP VLV TRIP                                                       Mercoid DS-7043-BFN-1-XX-071-0009                                       PS-71-13A            BFN-1-PS-071-0013A                     SEIS_15-1R1-1 SOLENOID (SPECIAL                                                              804 POWER SUPPLY)
BFN-1-XX-071-0009, RCIC TURB STOP VLV TRIP                                                       Mercoid DS-7043-BFN-1-XX-071-0009                                       PS-71-13B            BFN-1-PS-071-0013B                     SEIS_15-1R1-1 SOLENOID (SPECIAL                                                              804 POWER SUPPLY)
BFN-1-XX-071-0009, RCIC Static-O-ring TURB STOP VLV TRIP BFN-1-XX-071-0009                                       PS-71-21A            BFN-1-PS-071-0021A   54N4-GG118-M4-     SEIS_15-1R2 SOLENOID (SPECIAL C1A-TT POWER SUPPLY)
BFN-2-XX-071-0009, RCIC BFN-2-XX-071-0009        TURB STOP VLV TRIP               13A-K8            BFN-2-RLY-071-13A-K8   12HFA51A41F       SEIS_15-1R1 SOLENOID BFN-2-XX-071-0009, RCIC BFN-2-XX-071-0009        TURB STOP VLV TRIP              13A-K15            BFN-2-RLY-071-13A-K15  12HFA51A41F       SEIS_15-1R1 SOLENOID BFN-2-XX-071-0009, RCIC BFN-2-XX-071-0009        TURB STOP VLV TRIP              13A-K39            BFN-2-RLY-071-13A-K39  12HFA51A41F       SEIS_15-1R1 SOLENOID BFN-2-XX-071-0009, RCIC BFN-2-XX-071-0009        TURB STOP VLV TRIP               13A-K6            BFN-2-RLY-071-13A-K6   12HGA11A51F       SEIS_15-1R2 SOLENOID BFN-2-XX-071-0009, RCIC BFN-2-XX-071-0009        TURB STOP VLV TRIP               13A-K7            BFN-2-RLY-071-13A-K7   12HGA11A51F       SEIS_15-1R2 SOLENOID Page 43 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                                 Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                  Disposition BFN-2-XX-071-0009, RCIC BFN-2-XX-071-0009        TURB STOP VLV TRIP              2-71-9C                   2-71-9C*          Square D        SEIS_15-1R2 SOLENOID BFN-2-XX-071-0009, RCIC Mercoid DS-7043-BFN-2-XX-071-0009        TURB STOP VLV TRIP             PS-71-13A            BFN-2-PS-071-0013A                     SEIS_15-1R1-1 804 SOLENOID BFN-2-XX-071-0009, RCIC Mercoid DS-7043-BFN-2-XX-071-0009        TURB STOP VLV TRIP             PS-71-13B            BFN-2-PS-071-0013B                     SEIS_15-1R1-1 804 SOLENOID BFN-2-XX-071-0009, RCIC                                                      Static-O-ring BFN-2-XX-071-0009        TURB STOP VLV TRIP             PS-71-21A            BFN-2-PS-071-0021A   54N4-GG118-M4-     SEIS_15-1R2 SOLENOID                                                              C1A-TT BFN-3-XX-071-0009, RCIC BFN-3-XX-071-0009        TURB STOP VLV TRIP               13A-K8            BFN-3-RLY-071-13A-K8   12HFA51A41F       SEIS_13-2R1 SOLENOID BFN-3-XX-071-0009, RCIC BFN-3-XX-071-0009        TURB STOP VLV TRIP              13A-K15            BFN-3-RLY-071-13A-K15  12HFA51A41F       SEIS_15-1R1 SOLENOID BFN-3-XX-071-0009, RCIC BFN-3-XX-071-0009        TURB STOP VLV TRIP              13A-K39            BFN-3-RLY-071-13A-K39  12HFA51A41F       SEIS_13-2R1 SOLENOID BFN-3-XX-071-0009, RCIC BFN-3-XX-071-0009        TURB STOP VLV TRIP               13A-K6            BFN-3-RLY-071-13A-K6   12HGA11A51F       SEIS_13-2R2 SOLENOID BFN-3-XX-071-0009, RCIC BFN-3-XX-071-0009        TURB STOP VLV TRIP               13A-K7            BFN-3-RLY-071-13A-K7   12HGA11A51F       SEIS_13-2R2 SOLENOID Page 44 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                                     Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                                      Disposition BFN-3-XX-071-0009, RCIC BFN-3-XX-071-0009        TURB STOP VLV TRIP              3-71-9C                   3-71-9C*           Square D           SEIS_13-2R2 SOLENOID BFN-3-XX-071-0009, RCIC Mercoid DS-7043-BFN-3-XX-071-0009        TURB STOP VLV TRIP             PS-71-13A            BFN-3-PS-071-0013A                         SEIS_13-2R1-1 804 SOLENOID BFN-3-XX-071-0009, RCIC Mercoid DS-7043-BFN-3-XX-071-0009        TURB STOP VLV TRIP             PS-71-13B            BFN-3-PS-071-0013B                         SEIS_13-2R1-1 804 SOLENOID BFN-3-XX-071-0009, RCIC                                                      Static-O-ring BFN-3-XX-071-0009        TURB STOP VLV TRIP             PS-71-21A            BFN-3-PS-071-0021A   54N4-GG118-M4-         SEIS_13-2R2 SOLENOID                                                                C1A-TT BFN-0-GEN-082-000A, BFN-0-GEN-082-000A                                         86GA              BFN-0-86-082-2547A/GA  12HEA61C238           SEIS_1C-1R5 DIESEL GENERATOR A Exciter Breaker BFN-0-GEN-082-000A,                                                       Shunt trip relay BFN-0-GEN-082-000A                                           41               BFN-0-41-082-000A/1                          SEIS_9-1R4 DIESEL GENERATOR A                                                        (coil is internal to the bkr)
BFN-0-GEN-082-000A, BFN-0-GEN-082-000A                                     87GA phase A         BFN-0-87G-082-2547A/AA GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR A BFN-0-GEN-082-000A, BFN-0-GEN-082-000A                                     87GA phase B        BFN-0-87G-082-2547A/AB  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR A BFN-0-GEN-082-000A, BFN-0-GEN-082-000A                                     87GA phase C        BFN-0-87G-082-2547A/AC  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR A Page 45 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                                     Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                      Disposition BFN-0-GEN-082-000A, BFN-0-GEN-082-000A                                        A/OTX               BFN-0-RLY-082-A/OTX   12HFA51A42F           SEIS_1C-1R2 DIESEL GENERATOR A BFN-0-GEN-082-000A,                                                          Square D BFN-0-GEN-082-000A                                          OTR               BFN-0-RLY-082-A/OTR                         SEIS_1C-1R1 DIESEL GENERATOR A                                                             DO/7001 BFN-0-GEN-082-000A, BFN-0-GEN-082-000A                                          CRA                BFN-0-RLY-082-CRA    12HFA51A42F            SEIS_9-1R2 DIESEL GENERATOR A BFN-0-GEN-082-000A, BFN-0-GEN-082-000A                                        GRRA                BFN-0-RLY-082-GRRA    12HFA51A42F            SEIS_9-1R2 DIESEL GENERATOR A BFN-0-GEN-082-000A, BFN-0-GEN-082-000A                                         GRLA               BFN-0-RLY-082-GRLA    12HFA51A42F            SEIS_9-1R2 DIESEL GENERATOR A BFN-0-GEN-082-000A, BFN-0-GEN-082-000A                                         VLRA               BFN-0-RLY-082-VLRA    12HFA51A42F           SEIS_1C-1R2 DIESEL GENERATOR A BFN-0-GEN-082-000A, BFN-0-GEN-082-000A                                         VRRA               BFN-0-RLY-082-VRRA    12HFA51A42F           SEIS_1C-1R2 DIESEL GENERATOR A BFN-0-GEN-082-000B, BFN-0-GEN-082-000B                                         86GB              BFN-0-86-082-2547B/GB  12HEA61C238           SEIS_1C-3R5 DIESEL GENERATOR B Exciter Breaker BFN-0-GEN-082-000B,                                                       Shunt trip relay BFN-0-GEN-082-000B                                           41                BFN-0-41-082-000B/1                         SEIS_9-1R4 DIESEL GENERATOR B                                                        (coil is internal to the bkr)
Page 46 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-0-GEN-082-000B, BFN-0-GEN-082-000B                                     87GB phase A        BFN-0-87G-082-2547B/BA GE 12CFD12B1A      SEIS_9-1R3 DIESEL GENERATOR B BFN-0-GEN-082-000B, BFN-0-GEN-082-000B                                     87GB phase B        BFN-0-87G-082-2547B/BB GE 12CFD12B1A      SEIS_9-1R3 DIESEL GENERATOR B BFN-0-GEN-082-000B, BFN-0-GEN-082-000B                                     87GB phase C        BFN-0-87G-082-2547B/BC GE 12CFD12B1A      SEIS_9-1R3 DIESEL GENERATOR B BFN-0-GEN-082-000B, BFN-0-GEN-082-000B                                        B/OTX              BFN-0-RLY-082-B/OTX   12HFA51A42F      SEIS_1C-3R2 DIESEL GENERATOR B Square D Class BFN-0-GEN-082-000B, BFN-0-GEN-082-000B                                         OTR               BFN-0-RLY-082-B/OTR  8501 Type XUD0-   SEIS_1C-3R1 DIESEL GENERATOR B 1200 BFN-0-GEN-082-000B, BFN-0-GEN-082-000B                                         CRB                BFN-0-RLY-082-CRB    12HFA51A42F       SEIS_9-1R2 DIESEL GENERATOR B BFN-0-GEN-082-000B, BFN-0-GEN-082-000B                                        GRRB               BFN-0-RLY-082-GRRB    12HFA51A42F       SEIS_9-1R2 DIESEL GENERATOR B BFN-0-GEN-082-000B, BFN-0-GEN-082-000B                                         GRLB              BFN-0-RLY-082-GRLB    12HFA51A42F       SEIS_9-1R2 DIESEL GENERATOR B BFN-0-GEN-082-000B, BFN-0-GEN-082-000B                                         VLRB              BFN-0-RLY-082-VLRB    12HFA51A42F      SEIS_1C-3R2 DIESEL GENERATOR B Page 47 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                                     Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                      Disposition BFN-0-GEN-082-000B, BFN-0-GEN-082-000B                                         VRRB               BFN-0-RLY-082-VRRB    12HFA51A42F           SEIS_1C-3R2 DIESEL GENERATOR B BFN-0-GEN-082-000C, BFN-0-GEN-082-000C                                         86GC              BFN-0-86-082-2547C/GC  12HEA61C238           SEIS_1C-6R5 DIESEL GENERATOR C Exciter Breaker BFN-0-GEN-082-000C,                                                       Shunt trip relay BFN-0-GEN-082-000C                                           41                BFN-0-41-082-000C/1                         SEIS_9-1R4 DIESEL GENERATOR C                                                        (coil is internal to the bkr)
BFN-0-GEN-082-000C, BFN-0-GEN-082-000C                                     87GC phase A        BFN-0-87G-082-2547C/CA  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR C BFN-0-GEN-082-000C, BFN-0-GEN-082-000C                                     87GC phase B        BFN-0-87G-082-2547C/CB  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR C BFN-0-GEN-082-000C, BFN-0-GEN-082-000C                                     87GC phase C        BFN-0-87G-082-2547C/CC  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR C BFN-0-GEN-082-000C, BFN-0-GEN-082-000C                                        C/OTX               BFN-0-RLY-082-C/OTX   12HFA51A42F           SEIS_1C-6R2 DIESEL GENERATOR C Square D Class BFN-0-GEN-082-000C, BFN-0-GEN-082-000C                                          OTR               BFN-0-RLY-082-C/OTR  8501 Type XUD0-        SEIS_1C-6R1 DIESEL GENERATOR C 1200 BFN-0-GEN-082-000C, BFN-0-GEN-082-000C                                         CRC                 BFN-0-RLY-082-CRC    12HFA51A42F            SEIS_9-1R2 DIESEL GENERATOR C Page 48 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                                     Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                      Disposition BFN-0-GEN-082-000C, BFN-0-GEN-082-000C                                        GRRC                BFN-0-RLY-082-GRRC    12HFA51A42F            SEIS_9-1R2 DIESEL GENERATOR C BFN-0-GEN-082-000C, BFN-0-GEN-082-000C                                         GRLC               BFN-0-RLY-082-GRLC    12HFA51A42F            SEIS_9-1R2 DIESEL GENERATOR C BFN-0-GEN-082-000C, BFN-0-GEN-082-000C                                         VLRC               BFN-0-RLY-082-VLRC    12HFA51A42F           SEIS_1C-6R2 DIESEL GENERATOR C BFN-0-GEN-082-000C, BFN-0-GEN-082-000C                                         VRRC               BFN-0-RLY-082-VRRC    12HFA51A42F           SEIS_1C-6R2 DIESEL GENERATOR C BFN-0-GEN-082-000D, BFN-0-GEN-082-000D                                         86GD              BFN-0-86-082-2547D/GD 12HEA61C238X2          SEIS_1C-3R5 DIESEL GENERATOR D Exciter Breaker BFN-0-GEN-082-000D,                                                       Shunt trip relay BFN-0-GEN-082-000D                                           41               BFN-0-41-082-000D/1                          SEIS_9-1R4 DIESEL GENERATOR D                                                        (coil is internal to the bkr)
BFN-0-GEN-082-000D, BFN-0-GEN-082-000D                                     87GD phase A        BFN-0-87G-082-2547D/DA  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR D BFN-0-GEN-082-000D, BFN-0-GEN-082-000D                                     87GD phase B        BFN-0-87G-082-2547D/DB  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR D BFN-0-GEN-082-000D, BFN-0-GEN-082-000D                                     87GD phase C        BFN-0-87G-082-2547D/DC  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR D Page 49 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                                     Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                      Disposition BFN-0-GEN-082-000D, BFN-0-GEN-082-000D                                        D/OTX               BFN-0-RLY-082-D/OTX   12HFA51A42F           SEIS_1C-3R2 DIESEL GENERATOR D Square D Class BFN-0-GEN-082-000D, BFN-0-GEN-082-000D                                          OTR               BFN-0-RLY-082-D/OTR  8501 Type XUD0-        SEIS_1C-3R1 DIESEL GENERATOR D 1200 BFN-0-GEN-082-000D, BFN-0-GEN-082-000D                                         CRD                 BFN-0-RLY-082-CRD    12HFA51A42F            SEIS_9-1R2 DIESEL GENERATOR D BFN-0-GEN-082-000D, BFN-0-GEN-082-000D                                        GRRD                BFN-0-RLY-082-GRRD    12HFA51A42F            SEIS_9-1R2 DIESEL GENERATOR D BFN-0-GEN-082-000D, BFN-0-GEN-082-000D                                         GRLD               BFN-0-RLY-082-GRLD    12HFA51A42F            SEIS_9-1R2 DIESEL GENERATOR D BFN-0-GEN-082-000D, BFN-0-GEN-082-000D                                         VLRD               BFN-0-RLY-082-VLRD    12HFA51A42F           SEIS_1C-3R2 DIESEL GENERATOR D BFN-0-GEN-082-000D, BFN-0-GEN-082-000D                                         VRRD               BFN-0-RLY-082-VRRD    12HFA51A42F           SEIS_1C-3R2 DIESEL GENERATOR D BFN-3-GEN-082-0003A, BFN-3-GEN-082-0003A                                        86G3A              BFN-3-86-082-2547A/GA 12HEA61C238X2          SEIS_1C-4R6 DIESEL GENERATOR 3A Exciter Breaker BFN-3-GEN-082-0003A,                                                      Shunt trip relay BFN-3-GEN-082-0003A                                           41                BFN-3-410-82-000A/1                         SEIS_9-1R4 DIESEL GENERATOR 3A                                                       (coil is internal to the bkr)
Page 50 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                                Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-3-GEN-082-0003A, BFN-3-GEN-082-0003A                                   87G3A phase A         BFN-3-87G-082-2547A/AA GE 12CFD12B1A      SEIS_9-1R3 DIESEL GENERATOR 3A BFN-3-GEN-082-0003A, BFN-3-GEN-082-0003A                                   87G3A phase B        BFN-3-87G-082-2547A/AB  GE 12CFD12B1A      SEIS_9-1R3 DIESEL GENERATOR 3A BFN-3-GEN-082-0003A, BFN-3-GEN-082-0003A                                   87G3A phase C        BFN-3-87G-082-2547A/AC  GE 12CFD12B1A      SEIS_9-1R3 DIESEL GENERATOR 3A BFN-3-GEN-082-0003A, BFN-3-GEN-082-0003A                                       A/OTX              BFN-3-RLY-082-A/OTX    12HFA51A42F      SEIS_1C-4R2 DIESEL GENERATOR 3A Square D Class BFN-3-GEN-082-0003A, BFN-3-GEN-082-0003A                                        OTR               BFN-3-RLY-082-A/OTR   8501 Type XUD0-   SEIS_1C-4R1 DIESEL GENERATOR 3A 1200 BFN-3-GEN-082-0003A, BFN-3-GEN-082-0003A                                        CRA                BFN-3-RLY0-82-CRA     12HFA51A42F      SEIS_9-1R2U3 DIESEL GENERATOR 3A BFN-3-GEN-082-0003A, BFN-3-GEN-082-0003A                                       GRRA               BFN-3-RLY-082-GRRA     12HFA51A42F      SEIS_9-1R2U3 DIESEL GENERATOR 3A BFN-3-GEN-082-0003A, BFN-3-GEN-082-0003A                                        GRLA              BFN-3-RLY0-82-GRLA     12HFA51A42F      SEIS_9-1R2U3 DIESEL GENERATOR 3A BFN-3-GEN-082-0003A, BFN-3-GEN-082-0003A                                        VLRA              BFN-3-RLY0-82-VLRA     12HFA51A42H      SEIS_9-1R2U3 DIESEL GENERATOR 3A Page 51 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                                     Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                      Disposition BFN-3-GEN-082-0003A, BFN-3-GEN-082-0003A                                        VRRA               BFN-3-RLY-082-VRRA    12HFA51A42H           SEIS_9-1R2U3 DIESEL GENERATOR 3A BFN-3-GEN-082-0003B, BFN-3-GEN-082-0003B                                       86G3B              BFN-3-86-082-2547B/GB 12HEA61C238X2          SEIS_1C-5R6 DIESEL GENERATOR 3B Exciter Breaker BFN-3-GEN-082-0003B,                                                      Shunt trip relay BFN-3-GEN-082-0003B                                          41               BFN-3-41-082-000B/1                          SEIS_9-1R4 DIESEL GENERATOR 3B                                                       (coil is internal to the bkr)
BFN-3-GEN-082-0003B, BFN-3-GEN-082-0003B                                   87G3B phase A         BFN-3-87G-082-2547B/BA GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR 3B BFN-3-GEN-082-0003B, BFN-3-GEN-082-0003B                                   87G3B phase B        BFN-3-87G-082-2547B/BB  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR 3B BFN-3-GEN-082-0003B, BFN-3-GEN-082-0003B                                   87G3B phase C        BFN-3-87G-082-2547B/BC  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR 3B BFN-3-GEN-082-0003B, BFN-3-GEN-082-0003B                                       B/OTX               BFN-3-RLY-082-B/OTX   12HFA51A42F           SEIS_1C-5R3 DIESEL GENERATOR 3B Square D Class BFN-3-GEN-082-0003B, BFN-3-GEN-082-0003B                                         OTR               BFN-3-RLY-082-B/0TR  8501 Type XUD0-        SEIS_1C-5R1 DIESEL GENERATOR 3B 1200 BFN-3-GEN-082-0003B, BFN-3-GEN-082-0003B                                         CRB                BFN-3-RLY-082-CRB    12HFA51A42F          SEIS_9-1R2U3-1 DIESEL GENERATOR 3B Page 52 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                                     Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                      Disposition BFN-3-GEN-082-0003B, BFN-3-GEN-082-0003B                                       GRRB               BFN-3-RLY-082-GRRB     12HFA51A42F          SEIS_9-1R2U3-1 DIESEL GENERATOR 3B BFN-3-GEN-082-0003B, BFN-3-GEN-082-0003B                                        GRLB              BFN-3-RLY-082-GRLB     12HFA51A42F          SEIS_9-1R2U3-1 DIESEL GENERATOR 3B BFN-3-GEN-082-0003B, BFN-3-GEN-082-0003B                                        VLRB               BFN-3-RLY-082-VLRB    12HFA51A42H          SEIS_9-1R2U3-1 DIESEL GENERATOR 3B BFN-3-GEN-082-0003B, BFN-3-GEN-082-0003B                                        VRRB              BFN-3-RLY-082-VRRB     12HFA51A42H          SEIS_9-1R2U3-1 DIESEL GENERATOR 3B BFN-3-GEN-082-0003C, BFN-3-GEN-082-0003C                                       86G3C              BFN-3-86-82-2547C/GC  12HEA61C238X2          SEIS_1C-4R6 DIESEL GENERATOR 3C Exciter Breaker BFN-3-GEN-082-0003C,                                                      Shunt trip relay BFN-3-GEN-082-0003C                                          41               BFN-3-41-082-000C/1                          SEIS_9-1R4 DIESEL GENERATOR 3C                                                       (coil is internal to the bkr)
BFN-3-GEN-082-0003C, BFN-3-GEN-082-0003C                                   87G3C phase A         BFN-3-87G-082-2547C/CA GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR 3C BFN-3-GEN-082-0003C, BFN-3-GEN-082-0003C                                   87G3C phase B        BFN-3-87G-082-2547C/CB  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR 3C BFN-3-GEN-082-0003C, BFN-3-GEN-082-0003C                                   87G3C phase C        BFN-3-87G-082-2547C/CC  GE 12CFD12B1A           SEIS_9-1R3 DIESEL GENERATOR 3C Page 53 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                                     Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                      Disposition BFN-3-GEN-082-0003C, BFN-3-GEN-082-0003C                                       C/OTX               BFN-3-RLY-082-C/OTX   12HFA51A42F           SEIS_1C-4R2 DIESEL GENERATOR 3C Square D Class BFN-3-GEN-082-0003C, BFN-3-GEN-082-0003C                                         OTR               BFN-3-RLY-082-C/0TR  8501 Type XUD0-        SEIS_1C-4R1 DIESEL GENERATOR 3C 1200 BFN-3-GEN-082-0003C, BFN-3-GEN-082-0003C                                        CRC                 BFN-3-RLY-082-CRC    12HFA51A42F           SEIS_9-1R2U3 DIESEL GENERATOR 3C BFN-3-GEN-082-0003C, BFN-3-GEN-082-0003C                                       GRRC                BFN-3-RLY-082-GRRC    12HFA51A42F           SEIS_9-1R2U3 DIESEL GENERATOR 3C BFN-3-GEN-082-0003C, BFN-3-GEN-082-0003C                                        GRLC               BFN-3-RLY-082-GRLC    12HFA51A42F           SEIS_9-1R2U3 DIESEL GENERATOR 3C BFN-3-GEN-082-0003C, BFN-3-GEN-082-0003C                                        VLRC               BFN-3-RLY-082-VLRC    12HFA51A42H           SEIS_9-1R2U3 DIESEL GENERATOR 3C BFN-3-GEN-082-0003C, BFN-3-GEN-082-0003C                                        VRRC               BFN-3-RLY-082-VRRC    12HFA51A42H           SEIS_9-1R2U3 DIESEL GENERATOR 3C BFN-3-GEN-082-0003D, BFN-3-GEN-082-0003D                                       86G3D              BFN-3-86-082-2547D/GD 12HEA61C238X2          SEIS_1C-5R6 DIESEL GENERATOR 3D Exciter Breaker BFN-3-GEN-082-0003D,                                                      Shunt trip relay BFN-3-GEN-082-0003D                                          41               BFN-3-41-082-000D/1                          SEIS_9-1R4 DIESEL GENERATOR 3D                                                       (coil is internal to the bkr)
Page 54 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                                Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-3-GEN-082-0003D, BFN-3-GEN-082-0003D                                   87G3D phase A         BFN-3-87G-082-2547D/DA GE 12CFD22B1A      SEIS_9-1R3 DIESEL GENERATOR 3D BFN-3-GEN-082-0003D, BFN-3-GEN-082-0003D                                   87G3D phase B        BFN-3-87G-082-2547D/DB  GE 12CFD22B1A      SEIS_9-1R3 DIESEL GENERATOR 3D BFN-3-GEN-082-0003D, BFN-3-GEN-082-0003D                                   87G3D phase C        BFN-3-87G-082-2547D/DC  GE 12CFD22B1A      SEIS_9-1R3 DIESEL GENERATOR 3D BFN-3-GEN-082-0003D, BFN-3-GEN-082-0003D                                       D/OTX              BFN-3-RLY-082-D/OTX    12HFA51A42H      SEIS_1C-5R3 DIESEL GENERATOR 3D Square D Class BFN-3-GEN-082-0003D, BFN-3-GEN-082-0003D                                        OTR               BFN-3-RLY-082-D/0TR   8501 Type XUD0-   SEIS_1C-5R1 DIESEL GENERATOR 3D 1200 BFN-3-GEN-082-0003D, BFN-3-GEN-082-0003D                                        CRD                BFN-3-RLY-082-CRD     12HFA51A42H     SEIS_9-1R2U3-1 DIESEL GENERATOR 3D BFN-3-GEN-082-0003D, BFN-3-GEN-082-0003D                                       GRRD               BFN-3-RLY-082-GRRD     12HFA51A42H     SEIS_9-1R2U3-1 DIESEL GENERATOR 3D BFN-3-GEN-082-0003D, BFN-3-GEN-082-0003D                                        GRLD              BFN-3-RLY-082-GRLD     12HFA51A42H     SEIS_9-1R2U3-1 DIESEL GENERATOR 3D BFN-3-GEN-082-0003D, BFN-3-GEN-082-0003D                                        VLRD              BFN-3-RLY-082-VLRD     12HFA51A42H     SEIS_9-1R2U3-1 DIESEL GENERATOR 3D Page 55 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                Disposition BFN-3-GEN-082-0003D, BFN-3-GEN-082-0003D                                        VRRD              BFN-3-RLY-082-VRRD     12HFA51A42H    SEIS_9-1R2U3-1 DIESEL GENERATOR 3D BFN-0-BKR-211-000A/022, BFN-0-RLY-082-BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,           CAR                                     12HFA51A42F     SEIS_1C-1R2 2547ACAR TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,         51X1-A                51X-82-2547A/1      Q12HGA111J2   Chatter Acceptable TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, GE BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,         51VX-A               51V-82-2547A/X                     Chatter Acceptable 12PJV11AM2A TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,         51VZ-A                51V-82-2547A/Z      12HFA51A42F   Chatter Acceptable TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,           OTX              BFN-0-RLY-082-A/OTX    12HFA51A42F     SEIS_1C-1R2 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, Square D BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,           OTR              BFN-0-RLY-082-A/OTR                    SEIS_1C-1R1 DO/7001 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,          VLRA              BFN-0-RLY-082-VLRA     12HFA51A42F     SEIS_1C-1R2 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,          VRRA              BFN-0-RLY-082-VRRA     12HFA51A42F     SEIS_1C-1R2 TIE TO DIESEL GEN A Page 56 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                Disposition BFN-0-BKR-211-000A/022, Square D BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,         A/ESTR             BFN-0-RLY-082-A/ESTR                   SEIS_1C-1R1 DO/7001 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,         CASA-2             BFN-0-RLY-211-CASA-2   12HFA51A41H     SEIS_1C-1R8 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,        86-GA-818           BFN-0-86-082-2547A/GA  12HEA61C238     SEIS_1C-1R5 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,         86-1-614         BFN-0-86-211-000A/003     GE HEA-61B     SEIS_1C-1R4 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,     51-614 Phase A         BFN-0-51-211-000A/03A   GE IAC-51A     SEIS_1C-1R3 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,     51-614 Phase C         BFN-0-51-211-000A/03C   GE IAC-51A     SEIS_1C-1R3 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,         86-2-716           BFN-0-86-211-000A/024   GE HEA-61B     SEIS_1C-1R4 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,     51-716 Phase A         BFN-0-51-211-000A/24A   GE IAC-51A     SEIS_1C-1R3 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,     51-716 Phase C         BFN-0-51-211-000A/24C   GE IAC-51A     SEIS_1C-1R3 TIE TO DIESEL GEN A Page 57 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                 Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                  Disposition BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,          86/SA             BFN-0-86-211-000A/023   GE HEA-61B       SEIS_1C-1R4 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,           R1A               BFN-0-RLY-082-R1A     12HFA51A42F       SEIS_1C-1R2 TIE TO DIESEL GEN A BFN-0-BKR-211-000A/022, GE Hitachi NE - Modeled as DG fails BFN-0-BKR-211-000A/022   4KV SD BD A/22, BKR 1818,         Breaker           BFN-0-BKR-211-000A/022 317A7502P005          to start TIE TO DIESEL GEN A BFN-0-BKR-211-000B/004, BFN-0-RLY-082-BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,            CAR                                     12HFA51A42F       SEIS_1C-3R2 2547BCAR TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,          51X1-B                51X-82-2547B/1      Q12HGA111J2     Chatter Acceptable TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,          51VX-B                51V-82-2547B/X     GE12PJV11AM2A    Chatter Acceptable TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,          51VZ-B                51V-82-2547B/Z      12HFA51A42F     Chatter Acceptable TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,            OTX              BFN-0-RLY-082-B/OTX    12HFA51A42F       SEIS_1C-3R2 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004,                                                    Square D Class BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,            OTR              BFN-0-RLY-082-B/OTR   8501 Type XUD0-    SEIS_1C-3R1 TIE TO DIESEL GEN B                                                           1200 Page 58 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                Disposition BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,           VLRB               BFN-0-RLY-082-VLRB    12HFA51A42F     SEIS_1C-3R2 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,           VRRB               BFN-0-RLY-082-VRRB    12HFA51A42F     SEIS_1C-3R2 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, Square D BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,          B/ESTR             BFN-0-RLY-082-B/ESTR                   SEIS_1C-3R1 DO/7001 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,          CASA-5             BFN-0-RLY-211-CASA-5   12HFA51A41F     SEIS_1C-3R8 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, GE BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,         86-GA-822           BFN-0-86-211-000B/004                  SEIS_1C-3R5 12HEA61A213X2 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,           86-616            BFN-0-86-211-000B/002    HEA-61B       SEIS_1C-3R4 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,      51-616 Phase A         BFN-0-51-211-000B/02A   GE IAC-51A     SEIS_1C-3R3 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,      51-616 Phase C         BFN-0-51-211-000B/02C   GE IAC-51A     SEIS_1C-3R3 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,           86-714            BFN-0-86-211-000B/020    HEA-61B       SEIS_1C-3R4 TIE TO DIESEL GEN B Page 59 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,      51-714 Phase A         BFN-0-51-211-000B/20A   GE IAC-51A      SEIS_1C-3R3 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,      51-714 Phase c         BFN-0-51-211-000B/20C   GE IAC-51A      SEIS_1C-3R3 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,           86/SB             BFN-0-86-211-000B/003   GE HEA-61B      SEIS_1C-3R4 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004, BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,            R1B               BFN-0-RLY-082-R1B     12HFA51A42F      SEIS_1C-3R2 TIE TO DIESEL GEN B BFN-0-BKR-211-000B/004,                                                     Wylie-Siemens Modeled as DG fails BFN-0-BKR-211-000B/004   4KV SD BD B/4, BKR 1822,          Breaker           BFN-0-BKR-211-000B/004 Type 5-3AK-GEH-to start TIE TO DIESEL GEN B                                                       250-1200-58 BFN-0-BKR-211-000C/004, BFN-0-RLY-082-BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,            CAR                                     12HFA51A42F      SEIS_1C-6R2 2547CCAR TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,          51X1-C                51X-82-2547C/1       12HGA11J52    Chatter Acceptable TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, GE BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,          51VX-C                51V-82-2547C/X                     Chatter Acceptable 12PJV11AM2A TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, GE BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,          51VZ-C                51V-82-2547C/Z                     Chatter Acceptable 12PJV11AM2A TIE TO DIESEL GEN C Page 60 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,            OTX              BFN-0-RLY-082-C/OTX    12HFA51A42F      SEIS_1C-6R2 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004,                                                     Square D Class BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,            OTR              BFN-0-RLY-082-C/OTR   8501 Type XUD0-   SEIS_1C-6R1 TIE TO DIESEL GEN C                                                            1200 BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,           VLRC               BFN-0-RLY-082-VLRC    12HFA51A42F      SEIS_1C-6R2 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,           VRRC               BFN-0-RLY-082-VRRC    12HFA51A42F      SEIS_1C-6R2 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, Square D BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,          C/ESTR             BFN-0-RLY-082-C/ESTR                    SEIS_1C-6R1 DO/7001 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,          CASB-2             BFN-0-RLY-211-CASB-2   12HFA51A41H      SEIS_1C-6R8 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,         86-GA-812           BFN-0-86-211-000C/004 12HEA61A213X2     SEIS_1C-6R5 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,           86-718            BFN-0-86-211-000C/022    HEA-61B        SEIS_1C-6R4 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,      51-718 Phase A         BFN-0-51-211-000C/22A   GE IAC-51A      SEIS_1C-6R3 TIE TO DIESEL GEN C Page 61 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                Disposition BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,      51-718 Phase C         BFN-0-51-211-000C/22C   GE IAC-51A     SEIS_1C-6R3 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,          86-624             BFN-0-86-211-000C/002     HEA-61B      SEIS_1C-6R4 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,      51-624 Phase A         BFN-0-51-211-000C/02A   GE IAC-51A     SEIS_1C-6R3 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,      51-624 Phase C         BFN-0-51-211-000C/02C   GE IAC-51A     SEIS_1C-6R3 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,           86/SC             BFN-0-86-211-000C/003   GE HEA-61B     SEIS_1C-6R4 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,            R1C               BFN-0-RLY-082-R1C     12HFA51A42F     SEIS_1C-6R2 TIE TO DIESEL GEN C BFN-0-BKR-211-000C/004, GE -     Modeled as DG fails BFN-0-BKR-211-000C/004   4KV SD BD C/4, BKR 1812,          Breaker           BFN-0-BKR-211-000C/004 10AX012G10         to start TIE TO DIESEL GEN C BFN-0-BKR-211-000D/020, BFN-0-RLY-082-BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,            CAR                                     12HFA51A42F     SEIS_1C-3R2 2547DCAR TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, GE BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,          51X1-D                51X-82-2547D/1                    Chatter Acceptable Q12HGA111J2 TIE TO DIESEL GEN D Page 62 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                                Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-0-BKR-211-000D/020, GE BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,          51VX-D                51V-82-2547D/X                     Chatter Acceptable 12PJV11AM2A TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,          51VZ-D                51V-82-2547D/Z      Q12HGA111J2    Chatter Acceptable TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,            OTX              BFN-0-RLY-082-D/OTX    12HFA51A42F      SEIS_1C-3R2 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020,                                                     Square D Class BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,            OTR              BFN-0-RLY-082-D/OTR   8501 Type XUD0-   SEIS_1C-3R1 TIE TO DIESEL GEN D                                                            1200 BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,           VLRD               BFN-0-RLY-082-VLRD    12HFA51A42F      SEIS_1C-3R2 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,           VRRD               BFN-0-RLY-082-VRRD    12HFA51A42F      SEIS_1C-3R2 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, Square D BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,          D/ESTR             BFN-0-RLY-082-D/ESTR                    SEIS_1C-3R1 DO/7001 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,          CASB-5             BFN-0-RLY-211-CASB-5   12HFA51A41H      SEIS_1C-3R8 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,         86-GA-816           BFN-0-86-211-000D/020 12HEA61A213X2     SEIS_1C-3R5 TIE TO DIESEL GEN D Page 63 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                Disposition BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,          86-724             BFN-0-86-211-000D/022     HEA-61B      SEIS_1C-3R4 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,      51-724 Phase A         BFN-0-51-211-000D/22A   GE IAC-51A     SEIS_1C-3R3 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,      51-724 Phase C         BFN-0-51-211-000D/22C  12IAC51A101A    SEIS_1C-3R3 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,          86-618             BFN-0-86-211-000D/005     HEA-61B      SEIS_1C-3R4 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,      51-618 Phase A         BFN-0-51-211-000D/05A  12IAC51A101A    SEIS_1C-3R3 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,      51-618 Phase C         BFN-0-51-211-000D/05C   GE IAC-51A     SEIS_1C-3R3 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,           86/SD              BFN-0-86-211-000D/21   GE HEA-61B     SEIS_1C-3R4 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,            R1D                BFN-0-RLY-082-R1D    12HFA51A42F     SEIS_1C-3R2 TIE TO DIESEL GEN D BFN-0-BKR-211-000D/020, GE -      Modeled as DG fails BFN-0-BKR-211-000D/020  4KV SD BD D/20, BKR 1816,          Breaker           BFN-0-BKR-211-000D/020 10AX012G10         to start TIE TO DIESEL GEN D Page 64 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description          Circuit Drawing                                                 Disposition BFN-0-BKR-211-000A/005, BFN-0-BKR-211-000A/005   4KV SD BD A CMPT 5, NOR           50G-A/5          BFN-0-50G-211-000A/005    GE PJC-11A      SEIS_1C-1R6 FDR TO TRANS TS1A BFN-0-BKR-211-000A/005,                                                    Wylie-Siemens BFN-0-BKR-211-000A/005   4KV SD BD A CMPT 5, NOR           Breaker           BFN-0-BKR-211-000A/005 Type 5-3AK-GEH-    SEIS_1C-1 FDR TO TRANS TS1A                                                         250-1200-58 BFN-0-BKR-211-000A/021, BFN-0-BKR-211-000A/021   4KV SD BD A/21, NOR FDR          50G-A/21          BFN-0-50G-211-000A/020    GE PJC-11A      SEIS_1C-1R6 TO TRANS TDA BFN-0-BKR-211-000A/021,                                                    Wylie-Siemens BFN-0-BKR-211-000A/021   4KV SD BD A/21, NOR FDR           Breaker           BFN-0-BKR-211-000A/021 Type 5-3AK-GEH-    SEIS_1C-1 TO TRANS TDA                                                           250-1200-58 BFN-0-BKR-211-000B/005, BFN-0-BKR-211-000B/005  4KV SD BD B/5, NOR FDR TO          50G-B/5          BFN-0-50G-211-000B/005    GE PJC-11A      SEIS_1C-3R6 TRANS TS2A BFN-0-BKR-211-000B/005,                                                    Wylie-Siemens BFN-0-BKR-211-000B/005  4KV SD BD B/5, NOR FDR TO          Breaker           BFN-0-BKR-211-000B/005 Type 5-3AK-GEH-    SEIS_1C-3 TRANS TS2A                                                           250-1200-58 BFN-0-BKR-211-000B/014, BFN-0-BKR-211-000B/014   4KV SD BD B/14, NOR FDR          50G-B/14          BFN-0-50G-211-000B/014    GE PJC-11A      SEIS_1C-3R6 TO TRANS TS1E AND TDE BFN-0-BKR-211-000B/014,                                                    Wylie-Siemens BFN-0-BKR-211-000B/014   4KV SD BD B/14, NOR FDR           Breaker           BFN-0-BKR-211-000B/014 Type 5-3AK-GEH-    SEIS_1C-3 TO TRANS TS1E AND TDE                                                        250-1200-58 BFN-0-BKR-211-000C/020, BFN-0-BKR-211-000C/020   4KV SD BD C/20, NOR FDR          50G-C/20          BFN-0-50G-211-000C/020    GE PJC-11A      SEIS_1C-6R6 FOR TRANS TS1B Page 65 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                 Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                  Disposition BFN-0-BKR-211-000C/020,                                                     Wylie-Siemens BFN-0-BKR-211-000C/020   4KV SD BD C/20, NOR FDR           Breaker           BFN-0-BKR-211-000C/020 Type 5-3AK-GEH-     SEIS_1C-1 FOR TRANS TS1B                                                           250-1200-58 BFN-0-BKR-211-000C/005, BFN-0-BKR-211-000C/005    4KV SD BD C/5 NOR FDR            50G-C/5          BFN-0-50G-211-000C/005    GE PJC-11A       SEIS_1C-6R6 FOR TRANS TS2E BFN-0-BKR-211-000C/005,                                                     Wylie-Siemens BFN-0-BKR-211-000C/005    4KV SD BD C/5 NOR FDR            Breaker           BFN-0-BKR-211-000C/005 Type 5-3AK-GEH-     SEIS_1C-1 FOR TRANS TS2E                                                           250-1200-58 BFN-0-BKR-211-000D/013, BFN-0-BKR-211-000D/013   4KV SD BD D/13, NOR FDR          50G-D/13          BFN-0-50G-211-000D/013    GE PJC-11A       SEIS_1C-3R6 FOR TRANS TDB BFN-0-BKR-211-000D/013,                                                     Wylie-Siemens BFN-0-BKR-211-000D/013   4KV SD BD D/13, NOR FDR           Breaker           BFN-0-BKR-211-000D/013 Type 5-3AK-GEH-     SEIS_1C-3 FOR TRANS TDB                                                           250-1200-58 BFN-0-BKR-211-000D/019, BFN-0-BKR-211-000D/019   4KV SD BD D/19, NOR FDR          50G-D/19          BFN-0-50G-211-000D/019    GE PJC-11A       SEIS_1C-3R6 FOR TRANS TS2B BFN-0-BKR-211-000D/019, GE Hitachi NE -
BFN-0-BKR-211-000D/019   4KV SD BD D/19, NOR FDR           Breaker           BFN-0-BKR-211-000D/019                     SEIS_1C-3 317A7502P005 FOR TRANS TS2B BFN-0-BKR-211-000A/024, BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR           86-2-716           BFN-0-86-211-000A/024   GE HEA-61B       SEIS_1C-1R4 1716 BFN-0-BKR-211-000A/024, BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR            86-716          BFN-0-86-211-000A/024       HEA-61B        SEIS_1C-1R4 1716 Page 66 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                Disposition BFN-0-BKR-211-000A/024, BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR       51-716 Phase A         BFN-0-51-211-000A/24A   GE IAC-51A     SEIS_1C-1R3 1716 BFN-0-BKR-211-000A/024, BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR       51-716 Phase C         BFN-0-51-211-000A/24C   GE IAC-51A     SEIS_1C-1R3 1716 BFN-0-BKR-211-000A/024, BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR           86-614           BFN-0-86-211-000A/003      HEA-61B       SEIS_1C-1R4 1716 BFN-0-BKR-211-000A/024, BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR       51-614 Phase A         BFN-0-51-211-000A/03A   GE IAC-51A     SEIS_1C-1R3 1716 BFN-0-BKR-211-000A/024, BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR       51-614 Phase C         BFN-0-51-211-000A/03C   GE IAC-51A     SEIS_1C-1R3 1716 BFN-0-BKR-211-000A/024, BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR            86-SA              BFN-0-86-211-000A/23    HEA-61B       SEIS_1C-1R4 1716 BFN-0-BKR-211-000A/024, BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR           27SAX            BFN-0-27-211-000A/12J    12HFA51A41F     SEIS_1C-1R7 1716 BFN-0-BKR-211-000A/024, BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR           CASA-2             BFN-0-RLY-211-CASA-2   12HFA51A41H     SEIS_1C-1R8 1716 BFN-0-BKR-211-000A/024, BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR           CASB-2             BFN-0-RLY-211-CASB-2   12HFA51A41H     SEIS_1C-6R8 1716 Page 67 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                 Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                  Disposition BFN-0-BKR-211-000A/024,                                                                     Alternate feeder GE Hitachi NE -
BFN-0-BKR-211-000A/024  4KV SD BD A, ALT FDR BKR           Breaker           BFN-0-BKR-211-000A/024                     breakers not Q10AX012G6 1716                                                                                  modeled BFN-0-BKR-211-000B/020, BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,         86-2-714           BFN-0-86-211-000B/020   GE HEA-61B       SEIS_1C-3R4 ALT FDR FROM SD BUS 2 BFN-0-BKR-211-000B/020, BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,          86-714            BFN-0-86-211-000B/020    HEA-61B         SEIS_1C-3R4 ALT FDR FROM SD BUS 2 BFN-0-BKR-211-000B/020, BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,     51-714 Phase A         BFN-0-51-211-000B/20A   GE IAC-51A       SEIS_1C-3R3 ALT FDR FROM SD BUS 2 BFN-0-BKR-211-000B/020, BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,     51-714 Phase C         BFN-0-51-211-000B/20C   GE IAC-51A       SEIS_1C-3R3 ALT FDR FROM SD BUS 2 BFN-0-BKR-211-000B/020, BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,          86-616            BFN-0-86-211-000B/002    HEA-61B         SEIS_1C-3R4 ALT FDR FROM SD BUS 2 BFN-0-BKR-211-000B/020, BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,     51-616 Phase A         BFN-0-51-211-000B/02A   GE IAC-51A       SEIS_1C-3R3 ALT FDR FROM SD BUS 2 BFN-0-BKR-211-000B/020, BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,     51-616 Phase C         BFN-0-51-211-000B/02C   GE IAC-51A       SEIS_1C-3R3 ALT FDR FROM SD BUS 2 BFN-0-BKR-211-000B/020, BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,          86-SB             BFN-0-86-211-000B/003   GE HEA-61B       SEIS_1C-3R4 ALT FDR FROM SD BUS 2 Page 68 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-0-BKR-211-000B/020, BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,          27SBX           BFN-0-27-211-000B/12J    12HFA51A41F      SEIS_1C-3R7 ALT FDR FROM SD BUS 2 BFN-0-BKR-211-000B/020, BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,         CASA-5             BFN-0-RLY-211-CASA-5   12HFA51A41H      SEIS_1C-3R8 ALT FDR FROM SD BUS 2 BFN-0-BKR-211-000B/020, BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,         CASB-5             BFN-0-RLY-211-CASB-5   12HFA51A41H      SEIS_1C-3R8 ALT FDR FROM SD BUS 2 BFN-0-BKR-211-000B/020,                                                    Wylie-Siemens   Alternate feeder BFN-0-BKR-211-000B/020   4KV SD BD B/20, BKR 1714,         Breaker           BFN-0-BKR-211-000B/020 Type 5-3AK-GEH-    breakers not ALT FDR FROM SD BUS 2                                                       250-1200-58        modeled BFN-0-BKR-211-000D/005, BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,          86-2-618           BFN-0-86-211-000D/005   GE HEA-61B      SEIS_1C-3R4 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000D/005, BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,           86-618            BFN-0-86-211-000D/005     HEA-61B       SEIS_1C-3R4 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000D/005, BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,      51-618 Phase A         BFN-0-51-211-000D/05A  12IAC51A101A     SEIS_1C-3R3 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000D/005, BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,      51-618 Phase C         BFN-0-51-211-000D/05C   GE IAC-51A      SEIS_1C-3R3 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000D/005, BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,           86-718            BFN-0-86-211-000C/022     HEA-61B       SEIS_1C-6R4 ALT FDR FROM SD BUS 1 Page 69 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                 Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                  Disposition BFN-0-BKR-211-000D/005, BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,      51-718 Phase A         BFN-0-51-211-000C/22A   GE IAC-51A       SEIS_1C-6R3 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000D/005, BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,      51-718 Phase C         BFN-0-51-211-000C/22C   GE IAC-51A       SEIS_1C-6R3 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000D/005, BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,           86-SC             BFN-0-86-211-000C/003    HEA-61B         SEIS_1C-6R4 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000D/005, BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,           27SCX           BFN-0-27-211-000C/11H    12HFA51A41F      SEIS_11-1R1-1 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000D/005, BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,          CASA-5             BFN-0-RLY-211-CASA-5   12HFA51A41H       SEIS_1C-3R8 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000D/005, BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,          CASB-5             BFN-0-RLY-211-CASB-5   12HFA51A41H       SEIS_1C-3R8 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000D/005,                                                                      Alternate feeder GE Hitachi NE -
BFN-0-BKR-211-000D/005   4KV SD BD D/5, BKR 1618,          Breaker           BFN-0-BKR-211-000D/005                     breakers not Q10AX012G6 ALT FDR FROM SD BUS 1                                                                            modeled BFN-0-BKR-211-000C/002, BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,          86-2-624           BFN-0-86-211-000C/002   GE HEA-61B       SEIS_1C-6R4 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000C/002, BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,           86-624            BFN-0-86-211-000C/002    HEA-61B         SEIS_1C-6R4 ALT FDR FROM SD BUS 1 Page 70 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                Disposition BFN-0-BKR-211-000C/002, BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,      51-624 Phase A         BFN-0-51-211-000C/02A   GE IAC-51A     SEIS_1C-6R3 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000C/002, BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,      51-624 Phase C         BFN-0-51-211-000C/02C   GE IAC-51A     SEIS_1C-6R3 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000C/002, BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,          86-724             BFN-0-86-211-000D/022     HEA-61B      SEIS_1C-3R4 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000C/002, BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,      51-724 Phase A         BFN-0-51-211-000D/22A   GE IAC-51A     SEIS_1C-3R3 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000C/002, BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,      51-724 Phase C         BFN-0-51-211-000D/22C  12IAC51A101A    SEIS_1C-3R3 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000C/002, BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,           86-SD             BFN-0-86-211-000D/021     HEA-61B      SEIS_1C-3R4 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000C/002, BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,          27SDX            BFN-0-27-211-000D/11H    12HFA51A41F    SEIS_11-1R1-2 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000C/002, BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,          CASA-2             BFN-0-RLY-211-CASA-2   12HFA51A41H     SEIS_1C-1R8 ALT FDR FROM SD BUS 1 BFN-0-BKR-211-000C/002, BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,          CASB-2             BFN-0-RLY-211-CASB-2   12HFA51A41H     SEIS_1C-6R8 ALT FDR FROM SD BUS 1 Page 71 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                 Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                  Disposition BFN-0-BKR-211-000C/002,                                                                      Alternate feeder GE Hitachi NE -
BFN-0-BKR-211-000C/002   4KV SD BD C/2, BKR 1624,          Breaker           BFN-0-BKR-211-000C/002                     breakers not Q10AX012G6 ALT FDR FROM SD BUS 1                                                                            modeled BFN-0-BKR-211-000A/003, BFN-0-BKR-211-000A/003   4KV SD BD A/3, BKR 1614,          86-1-614           BFN-0-86-211-000A/003   GE HEA-61B       SEIS_1C-1R4 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000A/003, BFN-0-BKR-211-000A/003   4KV SD BD A/3, BKR 1614,           86-614          BFN-0-86-211-000A/003      HEA-61B         SEIS_1C-1R4 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000A/003, BFN-0-BKR-211-000A/003   4KV SD BD A/3, BKR 1614,      51-614 Phase A         BFN-0-51-211-000A/03A   GE IAC-51A       SEIS_1C-1R3 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000A/003, BFN-0-BKR-211-000A/003   4KV SD BD A/3, BKR 1614,      51-614 Phase C         BFN-0-51-211-000A/03C   GE IAC-51A       SEIS_1C-1R3 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000A/003, BFN-0-BKR-211-000A/003   4KV SD BD A/3, BKR 1614,           86-716          BFN-0-86-211-000A/024      HEA-61B         SEIS_1C-1R4 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000A/003, BFN-0-BKR-211-000A/003   4KV SD BD A/3, BKR 1614,      51-716 Phase A         BFN-0-51-211-000A/24A   GE IAC-51A       SEIS_1C-1R3 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000A/003, BFN-0-BKR-211-000A/003   4KV SD BD A/3, BKR 1614,      51-716 Phase C         BFN-0-51-211-000A/24C   GE IAC-51A       SEIS_1C-1R3 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000A/003, BFN-0-BKR-211-000A/003   4KV SD BD A/3, BKR 1614,           86-SA             BFN-0-86-211-000A/023    HEA-61B         SEIS_1C-1R4 NOR FDR FROM SD BUS 1 Page 72 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-0-BKR-211-000A/003, BFN-0-BKR-211-000A/003   4KV SD BD A/3, BKR 1614,           27SAX           BFN-0-27-211-000A/12J    12HFA51A41F      SEIS_1C-1R7 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000A/003,                                                     Wylie-Siemens BFN-0-BKR-211-000A/003   4KV SD BD A/3, BKR 1614,          Breaker           BFN-0-BKR-211-000A/003 Type 5-3AK-GEH-    SEIS_1C-1 NOR FDR FROM SD BUS 1                                                        250-1200-58 BFN-0-BKR-211-000B/002, BFN-0-BKR-211-000B/002   4KV SD BD B/2, BKR 1616,          86-1-616           BFN-0-86-211-000B/002   GE HEA-61B      SEIS_1C-3R4 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000B/002, BFN-0-BKR-211-000B/002   4KV SD BD B/2, BKR 1616,           86-616            BFN-0-86-211-000B/002     HEA-61B       SEIS_1C-3R4 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000B/002, BFN-0-BKR-211-000B/002   4KV SD BD B/2, BKR 1616,      51-616 Phase A         BFN-0-51-211-000B/02A   GE IAC-51A      SEIS_1C-3R3 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000B/002, BFN-0-BKR-211-000B/002   4KV SD BD B/2, BKR 1616,      51-616 Phase C         BFN-0-51-211-000B/02C   GE IAC-51A      SEIS_1C-3R3 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000B/002, BFN-0-BKR-211-000B/002   4KV SD BD B/2, BKR 1616,           86-714            BFN-0-86-211-000B/020     HEA-61B       SEIS_1C-3R4 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000B/002, BFN-0-BKR-211-000B/002   4KV SD BD B/2, BKR 1616,      51-714 Phase A         BFN-0-51-211-000B/20A   GE IAC-51A      SEIS_1C-3R3 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000B/002, BFN-0-BKR-211-000B/002   4KV SD BD B/2, BKR 1616,      51-714 Phase C         BFN-0-51-211-000B/20C   GE IAC-51A      SEIS_1C-3R3 NOR FDR FROM SD BUS 1 Page 73 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-0-BKR-211-000B/002, BFN-0-BKR-211-000B/002   4KV SD BD B/2, BKR 1616,           86-SB             BFN-0-86-211-000B/003     HEA-61B       SEIS_1C-3R4 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000B/002, BFN-0-BKR-211-000B/002   4KV SD BD B/2, BKR 1616,           27SBX           BFN-0-27-211-000B/12J    12HFA51A41F      SEIS_1C-3R7 NOR FDR FROM SD BUS 1 BFN-0-BKR-211-000B/002,                                                     Wylie-Siemens BFN-0-BKR-211-000B/002   4KV SD BD B/2, BKR 1616,          Breaker           BFN-0-BKR-211-000B/002 Type 5-3AK-GEH-    SEIS_1C-3 NOR FDR FROM SD BUS 1                                                        250-1200-58 BFN-0-BKR-211-000C/022, BFN-0-BKR-211-000C/022  4KV SD BD C/22, BKR 1718,          86-1-718           BFN-0-86-211-000C/022   GE HEA-61B      SEIS_1C-6R4 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000C/022, BFN-0-BKR-211-000C/022  4KV SD BD C/22, BKR 1718,           86-718            BFN-0-86-211-000C/022     HEA-61B       SEIS_1C-6R4 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000C/022, BFN-0-BKR-211-000C/022  4KV SD BD C/22, BKR 1718,      51-718 Phase A         BFN-0-51-211-000C/22A   GE IAC-51A      SEIS_1C-6R3 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000C/022, BFN-0-BKR-211-000C/022  4KV SD BD C/22, BKR 1718,      51-718 Phase C         BFN-0-51-211-000C/22C   GE IAC-51A      SEIS_1C-6R3 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000C/022, BFN-0-BKR-211-000C/022  4KV SD BD C/22, BKR 1718,           86-624            BFN-0-86-211-000C/002     HEA-61B       SEIS_1C-6R4 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000C/022, BFN-0-BKR-211-000C/022  4KV SD BD C/22, BKR 1718,      51-624 Phase A         BFN-0-51-211-000C/02A   GE IAC-51A      SEIS_1C-6R3 NOR FDR FROM SD BUS 2 Page 74 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                 Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                  Disposition BFN-0-BKR-211-000C/022, BFN-0-BKR-211-000C/022  4KV SD BD C/22, BKR 1718,      51-624 Phase C         BFN-0-51-211-000C/02C   GE IAC-51A       SEIS_1C-6R3 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000C/022, BFN-0-BKR-211-000C/022  4KV SD BD C/22, BKR 1718,           86-SC             BFN-0-86-211-000C/003     HEA-61B        SEIS_1C-6R4 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000C/022, BFN-0-BKR-211-000C/022  4KV SD BD C/22, BKR 1718,           27SCX           BFN-0-27-211-000C/11H    12HFA51A41F      SEIS_11-1R1-1 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000C/022, GE Hitachi NE -
BFN-0-BKR-211-000C/022  4KV SD BD C/22, BKR 1718,          Breaker           BFN-0-BKR-211-000C/022                     SEIS_1C-1 Q10AX012G6 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000D/022, BFN-0-BKR-211-000D/022  4KV SD BD D/22, BKR 1724,          86-1-724           BFN-0-86-211-000D/022   GE HEA-61B       SEIS_1C-3R4 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000D/022, BFN-0-BKR-211-000D/022  4KV SD BD D/22, BKR 1724,           86-724            BFN-0-86-211-000D/022     HEA-61B        SEIS_1C-3R4 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000D/022, BFN-0-BKR-211-000D/022  4KV SD BD D/22, BKR 1724,      51-724 Phase A         BFN-0-51-211-000D/22A   GE IAC-51A       SEIS_1C-3R3 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000D/022, BFN-0-BKR-211-000D/022  4KV SD BD D/22, BKR 1724,      51-724 Phase C         BFN-0-51-211-000D/22C  12IAC51A101A      SEIS_1C-3R3 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000D/022, BFN-0-BKR-211-000D/022  4KV SD BD D/22, BKR 1724,           86-618            BFN-0-86-211-000D/005     HEA-61B        SEIS_1C-3R4 NOR FDR FROM SD BUS 2 Page 75 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                 Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                  Disposition BFN-0-BKR-211-000D/022, BFN-0-BKR-211-000D/022   4KV SD BD D/22, BKR 1724,     51-618 Phase A         BFN-0-51-211-000D/05A  12IAC51A101A      SEIS_1C-3R3 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000D/022, BFN-0-BKR-211-000D/022   4KV SD BD D/22, BKR 1724,     51-618 Phase B         BFN-0-51-211-000D/05C   GE IAC-51A       SEIS_1C-3R3 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000D/022, BFN-0-BKR-211-000D/022   4KV SD BD D/22, BKR 1724,          86-SD             BFN-0-86-211-000D/021     HEA-61B        SEIS_1C-3R4 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000D/022, BFN-0-BKR-211-000D/022   4KV SD BD D/22, BKR 1724,         27SDX            BFN-0-27-211-000D/11H    12HFA51A41F      SEIS_11-1R1-2 NOR FDR FROM SD BUS 2 BFN-0-BKR-211-000D/022, GE Hitachi NE -
BFN-0-BKR-211-000D/022   4KV SD BD D/22, BKR 1724,         Breaker           BFN-0-BKR-211-000D/022                     SEIS_1C-3 317A7502P005 NOR FDR FROM SD BUS 2 BFN-3-BKR-211-03EA/009, Square D BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,         A/ESTR             BFN-3-RLY-082-A/ESTR                     SEIS_1C-4R1 DO/7001 TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,         51X1-A             BFN-3-RLY-082-2547A/1   12HGA11J51     Chatter Acceptable TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, BFN-3-RLY-082-BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,           CAR                                     12HFA51A42F       SEIS_1C-4R2 2547ACAR TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, GE BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,         51VX-A             BFN-3-51V-082-2547A/X                  Chatter Acceptable 12PJV11AM2A TIE TO DIESEL GEN 3A Page 76 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID   Component Type UNID                    Description           Circuit Drawing                                                Disposition BFN-3-BKR-211-03EA/009, BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,         51VZ-A             BFN-3-51V-082-2547A/Z  12HFA51A42F   Chatter Acceptable TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,           R3A               BFN-3-RLY-082-R3A     12HFA51A42H     SEIS_1C-4R3 TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,         CASA-2             BFN-3-RLY-211-CASA-2   12HFA51A41F     SEIS_1C-4R8 TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,         CASB-2             BFN-3-RLY-211-CASB-2   12HFA151A1F     SEIS_1C-4R8 TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,       86-G3A-838           BFN-3-86-211-03EA/09  12HEA61A213X2    SEIS_1C-4R6 TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,         86-334             BFN-3-86-211-03EA/07   GE HEA61C218    SEIS_1C-4R4 TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, GE BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,     51-334 Phase A         BFN-3-51-211-03EA/07A                 SEIS_1C-4R5-1 12IAC51A101A TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, GE BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,     51-334 Phase B       BFN-3-51-211-03EA/07C                   SEIS_1C-4R5-1 12IAC51A101A TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,         86-726             BFN-3-86-211-03EA/04  12HEA61C218X2    SEIS_1C-4R4 TIE TO DIESEL GEN 3A Page 77 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-3-BKR-211-03EA/009, GE BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,     51-726 Phase A         BFN-3-51-211-03EA/04A                  SEIS_1C-4R5-1 12IAC51A101A TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, GE BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,     51-726 Phase B         BFN-3-51-211-03EA/04C                  SEIS_1C-4R5-1 12IAC51A101A TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,         86-S3A              BFN-3-86-211-03EA/08 12HEA61C220X2    SEIS_1C-4R5-2 TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009, BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,          OTX                BFN-3-RLY-082-A/OTX   12HFA51A42H      SEIS_1C-4R2 TIE TO DIESEL GEN 3A BFN-3-BKR-211-03EA/009,                                                     Square D Class BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,          OTR                BFN-3-RLY-082-A/OTR  8501 Type XUD0-   SEIS_1C-4R1 TIE TO DIESEL GEN 3A                                                           1200 BFN-3-BKR-211-03EA/009,                                                     Wylie-Siemens Modeled as DG fails BFN-3-BKR-211-03EA/009  4KV SD BD 3EA/9, BKR 1838,         Breaker           BFN-3-BKR-211-03EA/009 Type 5-3AK-GEH-to start TIE TO DIESEL GEN 3A                                                       250-1200-58 BFN-3-BKR-211-03EB/011, Square D BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR            B/ESTR             BFN-3-RLY-082-B/ESTR                    SEIS_1C-5R1 DO/7001 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR            51X1-B             BFN-3-RLY-082-2547B/1   12HGA11J51    Chatter Acceptable 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, BFN-3-RLY-082-BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR             CAR                                      12HFA51A42F      SEIS_1C-5R2 2547BCAR 1842, TIE TO DIESEL GEN 3B Page 78 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                Disposition BFN-3-BKR-211-03EB/011, GE BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR            51VX-B             BFN-3-51V-082-2547B/X                Chatter Acceptable 12PJV11AM2A 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR            51VZ-B             BFN-3-51V-082-2547B/Z  12HFA51A42F   Chatter Acceptable 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR              R3B                BFN-3-RLY-082-R3B    12HFA51A42H     SEIS_1C-5R3 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR            CASA-5             BFN-3-RLY-211-CASA-5   12HFA51A41F     SEIS_1C-5R8 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR            CASB-5             BFN-3-RLY-211-CASB-5   12HFA151A1F     SEIS_1C-5R8 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR          86-G3B-842            BFN-3-86-211-03EB/11 12HEA61A213X2    SEIS_1C-5R6 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR            86-336              BFN-3-86-211-03EB/14 12HEA61C218X2   SEIS_1C-5R4-1 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, GE BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR        51-336 Phase A         BFN-3-51-211-03EB/14A                  SEIS_1C-5R5 12IAC51A101A 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, GE BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR        51-336 Phase C         BFN-3-51-211-03EB/14C                  SEIS_1C-5R5 12IAC51A101A 1842, TIE TO DIESEL GEN 3B Page 79 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-3-BKR-211-03EB/011, BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR            86-728              BFN-3-86-211-03EB/08 12HEA61C218X2    SEIS_1C-5R4-1 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, GE BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR        51-728 Phase A         BFN-3-51-211-03EB/08A                   SEIS_1C-5R5 12IAC51A101A 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, GE BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR        51-728 Phase C         BFN-3-51-211-03EB/08C                   SEIS_1C-5R5 12IAC51A101A 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR            86-S3B              BFN-3-86-211-03EB/07 12HEA61C219X2    SEIS_1C-5R4-1 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011, BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR            B/OTX               BFN-3-RLY-082-B/OTX   12HFA51A42H      SEIS_1C-5R3 1842, TIE TO DIESEL GEN 3B BFN-3-BKR-211-03EB/011,                                                     Square D Class BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR             OTR                BFN-3-RLY-082-B/0TR  8501 Type XUD0-   SEIS_1C-5R1 1842, TIE TO DIESEL GEN 3B                                                        1200 BFN-3-BKR-211-03EB/011,                                                     Wylie-Siemens Modeled as DG fails BFN-3-BKR-211-03EB/011    4KV SD BD 3EB/11, BKR            Breaker           BFN-3-BKR-211-03EB/011 Type 5-3AK-GEH-to start 1842, TIE TO DIESEL GEN 3B                                                    250-1200-58 BFN-3-BKR-211-03EC/010, Square D BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR            C/ESTR             BFN-3-RLY-082-C/ESTR                    SEIS_1C-4R1 DO/7001 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR            51X1-C             BFN-3-RLY-082-2547C/1   12HGA11J51    Chatter Acceptable 1832, TIE TO DG 3C Page 80 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                Disposition BFN-3-BKR-211-03EC/010, BFN-3-RLY-082-BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR             CAR                                      12HFA51A42F     SEIS_1C-4R2 2547CCAR 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, GE BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR            51VX-C             BFN-3-51V-082-2547C/X                Chatter Acceptable 12PJV11AM2A 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR            51VZ-C             BFN-3-51V-082-2547C/X  12HFA51A42F   Chatter Acceptable 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR             R3C                 BFN-3-RLY-082-R3C    12HFA51A42H     SEIS_1C-4R3 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR            CASA-2             BFN-3-RLY-211-CASA-2   12HFA51A41F     SEIS_1C-4R8 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR            CASB-2             BFN-3-RLY-211-CASB-2   12HFA151A1F     SEIS_1C-4R8 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR          86-G3C-832            BFN-3-86-211-03EC/10 12HEA61A213X2    SEIS_1C-4R6 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR            86-338              BFN-3-86-211-03EC/12 12HEA61C218X2    SEIS_1C-4R4 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, GE BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR        51-338 Phase A         BFN-3-51-211-03EC/12A                 SEIS_1C-4R5-1 12IAC51A101A 1832, TIE TO DG 3C Page 81 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                 Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                  Disposition BFN-3-BKR-211-03EC/010, GE BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR        51-338 Phase C         BFN-3-51-211-03EC/12C                   SEIS_1C-4R5-1 12IAC51A101A 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR            86-626              BFN-3-86-211-03EC/03 12HEA61C218X2      SEIS_1C-4R4 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, GE BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR        51-626 Phase A         BFN-3-51-211-03EC/03A                   SEIS_1C-4R5-1 12IAC51A101A 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, GE BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR        51-626 Phase C         BFN-3-51-211-03EC/03C                   SEIS_1C-4R5-1 12IAC51A101A 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR            86-S3C              BFN-3-86-211-03EC/11 12HEA61C219X2     SEIS_1C-4R5-2 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010, BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR            C/OTX               BFN-3-RLY-082-C/OTX   12HFA51A42H       SEIS_1C-4R2 1832, TIE TO DG 3C BFN-3-BKR-211-03EC/010,                                                     Square D Class BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR             OTR                BFN-3-RLY-082-C/0TR  8501 Type XUD0-    SEIS_1C-4R1 1832, TIE TO DG 3C                                                            1200 BFN-3-BKR-211-03EC/010, GE Hitachi NE - Modeled as DG fails BFN-3-BKR-211-03EC/010    4KV SD BD 3EC/10, BKR            Breaker           BFN-3-BKR-211-03EC/010 317A7502P005          to start 1832, TIE TO DG 3C BFN-3-BKR-211-03ED/010, Square D BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR           D/ESTR             BFN-3-RLY-082-D/ESTR                     SEIS_1C-5R1 DO/7001 1836, TIE TO DIESEL GEN 3D Page 82 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                Disposition BFN-3-BKR-211-03ED/010, BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR           51X1-D             BFN-3-RLY-082-2547D/1   12HGA11J51   Chatter Acceptable 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, BFN-3-RLY-082-BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR            CAR                                      12HFA51A42F     SEIS_1C-5R2 2547DCAR 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, GE BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR           51VX-D             BFN-3-51V-082-2547D/X                Chatter Acceptable 12PJV11AM2A 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR           51VZ-D             BFN-3-51V-082-2547D/Z  12HFA51A42F   Chatter Acceptable 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR            R3D                 BFN-3-RLY-082-R3D    12HFA51A42H     SEIS_1C-5R3 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR           CASA-5             BFN-3-RLY-211-CASA-5   12HFA51A41F     SEIS_1C-5R8 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR           CASB-5             BFN-3-RLY-211-CASB-5   12HFA151A1F     SEIS_1C-5R8 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR         86-G3D-836            BFN-3-86-211-03ED/10 12HEA61A213X2    SEIS_1C-5R6 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR           86-342              BFN-3-86-211-03ED/08 12HEA61C218X2   SEIS_1C-5R4-1 1836, TIE TO DIESEL GEN 3D Page 83 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-3-BKR-211-03ED/010, GE BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR       51-342 Phase A         BFN-3-51-211-03ED/08A                   SEIS_1C-5R5 12IAC51A101A 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, GE BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR       51-342 Phase C         BFN-3-51-211-03ED/08C                   SEIS_1C-5R5 12IAC51A101A 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR           86-628              BFN-3-86-211-03ED/01 12HEA61C218X2    SEIS_1C-5R4-1 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, GE BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR       51-628 Phase A         BFN-3-51-211-03ED/01A                   SEIS_1C-5R5 12IAC51A101A 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, GE BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR       51-628 Phase C         BFN-3-51-211-03ED/01C                   SEIS_1C-5R5 12IAC51A101A 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR           86-S3D              BFN-3-86-211-03ED/09 12HEA61C219X2    SEIS_1C-5R4-2 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010, BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR           D/OTX               BFN-3-RLY-082-D/OTX   12HFA51A42H      SEIS_1C-5R3 1836, TIE TO DIESEL GEN 3D BFN-3-BKR-211-03ED/010,                                                     Square D Class BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR            OTR                BFN-3-RLY-082-D/0TR  8501 Type XUD0-   SEIS_1C-5R1 1836, TIE TO DIESEL GEN 3D                                                         1200 BFN-3-BKR-211-03ED/010,                                                     Wylie-Siemens Modeled as DG fails BFN-3-BKR-211-03ED/010   4160V SD BD 3ED/10, BKR           Breaker           BFN-3-BKR-211-03ED/010 Type 5-3AK-GEH-to start 1836, TIE TO DIESEL GEN 3D                                                     250-1200-58 Page 84 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-3-BKR-211-03EA/010, 4KV SD BD 3EA/10, NORMAL BFN-3-BKR-211-03EA/010                                   50G-3EA/10         BFN-3-50G-211-03EA/010    GE PJC11A       SEIS_1C-4R7 FEEDER TO 480V SD BD 3A, XFMR TS3A BFN-3-BKR-211-03EA/010, Wylie-Siemens 4KV SD BD 3EA/10, NORMAL BFN-3-BKR-211-03EA/010                                     Breaker           BFN-3-BKR-211-03EA/010 Type 5-3AK-GEH-    SEIS_1C-4 FEEDER TO 480V SD BD 3A, 250-1200-58 XFMR TS3A BFN-3-BKR-211-03EB/009, BFN-3-BKR-211-03EB/009  4KV SD BD 3EB/9, NOR FDR        50G-3EB/9C          BFN-3-50G-211-03EB/009    GE PJC11A       SEIS_1C-5R7 FOR TRANS TS3E BFN-3-BKR-211-03EB/009,                                                    Wylie-Siemens BFN-3-BKR-211-03EB/009  4KV SD BD 3EB/9, NOR FDR           Breaker           BFN-3-BKR-211-03EB/009 Type 5-3AK-GEH-    SEIS_1C-5 FOR TRANS TS3E                                                           250-1200-58 BFN-3-BKR-211-03EC/007, BFN-3-BKR-211-03EC/007  4KV SD BD 3EC/7, NOR FDR         50G-3E7C/7         BFN-3-50G-211-03EC/007    GE PJC11A       SEIS_1C-4R7 FOR TRANS TS3B BFN-3-BKR-211-03EC/007,                                                    Wylie-Siemens BFN-3-BKR-211-03EC/007  4KV SD BD 3EC/7, NOR FDR           Breaker           BFN-3-BKR-211-03EC/007 Type 5-3AK-GEH-    SEIS_1C-4 FOR TRANS TS3B                                                           250-1200-58 BFN-3-BKR-211-03ED/13, BFN-3-BKR-211-03ED/13    FEEDER TO 480V HVAC BD          50G-3ED/13         BFN-3-50GS-211-03ED/13      GE HFC        SEIS_1C-5R7 B TRANSFORMER THB BFN-3-BKR-211-03ED/13,                                                     Wylie-Siemens BFN-3-BKR-211-03ED/13    FEEDER TO 480V HVAC BD            Breaker            BFN-3-BKR-211-03ED/13 Type 5-3AK-GEH-    SEIS_1C-5 B TRANSFORMER THB                                                          250-1200-58 Page 85 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                 Disposition BFN-3-BKR-211-03EA/004, 4KV SD BD 3EA/4 BKR 1726 BFN-3-BKR-211-03EA/004                                     86-334              BFN-3-86-211-03EA/07  GE HEA61C218     SEIS_1C-4R4 ALT FDR FROM 4KV BUS TIE BD BFN-3-BKR-211-03EA/004, 4KV SD BD 3EA/4 BKR 1726                                                          GE BFN-3-BKR-211-03EA/004                                 51-334 Phase A         BFN-3-51-211-03EA/07A                 SEIS_1C-4R5-1 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD BFN-3-BKR-211-03EA/004, 4KV SD BD 3EA/4 BKR 1726                                                          GE BFN-3-BKR-211-03EA/004                                 51-334 Phase C       BFN-3-51-211-03EA/07C                   SEIS_1C-4R5-1 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD BFN-3-BKR-211-03EA/004, 4KV SD BD 3EA/4 BKR 1726 BFN-3-BKR-211-03EA/004                                     86-726              BFN-3-86-211-03EA/04 12HEA61C218X2     SEIS_1C-4R4 ALT FDR FROM 4KV BUS TIE BD BFN-3-BKR-211-03EA/004, 4KV SD BD 3EA/4 BKR 1726                                                          GE BFN-3-BKR-211-03EA/004                                 51-726 Phase A         BFN-3-51-211-03EA/04A                 SEIS_1C-4R5-1 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD BFN-3-BKR-211-03EA/004, 4KV SD BD 3EA/4 BKR 1726                                                          GE BFN-3-BKR-211-03EA/004                                 51-726 Phase C         BFN-3-51-211-03EA/04C                 SEIS_1C-4R5-1 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD BFN-3-BKR-211-03EA/004, 4KV SD BD 3EA/4 BKR 1726 BFN-3-BKR-211-03EA/004                                     86-S3A              BFN-3-86-211-03EA/08 12HEA61C220X2   SEIS_1C-4R5-2 ALT FDR FROM 4KV BUS TIE BD Trips and Locks out BFN-3-BKR-211-03EA/004, normally open tie 4KV SD BD 3EA/4 BKR 1726 BFN-3-BKR-211-03EA/004                                     86-EAX               BFN-3-86-210-0001A   12HFA51A41F      board Feeder ALT FDR FROM 4KV BUS TIE Breaker. Not a BD chatter concern.
Page 86 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                 Disposition Trips Tie Board BFN-3-BKR-211-03EA/004, FDR BKR 4KV SD BD 3EA/4 BKR 1726                                                           GE BFN-3-BKR-211-03EA/004                                      R-726            BFN-3-RLY-211-03EA/04C                  Emergency use ALT FDR FROM 4KV BUS TIE                                                     12HFA51A41H only. Not a chatter BD concern.
BFN-3-BKR-211-03EA/004, Wylie-Siemens   Alternate feeder 4KV SD BD 3EA/4 BKR 1726 BFN-3-BKR-211-03EA/004                                     Breaker           BFN-3-BKR-211-03EA/004 Type 5-3AK-GEH-    breakers not ALT FDR FROM 4KV BUS TIE 250-1200-58        modeled BD BFN-3-BKR-211-03EB/008, 4KV SD BD 3EB/8 BKR 1728 BFN-3-BKR-211-03EB/008                                     86-336              BFN-3-86-211-03EB/14 12HEA61C218X2    SEIS_1C-5R4-1 ALT FDR FROM 4KV SD BD 3EA BFN-3-BKR-211-03EB/008, 4KV SD BD 3EB/8 BKR 1728                                                           GE BFN-3-BKR-211-03EB/008                                 51-336 Phase A         BFN-3-51-211-03EB/14A                   SEIS_1C-5R5 ALT FDR FROM 4KV SD BD                                                      12IAC51A101A 3EA BFN-3-BKR-211-03EB/008, 4KV SD BD 3EB/8 BKR 1728                                                           GE BFN-3-BKR-211-03EB/008                                 51-336 Phase C         BFN-3-51-211-03EB/14C                   SEIS_1C-5R5 ALT FDR FROM 4KV SD BD                                                      12IAC51A101A 3EA BFN-3-BKR-211-03EB/008, 4KV SD BD 3EB/8 BKR 1728 BFN-3-BKR-211-03EB/008                                     86-728              BFN-3-86-211-03EB/08 12HEA61C218X2    SEIS_1C-5R4-1 ALT FDR FROM 4KV SD BD 3EA BFN-3-BKR-211-03EB/008, 4KV SD BD 3EB/8 BKR 1728                                                           GE BFN-3-BKR-211-03EB/008                                 51-728 Phase A         BFN-3-51-211-03EB/08A                   SEIS_1C-5R5 ALT FDR FROM 4KV SD BD                                                      12IAC51A101A 3EA BFN-3-BKR-211-03EB/008, 4KV SD BD 3EB/8 BKR 1728                                                           GE BFN-3-BKR-211-03EB/008                                 51-728 Phase C         BFN-3-51-211-03EB/08C                   SEIS_1C-5R5 ALT FDR FROM 4KV SD BD                                                      12IAC51A101A 3EA Page 87 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                  Disposition BFN-3-BKR-211-03EB/008, 4KV SD BD 3EB/8 BKR 1728 BFN-3-BKR-211-03EB/008                                     86-S3B              BFN-3-86-211-03EB/07 12HEA61C219X2     SEIS_1C-5R4-1 ALT FDR FROM 4KV SD BD 3EA Trips and Locks out BFN-3-BKR-211-03EB/008, normally open tie 4KV SD BD 3EB/8 BKR 1728 BFN-3-BKR-211-03EB/008                                     86-EAX               BFN-3-86-210-0001A   12HFA51A41F       board Feeder ALT FDR FROM 4KV SD BD Breaker. Not a 3EA chatter concern.
Trips Tie Board BFN-3-BKR-211-03EB/008, FDR BKR 4KV SD BD 3EB/8 BKR 1728                                                           GE BFN-3-BKR-211-03EB/008                                      R-728            BFN-3-RLY-211-03EB/08C                   Emergency use ALT FDR FROM 4KV SD BD                                                      12HFA51A41H only. Not a chatter 3EA concern.
BFN-3-BKR-211-03EB/008, Wylie-Siemens    Alternate feeder 4KV SD BD 3EB/8 BKR 1728 BFN-3-BKR-211-03EB/008                                     Breaker           BFN-3-BKR-211-03EB/008 Type 5-3AK-GEH-     breakers not ALT FDR FROM 4KV SD BD 250-1200-58         modeled 3EA BFN-3-BKR-211-03EC/003, 4KV SD BD 3EC BKR 1626 BFN-3-BKR-211-03EC/003                                     86-338              BFN-3-86-211-03EC/12 12HEA61C218X2      SEIS_1C-4R4 ALT FDR FROM 4KV BUS TIE BD BFN-3-BKR-211-03EC/003, 4KV SD BD 3EC BKR 1626                                                            GE BFN-3-BKR-211-03EC/003                                 51-338 Phase A         BFN-3-51-211-03EC/12A                   SEIS_1C-4R5-1 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD BFN-3-BKR-211-03EC/003, 4KV SD BD 3EC BKR 1626                                                            GE BFN-3-BKR-211-03EC/003                                 51-338 Phase C         BFN-3-51-211-03EC/12C                   SEIS_1C-4R5-1 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD BFN-3-BKR-211-03EC/003, 4KV SD BD 3EC BKR 1626 BFN-3-BKR-211-03EC/003                                     86-626              BFN-3-86-211-03EC/03 12HEA61C218X2      SEIS_1C-4R4 ALT FDR FROM 4KV BUS TIE BD Page 88 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                                Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                  Disposition BFN-3-BKR-211-03EC/003, 4KV SD BD 3EC BKR 1626                                                            GE BFN-3-BKR-211-03EC/003                                 51-626 Phase A         BFN-3-51-211-03EC/03A                   SEIS_1C-4R5-1 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD BFN-3-BKR-211-03EC/003, 4KV SD BD 3EC BKR 1626                                                            GE BFN-3-BKR-211-03EC/003                                 51-626 Phase C         BFN-3-51-211-03EC/03C                   SEIS_1C-4R5-1 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD BFN-3-BKR-211-03EC/003, 4KV SD BD 3EC BKR 1626 BFN-3-BKR-211-03EC/003                                     86-S3C             BFN-3-86-211-03EC/011 12HEA61C219X2     SEIS_1C-4R5-2 ALT FDR FROM 4KV BUS TIE BD Trips and Locks out BFN-3-BKR-211-03EC/003, normally open tie 4KV SD BD 3EC BKR 1626 BFN-3-BKR-211-03EC/003                                     86-ECX               BFN-3-86-210-0001C   12HFA51A41F       board Feeder ALT FDR FROM 4KV BUS TIE Breaker. Not a BD chatter concern.
BFN-3-BKR-211-03EC/003,                                                                      Trips Tie Board 4KV SD BD 3EC BKR 1626                                                            GE            FDR BKR BFN-3-BKR-211-03EC/003                                      R-626            BFN-3-RLY-211-03EC/03C ALT FDR FROM 4KV BUS TIE                                                     12HFA51A41H      Emergency use BD                                                                             only. Not a concern.
BFN-3-BKR-211-03EC/003, Wylie-Siemens    Alternate feeder 4KV SD BD 3EC BKR 1626 BFN-3-BKR-211-03EC/003                                     Breaker           BFN-3-BKR-211-03EC/003 Type 5-3AK-GEH-     breakers not ALT FDR FROM 4KV BUS TIE 250-1200-58         modeled BD BFN-3-BKR-211-03ED/001, 4KV SD BD 3ED/1 BKR 1628 BFN-3-BKR-211-03ED/001                                     86-342              BFN-3-86-211-03ED/08 12HEA61C218X2     SEIS_1C-5R4-1 ALT FDR FROM 4KV BUS TIE BD BFN-3-BKR-211-03ED/001, 4KV SD BD 3ED/1 BKR 1628                                                           GE BFN-3-BKR-211-03ED/001                                 51-342 Phase A         BFN-3-51-211-03ED/08A                    SEIS_1C-5R5 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD Page 89 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component     Associated Component           Relay ID on                                                Fragility Group or Relay/Breaker UNID  Component Type UNID                    Description           Circuit Drawing                                                  Disposition BFN-3-BKR-211-03ED/001, 4KV SD BD 3ED/1 BKR 1628                                                           GE BFN-3-BKR-211-03ED/001                                 51-342 Phase C         BFN-3-51-211-03ED/08C                    SEIS_1C-5R5 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD BFN-3-BKR-211-03ED/001, 4KV SD BD 3ED/1 BKR 1628 BFN-3-BKR-211-03ED/001                                     86-628              BFN-3-86-211-03ED/01 12HEA61C218X2     SEIS_1C-5R4-1 ALT FDR FROM 4KV BUS TIE BD BFN-3-BKR-211-03ED/001, 4KV SD BD 3ED/1 BKR 1628                                                           GE BFN-3-BKR-211-03ED/001                                 51-628 Phase A         BFN-3-51-211-03ED/01A                    SEIS_1C-5R5 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD BFN-3-BKR-211-03ED/001, 4KV SD BD 3ED/1 BKR 1628                                                           GE BFN-3-BKR-211-03ED/001                                 51-628 Phase C         BFN-3-51-211-03ED/01C                    SEIS_1C-5R5 ALT FDR FROM 4KV BUS TIE                                                     12IAC51A101A BD BFN-3-BKR-211-03ED/001, 4KV SD BD 3ED/1 BKR 1628 BFN-3-BKR-211-03ED/001                                     86-S3D              BFN-3-86-211-03ED/09 12HEA61C219X2     SEIS_1C-5R4-2 ALT FDR FROM 4KV BUS TIE BD Trips and Locks out BFN-3-BKR-211-03ED/001, normally open tie 4KV SD BD 3ED/1 BKR 1628 BFN-3-BKR-211-03ED/001                                     86-ECX               BFN-3-86-210-0001C   12HFA51A41F       board Feeder ALT FDR FROM 4KV BUS TIE Breaker. Not a BD chatter concern.
Trips Tie Board BFN-3-BKR-211-03ED/001, FDR BKR 4KV SD BD 3ED/1 BKR 1628                                                           GE BFN-3-BKR-211-03ED/001                                      R-628            BFN-3-RLY-211-03ED/01C                   Emergency use ALT FDR FROM 4KV BUS TIE                                                     12HFA51A41H only. Not a chatter BD concern.
BFN-3-BKR-211-03ED/001, Wylie-Siemens    Alternate feeder 4KV SD BD 3ED/1 BKR 1628 BFN-3-BKR-211-03ED/001                                     Breaker           BFN-3-BKR-211-03ED/001 Type 5-3AK-GEH-     breakers not ALT FDR FROM 4KV BUS TIE 250-1200-58         modeled BD Page 90 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                                Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                                 Disposition BFN-0-PMP-023-0094, BFN-0-PMP-023-0094                                        27SDX              BFN-0-27-211-000D/11H  12HFA51A41F     SEIS_11-1R1-2 RHRSW PUMP D3 BFN-0-PMP-023-0094, BFN-0-PMP-023-0094                                        27SDY              BFN-0-27-211-000D/11M  12HFA51A41F     SEIS_11-1R1-2 RHRSW PUMP D3 BFN-0-PMP-023-0094, BFN-0-PMP-023-0094                                          50G               BFN-0-50G-023-0094     GE PJC-11A      SEIS_11-1R2 RHRSW PUMP D3 Wylie-Siemens BFN-0-PMP-023-0094,                                                                      Modeled as EECW BFN-0-PMP-023-0094                                        Breaker             BFN-0-BKR-023-0094   Type 5-3AK-GEH-RHRSW PUMP D3                                                                          pump fails to start 250-1200-58 BFN-0-PMP-023-0088, BFN-0-PMP-023-0088                                        27SCX              BFN-0-27-211-000C/11H  12HFA51A41F     SEIS_11-1R1-1 RHRSW PUMP B3 BFN-0-PMP-023-0088, BFN-0-PMP-023-0088                                        27SCY              BFN-0-27-211-000C/11M  12HFA51A41F     SEIS_11-1R1-1 RHRSW PUMP B3 BFN-0-PMP-023-0088, BFN-0-PMP-023-0088                                          50G               BFN-0-50G-023-0088     GE PJC-11A      SEIS_11-1R2 RHRSW PUMP B3 Wylie-Siemens BFN-0-PMP-023-0088,                                                                      Modeled as EECW BFN-0-PMP-023-0088                                        Breaker             BFN-0-BKR-023-0088   Type 5-3AK-GEH-RHRSW PUMP B3                                                                          pump fails to start 250-1200-58 BFN-0-PMP-023-0085, BFN-0-PMP-023-0085                                        27S3AX             BFN-3-27-211-03EA/08G  12HFA51A41F      SEIS_11-1R3 RHRSW PUMP A3 Page 91 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                                Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                                 Disposition BFN-0-PMP-023-0085, BFN-0-PMP-023-0085                                        27S3AY             BFN-3-27-211-03EA/08K  12HFA51A41F      SEIS_11-1R3 RHRSW PUMP A3 BFN-0-PMP-023-0085, BFN-0-PMP-023-0085                                          50G               BFN-0-50G-023-0085     GE PJC-11A      SEIS_11-1R4 RHRSW PUMP A3 Wylie-Siemens BFN-0-PMP-023-0085,                                                                      Modeled as EECW BFN-0-PMP-023-0085                                        Breaker             BFN-0-BKR-023-0085   Type 5-3AK-GEH-RHRSW PUMP A3                                                                          pump fails to start 250-1200-58 BFN-0-PMP-023-0091, BFN-0-PMP-023-0091                                        27S3BX             BFN-3-27-211-03EB/07G  12HFA51A41F      SEIS_11-1R5 RHRSW PUMP C3 BFN-0-PMP-023-0091, BFN-0-PMP-023-0091                                        27S3BY             BFN-3-27-211-03EB/07K  12HFA51A41F      SEIS_11-1R5 RHRSW PUMP C3 BFN-0-PMP-023-0091, BFN-0-PMP-023-0091                                          50G               BFN-0-50G-023-0091     GE PJC-11A      SEIS_11-1R6 RHRSW PUMP C3 Wylie-Siemens BFN-0-PMP-023-0091,                                                                      Modeled as EECW BFN-0-PMP-023-0091                                        Breaker             BFN-0-BKR-023-0091   Type 5-3AK-GEH-RHRSW PUMP C3                                                                          pump fails to start 250-1200-58 BFN-0-FAN-030-0064, DIESEL                                                    Cardox - Clark BFN-0-FAN-030-0064                                          GA                        GA*                            SEIS_17-1R1 GEN RM A EXH FAN A                                                          Cat # 5U8 BFN-0-FAN-030-0064, DIESEL BFN-0-FAN-030-0064                                         GATD                BFN-0-IC-039-0007      Cardox TM      SEIS_17-1R2 GEN RM A EXH FAN A Page 92 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-0-FAN-030-0065, DIESEL                                                   Cardox - Clark BFN-0-FAN-030-0065                                          GA                      GA*                             SEIS_17-1R1 GEN RM A EXH FAN B                                                         Cat # 5U8 BFN-0-FAN-030-0065, DIESEL BFN-0-FAN-030-0065                                         GATD              BFN-0-IC-039-0007      Cardox TM       SEIS_17-1R2 GEN RM A EXH FAN B BFN-0-FAN-030-0066, DIESEL                                                   Cardox - Clark BFN-0-FAN-030-0066                                          GB                      GB*                             SEIS_17-1R1 GEN RM B EXH FAN A                                                         Cat # 5U8 BFN-0-FAN-030-0066, DIESEL BFN-0-FAN-030-0066                                         GBTD              BFN-0-IC-039-0008      Cardox TM       SEIS_17-1R2 GEN RM B EXH FAN A BFN-0-FAN-030-0067, DIESEL                                                   Cardox - Clark BFN-0-FAN-030-0067                                          GB                      GB*                             SEIS_17-1R1 GEN RM B EXH FAN B                                                         Cat # 5U8 BFN-0-FAN-030-0067, DIESEL BFN-0-FAN-030-0067                                         GBTD              BFN-0-IC-039-0008      Cardox TM       SEIS_17-1R2 GEN RM B EXH FAN B BFN-0-FAN-030-0068, DIESEL                                                   Cardox - Clark BFN-0-FAN-030-0068                                          GC                      GC*                             SEIS_17-1R1 GEN RM C EXH FAN A                                                         Cat # 5U8 BFN-0-FAN-030-0068, DIESEL BFN-0-FAN-030-0068                                         GCTD              BFN-0-IC-039-0009      Cardox TM       SEIS_17-1R2 GEN RM C EXH FAN A BFN-0-FAN-030-0069, DIESEL                                                   Cardox - Clark BFN-0-FAN-030-0069                                          GC                      GC*                             SEIS_17-1R1 GEN RM C EXH FAN B                                                         Cat # 5U8 Page 93 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID    Component Type UNID                     Description           Circuit Drawing                                                 Disposition BFN-0-FAN-030-0069, DIESEL BFN-0-FAN-030-0069                                         GCTD              BFN-0-IC-039-0009       Cardox TM       SEIS_17-1R2 GEN RM C EXH FAN B BFN-0-FAN-030-0070, DIESEL                                                    Cardox - Clark BFN-0-FAN-030-0070                                          GD                      GD*                              SEIS_17-1R1 GEN RM D EXH FAN A                                                          Cat # 5U8 BFN-0-FAN-030-0070, DIESEL BFN-0-FAN-030-0070                                         GDTD              BFN-0-IC-039-0010       Cardox TM       SEIS_17-1R2 GEN RM D EXH FAN A BFN-0-FAN-030-0071, DIESEL                                                    Cardox - Clark BFN-0-FAN-030-0071                                          GD                      GD*                              SEIS_17-1R1 GEN RM D EXH FAN B                                                          Cat # 5U8 BFN-0-FAN-030-0071, DIESEL BFN-0-FAN-030-0071                                         GDTD              BFN-0-IC-039-0010       Cardox TM       SEIS_17-1R2 GEN RM D EXH FAN B BFN-0-FAN-030-0072, DG Cardox - Clark BFN-0-FAN-030-0072       AUX TRANS TDA RM EXH                EA                     EA*                              SEIS_17-1R1 Cat # 5U8 FAN BFN-0-FAN-030-0072, DG BFN-0-FAN-030-0072       AUX TRANS TDA RM EXH              EATD              BFN-0-IC-039-0005       Cardox TM       SEIS_17-1R2 FAN BFN-0-FAN-030-0073, 480V Cardox - Clark BFN-0-FAN-030-0073      AUX BD RM B EXH FAN DG               EB                     EB*                              SEIS_17-1R1 Cat # 5U8 TDB BFN-0-FAN-030-0073, 480V BFN-0-FAN-030-0073      AUX BD RM B EXH FAN DG             EBTD              BFN-0-IC-039-0006       Cardox TM       SEIS_17-1R2 TDB Page 94 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-3-FAN-030-0230, DIESEL                                                   Cardox - Clark BFN-3-FAN-030-0230                                          GA               BFN-3-RLY-039-0038A                    SEIS_17-1R1 GEN RM 3A EXH FAN A                                                         Cat # 5U8 BFN-3-FAN-030-0230, DIESEL BFN-3-FAN-030-0230                                         GATD               BFN-3-IC-039-0038     Cardox TM       SEIS_17-1R2 GEN RM 3A EXH FAN A BFN-3-FAN-030-0231, DIESEL                                                   Cardox - Clark BFN-3-FAN-030-0231                                          GA               BFN-3-RLY-039-0038A                    SEIS_17-1R1 GEN RM 3A EXH FAN B                                                         Cat # 5U8 BFN-3-FAN-030-0231, DIESEL BFN-3-FAN-030-0231                                         GATD               BFN-3-IC-039-0038     Cardox TM       SEIS_17-1R2 GEN RM 3A EXH FAN B BFN-3-FAN-030-0232, DIESEL                                                   Cardox - Clark BFN-3-FAN-030-0232                                          GB               BFN-3-RLY-039-0039A                    SEIS_17-1R1 GEN RM 3B EXH FAN A                                                         Cat # 5U8 BFN-3-FAN-030-0232, DIESEL BFN-3-FAN-030-0232                                         GBTD               BFN-3-IC-039-0039     Cardox TM       SEIS_17-1R2 GEN RM 3B EXH FAN A BFN-3-FAN-030-0233, DIESEL                                                   Cardox - Clark BFN-3-FAN-030-0233                                          GB               BFN-3-RLY-039-0039A                    SEIS_17-1R1 GEN RM 3B EXH FAN B                                                         Cat # 5U8 BFN-3-FAN-030-0233, DIESEL BFN-3-FAN-030-0233                                         GBTD               BFN-3-IC-039-0039     Cardox TM       SEIS_17-1R2 GEN RM 3B EXH FAN B BFN-3-FAN-030-0234, DIESEL                                                   Cardox - Clark BFN-3-FAN-030-0234                                          GC               BFN-3-RLY-039-0040A                    SEIS_17-1R1 GEN RM 3C EXH FAN A                                                         Cat # 5U8 Page 95 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID  Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-3-FAN-030-0234, DIESEL BFN-3-FAN-030-0234                                         GCTD               BFN-3-IC-039-0040     Cardox TM       SEIS_17-1R2 GEN RM 3C EXH FAN A BFN-3-FAN-030-0235, DIESEL                                                   Cardox - Clark BFN-3-FAN-030-0235                                          GC               BFN-3-RLY-039-0040A                    SEIS_17-1R1 GEN RM 3C EXH FAN B                                                         Cat # 5U8 BFN-3-FAN-030-0235, DIESEL BFN-3-FAN-030-0235                                         GCTD               BFN-3-IC-039-0040     Cardox TM       SEIS_17-1R2 GEN RM 3C EXH FAN B BFN-3-FAN-030-0236, DIESEL                                                   Cardox - Clark BFN-3-FAN-030-0236                                          GD               BFN-3-RLY-039-0041A                    SEIS_17-1R1 GEN RM 3D EXH FAN A                                                         Cat # 5U8 BFN-3-FAN-030-0236, DIESEL BFN-3-FAN-030-0236                                         GDTD               BFN-3-IC-039-0041     Cardox TM       SEIS_17-1R2 GEN RM 3D EXH FAN A BFN-3-FAN-030-0237, DIESEL                                                   Cardox - Clark BFN-3-FAN-030-0237                                          GD               BFN-3-RLY-039-0041A                    SEIS_17-1R1 GEN RM 3D EXH FAN B                                                         Cat # 5U8 BFN-3-FAN-030-0237, DIESEL BFN-3-FAN-030-0237                                         GDTD               BFN-3-IC-039-0041     Cardox TM       SEIS_17-1R2 GEN RM 3D EXH FAN B BFN-3-FAN-030-0243, DIESEL                                                   Cardox - Clark BFN-3-FAN-030-0243                                           EA                      EA*                            SEIS_17-1R1 AUX BD RM 3EA FAN                                                         Cat # 5U8 BFN-3-FAN-030-0243, DIESEL BFN-3-FAN-030-0243                                         EATD               BFN-3-IC-039-0036     Cardox TM       SEIS_17-1R2 AUX BD RM 3EA FAN Page 96 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-3-FAN-030-0244, DSL Cardox - Clark BFN-3-FAN-030-0244     AUX. BD. RM. 3EB & PIPE &             EB                     EB*                             SEIS_17-1R1 Cat # 5U8 ELEC. TUNNEL BFN-3-FAN-030-0244, DSL BFN-3-FAN-030-0244     AUX. BD. RM. 3EB & PIPE &           EBTD              BFN-3-IC-039-0037      Cardox TM       SEIS_17-1R2 ELEC. TUNNEL BFN-0-FCO-031-0088, UNITS                                                   Cardox - Clark BFN-0-FCO-031-0088                                           R1                      R1*                            SEIS_17-1R3 1 & 2 EL 593 AHU A                                                        Cat # 5U8 BFN-0-FCO-031-0088, UNITS BFN-0-FCO-031-0088                                         R1TD              BFN-1-IC-039-0015      Cardox TM       SEIS_17-1R4 1 & 2 EL 593 AHU A BFN-0-FCO-031-0088, UNITS                                                   Cardox - Clark BFN-0-FCO-031-0088                                           R2                      R2*                            SEIS_17-1R3 1 & 2 EL 593 AHU A                                                        Cat # 5U8 BFN-0-FCO-031-0088, UNITS BFN-0-FCO-031-0088                                         R2TD              BFN-2-IC-039-0017      Cardox TM       SEIS_17-1R4 1 & 2 EL 593 AHU A BFN-0-FCO-031-0089, UNITS Cardox - Clark BFN-0-FCO-031-0089     1 & 2 EL 593 AIR HANDLING             R1                      R1*                            SEIS_17-1R3 Cat # 5U8 UNIT 1B BFN-0-FCO-031-0089, UNITS BFN-0-FCO-031-0089     1 & 2 EL 593 AIR HANDLING           R1TD              BFN-1-IC-039-0015      Cardox TM       SEIS_17-1R4 UNIT 1B BFN-0-FCO-031-0089, UNITS Cardox - Clark BFN-0-FCO-031-0089     1 & 2 EL 593 AIR HANDLING             R2                      R2*                            SEIS_17-1R3 Cat # 5U8 UNIT 1B Page 97 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID    Component Type UNID                     Description           Circuit Drawing                                                 Disposition BFN-0-FCO-031-0089, UNITS BFN-0-FCO-031-0089     1 & 2 EL 593 AIR HANDLING           R2TD              BFN-2-IC-039-0017       Cardox TM       SEIS_17-1R4 UNIT 1B BFN-0-AHU-031-0088, U1 &                                                    Cardox - Clark BFN-0-AHU-031-0088                                           R1                      R1*                             SEIS_17-1R3 U2 EL 593 AHU A                                                          Cat # 5U8 BFN-0-AHU-031-0088, U1 &
BFN-0-AHU-031-0088                                         R1TD              BFN-1-IC-039-0015       Cardox TM       SEIS_17-1R4 U2 EL 593 AHU A BFN-0-AHU-031-0088, U1 &                                                    Cardox - Clark BFN-0-AHU-031-0088                                           R2                      R2*                             SEIS_17-1R3 U2 EL 593 AHU A                                                          Cat # 5U8 BFN-0-AHU-031-0088, U1 &
BFN-0-AHU-031-0088                                         R2TD              BFN-2-IC-039-0017       Cardox TM       SEIS_17-1R4 U2 EL 593 AHU A BFN-0-AHU-031-0089, U1 &                                                    Cardox - Clark BFN-0-AHU-031-0089                                           R1                      R1*                             SEIS_17-1R3 U2 EL 593 AHU 1B                                                          Cat # 5U8 BFN-0-AHU-031-0089, U1 &
BFN-0-AHU-031-0089                                         R1TD              BFN-1-IC-039-0015       Cardox TM       SEIS_17-1R4 U2 EL 593 AHU 1B BFN-0-AHU-031-0089, U1 &                                                    Cardox - Clark BFN-0-AHU-031-0089                                           R2                      R2*                             SEIS_17-1R3 U2 EL 593 AHU 1B                                                          Cat # 5U8 BFN-0-AHU-031-0089, U1 &
BFN-0-AHU-031-0089                                         R2TD              BFN-2-IC-039-0017       Cardox TM       SEIS_17-1R4 U2 EL 593 AHU 1B Page 98 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                               Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-0-FCO-031-0107, UNIT 3 Cardox - Clark BFN-0-FCO-031-0107       EL 593.0 AIR HANDLING             CR3                      CR3*                            SEIS_17-1R3 Cat # 5U8 UNIT 3A BFN-0-FCO-031-0107, UNIT 3 BFN-0-FCO-031-0107       EL 593.0 AIR HANDLING            CR3TD              BFN-3-IC-039-0018      Cardox TM       SEIS_17-1R4 UNIT 3A BFN-0-FCO-031-0107, UNIT 3 Cardox - Clark BFN-0-FCO-031-0107       EL 593.0 AIR HANDLING               R3                      R3*                            SEIS_17-1R3 Cat # 5U8 UNIT 3A BFN-0-FCO-031-0107, UNIT 3 BFN-0-FCO-031-0107       EL 593.0 AIR HANDLING             R3TD              BFN-3-IC-039-0019      Cardox TM       SEIS_17-1R4 UNIT 3A BFN-0-FCO-031-0108, UNIT 3 Cardox - Clark BFN-0-FCO-031-0108       EL 593.0 AIR HANDLING             CR3                      CR3*                            SEIS_17-1R3 Cat # 5U8 UNIT 3B BFN-0-FCO-031-0108, UNIT 3 BFN-0-FCO-031-0108       EL 593.0 AIR HANDLING            CR3TD              BFN-3-IC-039-0018      Cardox TM       SEIS_17-1R4 UNIT 3B BFN-0-FCO-031-0108, UNIT 3 Cardox - Clark BFN-0-FCO-031-0108       EL 593.0 AIR HANDLING               R3                      R3*                            SEIS_17-1R3 Cat # 5U8 UNIT 3B BFN-0-FCO-031-0108, UNIT 3 BFN-0-FCO-031-0108       EL 593.0 AIR HANDLING             R3TD              BFN-3-IC-039-0019      Cardox TM       SEIS_17-1R4 UNIT 3B BFN-3-AHU-031-0107, UNIT 3                                                  Cardox - Clark BFN-3-AHU-031-0107                                         CR3                      CR3*                            SEIS_17-1R3 MECH EQUIP RM AHU 3A                                                         Cat # 5U8 Page 99 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                                 Disposition BFN-3-AHU-031-0107, UNIT 3 BFN-3-AHU-031-0107                                         CR3TD              BFN-3-IC-039-0018       Cardox TM      SEIS_17-1R4 MECH EQUIP RM AHU 3A BFN-3-AHU-031-0107, UNIT 3                                                   Cardox - Clark BFN-3-AHU-031-0107                                           R3                       R3*                            SEIS_17-1R3 MECH EQUIP RM AHU 3A                                                         Cat # 5U8 BFN-3-AHU-031-0107, UNIT 3 BFN-3-AHU-031-0107                                         R3TD               BFN-3-IC-039-0019       Cardox TM      SEIS_17-1R4 MECH EQUIP RM AHU 3A BFN-3-AHU-031-0108, UNIT 3                                                   Cardox - Clark BFN-3-AHU-031-0108                                          CR3                       R3*                            SEIS_17-1R3 MECH EQUIP RM AHU 3B                                                         Cat # 5U8 BFN-3-AHU-031-0108, UNIT 3 BFN-3-AHU-031-0108                                         CR3TD              BFN-3-IC-039-0018       Cardox TM      SEIS_17-1R4 MECH EQUIP RM AHU 3B BFN-3-AHU-031-0108, UNIT 3                                                   Cardox - Clark BFN-3-AHU-031-0108                                           R3                       R3*                            SEIS_17-1R3 MECH EQUIP RM AHU 3B                                                         Cat # 5U8 BFN-3-AHU-031-0108, UNIT 3 BFN-3-AHU-031-0108                                         R3TD               BFN-3-IC-039-0019       Cardox TM      SEIS_17-1R4 MECH EQUIP RM AHU 3B BFN-1-FCV-073-0002, HPCI BFN-1-FCV-073-0002          STEAM LINE INBD               23A-K37           BFN-1-RLY-073-23A-K37   12HFA151A1F     SEIS_14-1R2-1 ISOLATION VLV BFN-1-FCV-073-0002, HPCI BFN-1-FCV-073-0002          STEAM LINE INBD               23A-K12           BFN-1-RLY-073-23A-K12   12HFA151A1F     SEIS_14-1R1-2 ISOLATION VLV Page 100 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-1-FCV-073-0002, HPCI BFN-1-FCV-073-0002          STEAM LINE INBD               23A-K27           BFN-1-RLY-073-23A-K27   12HFA151A1F    SEIS_14-1R1-2 ISOLATION VLV BFN-2-FCV-073-0002, HPCI BFN-2-FCV-073-0002          STEAM LINE INBD               23A-K37           BFN-2-RLY-073-23A-K37   12HFA151A1F    SEIS_14-1R2-1 ISOLATION VLV BFN-2-FCV-073-0002, HPCI BFN-2-FCV-073-0002          STEAM LINE INBD               23A-K12           BFN-2-RLY-073-23A-K12   12HFA151A1F    SEIS_14-1R2-1 ISOLATION VLV BFN-2-FCV-073-0002, HPCI BFN-2-FCV-073-0002          STEAM LINE INBD               23A-K27           BFN-2-RLY-073-23A-K27   12HFA151A1F    SEIS_14-1R1-2 ISOLATION VLV BFN-3-FCV-073-0002, HPCI BFN-3-FCV-073-0002          STEAM LINE INBD               23A-K37           BFN-3-RLY-073-23A-K37   12HFA151A1F    SEIS_14-1R2-1 ISOLATION VLV BFN-3-FCV-073-0002, HPCI BFN-3-FCV-073-0002          STEAM LINE INBD              23A-K12A          BFN-3-RLY-073-23A-K12A   12HFA151A1F    SEIS_14-1R2-1 ISOLATION VLV BFN-3-FCV-073-0002, HPCI BFN-3-FCV-073-0002          STEAM LINE INBD               63-73-2             BFN-3-63-073-0002     12HFA151A1F    SEIS_14-1R1-2 ISOLATION VLV BFN-1-FCV-073-0003, HPCI BFN-1-FCV-073-0003         STEAM LINE OUTBD               23A-K37           BFN-1-RLY-073-23A-K37   12HFA151A1F    SEIS_14-1R2-1 ISOLATION VLV BFN-1-FCV-073-0003, HPCI BFN-1-FCV-073-0003         STEAM LINE OUTBD               23A-K12           BFN-1-RLY-073-23A-K12   12HFA151A1F    SEIS_14-1R1-2 ISOLATION VLV Page 101 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-1-FCV-073-0003, HPCI BFN-1-FCV-073-0003         STEAM LINE OUTBD               23A-K27           BFN-1-RLY-073-23A-K27   12HFA151A1F    SEIS_14-1R1-2 ISOLATION VLV BFN-2-FCV-073-0003, HPCI BFN-2-FCV-073-0003         STEAM LINE OUTBD               23A-K37           BFN-2-RLY-073-23A-K37   12HFA151A1F    SEIS_14-1R2-1 ISOLATION VLV BFN-2-FCV-073-0003, HPCI BFN-2-FCV-073-0003         STEAM LINE OUTBD               23A-K12           BFN-2-RLY-073-23A-K12   12HFA151A1F    SEIS_14-1R2-1 ISOLATION VLV BFN-2-FCV-073-0003, HPCI BFN-2-FCV-073-0003         STEAM LINE OUTBD               23A-K27           BFN-2-RLY-073-23A-K27   12HFA151A1F    SEIS_14-1R1-2 ISOLATION VLV BFN-3-FCV-073-0003, HPCI BFN-3-FCV-073-0003         STEAM LINE OUTBD               23A-K37           BFN-3-RLY-073-23A-K37   12HFA151A1F    SEIS_14-1R2-1 ISOLATION VLV, BFN-3-FCV-073-0003, HPCI BFN-3-FCV-073-0003         STEAM LINE OUTBD              23A-K12A          BFN-3-RLY-073-23A-K12A   12HFA151A1F    SEIS_14-1R2-1 ISOLATION VLV, BFN-3-FCV-073-0003, HPCI BFN-3-FCV-073-0003         STEAM LINE OUTBD               63-73-2             BFN-3-63-073-0002     12HFA151A1F    SEIS_14-1R1-2 ISOLATION VLV, BFN-1-FCV-073-0044, HPCI BFN-1-FCV-073-0044                                         23A-K1            BFN-1-RLY-073-23A-K1   12HFA151A1F    SEIS_14-1R1-2 SYSTEM INBD DISCH VLV BFN-1-FCV-073-0044, HPCI BFN-1-FCV-073-0044                                         23A-K3            BFN-1-RLY-073-23A-K3   12HFA151A1F    SEIS_14-1R1-2 SYSTEM INBD DISCH VLV Page 102 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                               Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                                Disposition BFN-2-FCV-073-0044, HPCI BFN-2-FCV-073-0044                                        23A-K1             BFN-2-RLY-073-23A-K1   12HFA151A1F    SEIS_14-1R1-2 SYSTEM INBD DISCH VLV BFN-2-FCV-073-0044, HPCI BFN-2-FCV-073-0044                                        23A-K3             BFN-2-RLY-073-23A-K3   12HFA151A1F    SEIS_14-1R1-2 SYSTEM INBD DISCH VLV BFN-3-FCV-073-0044, HPCI BFN-3-FCV-073-0044                                        23A-K1             BFN-3-RLY-073-23A-K1   12HFA151A1F    SEIS_14-1R1-2 SYSTEM INBD DISCH VLV BFN-3-FCV-073-0044, HPCI BFN-3-FCV-073-0044                                        23A-K3             BFN-3-RLY-073-23A-K3   12HFA151A1F    SEIS_14-1R1-2 SYSTEM INBD DISCH VLV BFN-1-PMP-074-0005, BFN-1-PMP-074-0005     RESIDUAL HEAT REMOVAL             10A-K18A          BFN-1-RLY-074-10A-K18A   12HFA151A1F     SEIS_12-1R1 PUMP 1A BFN-1-PMP-074-0016, BFN-1-PMP-074-0016     RESIDUAL HEAT REMOVAL             10A-K18B          BFN-1-RLY-074-10A-K18B   12HFA151A1F     SEIS_12-1R1 PUMP 1C BFN-1-PMP-074-0028, BFN-1-PMP-074-0028     RESIDUAL HEAT REMOVAL             10A-K21A          BFN-1-RLY-074-10A-K21A   12HFA151A1F     SEIS_12-1R1 PUMP 1B BFN-1-PMP-074-0039, BFN-1-PMP-074-0039     RESIDUAL HEAT REMOVAL             10A-K21B          BFN-1-RLY-074-10A-K21B   12HFA151A1F     SEIS_12-1R1 PUMP 1D BFN-2-PMP-074-0005, BFN-2-PMP-074-0005     RESIDUAL HEAT REMOVAL             10A-K18A          BFN-2-RLY-074-10A-K18A   12HFA151A1F     SEIS_12-1R1 PUMP 2A Page 103 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                              Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                               Disposition BFN-2-PMP-074-0016, BFN-2-PMP-074-0016     RESIDUAL HEAT REMOVAL             10A-K18B          BFN-2-RLY-074-10A-K18B  12HFA151A1F     SEIS_12-1R1 PUMP 2C BFN-2-PMP-074-0028, BFN-2-PMP-074-0028     RESIDUAL HEAT REMOVAL             10A-K21A          BFN-2-RLY-074-10A-K21A  12HFA151A1F     SEIS_12-1R1 PUMP 2B BFN-2-PMP-074-0039, BFN-2-PMP-074-0039     RESIDUAL HEAT REMOVAL             10A-K21B          BFN-2-RLY-074-10A-K21B  12HFA151A1F     SEIS_12-1R1 PUMP 2D BFN-3-PMP-074-0005, BFN-3-PMP-074-0005     RESIDUAL HEAT REMOVAL             10A-K18A          BFN-3-RLY-074-10A-K18A  12HFA151A1F     SEIS_12-1R1 PUMP 3A BFN-3-PMP-074-0016, BFN-3-PMP-074-0016     RESIDUAL HEAT REMOVAL             10A-K18B          BFN-3-RLY-074-10A-K18B  12HFA151A1F     SEIS_12-1R1 PUMP 3C BFN-3-PMP-074-0028, BFN-3-PMP-074-0028     RESIDUAL HEAT REMOVAL             10A-K21A          BFN-3-RLY-074-10A-K21A  12HFA151A1F     SEIS_12-1R1 PUMP 3B BFN-3-PMP-074-0039, BFN-3-PMP-074-0039     RESIDUAL HEAT REMOVAL             10A-K21B          BFN-3-RLY-074-10A-K21B  12HFA151A1F     SEIS_12-1R1 PUMP 3D BFN-1-FCV-074-0053, RHR BFN-1-FCV-074-0053       SYS I LPCI INBD INJECT          10A-K67A          BFN-1-RLY-074-10A-K67A  12HGA11A51F    SEIS_14-1R3-1 VALVE BFN-2-FCV-074-0053, RHR BFN-2-FCV-074-0053       SYS I LPCI INBD INJECT          10A-K67A          BFN-2-RLY-074-10A-K67A  12HGA11A51F    SEIS_14-1R3-1 VALVE Page 104 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component      Associated Component           Relay ID on                                              Fragility Group or Relay/Breaker UNID   Component Type UNID                     Description           Circuit Drawing                                               Disposition BFN-3-FCV-074-0053, RHR BFN-3-FCV-074-0053       SYS I LPCI INBD INJECT          10A-K67A          BFN-3-RLY-074-10A-K67A  12HGA11A51F    SEIS_14-1R3-1 VALVE BFN-1-FCV-074-0067, RHR BFN-1-FCV-074-0067       SYS II LPCI INBD INJECT         10A-K66B          BFN-1-RLY-074-10A-K66B  12HGA11A51F    SEIS_14-1R3-1 VALVE BFN-2-FCV-074-0067, RHR BFN-2-FCV-074-0067       SYS II LPCI INBD INJECT         10A-K66B          BFN-1-RLY-074-10A-K66B  12HGA11A51F    SEIS_14-1R3-1 VALVE BFN-3-FCV-074-0067, RHR BFN-3-FCV-074-0067       SYS II LPCI INBD INJECT         10A-K66B          BFN-1-RLY-074-10A-K66B  12HGA11A51F    SEIS_14-1R3-1 VALVE BFN-1-PMP-075-0005, CORE BFN-1-PMP-075-0005                                       14A-K12A          BFN-1-RLY-075-14A-K12A  12HFA151A1F     SEIS_11-1R7 SPRAY PUMP 1A BFN-1-PMP-075-0014, CORE BFN-1-PMP-075-0014                                       14A-K14A          BFN-1-RLY-075-14A-K14A  12HFA151A1F     SEIS_11-1R7 SPRAY PUMP 1C BFN-1-PMP-075-0033, CORE BFN-1-PMP-075-0033                                       14A-K12B          BFN-1-RLY-075-14A-K12B  12HFA151A1F     SEIS_11-1R7 SPRAY PUMP 1B BFN-1-PMP-075-0042, CORE BFN-1-PMP-075-0042                                       14A-K14B          BFN-1-RLY-075-14A-K14B  12HFA151A1F     SEIS_11-1R7 SPRAY PUMP 1D BFN-2-PMP-075-0005, CORE BFN-2-PMP-075-0005                                       14A-K12A          BFN-2-RLY-075-14A-K12A  12HFA151A1F     SEIS_11-1R7 SPRAY PUMP 2A Page 105 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID    Component Type UNID                     Description           Circuit Drawing                                                  Disposition BFN-2-PMP-075-0014, CORE BFN-2-PMP-075-0014                                       14A-K14A          BFN-2-RLY-075-14A-K14A   12HFA151A1F       SEIS_11-1R7 SPRAY PUMP 2C BFN-2-PMP-075-0033, CORE BFN-2-PMP-075-0033                                       14A-K12B          BFN-2-RLY-075-14A-K12B   12HFA151A1F       SEIS_11-1R7 SPRAY PUMP 2B BFN-2-PMP-075-0042, CORE BFN-2-PMP-075-0042                                       14A-K14B          BFN-2-RLY-075-14A-K14B   12HFA151A1F       SEIS_11-1R7 SPRAY PUMP 2D BFN-3-PMP-075-0005, CORE BFN-3-PMP-075-0005     SPRAY PMP 3A, 4KV SD BD           14A-K12A          BFN-3-RLY-075-14A-K12A   12HFA151A1F       SEIS_11-1R7 3EA/6 BFN-3-PMP-075-0014, CORE BFN-3-PMP-075-0014     SPRAY PMP 3C, 4KV SD BD           14A-K14A          BFN-3-RLY-075-14A-K14A   12HFA151A1F       SEIS_11-1R7 3EB/5 BFN-3-PMP-075-0033, CORE BFN-3-PMP-075-0033     SPRAY PMP 3B, 4KV SD BD           14A-K12B          BFN-3-RLY-075-14A-K12B   12HFA151A1F       SEIS_11-1R7 3EC/13 BFN-3-PMP-075-0042, CORE BFN-3-PMP-075-0042     SPRAY PMP 3D, 4KV SD BD           14A-K14B          BFN-3-RLY-075-14A-K14B   12HFA51A41F       SEIS_11-1R7 3ED/II Not a chatter BFN-1-FCV-075-0037, concern. Valve auto BFN-1-FCV-075-0037      SYSTEM 2 MIN FLOW ISOL           FS-75-49            BFN-1-FS-075-0049       Static O Ring opens after chatter VALVE event.
Not a chatter BFN-2-FCV-075-0037, concern. Valve auto BFN-2-FCV-075-0037      SYSTEM 2 MIN FLOW ISOL           FS-75-49            BFN-2-FS-075-0049       Static O Ring opens after chatter VALVE event.
Page 106 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 4.1-2: Components of Chatter Concern Requiring Functional Fragility Analysis Associated Component       Associated Component          Relay ID on                                                Fragility Group or Relay/Breaker UNID    Component Type UNID                     Description           Circuit Drawing                                                 Disposition Not a chatter BFN-3-FCV-075-0037, concern. Valve auto BFN-3-FCV-075-0037      SYSTEM 2 MIN FLOW ISOL           FS-75-49            BFN-3-FS-075-0049       Static O Ring opens after chatter VALVE event.
Note: An asterisk (*) in Relay/Breaker UNID column indicates the component is part of a larger component and was not assigned an individual UNID. The relay ID on the control circuit dwg is used in this column.
Page 107 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                            December 2019 4.2   Walkdown Approach This section provides a summary of the methodology and scope of the seismic walkdowns performed for the SPRA. Walkdowns were performed by personnel with appropriate qualifications as defined in the SPID [2]. The walkdowns were performed on the non-screened components following screening described in Section 4.4.1.
Walkdowns of those SSCs included on the SEL were performed as part of the development of the SEL to assess the as-installed condition of these SSCs for use in determining their seismic capacity and performing initial screening.
Walkdowns were performed in accordance with guidance in the SPID, Section 6.5, and the associated requirements in the PRA Standard [7].
Several previous seismic walkdowns for BFN have been documented. The information gathered during these previous walkdowns and the results and conclusions contained in the walkdown information were used, where applicable, to supplement plant drawings and calculations. These previous walkdowns include:
* Unresolved Safety Issue (USI) A-46/IPEEE [21, 22, 23] - Performed seismic walkdowns and seismic evaluations following the guidelines in EPRI NP-6041-SL
[24] for all three units in support of USI A-46 and subsequently for IPEEE. The seismic evaluations for Units 2 and 3 were completed in 1996, and those for Unit 1 were completed in 2005.
* NTTF 2.3 Seismic [25, 26, 27] - Performed in response to NTTF Recommendation 2.3 Seismic to identify and address degraded, nonconforming, or unanalyzed conditions, and to verify the current plant configuration with the current seismic licensing basis.
* Expedited Seismic Evaluation Process (ESEP) [28] - Performed to focus the initial industry efforts on short-term evaluations to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events, including FLEX equipment installations. This included walkdowns and calculations to demonstrate that the high confidence low probability of failure (HCLPF) seismic capacity for the ESEP subset of plant equipment exceeded the Review Level Ground Motion (RLGM). The RLGM was set to 2xSSE (0.4g) for this purpose.
Information from these walkdowns was gathered and reviewed to obtain inputs and insights for the development of component fragilities. To ensure that the information remained valid and to include components that had not been previously walked down, all components on the SEL except those that are inherently rugged and screened out for SPRA application, including those previously walked down, were included in the scope of the current SPRA walkdowns. However, for components that had been previously walked down and for which sufficient information was available to permit development of a fragility, the walkdown was limited to a walk-by of the individual components.
Page 108 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 Detailed walkdowns were performed for components that had not been previously walked down after the list was reduced by using screening approach described in Section 4.4.1. During a detailed walkdown, the caveats from the Seismic Qualification Utility Group (SQUG) Generic Information Procedure (GIP) [29] were verified and sufficient information was gathered to permit development of a fragility. This included information on anchorage, configuration, weight, dimensions, load path, and other structural information. In addition, the walkdown team focused on potential adverse seismic interaction issues, including the potential for seismically induced fire and flood and seismic II/I concerns, such as masonry block walls in the vicinity of the components.
More simplified walk-bys were performed for components that had been previously walked down. During walk-bys, the walkdown team inspected these components to ensure that there were no obvious changes that might adversely impact their seismic capacity. In particular, the walkdown team focused on potential seismic interaction concerns and conditions. In general, walk-bys were less detailed and less intrusive than walkdowns.
Components that were not accessible during plant operation were walked down during plant outages. Separate walkdowns were performed to assess operator pathways used to perform operator actions, to assess implementation of FLEX, to obtain detailed information related to in-cabinet amplification factors for relays, and to provide specific inputs to the fragility team, such as nozzle loads. In addition, even though the walkdown team focused on the potential for seismically induced fire and flood during the walkdowns, a separate walkdown was conducted to specifically evaluate the potential for seismically induced fires due to electrical faults.
Walkdown documentation for equipment and structures consisted of noting the existing conditions, taking photographs, and recording any findings.
4.2.1 Significant Walkdown Results and Insights Consistent with the guidance from EPRI NP-6041-SL [24], no significant findings or adverse conditions were noted during the BFN seismic walkdowns. Observations made during the walkdowns are documented in the walkdown report [30].
Components on the SEL were evaluated for seismic anchorage, interaction effects (including block walls and other items that might cause a reduction in seismic capacity),
and effects of component degradation, such as corrosion and concrete cracking, for consideration in the development of SEL fragilities. In addition, walkdowns were performed on operator pathways, and the potential for seismic-induced fire and flooding scenarios was assessed. Potential internal flood scenarios were incorporated into the BFN SPRA model. The walkdown observations were judged to be adequate for use in developing the SSC fragilities for the SPRA.
4.2.2 Seismic Equipment List and Seismic Walkdowns Technical Adequacy The BFN SPRA SEL development [19] and walkdowns [30] were subjected to an independent peer review against the pertinent requirements in the PRA Standard. The Page 109 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                           December 2019 SEL development and walkdowns were peer reviewed relative to Capability Category II for the full set of SRs in the PRA Standard. After completion of the subsequent independent assessment, the full set of SRs was met, and the SEL and walkdowns were determined to be acceptable for use in the SPRA.
The peer review assessment [6] and subsequent disposition of peer review findings through an independent assessment [16] are further described in Appendix A and establish that the BFN SPRA SEL and seismic walkdowns are suitable for this SPRA application.
4.3   Dynamic Analysis of Structures This section summarizes the dynamic analyses of structures that contain systems and components important to achieving a safe shutdown, using fixed-base and/or soil-structure interaction (SSI) analyses (as applicable). The section describes the methodologies used, responses at various locations within the structures and relevant outputs, important assumptions, and sources of uncertainty. A list of structures and description of relevant parameters is provided in Table 4.3-1.
4.3.1 Fixed-base Analysis BFN is a firm rock site; SSI was performed for each of the major structures analyzed for the SPRA. Note that fixed-base analyses were performed as a verification step in development of the RB, DGB and IPS SSI models [31].
4.3.2 Soil-Structure Interaction (SSI) Analysis Multi-case deterministic SSI analyses considering ground motion incoherence were performed for the RB, DGB, and IPS. The SSI between the structures and the surrounding soil medium is considered by System for Analysis for Soil-Structure Interaction (SASSI) at defined interaction nodes. Cutoff frequency for the SSI analyses was chosen to be 50 Hz, and the SSI models were sufficiently refined to transmit frequencies of at least 50 Hz through the soil/rock-foundation interface. RB and DGB SSI analyses utilized the SASSI Modified Subtraction Method (MSM). For RB SSI analysis, the nodes at all four sides of the excavated soil and the nodes within three soil layers (at EL 515 ft, EL 542.63 ft, and EL 565 ft) were considered as interaction node. A sensitivity study was performed to verify that using MSM with three intermediate interaction node layers is adequate to generate accurate In-Structure Response Spectra (ISRS) results. For the DGB, all interaction nodes attached to the excavated soil model directly underneath the DGB units were considered for analysis, while for portions of the soil associated with the RB model, the interaction nodes on the RB basement-soil interface (i.e., RB basement walls and basemat) in addition to the top layer of interaction nodes of the excavated soil were considered for analysis. A sensitivity study was performed to verify the adequacy of the interaction node selection for analysis. IPS SSI analysis utilized the SASSI Direct Method (DM), where all soil layer interfaces and excavated soil nodes are defined as interaction nodes. SSI analyses in the three spatial directions were performed simultaneously.
Page 110 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 The site conditions in the SSI models are represented by uniform horizontal soil layers with equivalent linear soil properties and by an underlying half-space layer.
Median soil profiles were defined with hazard-compatible soil properties based on those from the BFN PSHA report [11]. The soil properties include shear-wave velocity (VS),
compression-wave velocity (VP), corresponding damping (DS and DP), and unit weight.
These properties and values are provided by the PSHA for a range of hazard levels.
The hazard level of interest corresponding to reference earthquake used in SPRA fragility analysis for BFN is 10-5 AFE. The properties and values at AFE of 10-5 represent the hazard-consistent median soil profile for each structure based on its applicable FIRS. The soil layering profiles (i.e., layer thicknesses) for SSI analysis are refined from that of the PSHA to meet passing frequency requirements.
SSI analyses considered soil and structural property variation via use of Best Estimate (BE), Lower Bound (LB), and Upper Bound (UB) structure models and BE, LB, and UB soil models such that five analysis cases were developed: BEsoil-BEstructure, LBsoil-BEstructure, UBsoil-BEstructure, BEsoil-LBstructure, and BEsoil-UBstructure. Ground motion variability was considered via use of five independent sets of time histories (THs) for each analysis case. Soil properties for each layer of each variable soil case (BE, LB, and UB) were defined consistent with the results of the probabilistic site response analysis performed with the PSHA.
The following is a summary of the main steps.
: 1. Develop BE structural model. A cracking assessment of the structural model is performed at the selected hazard level of interest.
: 2. Develop median-centered SSI model by considering BE soil strain compatible soil properties.
: 3. Generate five sets of ground motions by spectrally matching five seed motions to the FIRS at the selected hazard level of interest.
: 4. Develop BE SSI models using the BE structural models with effective stiffness and damping consistent with 10-5 AFE hazard level ground motion.
: 5. Uniquely pair each set of ground motions with the five SSI models, generating a total of 25 SSI analyses.
: 6. Extract results from the 25 SSI analyses and generate results, including the median and 84% ISRS as well as their variability.
For each simulation, structural and soil properties were defined consistent with their response at a hazard level of interest selected via coordination with fragility and PRA analysts. This hazard level of interest was selected to be the 10-5 AFE hazard level based on insights from incremental risk quantifications, especially regarding the relative risk-significance of different acceleration intervals and individual components. A list of structures and description of relevant parameters are provided in Table 4.3-1.
Page 111 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                            December 2019 4.3.3 Structure Response Models The purpose of the mathematical models, which are the Finite Element Model (FEM) or the Lumped Mass Stick Model (LMSM), is to adequately determine the response of the structure in the frequency range of interest consistent with the seismic hazard. The mathematical models include structural elements that form the load-resisting system and appropriately represent the locations of mass and stiffness, thereby accounting for eccentric torsional effects. Dynamic analysis for both LMSMs and FEMs is performed in SASSI to capture structural response due to both horizontal and vertical motions.
The following subsections provide the modeling approach and general input properties used for the development of the FEMs and LMSMs.
4.3.3.1 Lumped Mass Stick Models The only LMSM used in the BFN SPRA is the NSSS as part of the RB FEM. Consistent with the design basis reference models, matrix elements were used in some parts of the NSSS LMSM. The NSSS 3D LMSM was developed by combining the horizontal and vertical 2D NSSS models. The 3D NSSS horizontal properties were selected by assigning the 2D NSSS horizontal mass and stiffness properties to both horizontal directions of the 3D model. Similarly, the vertical properties were selected by assigning the vertical 2D NSSS model properties to the vertical direction of the 3D model. Due to the symmetrical shape of the NSSS, a single 3D LMSM adequately represents the dynamic properties of the NSSS.
The LMSM of the NSSS meets or exceeds the seven criteria listed in Section 6.3.1 of the SPID [2] as minimum requirements, which are paraphrased as follows:
: 1. The structural models should be capable of capturing the overall structural responses for both the horizontal and vertical components of ground motion.
: 2. One combined model should be used if there is significant coupling between the horizontal and vertical responses.
: 3. The structural mass should be lumped so that the total mass, as well as the center of gravity (CG), is preserved.
: 4. The number of nodal or dynamic degrees of freedom should be sufficient to represent significant structural modes up to 20 Hz.
: 5. The torsional effects resulting from eccentricities between the CG and the center of rigidity (CR) should be included.
: 6. The multi-stick model should be used if the one-stick model is insufficient to represent the structure.
: 7. The in-plane floor flexibility (and subsequent amplified seismic response) should be captured appropriately for developing ISRS accurate up to 15 Hz.
Page 112 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 4.3.3.2 3D Finite Element Models As shown in Table 4.3-1, the RB, DGB, and IPS SSI models were developed using detailed 3D FEMs. Additionally, the substructure portions of all the models, regardless of being LMSM or 3D FEM, were developed using 3D solid or shell elements.
4.3.3.2.1 Reinforced Concrete Walls and Slabs Shell elements representing the floors were modeled at the center of the slab thickness.
However, for the foundation slabs that were modeled with shell elements, the shell elements were placed at the bottom of the slab to be consistent with the soil profile layering elevations and to maintain consistency with the FIRS definition elevations.
The walls were also explicitly modeled with shell elements. The walls were modeled from CG to CG of the slab shell elements. Openings in walls and slabs that were judged to not influence dynamic behavior were neglected. Typically, an opening smaller than about 10% of the wall is considered to have insignificant influence on the overall dynamic characteristics of the structure; therefore, these small openings could be neglected in the FEMs. Most of the floor slabs and walls were modeled with 4-node shell elements, although 3-node shell elements were used for mesh compatibility.
4.3.3.2.2 Foundations The RB and IPS foundation slabs were modeled with concrete shell elements, whereas the DGB foundations were modeled with 3D solid elements due to their significant thickness. This is because the DGB foundations house fuel tanks. The effect of the tanks inside of the DGB foundations is considered through proportionally reducing the elastic modulus of the DGB foundation material.
4.3.3.2.3 Concrete Block Walls The concrete block walls are considered to crack before the concrete walls and, therefore, not significantly contribute to the stiffness of the structural system. Therefore, the modulus of elasticity of these walls was considered as 10% of the value for concrete where they are explicitly modelled.
4.3.3.3 Structural Damping Material damping is considered using the guidance of Sections 3.1.2.2 of American Society of Civil Engineers (ASCE) 4-98 [34] as well as Section 3.2.2 of ASCE 4-16 [35],
consistent with the damping ratios used in other SPRAs.
Damping is a function of strain response (i.e., the larger the strain response, the larger the damping value). This is reflected in the ASCE 4-98 Table 3.1-1, which provides damping values for different response levels. For reinforced-concrete elements, the median damping ratios are 4% and 7% of critical damping for response level 1 and response level 2, respectively.
For steel structures, the median damping ratios are 2% and 4% of critical damping for response level 1 and response level 2, respectively, based on ASCE 4-98 Table 3.1-1.
Page 113 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 The justification for the response level used is provided in the SSI model documentation of the applicable structures [31].
For the reinforced-concrete shear walls in the median models, the response levels and corresponding damping ratios were selected based on the in-plane shear stress and out-of-plane bending stress of the wall. If the average shear and/or bending stresses in the walls at any given time step exceed the stress limits provided in Section C3.2.2 of ASCE 4-16 [35], response level 2 is considered and 7% damping ratio is assigned. If the average shear and/or bending stress in the walls at any given time step do not exceed the stress limits that are provided in Section C3.2.2 of ASCE 4-16, response level 1 is considered and 4% damping ratio is assigned. The concrete stress limits for response level and damping determination are 3f'C for shear and 7.5f'C for bending.
For reinforced-concrete slabs and beams, they are considered as cracked due to the addition of dead and live load bending stresses to the seismic bending stresses, and response level 2 (7% damping) is assigned.
For the steel beams and columns (both steel and concrete), response level 1 is considered without further investigation. This is because these members are secondary members, and the selection of their damping through detailed evaluation of the stress is not expected to significantly change the overall response.
4.3.3.4 Concrete cracking Consistent with the methodology described in ASCE 4-16 [35], the level of concrete cracking was considered based on the state of stress identified from the SSI model. Key lateral-load-resisting concrete features at each major elevation were checked for cracking through evaluation of in-plane shear stress. This assessment considered only seismic loads and did not consider gravity load or dead load. Therefore, no cracking assessment for bending stress was considered. The maximum (over the time series) average shear stresses at the base of the lateral load resisting systems were evaluated to calculate the BE demands.
The BE (median) stiffnesses of concrete structures were determined to be consistent with consistent with the stress state in the structure. This was accomplished by verifying that the stress state in the main load carrying elements (i.e., concrete shear walls) is consistent with the expected response level, as documented for each structure in its corresponding SSI model documentation. Determination of the effective stiffness of the reinforced-concrete members follows the guidance of ASCE 43-05 [36]. The adjustment of the stiffness is achieved by changing the cross-section properties (i.e., thicknesses) rather than the elastic and shear moduli. The changes in cross-section thicknesses are applied to the specific direction that is cracked (i.e., membrane vs. bending). The reduced section thicknesses were not considered in mass calculation and were only used in stiffness calculations.
Page 114 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                      December 2019 Table 4.3-1: Description of Structures and Analysis Methods for BFN SPRA Structure          Foundation      Type of Analysis Method         Comments/Other Information Condition       Model Reactor Building (RB) Rock               3D FEM      Multi-case        Shear-wave velocity 7,000 ft/sec deterministic SSI at foundation level; SSI analysis performed with incoherence, 5 soil-structure cases used (BE, UB, LB cases for structure and soil), 5 THs for each case.
Diesel Generator        Compacted        3D FEM      Multi-case        Average shear-wave velocity for Building (DGB)          earth                        deterministic SSI the structural backfill 740 ft/sec for backfill (12 ft)                               top ~50 ft. below grade and ~7,000 over crushed                                   ft/sec below it; SSI analysis stone backfill                                 performed with incoherence, 5 soil-(38 ft)                                        structure cases used (BE, UB, LB cases for structure and soil), 5 THs for each case.
Intake Pumping          Rock             3D FEM      Multi-case        Shear-wave velocity 7,000 ft/sec Station (IPS)                                        deterministic SSI at foundation level; SSI analysis performed with incoherence, 5 soil-structure cases used (BE, UB, LB cases for structure and soil), 5 THs for each case.
4.3.3.5 Structural Impact between Reactor Building and Turbine Building An investigation to evaluate the effect of the impact between Turbine Building (TB)
Unit 1 and RB Unit 1 through nonlinear building impact analysis was performed using impact elements. The analyses were not intended to provide refined results for local responses, but rather to provide an indication of the effects of such impact on general response of interest. The results show that the effect of the impact on the ISRS results in the RB is negligible. This is because the impacts occur at slow velocities and that the nature of the impacts is not impulsive. In other words, the TB does not bounce and does not apply an impulse load on RB; rather, it gets pushed by the RB at the onset of impact. On the other hand, since the TB foundation is very flexible and has very low stiffness, the impact between RB and TB significantly increases the acceleration response of the TB at even low frequencies since it gets pushed by the RB upon impact. However, the fragility of the RB structure considered the effects of the RB-TB impact. Detailed discussion is documented in [31].
4.3.4 Seismic Structure Response Analysis Technical Adequacy The BFN SPRA Seismic Structure Response and Soil-Structure Interaction Analysis
[31] were subjected to an independent peer review against the pertinent requirements in the PRA Standard. The seismic structure response and SSI were peer reviewed relative to Capability Category II for the full set of requirements in the PRA Standard. After completion of the subsequent independent assessment, the full set of requirements was met, and the seismic structure response and SSI were determined to be acceptable for use in the SPRA.
Page 115 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 The peer review assessment [6] and subsequent disposition of peer review findings through an independent assessment [16] are further described in Appendix A.
4.4    SSC Fragility Analysis The SSC seismic fragility analysis considers the impact of seismic events on the probability of SSC failures at a given value of a seismic motion parameter defined as PGA. The fragilities of the SSCs that participate in the SPRA accident sequences (i.e.,
those included on the SEL) were addressed in the model. Seismic fragilities for the significant risk contributors (i.e., those that have an important contribution to plant risk, are realistic, and plant-specific based on actual current conditions of the SSCs in the plant) were confirmed through the detailed walkdown of the plant.
This section summarizes the fragility analysis methodology and presents a tabulation of the fragilities with appropriate parameters for those SSCs determined to be sufficiently risk important based on the final SPRA quantification (as summarized in Section 5).
This section also discusses important assumptions and important sources of uncertainty, and any fragility-related insights identified.
4.4.1 SSC Screening Approach The BFN SEL, consisting of approximately 6,800 components combined for all three units, was reviewed, analyzed, and then reduced to about 1,600 components prior to walkdowns. The process of reducing the SEL is an iterative and multi-step process as summarized below.
First, the SEL provided to the Seismic Review Team (SRT) was reduced by removing components judged to be non-contributors to the overall response of the SPRA. It was identified that all components that are not in a Category I Building (not counting tanks in the Yard) are not contributors to the SPRA and can be screened as not necessary.
These components include anything not within the RB, DGB, IPS or Yard. No fragility value is required for these components.
Components that are judged inherently rugged were also screened out from needing a walkdown. These items included check valves, manual valves, throttle valves, small safety valves, small relief valves, solenoid valves, temperature elements, hand switches, small in-line strainers, and small in-line filters. These components are driven by the system they are mounted on as they are typically more rugged. Passive valves are small, lightweight, robust, and are typically mounted in line with piping. They do not need to change state during or after an event and have no external vulnerabilities. While the failure of one of these valves can contribute to the results of the SPRA, they are bound by the fragility of the distribution system to which they are attached. No fragility value is specifically developed for passive valves, but fragility for piping is developed.
Piping is walked by as part of the distribution system walkdown. This same methodology applies to filters. Dampers are made of robust steel and are typically thick in gauge compared to the duct system to which they are mounted. While they may have to change state after an earthquake, they do not need to change state during the seismic event. As was the case with passive valves, the fragility of the damper is driven Page 116 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 by the duct system to which they are mounted. While duct systems are walked down as part of the distribution system walkdown, it is understood that, in general, the failure mode of ducting is usually the supports of that duct. Ducts are either designed to handle tornado vacuum loads, which create more stress in the duct than earthquake loads, or are protected by tornado dampers. The dampers that were not in-line dampers were part of the SEL list. These include the fire dampers in the DG fan room, exhaust dampers in the DGB, the shutdown board room fire dampers in the Unit 1 and 2 RB, and fire dampers in Control Bay (CB) air conditioning room. These dampers were checked for interaction concern during the walkdowns.
The seismic walkdown team reviewed a sample of rugged components on the SEL and did not find any seismic interaction concerns that will invalidate rugged consideration.
These components were removed from the SEL by the SPRA systems analysts. Other components were considered to be less rugged but would still have sufficient capacity such that their failures would be unlikely to contribute significantly to the SCDF/SLERF in a SPRA. These components are retained on the SEL. The rugged components retained on the SEL were assigned high HCLPF capacity.
The components that reside inside other components are screened by the rule-of-the-box. Examples include level indicators inside tanks and switches inside a panel. These components are still addressed in the fragility analysis, but a walkdown of the box component is all that is necessary. These devices were modeled in the SPRA with the fragility value of their box assigned to them. It was assured that boxes containing devices are included in the SEL.
4.4.2 SSC Fragility Analysis Methodology For the BFN SPRA, the following methods were used to determine seismic fragilities for SSCs included in the SPRA:
Consistent with the requirements in the PRA Standard, the fragility analysis for the selected SSCs is based on the methodology in EPRI guidelines. The strategy for developing the fragilities for the complete set of SSCs on the SPRA SEL follows the recommendations of EPRI NP-6041-SL [24], EPRI TR-103959 [37],
EPRI TR-1019200 [38] and EPRI 3002000709 [18], EPRI 3002012994 [39], and proceeds progressively from using experienced-based capacities to component-specific evaluations. Regardless of the method, the development of fragility estimates uses plant-specific information based on SSC conditions, as confirmed through detailed walkdowns.
Components are first binned into equipment classes according to SQUG classes
[29] and then grouped according to similarity and location. Representative samples in each equipment group are then evaluated to obtain fragility estimates for all the items in the group.
The SPRA approach used at BFN initially utilized three quantifications. In addition to these formal quantifications, various sensitivity studies were performed during the effort to help identify important risk contributors. After each quantification and completion of the sensitivity studies, components identified as risk-significant were selected and Page 117 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                  December 2019 evaluated further to improve their calculated fragilities in order to reduce their risk significance. This approach has been successfully implemented at several plants and complies with the PRA Standard and the SPID. All three quantifications and numerous sensitivity studies were performed prior to the peer review. Subsequent to the peer review and to address peer review findings, additional quantifications were performed.
After each quantification, the results were reviewed to determine whether additional insights were obtained and to determine whether further refinement of fragilities associated with top risk contributors would improve the results and yield a more realistic model.
For the first quantification, site-specific representative fragilities (referred to as representative throughout) were typically developed by scaling from prior work performed for resolution of USI A-46, the Seismic IPEEE at BFN, and existing design basis calculations to account for available margins in the design. This is the margin between allowable values associated with design requirements and values associated with HCLPF evaluations. These margins were used to develop a Safety Factor, which is anchored to the PGA of the 1E-05 UHRS to estimate a HCLPF fragility value. The generic values of aleatory variability and epistemic uncertainty from the SPID were applied to the HCLPF to obtain the median fragility value.
For the second quantification, enhanced fragilities were provided for top risk contributors to both SCDF and SLERF. The top risk contributors were determined based on the F-V numbers from the initial quantification and subsequent sensitivity studies. The cutoff F-V value for selecting components from the first quantification was 5E-05 for both SCDF and SLERF. This is well below the threshold from the PRA Standard of 5E-03. The fragilities were calculated using the Conservative Deterministic Failure Model (CDFM) method to determine the HCLPF. The generic uncertainty values, as recommended in Table 6.2 of the SPID for various SSCs, were used to estimate the median fragility value, with the generic uncertainty values adjusted if needed to account for specific conditions. Site-specific information obtained from walkdowns and plant documentation, including actual anchorage and configuration details, were used along with ISRS at the location of the individual components.
Fragilities for the third quantification were developed for the dominant risk contributors (components with F-V greater than 5E-03) as identified during the second SPRA quantification. When beneficial, the fragilities for the final quantification were computed using the Separation of Variable (SoV) approach, where the median capacity and the associated variabilities are calculated rigorously, and then the HCLPF capacity is back-calculated using the median capacity and the variabilities. The SoV approach provides more realistic fragilities.
Critical failure modes, such as structure/anchorage or functionality or block wall, were identified and fragility calculations were performed for the median capacity Am for each of the failure modes. The lowest, governing Am was selected, and when two or more failure modes were close (i.e., their median capacities within 20% of each other), the governing median capacity was computed for combined failure.
Subsequent to the peer review, additional quantifications were performed to further refine the SPRA model and to respond to peer review findings. These quantifications Page 118 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 are described in Section 5 of this report. To support these quantifications, additional refined fragilities were developed using either the CDFM or SoV approach as appropriate. Overall, fragilities for three structures, 124 electrical and mechanical components, 29 block walls and all relays were refined following the Hybrid method, and 11 electrical components, 23 block walls and 20 relays were refined following the SoV approach. The representative fragilities are retained for the SSCs, block walls, and relays that are determined to be non-dominant risk contributors. In some cases, refined fragilities were provided for certain SSCs for use in various sensitivity studies. These refined fragilities were developed based on estimates and maximum potential improvements to determine the impact and benefit of developing more detailed fragilities for these items based on the results of the sensitivity studies. No components were screened out based on capacity, other than the inherently rugged components.
4.4.3 SSC Fragility Analysis Results and Insights The final set of fragilities for the risk-important contributors to SCDF and SLERF are summarized in Section 5. Refer to Tables 5.4-4 through 5.4-6 for SCDF and Tables 5.5-4 and 5.5-6 for SLERF. Detailed (SoV) calculations have generally been performed for the highest risk-significant SSCs, as well as for selected other components.
Consistent with the three-step graded approach for risk quantification, components for refinement were selected based on interim sensitivity studies and previously completed risk quantifications. The fragilities of selected components that were identified to be risk-significant were previously refined using the CDFM-based Hybrid Method, and several were refined using the SoV Approach. Using the refined fragilities in the subsequent risk quantifications resulted in either the refined fragility group becoming less risk-significant or new fragility groups (with CDFM-based fragility) becoming more risk-significant.
According to Section 6.4.1, EPRI SPID, The CDFM approach for developing fragilities is a simpler method that can be performed consistently by more analysts and is an acceptable approach for generating fragilities within an SPRA for the majority of components for which a less detailed assessment is necessary. Because only a handful of components are risk-significant enough to justify the additional effort required by the separation of variables method, the CDFM method can provide efficiencies in the overall effort. Therefore, use of the CDFM approach is useful and beneficial for calculating fragilities of SSCs for use in seismic PRAs conducted to address the 50.54(f) letter.
After the final risk quantification, as previously described, many of the SSCs with refined fragilities based on the SoV approach dropped off the risk-significant list, and other SSCs with refined fragilities based on CDFM approach appeared on the risk-significant list. Sensitivity studies were conducted after the final risk quantification by varying the fragilities of risk-significant SSCs to ensure that the overall risk profile remains stable.
Details of the sensitivity studies are provided in Section 5.7.
Page 119 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 4.4.4 SSC Fragility Analysis Technical Adequacy The BFN SPRA SSC Fragility Analysis [40] was subjected to an independent peer review against the pertinent requirements in the PRA Standard. The SSC fragility analysis was peer reviewed relative to Capability Category II for the full set of SRs in the PRA Standard [7]. After completion of the subsequent independent assessment [16],
the full set of SRs were met, and the SSC fragility analysis was determined to be acceptable for use in the SPRA.
The peer review assessment [6], and subsequent disposition and closure of peer review findings through an independent assessment [16], are further described in Appendix A.
Page 120 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                           December 2019 5.0 Plant Seismic Logic Model The seismic plant response analysis models the various combinations of structural, equipment, and human failures given the occurrence of a seismic event that could initiate and propagate a seismic core damage or large early release sequence. This model was quantified to determine the overall SCDF and SLERF and to identify the important contributors, e.g., important accident sequences, SSC failures, and human actions. The quantification process also includes an evaluation of sources of uncertainty and provides a perspective on how such sources of uncertainty affect SPRA insights.
5.1    Development of the SPRA Plant Seismic Logic Model The BFN seismic response model was developed by starting with the BFN internal events at-power Level 1/Level 2 PRA model of record as of February 2018 [62], and then adapting the model in accordance with guidance in the SPID and the PRA Standard. This includes the addition of seismic initiating events (IEs) based on the plant-specific seismic hazard curve and seismic fragility-related basic events to the appropriate portions of the IEPRA, eliminating some parts of the internal events model that do not apply, and adjusting the IEPRA model HRA to account for response during and following a seismic event. This modeling approach leaves the IEPRA system logic intact while incorporating the necessary additions required for the SPRA.
The BFN SPRA model was developed using the EPRI Risk and Reliability Workstation software suite (CAFTA, FRANX, HRA Calculator, ACUBE, SYSIMP and UNCERT).
Both random and seismic-induced failures of modeled SSCs were included. Seismic-induced fire and flooding were also evaluated.
5.1.1 Seismic Initiating Event The seismic IE was modeled using nine discrete hazard bins based on increasing PGA.
The seismic hazard bins are listed in Table 5.1-1. Each bin is treated as a seismic initiator, and the SCDF and SLERF results are summed over all the bins to obtain the total SCDF and SLERF.
The bin ranges were chosen such that the first bin covers the PGA range from the Operating Basis Earthquake (OBE) to the Safe Shutdown Earthquake (SSE), while the second covers the range from the SSE to a common Review Level Earthquake (RLE) from the IPEEE (0.3g).
The OBE, the strongest earthquake at which the plant is designed to be able to continue normal operation, is defined as 0.1g. Below 0.1g, no significant seismic impacts are expected. The SSE is defined as an acceleration of 0.2g. The plant is seismically designed such that safety-related equipment should not fail given an SSE.
Page 121 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                    December 2019 Table 5.1-1: Seismic Hazard Bins Lower        Upper      Bin Mean           Bin Mean Seismic Bound        Bound         PGA            Frequency             Notes Bin (g)         (g)          (g)               (1/y)
      %G01        0.1          0.2         0.14            2.68E-04          OBE to SSE
      %G02        0.2          0.3         0.25            5.56E-05        SSE to 0.3g RLE
      %G03        0.3          0.6         0.42            2.98E-05
      %G04        0.6          0.7         0.65            2.17E-06
      %G05        0.7          0.9         0.79            2.19E-06
      %G06        0.9          1.5         1.16            1.82E-06
      %G07        1.5          1.7         1.60            1.80E-07
      %G08        1.7          3.0         2.26            3.32E-07
      %G09        3.0                       3.3            5.96E-08         Unbounded bin Total=3.6E-04 Note: For %G09, FRANX calculates the representative ground motion as the addition of 10% to the lowest PGA of the bin, 1.1*3.0g = 3.3g.
5.1.2 Accident Sequences The IEPRA uses event trees (ETs) to model the potential plant responses to IEs. The SPRA uses the same approach. The SPRA uses a seismic initiating event tree (SIET) to partition the seismic IE into accident sequence types typically modeled in the IEPRA.
Transfers can then be made from the SIET to the corresponding IEPRA ETs to model plant response.
The SIET top events include the recommended minimum set of IEs listed in NUREG/CR-4840 [41] except for the initial status of the power conversion system. No credit is taken for non-safety-related equipment such as the power conversion system in the BFN SPRA base case.
An additional top event involving seismically induced direct core damage is included in the SIET. The sequence leads directly to core damage and, therefore, does not transfer to an IEPRA ET. Structural failures of the RB, CB, IPS are assumed to lead directly to core damage. In addition, structural failure of the DGB combined with a loss of offsite power (LOOP) is assumed to lead directly to core damage. Reactor vessel ruptures or other excessive loss-of-coolant accidents (LOCAs), and structural support failures of the reactor pressure vessel are assumed to lead directly to core damage. Finally, seismic failure of the control room ceiling resulting in operator abandonment and failure to shut down the plant remotely is assumed to lead to core damage.
5.1.3 Loss of Offsite Power The fragility of seismically induced LOOP resulting from switchyard or grid failures was obtained from Table 6-1 in NUREG/CR-6544 [42]. Seismic-induced LOOP is predicted to occur with a median ground acceleration of 0.3g. The predicted failure mode is failure of ceramic insulators in the switchyard. Use of this fragility for seismically induced Page 122 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 LOOP is a standard industry practice for plants in the eastern portion of the U.S. The path for transmission of offsite power to safety-related equipment and non-safety-related equipment within the plant was considered to be governed by the fragility for seismically induced offsite power, including any paths through the TB. Note that seismically induced LOOP is assumed to fail both switchyards (complete seismic correlation). The SPRA takes no credit for recovery of offsite power.
5.1.4 Very Small LOCA (VSLOCA)
SPRAs need to consider whether a coincident VSLOCA needs to be modeled for other SIET sequences. For other LOCA sequences, which are small LOCA (SLOCA), medium LOCA (MLOCA), large LOCA (LLOCA) and interfacing system LOCA (ISLOCA), the addition of a coincident VSLOCA would have no impact because the other LOCAs modeled are already larger than a VSLOCA. Also, the direct core damage events modeled are not impacted by a VSLOCA because they are assumed to go directly to core damage and early release. Inclusion of a coincident VSLOCA might potentially impact accident progression and success for SIET sequences of general transient (GTRAN), break inside containment, and break outside containment. However, the plant-specific fragility analysis determined that the seismic fragility for the VSLOCA was high (Am=3.68 g).
5.1.5 SLERF Analysis The seismic Level 2 PRA analysis includes an accident event progression following core melt that is similar to the event progression initiated by an internal events initiator.
The BFN IEPRA developed an SLERF model consisting of containment event trees (CETs) and supporting fault trees. The SPRA used the SLERF model and incorporated the impact of seismic events into it.
The process of performing the containment analysis begins with an evaluation of the BFN SPRA Level 1 sequences. These sequences are categorized in terms of the type of challenge to containment posed by each sequence and the operability of systems that could mitigate these effects. These Plant Damage States (PDSs) are used to assist in the linking of seismic Level 1 sequences to the appropriate SLERF sequences. While each seismic Level 1 accident sequence is explicitly treated in the CAFTA computer model of the BFN plant, the Level 1 sequence logic is transferred into the CETs to take advantage of the similarities in accident challenges from the Level 1 analysis and to streamline the quantification of the core melt progression CETs. The PDS grouping and the CETs identify the general course of the accident sequence, including which systems are operating and the specific phenomena that may occur.
Development of the SLERF model for the seismic sequences was performed in the same manner as for the IEPRA. In the IEPRA, each sequence in the ET that results in core damage is sufficiently subdivided to indicate the type of event, the state of the primary system, and the state of containment protection systems. Each Level 1 end state has a designated PDS. The state of these PDSs determines which CET the sequence is input into in the SLERF analysis. For sequences that can result in a large Page 123 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 early release, the SLERF or the Conditional Large Early Release Probability (CLERP) can be determined.
5.1.6 Summary of Resulting Correlated Component Groupings Correlation of components (or common cause failure (CCF)) is considered in accordance with the PRA Standard. There are insufficient data on partial or full correlation of seismic failures of similar components in similar locations and alignments to perform sophisticated seismic correlations in SPRAs. Instead, a common practice is to assume complete seismic correlation for these groups of similar components, locations, and alignments. The BFN SPRA results involve complete seismic correlation within fragility groups.
5.1.7 Summary of HRA methodology Operator actions that are modeled in the SPRA are either pre-initiator or post-initiator.
Pre-initiator Human Failure Events (HFEs) are events that represent the impact of human failures committed prior to the initiation of an accident sequence (e.g., during test or maintenance or the use of calibration procedures). Pre-initiator actions are latent and not affected by seismic events, so their assessments are not changed from the IEPRA model.
The list of post-initiator human actions for the internal events model is the starting point of the seismic HRA, and all existing HRAs are analyzed for modification due to seismic effects. The HFEs associated with the existing accident sequence models were retained in the SPRA model. The model was also examined for any potential human actions unique to the seismic analysis, and any new operator actions identified were added to the SPRA. Any new operator actions added to the seismic model are discussed further in the SPRA HRA Notebook [43].
Since the potential earthquakes examined vary in magnitude, as does the on-site acceleration, the level of plant damage varies accordingly due to the impacts of the different seismic events. Post-initiator HFEs retained in the SPRA model were evaluated for seismic impacts. The degree of impact is dependent on the seismic acceleration level. The seismic impacts on every post-initiator HFE in the SPRA models were accounted for by the HFE-specific performance shaping factors and selected minimal values that increase with acceleration as a function of the PDS. Following the EPRI SPRA guideline [18], the seismically adjusted HFEs use the internal events HFE nomenclature, with a suffix of _Sn, where n ranges from 1 to 4; i.e., four separate seismic acceleration ranges were evaluated for varying seismic impacts. The SPRA HRA Notebook discusses which HFE bins correspond to which seismic acceleration levels. For bin S4 (which includes the highest acceleration seismic initiators), it was conservatively assumed that all post-trip actions are set to failed (1.0).
The use of the same method from the internal events model for the HRA dependency analysis is valid for the SPRA HRA. The SPRA HRA Notebook discusses the method used to assess HFE dependency. The SPRA Quantification Notebook [44] also provides details of how the HRA dependency analysis was performed.
Page 124 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 Accessibility for HFEs performed outside the control room was addressed by walkdowns.
5.1.8 Seismic-Fire Seismic-fire interaction events have the potential to contribute significantly to core damage or large early release. The guidelines in Appendix G of EPRI 3002000709 [18]
were followed in the identification and assessment of potential seismic-fire interaction events. That effort included an assessment of fire ignition sources categorized as medium or higher and additional sources identified in the IPEEE and Final Safety Analysis Report (FSAR). The seismic-fire interaction sources and the affected SSCs were walked down by a team of fragility and PRA engineers. The walkdown findings are documented in the walkdown report [30].
Seismic-fire interaction event identification and assessment is included in the BFN SPRA Seismic-Fire Interaction evaluation [46]. This evaluation was identified as a Best Practice by the peer review team. The implementation of the EPRI 3002012980 [47]
process for seismic-fire was comprehensive and complete. Every fire source was considered and clearly dispositioned. Qualitative screening was carefully applied.
Quantitative screening was done correctly and clearly. Everything was documented in a clear, comprehensive, and traceable manner.
The results of the evaluation indicated no seismically induced fire events need to be included in the SPRA.
5.1.9 Seismic-Flood Seismic-flood interaction events have the potential to contribute significantly to core damage or large early release. A two-step process was used to identify such events at BFN. The first step was to review internal flood scenarios modeled in the internal flooding portion of the IEPRA [48]. All scenarios from the IEPRA were identified and were subject to further evaluation, including the scenarios that were screened out in the internal events model. The screening in the IEPRA is based on the frequency (SCDF or SLERF), which is the product of the pipe-break frequency and the conditional core damage probability (CCDP) or CLERP. For a seismic event, the pipe-break frequencies directly depend on the occurrence frequency of a certain level of earthquake and the pipe fragility. The flooding scenarios that were screened out due to the very low internal event pipe-break frequency may have considerably high CCDP or CLERP and become seismically risk-significant in combination with the potential seismic failure of other equipment. Therefore, all internal flooding scenarios, screened or not, are included in the seismic-flood interaction evaluation. Second, the scenario with the highest CCDP was then chosen as the seismic flooding scenario for a flooding source that may be from various piping or tanks. The seismic-flood interaction sources and the affected SSCs were walked down by a team of fragility and PRA engineers. The walkdown findings are documented in the walkdown report [30].
There are three types of scenarios in the IEPRA internal flooding analysis: spray, flood, and major flood. The major flood scenario typically has the highest failure probability of Page 125 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                 December 2019 the three and is kept for the seismic-flood interaction model. The seismic flooding model is built in FRANX, and no operator flooding recovery actions are credited for the seismic-induced flooding. Finally, the seismic-flooding model was injected into the CAFTA fault tree via the FRANX tool XINIT for the BFN SPRA quantification.
5.2   SPRA Plant Seismic Logic Model Technical Adequacy The BFN SPRA seismic plant response methodology and analysis were subjected to an independent peer review against the pertinent requirements in the PRA Standard [7].
The seismic plant response methodology and analysis were peer reviewed relative to Capability Category II for the full set of SRs in the PRA Standard. After completion of the subsequent independent assessment, the full set of SRs was met, and the seismic plant response methodology and analysis were determined acceptable for use in the SPRA.
The peer review assessment [6] and subsequent disposition of peer review findings through an independent closure peer review assessment [16] are further described in Appendix A.
5.3   Seismic Risk Quantification In the SPRA risk quantification, the seismic hazard is integrated with the seismic response analysis model to calculate the frequencies of core damage and large early release of radioactivity to the environment. This section describes the SPRA quantification methodology and important modeling assumptions.
5.3.1 SPRA Quantification Methodology Several ACCESS tables within FRANX were used to define the seismic hazard bins, assign seismic fragilities to basic events with the logic model, calculate fragilities associated with each of the seismic hazard bins, and assign a Human Error Probability (HEP) by seismic bin for each HFE. Note that the seismic HRA module with FRANX 4.2 was not used to determine the HEPs because that module is presently inconsistent with seismic HRA guidelines presented in the EPRI report [18]. The module was developed based on an earlier draft of that report.
The following steps were used to perform the SPRA model quantification for both SCDF and SLERF for each unit:
(1)    Obtain CCDP or CLERP cutsets for each seismic bin using FRANX 4.2 and ACUBE with initial fragility and HEP values and generally assuming complete seismic correlation within fragility groups.
(2)    Identify fragilities and HEPs to be refined.
(3)    Refine fragility groups for complete seismic correlation modeling.
(4)    Identify final set of fragilities to be inserted into the model (because of model size limitations and software constraints).
Page 126 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 (5)  Perform truncation sensitivity to determine final truncation level; quantify the models (six FRANX files) with initial HEP values.
(6)  Assemble bin cutsets into combined cutset files (one for SCDF and one for SLERF for each unit).
(7)  Perform HFE detailed HRA analysis and HFE dependency analysis; incorporate new HRA values into the model.
(8)  Finalize quantification of SCDF and SLERF (ACUBE analysis).
(9)  Evaluate basic event importances (SYSIMP/ACUBE analysis supplemented by selected sensitivity analyses).
(10) Perform uncertainty analysis (UNCERT).
(11) Evaluate sensitivity cases.
Specific issues related to quantification are discussed in the following sections addressing SCDF and SLERF results.
5.3.2 SPRA Model and Quantification Assumptions Hazard/fragilities/structures analyses assumptions:
: 1. Refer to Section 3 of this submittal for a discussion of assumptions and uncertainties associated with the hazard analysis.
: 2. Refer to Section 4 of this submittal for a discussion of assumptions and uncertainties associated with the fragilities/structures analyses.
Key plant response modeling assumptions:
: 1. Structural failures of the RB, CB, or IPS, or DGB combined with LOOP, are assumed to fail sufficient equipment within the structure to lead directly to core damage and large early release.
: 2. In addition to these large structure failures, seismic failures of the reactor vessel and its supports are also considered to lead directly to core damage and large early release.
: 3. Finally, the combination of a seismically induced failure of the control room (ceiling collapse) and failure of the operators to safely shut down the plant remotely is also assumed to lead directly to core damage and large early release.
These are potentially conservative assumptions.
5.4   SCDF Results 5.4.1 Overall SCDF The SPRA shows that the point estimate SCDF is 6.30E-06 per reactor year (/ry) for Unit 1, 6.40E-06 /ry for Unit 2, and 7.13E-06 /ry for Unit 3.
Page 127 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                  December 2019 5.4.2 SCDF as a Function of Hazard Interval A summary of the SCDF results for each seismic hazard interval is presented in Table 5.4-1 for Unit 1 SCDF, Table 5.4-2 for Unit 2 SCDF, and Table 5.4-3 for Unit 3 SCDF.
Table 5.4-1: Unit 1 SCDF Contribution by Initiating Event Earthquake                          Percent Truncation    Scenario        Description                         CCDP         SCDF Frequency                         Contribution 1.0E-12      %G01      Seismic Initiating Event       2.68E-04 7.9E-04     2.11E-07       3.4%
(0.1g to <0.2g) 1.0E-12      %G02      Seismic Initiating Event       5.56E-05 9.0E-03     5.03E-07       8.0%
(0.2g to <0.3g) 1.0E-11      %G03      Seismic Initiating Event       2.98E-05 4.8E-02     1.44E-06      22.8%
(0.3g to <0.6g) 1.0E-11      %G04      Seismic Initiating Event       2.17E-06 3.3E-01     7.15E-07      11.3%
(0.6g to <0.7g) 3.0E-10      %G05      Seismic Initiating Event       2.19E-06 5.9E-01     1.29E-06      20.5%
(0.7g to <0.9g) 3.0E-08      %G06      Seismic Initiating Event       1.82E-06 8.9E-01     1.63E-06      25.8%
(0.9g to <1.5g) 3.0E-08      %G07      Seismic Initiating Event       1.80E-07 9.0E-01     1.62E-07       2.6%
(1.5g to <1.7g) 2.0E-08      %G08      Seismic Initiating Event       3.32E-07 9.1E-01     3.03E-07       4.8%
(1.7g to <3g) 2.0E-08      %G09      Seismic Initiating Event       5.96E-08 9.2E-01     5.48E-08       0.9%
(>3g)
Total SCDF=              6.30E-06 Page 128 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                  December 2019 Table 5.4-2: Unit 2 SCDF Contribution by Initiating Event Earthquake                          Percent Truncation    Scenario         Description                         CCDP         SCDF Frequency                         Contribution 1.0E-12      %G01       Seismic Initiating Event      2.68E-04  8.4E-04     2.26E-07       3.5%
(0.1g to <0.2g) 1.0E-12      %G02       Seismic Initiating Event      5.56E-05  9.7E-03     5.42E-07       8.5%
(0.2g to <0.3g) 1.0E-11      %G03       Seismic Initiating Event      2.98E-05  5.1E-02     1.52E-06      23.7%
(0.3g to <0.6g) 2.0E-11      %G04       Seismic Initiating Event      2.17E-06  3.3E-01     7.10E-07      11.1%
(0.6g to <0.7g) 4.0E-10      %G05       Seismic Initiating Event      2.19E-06  5.7E-01     1.26E-06      19.6%
(0.7g to <0.9g) 3.0E-08      %G06       Seismic Initiating Event      1.82E-06  8.9E-01     1.63E-06      25.4%
(0.9g to <1.5g) 3.0E-08      %G07       Seismic Initiating Event      1.80E-07  9.0E-01     1.62E-07       2.5%
(1.5g to <1.7g) 2.0E-08      %G08       Seismic Initiating Event      3.32E-07  9.1E-01     3.03E-07       4.7%
(1.7g to <3g) 2.0E-08      %G09       Seismic Initiating Event      5.96E-08  9.2E-01     5.48E-08       0.9%
(>3g)
Total SCDF=              6.40E-06 Page 129 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                  December 2019 Table 5.4-3: Unit 3 SCDF Contribution by Initiating Event Earthquake                          Percent Truncation    Scenario         Description                         CCDP         SCDF Frequency                         Contribution 1.0E-12      %G01       Seismic Initiating Event      2.68E-04  9.8E-04     2.62E-07       4.1%
(0.1g to <0.2g) 1.0E-12      %G02       Seismic Initiating Event      5.56E-05  1.0E-02     5.74E-07       9.0%
(0.2g to <0.3g) 1.0E-11      %G03       Seismic Initiating Event      2.98E-05  6.2E-02     1.85E-06      28.9%
(0.3g to <0.6g) 2.0E-11      %G04       Seismic Initiating Event      2.17E-06  3.9E-01     8.55E-07      13.4%
(0.6g to <0.7g) 7.0E-07      %G05       Seismic Initiating Event      2.19E-06  6.5E-01     1.42E-06      22.2%
(0.7g to <0.9g) 4.0E-08      %G06       Seismic Initiating Event      1.82E-06  9.0E-01     1.64E-06      25.7%
(0.9g to <1.5g) 4.0E-08      %G07       Seismic Initiating Event      1.80E-07  9.0E-01     1.62E-07       2.5%
(1.5g to <1.7g) 2.0E-08      %G08       Seismic Initiating Event      3.32E-07  9.1E-01     3.03E-07       4.7%
(1.7g to <3g) 2.0E-08      %G09       Seismic Initiating Event      5.96E-08  9.2E-01     5.48E-08       0.9%
(>3g)
Total SCDF=              7.13E-06 5.4.3 Fragility Group Importance for SCDF The SSCs with the most significant seismic failure contributions to SCDF for Unit 1 are listed in Table 5.4-4, sorted by F-V. The seismic fragilities for each of the significant contributors are also provided in Table 5.4-4, along with the corresponding limiting seismic failure mode and method of fragility calculation. The corresponding measures for Unit 2 and Unit 3 are presented in Tables 5.4-5 and 5.4-6.
Page 130 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                   December 2019 Table 5.4-4: Unit 1 SCDF Fragility Group Importance Measures Ranked by F-V Am                                 Fragility Fragility Group                    Description                   F-V             u   r   Failure Mode (g)                                 Method SEIS_LOOP       LOOP (Loss of Offsite Power)                    0.885 0.30 0.45 0.30          Ceramic          Table 6-1 insulators      NUREG/CR-6544 SEIS_12-1P-1    RHRSW pumps based on pipe frag (Pipe            0.056      2.45 0.61 0.47   Soil (Buried         SoV Calc)                                                                     Piping) Failure SEIS_12-1P-2    Emergency Equipment Cooling Water               0.053      2.48 0.60 0.48   Soil (Buried         SoV (EECW) pumps based on pipe frag calc                                      Piping) Failure SEIS_2-1-1      Unit Batteries (Fragility Group 15-03)          0.047      1.32 0.38 0.24   Anchorage      Representative SEIS_1B-2-1a    480v BD 219 A, B (U1, U2) (Fragility Group      0.033      1.95 0.56 0.22   Anchorage            SoV 01-05-01) Anchorage Failure SEIS_1C-4R8     3EA,3EC SDBD Common Accident Signal             0.022      0.63 0.32 0.24  Functionality        CDFM Relays (CASA)
SEIS_1B-2-1f    480v BD 219 A, B (U1, U2) (Fragility Group      0.022      2.61 0.64 0.24  Functionality         SoV 01-05-01) Functional Failure SEIS_12-1b      RHRSW pumps and EECW Alternate                  0.017      1.45 0.32 0.24   Anchorage           CDFM (Fragility Group 06-03)
SEIS_5-2B       Initiation relays and panels                    0.016      0.97 0.32 0.24   Anchorage           CDFM SEIS_3-2        Batt CH 248-1 (Fragility Group 16-06)           0.015      1.47 0.32 0.24  Functionality        CDFM SEIS_11-1R3     Relay group 3 for group SEIS_11-1 (EECW         0.013      0.61 0.32 0.24  Functionality        CDFM Pp A3 UV device)
SEIS_11-1R1-1   Relay group 1-1 for group SEIS_11-1 (EECW       0.011      1.31 0.37 0.29  Functionality         SoV Pp B3 UV device)
SEIS_BLD-IPS    Intake Pumping Station                          0.010      1.90 0.26 0.24    Structural         CDFM Analysis SEIS_1C-        Wall 28 falls towards 480v SD BD 1A and 1B      0.010      2.18 0.54 0.27   Block Wall           SoV 1W28SD                                                                                         Failure Page 131 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                           December 2019 Table 5.4-4: Unit 1 SCDF Fragility Group Importance Measures Ranked by F-V Am                               Fragility Fragility Group                              Description                          F-V             u   r  Failure Mode (g)                              Method SEIS_4-5               4160-480V TRANSFORMER, CONTROL                            0.008 0.60 0.38 0.24      Functionality  Representative BAY (Fragility Group 4-12)
SEIS_2-1W8             Battery and Wall 8 or 47 interface (Block Wall            0.007      2.12 0.41 0.29  Block Wall          SoV Group 5)                                                                               Failure SEIS_11-1R2            Relay group 2 for group SEIS_11-1 (EECW                   0.007      1.29 0.38 0.24 Functionality       CDFM Pp B3&D3 OC device)
SEIS_1B-1              480V BD 1A, 2A, 3A, 1B, 2B, 3B (Fragility                 0.007      1.70 0.32 0.24 Functionality       CDFM Group 02-01)
SEIS_1C-3R6            B,D SDBD 480 Transformer Trip relays (50G)                0.007      1.32 0.32 0.24 Functionality       CDFM SEIS_11-1R4            Relay group 4 for group SEIS_11-1 (EECW                   0.005      0.94 0.51 0.27 Functionality        SoV Pp A3 OC device)
SEIS_1C-4R4            3EA,3EC SDBD Lockout relays (86)                          0.005      0.94 0.32 0.24 Functionality       CDFM SEIS_1C-4R5-2          3EA,3EC SDBD Lockout relays (86)                          0.005      0.94 0.32 0.24 Functionality       CDFM SEIS_HINST             Seismic failure of Main Control Room                      0.005      1.96 0.24 0.32 Functionality       CDFM instrumentation

See Section 5.7 for additional discussion on these representative fragilities Page 132 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                   December 2019 Table 5.4-5: Unit 2 SCDF Importance Measures Ranked by F-V Am                                  Fragility Fragility Group                   Description                   F-V             u   r   Failure Mode (g)                                  Method SEIS_LOOP       LOOP (Loss of Offsite Power)                   0.885      0.3  0.45 0.30      Ceramic          Table 6-1 insulators      NUREG/CR-6544 SEIS_12-1P-1    RHRSW pumps based on pipe frag (Pipe           0.056      2.45 0.61 0.47    Soil (Buried         SoV Calc)                                                                     Piping) Failure SEIS_12-1P-2    EECW pumps based on pipe frag calc             0.053      2.48 0.60 0.48    Soil (Buried         SoV Piping) Failure SEIS_2-1-1      Unit Batteries (Fragility Group 15-03)         0.047      1.32 0.38 0.24    Anchorage      Representative SEIS_1B-2-1a    480v BD 219 A, B (U1, U2) (Fragility Group     0.033      1.95 0.56 0.22    Anchorage            SoV 01-05-01) Anchorage Failure SEIS_1B-2-1f    480v BD 219 A, B (U1, U2) (Fragility Group     0.022      2.61 0.64 0.24   Functionality         SoV 01-05-01) Functional Failure SEIS_1C-4R8     3EA,3EC SDBD Common Accident Signal            0.019      0.63 0.32 0.24   Functionality        CDFM Relays (CASA)
SEIS_12-1b      RHRSW pumps and EECW Alternate                 0.017      1.45 0.32 0.24    Anchorage           CDFM (Fragility Group 06-03)
SEIS_3-2        Batt CH 248-1 (Fragility Group 16-06)          0.015      1.47 0.32 0.24   Functionality        CDFM SEIS_11-1R3     Relay group 3 for group SEIS_11-1 (EECW        0.015      0.61 0.32 0.24   Functionality        CDFM Pp A3 UV device)
SEIS_1A-4W64    250v DC Bus A interface with wall 64 (Block    0.013      2.18 0.54 0.27    Block Wall           SoV Wall Group 3)                                                                  Failure SEIS_BLD-IPS    Intake Pumping Station                         0.010      1.90 0.26 0.24     Structural         CDFM Analysis SEIS_1C-        Wall 63 falls towards 480v SD BD 2A and 2B     0.010      2.18 0.54 0.27    Block Wall           SoV 1W63SD                                                                                         Failure Page 133 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                           December 2019 Table 5.4-5: Unit 2 SCDF Importance Measures Ranked by F-V Am                               Fragility Fragility Group                              Description                          F-V             u   r  Failure Mode (g)                              Method SEIS_11-1R1-1          Relay group 1-1 for group SEIS_11-1 (EECW                 0.009      1.31 0.37 0.29 Functionality        SoV Pp B3 UV device)
SEIS_5-2B              Initiation relays and panels                              0.009      0.97 0.32 0.24  Anchorage          CDFM SEIS_2-1W8             Battery and Wall 8 or 47 interface (Block Wall            0.007      2.12 0.41 0.29  Block Wall          SoV Group 5)                                                                               Failure SEIS_1B-1              480V BD 1A, 2A, 3A, 1B, 2B, 3B (Fragility                 0.007      1.70 0.32 0.24 Functionality       CDFM Group 02-01)
SEIS_11-1R2            Relay group 2 for group SEIS_11-1 (EECW                   0.006      1.29 0.38 0.24 Functionality       CDFM Pp B3&D3 OC device)
SEIS_11-1R4            Relay group 4 for group SEIS_11-1 (EECW                   0.006      0.94 0.51 0.27 Functionality        SoV Pp A3 OC device)
SEIS_1C-               Wall 28 falls towards 480v SD BD 1A and 1B                0.006      2.18 0.54 0.27  Block Wall          SoV 1W28SD                                                                                                        Failure SEIS_14-1R1-2          Relays for Group SEIS_14-1, Relay Group 1                 0.006      1.62 0.32 0.24 Functionality       CDFM (HPCI/RCIC Isolations)
SEIS_HINST             Seismic failure of Main Control Room                      0.005      1.96 0.24 0.32 Functionality       CDFM instrumentation

See Section 5.7 for additional discussion on these representative fragilities Page 134 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                    December 2019 Table 5.4-6: Unit 3 SCDF Importance Measures Ranked by F-V Am                                  Fragility Fragility Group                    Description                   F-V             u   r   Failure Mode (g)                                  Method SEIS_LOOP       LOOP (Loss of Offsite Power)                    0.900 0.30 0.45 0.30           Ceramic          Table 6-1 insulators      NUREG/CR-6544 SEIS_1B-2-2     480v BD 219 3EA, 3EB (U3) (Fragility Group      0.049      1.14 0.32 0.24    Anchorage           CDFM 01-05-02)
SEIS_12-1P-1    RHRSW pumps based on pipe frag (Pipe            0.046      2.45 0.61 0.47    Soil (Buried         SoV Calc)                                                                      Piping) Failure SEIS_12-1P-2    EECW pumps based on pipe frag calc              0.044      2.48 0.60 0.48    Soil (Buried         SoV Piping) Failure SEIS_2-1-1      Unit Batteries (Fragility Group 15-03)          0.036      1.32 0.38 0.24    Anchorage      Representative SEIS_1C-4R8     3EA,3EC SDBD Common Accident Signal             0.032      0.63 0.32 0.24   Functionality        CDFM Relays (CASA)
SEIS_1B-2-1a    480v BD 219 A, B (U1, U2) (Fragility Group      0.026      1.95 0.56 0.22    Anchorage            SoV 01-05-01) Anchorage Failure SEIS_1C-4       U3 4kv SD BD EA and EC (Fragility Group         0.018      1.09 0.32 0.24   Functionality        CDFM 03-03)
SEIS_1B-2-1f    480v BD 219 A, B (U1, U2) (Fragility Group      0.018      2.61 0.64 0.24   Functionality         SoV 01-05-01) Functional Failure SEIS_12-1b      RHRSW Pumps and EECW Alternate                  0.012      1.45 0.32 0.24    Anchorage           CDFM (Fragility Group 06-03)
SEIS_3-2        Batt CH 248-1 (Fragility Group 16-06)           0.011      1.47 0.32 0.24   Functionality        CDFM SEIS_11-1R3     Relay group 3 for group SEIS_11-1 (EECW         0.009      0.61 0.32 0.24   Functionality        CDFM Pp A3 UV device)
SEIS_BLD-IPS    Intake Pumping Station                          0.009      1.90 0.26 0.24     Structural         CDFM Analysis SEIS_5-2B       Initiation relays and panels                    0.008      0.97 0.32 0.24    Anchorage           CDFM Page 135 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                           December 2019 Table 5.4-6: Unit 3 SCDF Importance Measures Ranked by F-V Am                               Fragility Fragility Group                              Description                          F-V             u   r  Failure Mode (g)                              Method SEIS_1C-               Wall 95 falls towards 480v SD BD 3A and 3B                0.008 2.18 0.54 0.27       Block Wall          SoV 1W95SD                                                                                                        Failure SEIS_1C-               Wall 95 falls towards 480v RMOV 3A and                    0.008      2.18 0.54 0.27  Block Wall          SoV 1W95MOV                250vdc RMOV BD 3A                                                                      Failure SEIS_1C-4R4            3EA,3EC SDBD Lockout relays (86)                          0.006      0.94 0.32 0.24 Functionality       CDFM SEIS_1C-4R5-2          3EA,3EC SDBD Lockout relays (86)                          0.006      0.94 0.32 0.24 Functionality       CDFM SEIS_1C-4R7            3EA,3EC SDBD 480 Transformer Trip relays                  0.005      0.82 0.32 0.24 Functionality       CDFM (50G)
SEIS_2-1W8             Battery and Wall 8 or 47 interface (Block Wall            0.005      2.12 0.41 0.29  Block Wall          SoV Group 5)                                                                               Failure

See Section 5.7 for additional discussion on these representative fragilities Page 136 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 The EPRI SYSIMP software was used to calculate the importance measure of each fragility group, considering the combined F-V importance across all the seismic initiator bins.
LOOP represents the most significant contributor, which is consistent with the results of previous SPRA studies across the nuclear industry that have found that extended LOOP events are dominant for seismic risk. The two fragility groups with the highest F-V values excluding LOOP are SEIS_12-1P-2 (EECW pumps based on pipe frag calc) and SEIS_12-1P-1 (RHRSW pumps based on pipe fragility calculation) for Unit 1 and Unit 2. SEIS_12-1P-1 and SEIS_1B-2-2 (480v BD 219 3EA, 3EB (U3) (Fragility Group 01-05-02)) are the most risk-significant groups for Unit 3. SEIS_12-1P-1 is important because the SEIS_12-1P-1 pipe rupture fails the RHRSW system, which is responsible for removing decay heat from the RHR system. SEIS_12-1P-2 fails the EECW pumps that supply emergency cooling water for the EDG engine coolers and Emergency Core Cooling System (ECCS) pump room coolers among other important pieces of equipment. SEIS_1B-2-2 is important because it leads to failure of the Unit 3 EDGs.
5.4.4 SCDF Component Importance (Non-Seismic Failures)
Components were determined to be significant if the components Risk Achievement Worth (RAW) is greater than or equal to two or its F-V is greater than 0.005, per the definition of significant basic event in Reference [8]. Components are considered risk-significant if the component has a F-V value greater than 0.005 or a RAW greater than or equal to two for either the SCDF or SLERF importance measures.
Table 5.4-7 contains the F-V importance measures for each risk-significant individual component, common-cause, or test and maintenance basic event that appears in the seismic cutsets for the Unit 1 SCDF. Table 5.4-8 and Table 5.4-9 contain similar information for Units 2 and 3.
Table 5.4-7: Unit 1 SCDF Risk-Significant Individual Component Importance by Fussell-Vesely Basic Event                   Description                                                  F-V DGGFR0EDG_082__DGA DG A FAILS TO RUN                                                    6.62E-03 DGGFR0EDG_082__DGC            DG C FAILS TO RUN                                         5.60E-03 TM_0BATA2480000A              SHUTDOWN BATTERY SB-A UNAVAILABLE DUE TO TEST             5.43E-03 AND MAINTENANCE TM_0PNLA2480000A              SHUTDOWN BATTERY BOARD SB-A UNAVAILABLE DUE               5.43E-03 TO TEST AND MAINTENANCE DGGFR0EDG_082__DGB            DG B FAILS TO RUN                                         5.39E-03 Table 5.4-8: Unit 2 SCDF Risk-Significant Component Importance by Fussell-Vesely Basic Event                    Description                                                 F-V DGGFR0EDG_082__DGD             DG D FAILS TO RUN                                        6.90E-03 DGGFR0EDG_082__DGB             DG B FAILS TO RUN                                        6.65E-03 DGGFR0EDG_082__DGC             DG C FAILS TO RUN                                        6.33E-03 Page 137 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 Table 5.4-9: Unit 3 SCDF Risk-Significant Component Importance by Fussell-Vesely Basic Event                  Description                                                 F-V TM_0BATA2480003              0-BATA-248-0003 (MAIN BATTERY 3) UNAVAILABLE DUE         1.28E-02 TO T&M DGGFR3EDG_082_DG3C           DG 3C FAILS TO RUN                                       7.11E-03 DGGFR3EDG_082_DG3A           DG 3A FAILS TO RUN                                       5.98E-03 Another method of risk ranking components is by RAW. For basic events, RAW is determined by setting the event to TRUE, solving the model, and dividing the recalculated SCDF by the original SCDF. Tables 5.4-10, 5.4-11, and 5.4-12 list the components, common cause groups, and test and maintenance events with a RAW value greater than or equal to two for the Unit 1, 2, and 3 CDF, respectively.
Table 5.4-10: Unit 1 SCDF Risk-Significant Individual Components, Common-Cause Groups, and Test/Maintenance Events Importance by RAW Basic Event                Description                                                  RAW U0_21104CBKFO_ALL          CCF of all components in group 'U0_21104CBKFO'                10.5 U0_08208DGGFD_ALL          CCF of all components in group 'U0_08208DGGFD'                10.5 U0_21104CBKFO_1_2_3        CCF of three components: CBKFO0BKR_211A_003 &                 10.5 CBKFO0BKR_211B_002 & CBKFO0BKR_211C_022 U0_21104CBKFO_1_2_4        CCF of three components: CBKFO0BKR_211A_003 &                 10.5 CBKFO0BKR_211B_002 & CBKFO0BKR_211D_022 U0_08208DGGFR_ALL          CCF of all components in group 'U0_08208DGGFR'                10.5 U0_03002FANFR1_1_2         CCF of two components: FANFR0FAN_0300072 &                    10.5 FANFR0FAN_0300073 U0_03008FANFD_ALL          CCF of all components in group 'U0_03008FANFD'                10.5 U1_CCFMECH                 CCF of mechanical equipment causes failure to scram           10.3 U0_03008FANFR_ALL          CCF of all components in group 'U0_03008FANFR'                10.3 U0_08208DGGFR_1_2_3        CCF of three components: DGGFR0EDG_082__DGA &                 10.3 DGGFR0EDG_082__DGB & DGGFR0EDG_082__DGC U0_08208DGGFR_1_2_4        CCF of three components: DGGFR0EDG_082__DGA &                 10.3 DGGFR0EDG_082__DGB & DGGFR0EDG_082__DGD U0_02308PRHFR_ALL          CCF of all components in group 'U0_02308PRHFR'                10.3 U0_02308PRHFD_ALL          CCF of all components in group 'U0_02308PRHFD'                10.3 U0_08208DGGFD_1_2_3        CCF of three components: DGGFD0EDG_082__DGA &                 10.3 DGGFD0EDG_082__DGB & DGGFD0EDG_082__DGC U0_08208DGGFD_1_2_4        CCF of three components: DGGFD0EDG_082__DGA &                 10.3 DGGFD0EDG_082__DGB & DGGFD0EDG_082__DGD U0_24804BCHFR_ALL          CCF of all components in group 'U0_24804BCHFR'                7.2 U0_24803BATFR_1_2_3        CCF of three components: BATFR0BATA2480001 &                  7.2 BATFR0BATA2480002 & BATFR0BATA2480003 U1_07102FANFR_1_2          CCF of two components: FANFR1FAN_0710601 &                    6.2 FANFR1FAN_0710602 U0_02308CKVFO_ALL          CCF of all components in group 'U0_02308CKVFO'                4.3 Page 138 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table 5.4-10: Unit 1 SCDF Risk-Significant Individual Components, Common-Cause Groups, and Test/Maintenance Events Importance by RAW Basic Event               Description                                                 RAW FANFR1FAN_0710602         SAI PANEL 9-82 FAN COOLER FAN-71-602 FAILED                   3.6 TM_1BDDD2810001A          250V RMOV BD 1A UNAVAILABLE DUE TO TEST AND                   3.6 MAINTENANCE BUSFR1BDDD2810001A        250V RMOV BD 1A BUS FAILED                                    3.5 HFL_1068CCFPTLOPR         LOW RX PRESSURE PERMISSIVE SIGNAL COMMON                      3.3 CAUSE MISCALIBRATION U1_00304BISFD3_ALL        CCF of all components in group 'U1_00304BISFD3'               3.2 U0_08208DGGFR_1_2         CCF of two components: DGGFR0EDG_082__DGA &                   2.7 DGGFR0EDG_082__DGB U0_21104CBKFO_1_2         CCF of two components: CBKFO0BKR_211A_003 &                   2.4 CBKFO0BKR_211B_002 U0_21104CBKFO_2_3_4       CCF of three components: CBKFO0BKR_211B_002 &                 2.0 CBKFO0BKR_211C_022 & CBKFO0BKR_211D_022 Table 5.4-11: Unit 2 SCDF Risk-Significant Individual Components, Common-Cause Groups, and Test/Maintenance Events Importance by RAW Basic Event                 Description                                               RAW U0_03008FANFD_ALL           CCF of all components in group 'U0_03008FANFD'             10.3 U0_03002FANFR1_1_2          CCF of two components: FANFR0FAN_0300072 &                 10.3 FANFR0FAN_0300073 U0_08208DGGFR_ALL           CCF of all components in group 'U0_08208DGGFR'             10.3 U2_CCFMECH                  CCF of mechanical equipment causes failure to scram        10.1 U0_02308PRHFD_ALL           CCF of all components in group 'U0_02308PRHFD'             10.1 U0_02308PRHFR_ALL           CCF of all components in group 'U0_02308PRHFR'             10.1 U0_03008FANFR_ALL           CCF of all components in group 'U0_03008FANFR'             10.1 U0_08208DGGFD_2_3_4         CCF of three components: DGGFD0EDG_082__DGB &              10.1 DGGFD0EDG_082__DGC & DGGFD0EDG_082__DGD U0_08208DGGFD_ALL           CCF of all components in group 'U0_08208DGGFD'             10.1 U0_08208DGGFR_2_3_4         CCF of three components: DGGFR0EDG_082__DGB &              10.1 DGGFR0EDG_082__DGC & DGGFR0EDG_082__DGD U0_21104CBKFO_2_3_4         CCF of three components: CBKFO0BKR_211B_002 &              10.1 CBKFO0BKR_211C_022 & CBKFO0BKR_211D_022 U0_21104CBKFO_ALL           CCF of all components in group 'U0_21104CBKFO'             10.1 U0_21104CBKFO_1_2_4         CCF of three components: CBKFO0BKR_211A_003 &              7.3 CBKFO0BKR_211B_002 & CBKFO0BKR_211D_022 U0_08208DGGFR_1_2_4         CCF of three components: DGGFR0EDG_082__DGA &              7.1 DGGFR0EDG_082__DGB & DGGFR0EDG_082__DGD U0_24803BATFR_1_2_3         CCF of three components: BATFR0BATA2480001 &               7.1 BATFR0BATA2480002 & BATFR0BATA2480003 U0_24804BCHFR_ALL           CCF of all components in group 'U0_24804BCHFR'             7.1 Page 139 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                 December 2019 Table 5.4-11: Unit 2 SCDF Risk-Significant Individual Components, Common-Cause Groups, and Test/Maintenance Events Importance by RAW Basic Event                  Description                                               RAW U0_08208DGGFD_1_2_4          CCF of three components: DGGFD0EDG_082__DGA &              6.9 DGGFD0EDG_082__DGB & DGGFD0EDG_082__DGD U2_07102FANFR_1_2            CCF of two components: FANFR2FAN_0710601 &                 6.0 FANFR2FAN_0710602 U0_02308CKVFO_ALL            CCF of all components in group 'U0_02308CKVFO'             4.2 FANFR2FAN_0710602            SAI PANEL 9-82 FAN COOLER FAN-71-602 FAILED                3.6 TM_2BDDD2810002A             250V RMOV BD 2A UNAVAILABLE DUE TO TEST AND                3.6 MAINTENANCE BUSFR2BDDD2810002A           250V RMOV BD 2A FAILS                                      3.5 U2_00304BISFD3_ALL           CCF of all components in group 'U2_00304BISFD3'            3.2 U0_08208DGGFR_2_4            CCF of two components: DGGFR0EDG_082__DGB &                2.3 DGGFR0EDG_082__DGD U0_21104CBKFO_2_4            CCF of two components: CBKFO0BKR_211B_002 &                2.2 CBKFO0BKR_211D_022 U0_08208DGGFR_2_3            CCF of two components: DGGFR0EDG_082__DGB &                2.1 DGGFR0EDG_082__DGC Table 5.4-12: Unit 3 SCDF Risk-Significant Components, Common-Cause Groups, and Test/Maintenance Events Importance by RAW Basic Event                  Description                                               RAW U0_03008FANFD_ALL            CCF of all components in group 'U0_03008FANFD'             9.3 U3_03008FANFD_ALL            CCF of all components in group 'U3_03008FANFD'             9.3 U0_03002FANFR1_1_2           CCF of two components: FANFR0FAN_0300072 &                 9.3 FANFR0FAN_0300073 U3_03002FANFR_1_2            CCF of two components: FANFR3FAN_0300243 &                 9.3 FANFR3FAN_0300244 U0_08208DGGFR_ALL            CCF of all components in group 'U0_08208DGGFR'             9.3 U3_CCFMECH                   CCF of mechanical equipment causes failure to scram        9.1 U0_02308PRHFD_ALL            CCF of all components in group 'U0_02308PRHFD'             9.1 U0_02308PRHFR_ALL            CCF of all components in group 'U0_02308PRHFR'             9.1 U0_08208DGGFD_1_2_4          CCF of three components: DGGFD0EDG_082__DGA &              9.1 DGGFD0EDG_082__DGB & DGGFD0EDG_082__DGD U0_08208DGGFD_5_6_7          CCF of three components: DGGFD3EDG_082_DG3A &              9.1 DGGFD3EDG_082_DG3B & DGGFD3EDG_082_DG3C U0_08208DGGFR_1_2_4          CCF of three components: DGGFR0EDG_082__DGA &              9.1 DGGFR0EDG_082__DGB & DGGFR0EDG_082__DGD U0_08208DGGFR_5_6_7          CCF of three components: DGGFR3EDG_082_DG3A &              9.1 DGGFR3EDG_082_DG3B & DGGFR3EDG_082_DG3C U0_03008FANFR_ALL            CCF of all components in group 'U0_03008FANFR'             9.1 U3_03008FANFR_ALL            CCF of all components in group 'U3_03008FANFR'             9.1 Page 140 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 Table 5.4-12: Unit 3 SCDF Risk-Significant Components, Common-Cause Groups, and Test/Maintenance Events Importance by RAW Basic Event                  Description                                            RAW U0_21104CBKFO_1_2_4          CCF of three components: CBKFO0BKR_211A_003 &            9.1 CBKFO0BKR_211B_002 & CBKFO0BKR_211D_022 U3_21104CBKFO_1_2_3          CCF of three components: CBKFO3BKR_211A_007 &            9.1 CBKFO3BKR_211B_014 & CBKFO3BKR_211C_012 U0_08208DGGFD_ALL            CCF of all components in group 'U0_08208DGGFD'           9.1 U0_21104CBKFO_ALL            CCF of all components in group 'U0_21104CBKFO'           9.1 U3_21104CBKFO_ALL            CCF of all components in group 'U3_21104CBKFO'           9.1 U0_24803BATFR_1_2_3          CCF of three components: BATFR0BATA2480001 &             6.4 BATFR0BATA2480002 & BATFR0BATA2480003 U0_24804BCHFR_ALL            CCF of all components in group 'U0_24804BCHFR'           6.4 U3_07102FANFR_1_2            CCF of two components: FANFR3FAN_0710601 &               5.4 FANFR3FAN_0710602 U0_02308CKVFO_ALL            CCF of all components in group 'U0_02308CKVFO'           3.9 HINST3                       Seismic - Instrument Impact on MCR HEPs - G03            3.6 HINST2                       Seismic - Instrument Impact on MCR HEPs - G02            3.5 TM_3BDDD2810003A             250V RMOV BD 3A UNAVAILABLE DUE TO TEST AND              3.4 MAINTENANCE FANFR3FAN_0710602            SAI PANEL 9-82 FAN COOLER FAN-71-602 FAILED              3.4 BUSFR3BDDD2810003A           250V RMOV BD 3A FAILS                                    3.2 U0_08208DGGFR_1_6_7          CCF of three components: DGGFR0EDG_082__DGA &            3.2 DGGFR3EDG_082_DG3B & DGGFR3EDG_082_DG3C HFL_3068CCFPTLOPR            LOW RX PRESSURE PERMISSIVE SIGNAL COMMON                 3.1 CAUSE MISCALIBRATION U3_00304BISFD3_ALL           CCF of all components in group 'U3_00304BISFD3'          2.9 TM_0BATA2480003              0-BATA-248-0003 (MAIN BATTERY 3) UNAVAILABLE DUE         2.4 TO T&M U0_21104CBKFO_1_2_3          CCF of three components: CBKFO0BKR_211A_003 &            2.2 CBKFO0BKR_211B_002 & CBKFO0BKR_211C_022 FUSSO0FU2_2803_111           FUSED SWITCH 111 FAILS                                   2.2 U0_08208DGGFD_1_6_7          CCF of three components: DGGFD0EDG_082__DGA &            2.2 DGGFD3EDG_082_DG3B & DGGFD3EDG_082_DG3C U3_21104CBKFO_1_3_4          CCF of three components: CBKFO3BKR_211A_007 &            2.2 CBKFO3BKR_211C_012 & CBKFO3BKR_211D_008 BATFR0BATA2480003            BATTERY 3 FAILS DURING OPERATION INCLUDING               2.0 COMMON CAUSE BUSFR0BDDD2800003            BATTERY BD. 3 FAILS                                      2.0 BUSFR3BDBB2680003B           480V RMOV BD 3B FAILS                                    2.0 Page 141 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                          December 2019 5.4.5 Significant Human Failure Events Significant post-initiator operator actions are defined as those operator action basic events that have an F-V value greater than 0.005 or a RAW greater than or equal to two. Note that the common methods of calculating RAW for basic events will not yield useful results for the HRA events, due to the processing of combination events. This is because the events are set to one during the quantification process and a recovery event representing the combination or single event is appended to the cutset.
Therefore, setting the event to one to determine the RAW value has no effect on CDF.
The F-V values of each operator action were determined in SYSIMP by defining groups where each particular operator action appearing in each seismic HRA bin (S1, S2, S3, and S4) was simultaneously set to FALSE in the combined cutset file to determine the combined importance across all seismic bins. Tables 5.4-13, 5.4-14, and 5.4-15 list the operator actions that were determined to be risk-significant for SCDF in the seismic model for Unit 1, Unit 2, and Unit 3, respectively.
Table 5.4-13: Risk-Significant Operator Actions for Unit 1 SCDF Operator Action                Description                                                         F-V HFA_OPS_FLEXN2ALIGN            Operator fails to align N2 FLEX backup to Drywell Control Air    2.07E-01 (Safety Relief Valves (SRVs))
HFA_0084CADALIGN               Operator fails to align Containment Air Dilution (CAD)           1.82E-01 backup to Drywell Control Air (DCA)
HFA_0074HPSPC2                 Failure to align RHR for suppression pool cooling                1.08E-01 (Anticipated Transient Without Scram (ATWS) or Inadvertent Opening of One Relief Valve(IOORV)) **
HFA_0074HPSPC1                 Failure to align RHR for suppression pool cooling (non-          6.59E-02 ATWS/IOORV)
HFA_0023ALIGNEECW_L            Operator fails to align backup EECW pump                         6.23E-02 HFA_OPS_4KVSDBDRESET           Operator reset of 4kV Shutdown Board lockout relays              5.40E-02 (seismic)
HFA_0SBDALTDC                  Operator fails to isolate SD BD and align alternate DC           5.35E-02 HFA_0248ALNALTCHG              Failure to align alternate battery charger                       3.08E-02 HFA_0280ALNALTBBD              Operator fails to align alternate feeder                         2.82E-02 HFA_0033HVACDOOR               Failure to open doors and install fans after HVAC failure        2.11E-02 HFA_0031STARTHVAC              Failure to start standby Control Bay HVAC                        1.83E-02 HFA_0231480SDBTIE              Failure to transfer 480V shutdown board to alternate source      1.49E-02 HFA_0268480CRSTIE              Failure to transfer deenergized 480v board to alternate          1.13E-02 supply HFA_0001HPRVD1_L               Failure to initiate reactor-vessel depressurization (transient   8.90E-03 or ATWS)
HFA_0073MANLEVEL               Operator fails to manually control level with High Pressure      8.30E-03 Coolant Injection (HPCI)
HFA_HINST*                     Seismic failure of Main Control Room instrumentation             5.10E-03 Page 142 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                             December 2019 Table 5.4-13: Risk-Significant Operator Actions for Unit 1 SCDF Operator Action                    Description                                                          F-V
*Although this is not an operator action but an equipment failure, it has an effect similar to that of multiple operator action failures.
Table 5.4-14: Risk-Significant Operator Actions for Unit 2 SCDF Operator Action                    Description                                                      F-V HFA_OPS_FLEXN2ALIGN                Operator fails to align N2 FLEX backup to Drywell             2.27E-01 Control Air (SRVs)
HFA_0084CADALIGN                   Operator fails to align CAD backup to DCA                     1.88E-01 HFA_0074HPSPC2                     Failure to align RHR for suppression pool cooling             1.12E-01 (ATWS or IORV) **
HFA_0074HPSPC1                     Failure to align RHR for suppression pool cooling             7.10E-02 (non-ATWS/IORV)
HFA_0023ALIGNEECW_L                Operator fails to align backup EECW pump                      5.83E-02 HFA_0SBDALTDC                      Operator fails to isolate SD BD and align alternate DC        4.71E-02 HFA_OPS_4KVSDBDRESET               Operator reset of 4kV Shutdown Board lockout relays           3.70E-02 (seismic)
HFA_0248ALNALTCHG                  Failure to align alternate battery charger                    3.02E-02 HFA_0280ALNALTBBD                  Operator fails to align alternate feeder                      2.73E-02 HFA_0033HVACDOOR                   Failure to open doors and install fans after HVAC             2.40E-02 failure HFA_0231480SDBTIE                  Failure to transfer 480V shutdown board to alternate          1.84E-02 source HFA_0031STARTHVAC                  Failure to start standby Control Bay HVAC                     1.82E-02 HFA_0073MANLEVEL                   Operator fails to manually control level with HPCI            1.76E-02 HFA_0268480CRSTIE                  Failure to transfer deenergized 480v board to                 1.24E-02 alternate supply HFA_0001HPRVD1                     Failure to initiate reactor-vessel depressurization           1.04E-02 (transient or ATWS)
HFA_0001HPRVD1_L                   Failure to initiate reactor-vessel depressurization           8.60E-03 (transient or ATWS)
HFA_HINST*                         Seismic failure of Main Control Room instrumentation          5.10E-03
*Although this is not an operator action but an equipment failure, it has an effect similar to that of multiple operator action failures.
Table 5.4-15: Risk-Significant Operator Actions for Unit 3 SCDF Operator Action                    Description                                                      F-V HFA_OPS_FLEXN2ALIGN                Operator fails to align N2 FLEX backup to Drywell             1.95E-01 Control Air (SRVs)
HFA_0084CADALIGN                   Operator fails to align CAD backup to DCA                     1.53E-01 Page 143 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                      December 2019 Table 5.4-15: Risk-Significant Operator Actions for Unit 3 SCDF Operator Action                Description                                                  F-V HFA_0074HPSPC2                 Failure to align RHR for suppression pool cooling         1.00E-01 (ATWS or IORV)
HFA_0074HPSPC1                 Failure to align RHR for suppression pool cooling         7.05E-02 (non-ATWS/IORV)
HFA_OPS_4KVSDBDRESET           Operator reset of 4kV Shutdown Board lockout relays       6.78E-02 (seismic)
HFA_0SBDALTDC                  Operator fails to isolate SD BD and align alternate DC    4.78E-02 HFA_0231480SDBTIE              Failure to transfer 480V shutdown board to alternate      4.53E-02 source HFA_0023ALIGNEECW_L            Operator fails to align backup EECW pump                  4.01E-02 HFA_0248ALNALTCHG              Failure to align alternate battery charger                1.50E-02 HFA_0268480CRSTIE              Failure to transfer deenergized 480v board to             1.33E-02 alternate supply HFA_0073MANLEVEL               Operator fails to manually control level with HPCI        1.08E-02 HFA_0280ALNALTBBD              Operator fails to align alternate feeder                  1.07E-02 HFA_0001HPRVD1_L               Failure to initiate reactor-vessel depressurization       7.70E-03 (transient or ATWS)
HFA_0033HVACDOOR               Failure to open doors and install fans after HVAC         6.30E-03 failure HFA_0001HPRVD1                 Failure to initiate reactor-vessel depressurization       5.70E-03 (transient or ATWS)
HFA_0031STARTHVAC              Failure to start standby Control Bay HVAC                 5.00E-03 5.4.5.1 Summary of the Approach used to Evaluate Human Error Probabilities The approach used to evaluate HEPs is based on EPRI 3002008093 [49]. Each HFE that was determined to be feasible for the SPRA was subject to a detailed analysis.
5.4.5.2 Detailed Analysis for HEPs All HFEs in the BFN SPRA were analyzed with detailed HRA [42], in accordance with the guidance in EPRI 3002008093 [49].
The EPRI approach for seismic HRA directs the detailed analysis of HFEs to be done in two parts: qualitative and quantitative analysis. In practice, these are done in tandem for each HFE, and the starting point for the BFN seismic HRA is the IEPRA HRA. Detailed analysis was performed for EPRI Bins 1 through 3. No detailed analysis was performed for EPRI Bin 4, as all HFEs are considered infeasible due to the damage state of this bin and the uncertainty of instrumentation availability.
Page 144 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 5.4.5.3 Operator action credit for FLEX The FLEX operator actions to align backup nitrogen to drywell control air for the SRVs were included in the model. No other FLEX systems were credited in the model.
5.4.6 Significant SCDF Accident Sequences The SPRA evaluated the GTRAN, Direct Core Damage, LLOCA, MLOCA, IOORV, ISLOCA, and ATWS ETs that were the same ET categories considered in the IEPRA, except for Direct Core Damage, which is unique to the SPRA. The results were dominated by seismically induced LOOP transient scenarios, which are treated in the GTRAN ET. Since FRANX eliminates events that are set TRUE in its final processing, it is difficult to obtain accurate results of the exact percentages due to each accident sequence since the sequence flags are absent from the FRANX group cutset results.
However, an examination of the results of a consolidated CAFTA cutset file without ACUBE processing and with the flags intact should show a relatively accurate representation of the percentages that each accident sequence contributes. The results show that the dominant sequences are GTRAN-005, GTRAN-006, GTRAN-002, and GTRAN-012 for Unit 1. Unit 2 is dominated by GTRAN-005, GTRAN-006, GTRAN-012, GTRAN-002 and GTRAN-011. Unit 3 is dominated by GTRAN-005, GTRAN-006, GTRAN-002, GTRAN-012 and GTRAN-001A. Figure 5.4-1, Figure 5.4-2, and Figure 5.4-3 show the relative percentage contribution for each dominant accident sequence based on a consolidated CAFTA cutset file for Unit 1, Unit 2, and Unit 3, respectively.
Significant accident sequences for SCDF are discussed in Table 5.4-16. The differences in results between units are primarily the result of power dependency differences for the SPRA.
Page 145 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                      December 2019 Unit 1 CDF FLG_GTRAN_003 FLG_GTRAN_004    4%
4%
FLG_GTRAN_001A 8%
FLG_GTRAN_005 FLG_GTRAN_005         FLG_GTRAN_006 FLG_GTRAN_011 8%                                     26%             FLG_GTRAN_002 FLG_GTRAN_012 FLG_GTRAN_011 FLG_GTRAN_012 14%                                       FLG_GTRAN_001A FLG_GTRAN_006 22%             FLG_GTRAN_004 FLG_GTRAN_003 FLG_GTRAN_002 14%
Figure 5.4-1: Unit 1 SCDF Accident Sequence Contribution Figure 5.4-2: Unit 2 SCDF Accident Sequence Contribution Page 146 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                     December 2019 Figure 5.4-3: Unit 3 SCDF Accident Sequence Contribution Page 147 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                            December 2019 Table 5.4-16: Dominant SCDF Accident Sequences in SPRA Accident Sequence                           Description                                                        Discussion GTRAN-001A      The scram successfully occurs, the Power Conversion Systems       The PCS is assumed failed for all seismic scenarios. The CST (PCS) fails, and there are no breaks outside containment or       is assumed failed for all seismic scenarios, which fails CRD stuck-open relief valves. HPCI or RCIC is successful for at least injection. Top contributors to this sequence involve common 4 hours. Alternate shutdown cooling fails. Control Rod Drive      accident signal relays fragility group (SEIS_1C-4R8) failing, (CRD) fails. Low-pressure injection via core spray or Low         which fails the diesels, and initiation relays and panels (SEIS_5-Pressure Coolant Injection (LPCI) fail.                           2B) failing, which fails standby coolant injection, LPCI, core spray, and shutdown cooling. The shutdown boards fail because of relay chatter, and the operator action to reset the relays also fails.
GTRAN-002      The scram successfully occurs, the PCS fails, and there are no    The PCS is assumed failed for all seismic scenarios. The CST breaks outside containment or stuck-open relief valves. HPCI      is assumed failed for all seismic scenarios, which fails CRD or RCIC is successful for at least 4 hours. Early suppression     injection. HPCI and RCIC fail after 4 hours due a failure to align pool cooling is not successful or initiated in time to prevent    suppression pool cooling. Late suppression pool cooling is not exceeding Heat Capacity Temperature Limit (HCTL) and 190          successful because the operator fails to align it, and drywell deg F in the suppression pool. Therefore, long-term HPCI or       sprays are also not available because of operator failure to align RCIC is not successful. The CRD and High Pressure Make-Up         it. All low-pressure injection sources have failed. Shutdown (HPMU) systems fail, and manual depressurization is               cooling is not available due to operator alignment failure. LPCI challenged and initiated successfully with 2 SRVs at HCTL         has failed because it cannot take suction from the CST, which about 4 hours after the scram. After depressurization, low-       is assumed failed for seismic scenarios. Core spray is failed for pressure injection by RHR in the LPCI mode or Core Spray (CS)     the same reason. Standby coolant injection has failed because is successful. Alternate Shutdown Cooling (ASDC) is               operators have failed to align it.
unsuccessful. Late suppression pool cooling fails, allowing the primary containment to pressurize to PCPL in about 13 hours, but the hardened wetwell vent successfully establishes Decay Heat Removal (DHR). After the successful vent, the suppression pool no longer provides a successful suction source for RHR or CS, and late injection is not re-established when all the available injection sources fail. These sources are condensate injection, standby coolant injection, shutdown cooling and RHR or CS with suction on the CST. Core damage is caused by loss of injection and the Reactor Pressure Vessel (RPV) is at low pressure.
GTRAN-005      Sequence GTRAN-005 is the same as sequence GTRAN-002              For Unit 1 and Unit 2, the driving seismic consideration is except all low-pressure injections fail. core damage occurs       generally failure of fragility group SEIS_5-2A2-2-C-G06 about 1.5 hours after scram with the RPV at low pressure.         (SEISMIC FRAGILITY FOR %G06: Initiation relays and panels Page 148 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                             December 2019 (20-08)), which fails accident initiation signals for multiple A GTRAN-005 cutset is the dominant cutset for Units 1, 2, and     systems. For Unit 3, it is failure of the 480V shutdown board via 3 CDF.                                                            fragility group SEIS_1B-2-2-C-G06, while cutsets involving fragility group SEIS_5-2A2-2-C-G06 are also top contributors similar to Units 1 and 2.
GTRAN-006 Sequence GTRAN-006 is the same as sequence GTRAN-002              For all units, the dominant cutsets for GTRAN-006 are not except depressurization fails. Without depressurization, there    directly related to seismic failures (other than systems like the are no other available injection sources, and core damage         CST that are assumed to be failed for seismic initiators and, occurs. CD occurs about 1.5 hours after scram with the RPV at     therefore, do not appear in the cutsets), but rather the fact that high pressure.                                                    all operator actions are assumed to be failed for seismic bin
                                                                            %G07-%G09. Top cutsets for Unit 1 involve simultaneous GTRAN-006 cutsets are in the top 10 most significant for all      failure to initiate depressurization via operator intervention and three units.                                                      to align RHR for suppression pool cooling, or simultaneous failure to establish alternate FLEX nitrogen supply for controlling the SRVs during emergency depressurization, along with the failure to align RHR for suppression pool cooling. For Unit 2 and Unit 3, these scenarios are also dominant.
GTRAN-011 Sequence GTRAN-011 is the same as sequence GTRAN-007              The PCS is assumed failed for all seismic scenarios. The CST except all low-pressure injections fail. CD occurs in about 30 to and the condensate system are assumed failed by the seismic 40 minutes with the RPV at low pressure.                          event. HPCI fails early because it receives an erroneous isolation signal due to the failure of fragility group SEIS_14-1R1 In sequence GTRAN-007, the scram successfully occurs, the         (HPCI/RCIC isolations relays). This also fails RCIC. Early low-PCS fails, and there are no breaks outside containment or         pressure injection fails because core spray and LPCI both fail.
stuck-open relief valves. Early HPCI or RCIC are unsuccessful     Core spray loop 1 fails because of loss of pump power from as an initial injection source. A cooldown is initiated but HPMU  both 4kV shutdown boards A and B from a LOOP event and a is unsuccessful. CRD may be available but is not challenged       loss of the diesels due to EECW failure due to RHRSW pump because it lacks sufficient capacity to be used as an initial     failure from fragility group SEIS_11-1R1 (EECW pump B&D injection source. When RPV level drops to Top of Active Fuel      relays). Core spray loop 2 is lost for similar reasons.
(TAF), manual depressurization is successful. Low-pressure        LPCI loop 1 fails for several reasons, both because the loop 1 injection by RHR in the LPCI mode or CS is successful as an       injection path fails and because the RHR loop 1 supply is not initial injection source but must be initiated within 30 minutes. available. The LPCI loop 1 injection path fails because of ASDC is unsuccessful. Late suppression pool cooling and           reactor motor-operated valve (RMOV) board failure caused by drywell spray are unsuccessful. Hardened Containment Vent         loss of the normal and alternate shutdown board supplies due (HCV) successfully accomplishes DHR. Without SPC, PCPL is         to a LOOP along with diesel generator failure as a result of reached in approximately 10 hours. Successful venting fails the   EECW failure due again to fragility group SEIS_11-1R1. LPCI suppression pool as a suction source, and RHR and CS with         loop 2 fails for similar reasons.
suction on the suppression pool are lost. A post-vent injection source of either RHR with suction on the CST, CS with suction on the CST, shutdown cooling, standby coolant injection, Page 149 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                         December 2019 condensate injection, or CRD (1 pump) is unsuccessful. CD occurs due to loss of injection, and the RPV fails at low pressure.
GTRAN-012 Sequence GTRAN-012 is the same as sequence GTRAN-007           The PCS is assumed failed for all seismic scenarios. RCIC and except depressurization fails. Without depressurization, there HPCI both fail. HPCI and RCIC both fail due to steam supply are no other available injection sources, and core damage      path failures because the RCIC steam supply line outboard occurs. Core damage occurs in about 30 to 40 minutes with the  isolation valve fails due to fragility group SEIS_14-1R1-2 RPV at high pressure.                                          (HPCI/RCIC isolations). Emergency depressurization fails due to operator failure to initiate depressurization.
Page 150 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                   December 2019 5.5    SLERF Results 5.5.1 Overall SLERF The baseline total LERF is 3.00E-6 /ry for Unit 1, 3.10E-6 /ry for Unit 2, and 3.31E-6 /ry for Unit 3.
5.5.2 SLERF as a Function of Hazard Interval A summary of the SLERF results for each seismic hazard interval is presented in Table 5.5-1 for Unit 1 SLERF, Table 5.5-2 for Unit 2 SLERF, and Table 5.5-3 for Unit 3 SLERF.
Table 5.5-1: Unit 1 SLERF Contribution by Initiating Event Earthquake                            Percent Truncation    Scenario         Description                         CLERP        SLERF Frequency                           Contribution 1.0E-12      %G01       Seismic Initiating Event       2.68E-04 2.8E-05      7.50E-09       0.2%
(0.1g to <0.2g) 1.0E-12      %G02       Seismic Initiating Event       5.56E-05 9.1E-04      5.05E-08       1.7%
(0.2g to <0.3g) 2.0E-12      %G03       Seismic Initiating Event       2.98E-05 8.5E-03      2.52E-07       8.4%
(0.3g to <0.6g) 7.0E-11      %G04       Seismic Initiating Event       2.17E-06 9.9E-02      2.15E-07       7.2%
(0.6g to <0.7g) 2.0E-09      %G05       Seismic Initiating Event       2.19E-06 2.5E-01      5.57E-07       18.5%
(0.7g to <0.9g) 6.0E-08      %G06       Seismic Initiating Event       1.82E-06 7.7E-01      1.40E-06       46.7%
(0.9g to <1.5g) 5.0E-08      %G07       Seismic Initiating Event       1.80E-07 9.0E-01      1.61E-07       5.4%
(1.5g to <1.7g) 2.0E-08      %G08       Seismic Initiating Event       3.32E-07 9.1E-01      3.03E-07       10.1%
(1.7g to <3g) 2.0E-08      %G09       Seismic Initiating Event       5.96E-08 9.2E-01      5.48E-08       1.8%
(>3g)
Total              3.00E-06 SLERF=
Page 151 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                   December 2019 Table 5.5-2: Unit 2 SLERF Contribution by Initiating Event Earthquake                           Percent Truncation    Scenario         Description                        CLERP        SLERF Frequency                          Contribution 1.0E-12      %G01       Seismic Initiating Event       2.68E-04 2.8E-05     7.62E-09        0.2%
(0.1g to <0.2g) 5.0E-11      %G02       Seismic Initiating Event       5.56E-05 8.7E-04     4.86E-08        1.6%
(0.2g to <0.3g) 5.0E-12      %G03       Seismic Initiating Event       2.98E-05 9.0E-03     2.68E-07        8.6%
(0.3g to <0.6g) 2.0E-10      %G04       Seismic Initiating Event       2.17E-06 9.8E-02     2.12E-07        6.9%
(0.6g to <0.7g) 2.0E-09      %G05       Seismic Initiating Event       2.19E-06 2.7E-01     5.88E-07       19.0%
(0.7g to <0.9g) 5.0E-08      %G06       Seismic Initiating Event       1.82E-06 8.0E-01     1.45E-06       46.9%
(0.9g to <1.5g) 4.0E-08      %G07       Seismic Initiating Event       1.80E-07 9.0E-01     1.61E-07        5.2%
(1.5g to <1.7g) 2.0E-08      %G08       Seismic Initiating Event       3.32E-07 9.1E-01     3.03E-07        9.8%
(1.7g to <3g) 2.0E-08      %G09       Seismic Initiating Event       5.96E-08 9.2E-01     5.48E-08        1.8%
(>3g)
Total              3.10E-06 SLERF=
Page 152 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                   December 2019 Table 5.5-3: Unit 3 SLERF Contribution by Initiating Event Earthquake                           Percent Truncation    Scenario         Description                        CLERP        SLERF Frequency                          Contribution 1.0E-12      %G01       Seismic Initiating Event       2.68E-04 2.9E-05     7.72E-09        0.2%
(0.1g to <0.2g) 5.0E-11      %G02       Seismic Initiating Event       5.56E-05 9.0E-04     5.00E-08        1.6%
(0.2g to <0.3g) 4.0E-12      %G03       Seismic Initiating Event       2.98E-05 1.1E-02     3.17E-07       10.2%
(0.3g to <0.6g) 4.0E-10      %G04       Seismic Initiating Event       2.17E-06 1.1E-01     2.43E-07        7.8%
(0.6g to <0.7g) 4.0E-09      %G05       Seismic Initiating Event       2.19E-06 3.0E-01     6.57E-07       21.2%
(0.7g to <0.9g) 5.0E-08      %G06       Seismic Initiating Event       1.82E-06 8.3E-01     1.52E-06       49.0%
(0.9g to <1.5g) 5.0E-08      %G07       Seismic Initiating Event       1.80E-07 9.0E-01     1.61E-07        5.2%
(1.5g to <1.7g) 2.0E-08      %G08       Seismic Initiating Event       3.32E-07 9.1E-01     3.03E-07        9.8%
(1.7g to <3g) 2.0E-08      %G09       Seismic Initiating Event       5.96E-08 9.2E-01     5.48E-08        1.8%
(>3g)
Total              3.31E-06 SLERF=
5.5.3 Fragility Group Importance for SLERF The SSCs with the most significant seismic failure contributions to SLERF for Unit 1 are listed in Table 5.5-4, sorted by F-V. The seismic fragilities for each of the significant contributors are also provided in Table 5.5-4, along with the corresponding limiting seismic failure mode and method of fragility calculation. The corresponding measures for Unit 2 and Unit 3 are presented in Table 5.5-5 and 5.5-6.
Page 153 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                  December 2019 Table 5.5-4: Unit 1 SLERF Fragility Group Importance Measures Ranked by F-V Am                Failure Fragility Group                          Description                           F-V        u   r                  Fragility Method (g)                Mode SEIS_LOOP                   LOOP (Loss of Offsite Power)                    0.820 0.30 0.45 0.30     Ceramic        Table 6-1 insulators   NUREG/CR-6544 SEIS_5-2B                     Initiation relays and panels                  0.210 0.97 0.32 0.24    Anchorage         CDFM SEIS_2-1-1               Unit Batteries (Fragility Group 15-03)             0.183 1.32 0.38 0.24    Anchorage     Representative SEIS_5-8           REACTOR PROTECTION & NSS PNL (18-02)                    0.056 1.22 0.32 0.24   Functionality  Representative SEIS_12-1P-2              EECW pumps based on pipe frag calc                 0.042 2.48 0.60 0.48   Soil (Buried         SoV Piping)
Failure SEIS_2-1W8       Battery and Wall 8 or 47 interface (Block Wall Group 5)     0.035 2.12 0.41 0.29    Block Wall          SoV Failure SEIS_1B-2-1a     480v BD 219 A, B (U1, U2) (Fragility Group 01-05-01)        0.034 1.95 0.56 0.22    Anchorage           SoV Anchorage Failure SEIS_BLD-IPS                      Intake Pumping Station                      0.021 1.90 0.26 0.24    Structural        CDFM Analysis SEIS_1B-2-1f     480v BD 219 A, B (U1, U2) (Fragility Group 01-05-01)        0.021 2.61 0.64 0.24   Functionality        SoV Functional Failure SEIS_12-1b     RHRSW pumps and EECW Alternate (Fragility Group 06-          0.016 1.45 0.32 0.24    Anchorage         CDFM 03)
SEIS_1B-1      480V BD 1A, 2A, 3A, 1B, 2B, 3B (Fragility Group 02-01)       0.015 1.70 0.32 0.24   Functionality      CDFM SEIS_12-1P-1         RHRSW pumps based on pipe frag (Pipe Calc)              0.012 2.45 0.61 0.47   Soil (Buried         SoV Piping)
Failure SEIS_11-1R1-1    Relay group 1-1 for group SEIS_11-1 (EECW Pp B3 UV           0.011 1.31 0.37 0.29   Functionality        SoV device)
SEIS_14-1R1-2    Relays for Group SEIS_14-1, Relay Group 1 (HPCI/RCIC         0.011 1.62 0.32 0.24   Functionality      CDFM Isolations)
Page 154 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                           December 2019 Table 5.5-4: Unit 1 SLERF Fragility Group Importance Measures Ranked by F-V Am                 Failure Fragility Group                                      Description                              F-V        u   r                  Fragility Method (g)                 Mode SEIS_11-1R3               Relay group 3 for group SEIS_11-1 (EECW Pp A3 UV                  0.011 0.61 0.32 0.24 Functionality        CDFM device)
SEIS_1C-                      Wall 28 falls towards 480v SD BD 1A and 1B                  0.011 2.18 0.54 0.27    Block Wall          SoV 1W28SD                                                                                                              Failure SEIS_5-4                Panels group 4-1 Control Room panels Lower Fragility              0.010 2.04 0.38 0.24    Anchorage     Representative (Fragility Group 20-02 and 20-03)
SEIS_11-1R2            Relay group 2 for group SEIS_11-1 (EECW Pp B3&D3 OC                  0.009 1.29 0.38 0.24   Functionality      CDFM device)
SEIS_1C-4R8            3EA,3EC SDBD Common Accident Signal Relays (CASA)                    0.006 0.63 0.32 0.24   Functionality      CDFM SEIS_5-2A2-2                          Initiation relays and panels (20-08)                  0.006 1.30 0.58 0.23    Anchorage           SoV SEIS_1C-                Wall 28 falls toward 480v RMOV 1A and 4kv SD BD A                 0.006 2.18 0.54 0.27     Blockwall          SoV 1W28MOV SEIS_12-1a                       EECW pumps (Fragility Group 06-03-01)                     0.005 1.34 0.32 0.24    Anchorage         CDFM

See Section 5.7 for additional discussion on these representative fragilities Page 155 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                  December 2019 Table 5.5-5: Unit 2 SLERF Fragility Group Importance Measures Ranked by F-V Am                Failure Fragility Group                           Description                          F-V        u   r                  Fragility Method (g)                Mode SEIS_LOOP                    LOOP (Loss of Offsite Power)                   0.813 0.30 0.45 0.30     Ceramic        Table 6-1 insulators   NUREG/CR-6544 SEIS_5-2B                     Initiation relays and panels                 0.189 0.97 0.32 0.24    Anchorage         CDFM SEIS_2-1-1               Unit Batteries (Fragility Group 15-03)             0.184 1.32 0.38 0.24    Anchorage     Representative SEIS_5-8          REACTOR PROTECTION & NSS PNL (18-02)                    0.070 1.22 0.32 0.24   Functionality  Representative SEIS_12-1P-2             EECW Pumps based on pipe frag calc                  0.044 2.48 0.60 0.48   Soil (Buried         SoV Piping)
Failure SEIS_1B-2-1a      480v BD 219 A, B (U1, U2) (Fragility Group 01-05-01)       0.039 1.95 0.56 0.22    Anchorage           SoV Anchorage Failure SEIS_2-1W8       Battery and Wall 8 or 47 interface (Block Wall Group 5)     0.039 2.12 0.41 0.29    Block Wall          SoV Failure SEIS_14-1R1-2    Relays for Group SEIS_14-1, Relay Group 1 (HPCI/RCIC         0.036 1.62 0.32 0.24   Functionality      CDFM Isolations)
SEIS_12-1b     RHRSW Pumps and EECW Alternate (Fragility Group 06-          0.033 1.45 0.32 0.24    Anchorage         CDFM 03)
SEIS_5-2A2-2               Initiation relays and panels (20-08)              0.031 1.30 0.58 0.23    Anchorage           SoV SEIS_11-1R1-1     Relay group 1-1 for group SEIS_11-1 (EECW Pp B3 UV          0.031 1.31 0.37 0.29   Functionality        SoV device)
SEIS_11-1R2    Relay group 2 for group SEIS_11-1 (EECW Pp B3&D3 OC           0.030 1.29 0.38 0.24   Functionality      CDFM device)
SEIS_11-1R3       Relay group 3 for group SEIS_11-1 (EECW Pp A3 UV           0.030 0.61 0.32 0.24   Functionality      CDFM device)
SEIS_BLD-IPS                       Intake Pumping Station                     0.025 1.90 0.26 0.24    Structural        CDFM Analysis SEIS_1C-4R8     3EA,3EC SDBD Common Accident Signal Relays (CASA)            0.024 0.63 0.32 0.24   Functionality      CDFM Page 156 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                  December 2019 Table 5.5-5: Unit 2 SLERF Fragility Group Importance Measures Ranked by F-V Am                Failure Fragility Group                          Description                           F-V        u   r                  Fragility Method (g)                Mode SEIS_11-1R4      Relay group 4 for group SEIS_11-1 (EECW Pp A3 OC            0.023 0.94 0.51 0.27 Functionality          SoV device)
SEIS_1B-2-1f      480v BD 219 A, B (U1, U2) (Fragility Group 01-05-01)       0.022 2.61 0.64 0.24   Functionality        SoV Functional Failure SEIS_1C-6R6           C SDBD 480 Transformer Trip relays (50G)               0.021 1.29 0.38 0.24   Functionality      CDFM SEIS_1C-6R2        C DG BKR Trip Relays (CAR, OTX, RI, VRL, VRR)             0.021 1.43 0.32 0.24   Functionality      CDFM SEIS_1C-4R7        3EA,3EC SDBD 480 Transformer Trip relays (50G)            0.020 0.82 0.32 0.24   Functionality      CDFM SEIS_1C-4R4               3EA,3EC SDBD Lockout relays (86)                   0.020 0.94 0.32 0.24   Functionality      CDFM SEIS_1C-4R5-2              3EA,3EC SDBD Lockout relays (86)                   0.020 0.94 0.32 0.24   Functionality      CDFM SEIS_5-2A2-1               Initiation relays and panels (20-07)              0.020 1.96 0.32 0.24   Functionality      CDFM SEIS_1B-2-2      480v BD 219 3EA, 3EB (U3) (Fragility Group 01-05-02)        0.020 1.14 0.32 0.24    Anchorage         CDFM SEIS_1C-4          U3 4kv SD BD EA and EC (Fragility Group 03-03)           0.020 1.09 0.32 0.24   Functionality      CDFM SEIS_1A-4W64    250v DC Bus A interface with wall 64 (Block Wall Group 3)     0.019 2.18 0.54 0.27    Block Wall          SoV Failure SEIS_12-1a             EECW Pumps (Fragility Group 06-03-01)                0.018 1.34 0.32 0.24    Anchorage         CDFM SEIS_12-1P-1         RHRSW Pumps based on pipe frag (Pipe Calc)              0.018 2.45 0.61 0.47   Soil (Buried         SoV Piping)
Failure SEIS_1B-1      480V BD 1A, 2A, 3A, 1B, 2B, 3B (Fragility Group 02-01)       0.017 1.7  0.32 0.24   Functionality      CDFM SEIS_HINST        Seismic failure of Main Control Room instrumentation       0.016 1.96 0.24 0.32   Functionality      CDFM Page 157 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                           December 2019 Table 5.5-5: Unit 2 SLERF Fragility Group Importance Measures Ranked by F-V Am                 Failure Fragility Group                                      Description                              F-V        u   r                  Fragility Method (g)                 Mode SEIS_1C-6R8                 C SDBD Common Accident Signal Relays (CASA)                     0.013 1.62 0.38 0.24 Functionality        CDFM SEIS_1C-3R8                B,D SDBD Common Accident Signal Relays (CASA)                    0.013 1.62 0.38 0.24   Functionality      CDFM SEIS_1C-                      Wall 63 falls towards 480v SD BD 2A and 2B                  0.012 2.18 0.54 0.27     Blockwall          SoV 1W63SD SEIS_5-4                Panels group 4-1 Control Room panels Lower Fragility              0.011 2.04 0.38 0.24   Functionality  Representative (Fragility Group 20-02 and 20-03)
SEIS_1C-4R3                              3A,3C DG BKR Trip relay (R3)                       0.008 1.26 0.32 0.24   Functionality      CDFM SEIS_1C-1R6                      A SDBD 480 Transformer Trip relays (50G)                   0.008 1.29 0.38 0.24   Functionality      CDFM SEIS_1C-                Wall 63 falls toward 480v RMOV 2A and 4kv SD BD C                 0.007 2.18 0.54 0.27     Blockwall          SoV 1W63MOV SEIS_19-4             CAD NITROGEN STRG TNK (084) (Fragility Group 21-06)                  0.007 1.22 0.26 0.24    Anchorage         CDFM SEIS_9-1                                EDG (Fragility Group 17-01)                       0.006 2.03 0.32 0.24    Functional    Representative SEIS_1A-2                         250v DC bus B (Fragility Group 01-02)                    0.006 1.89 0.32 0.24    Functional    Representative SEIS_5-9-1                            HPCI/RCIC PNLS (925-0058&63)                         0.005 1.82 0.32 0.24    Anchorage     Representative

See Section 5.7 for additional discussion on these representative fragilities Page 158 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                  December 2019 Table 5.5-6: Unit 3 SLERF Fragility Group Importance Measures Ranked by F-V Am                Failure Fragility Group                          Description                           F-V        u   r                  Fragility Method (g)                Mode SEIS_LOOP                   LOOP (Loss of Offsite Power)                    0.813 0.30 0.45 0.30     Ceramic        Table 6-1 insulators   NUREG/CR-6544 SEIS_5-2B                     Initiation relays and panels                  0.195 0.97 0.32 0.24    Anchorage         CDFM SEIS_2-1-1               Unit Batteries (Fragility Group 15-03)             0.135 1.32 0.38 0.24    Anchorage     Representative SEIS_1B-2-2      480v BD 219 3EA, 3EB (U3) (Fragility Group 01-05-02)        0.066 1.14 0.32 0.24    Anchorage         CDFM SEIS_5-8           REACTOR PROTECTION & NSS PNL (18-02)                    0.054 1.22 0.32 0.24   Functionality  Representative SEIS_12-1P-2             EECW Pumps based on pipe frag calc                  0.031 2.48 0.60 0.48   Soil (Buried         SoV Piping)
Failure SEIS_1B-2-1a      480v BD 219 A, B (U1, U2) (Fragility Group 01-05-01)       0.029 1.95 0.56 0.22    Anchorage           SoV Anchorage Failure SEIS_2-1W8       Battery and Wall 8 or 47 interface (Block Wall Group 5)     0.025 2.12 0.41 0.29    Block Wall          SoV Failure SEIS_11-1R3      Relay group 3 for group SEIS_11-1 (EECW Pp A3 UV            0.024 0.61 0.32 0.24   Functionality      CDFM device)
SEIS_1C-4          U3 4kv SD BD EA and EC (Fragility Group 03-03)           0.017 1.09 0.32 0.24   Functionality      CDFM SEIS_1B-2-1f      480v BD 219 A, B (U1, U2) (Fragility Group 01-05-01)       0.017 2.61 0.64 0.24   Functionality        SoV Functional Failure SEIS_BLD-IPS                      Intake Pumping Station                      0.016 1.90 0.26 0.24    Structural        CDFM Analysis SEIS_1C-4R8     3EA,3EC SDBD Common Accident Signal Relays (CASA)            0.015 0.63 0.32 0.24   Functionality      CDFM SEIS_14-1R1-2    Relays for Group SEIS_14-1, Relay Group 1 (HPCI/RCIC         0.010 1.62 0.32 0.24   Functionality      CDFM Isolations)
Page 159 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                            December 2019 Table 5.5-6: Unit 3 SLERF Fragility Group Importance Measures Ranked by F-V Am                Failure Fragility Group                                      Description                               F-V        u   r                  Fragility Method (g)                Mode SEIS_11-1R4               Relay group 4 for group SEIS_11-1 (EECW Pp A3 OC                   0.010 0.94 0.51 0.27 Functionality          SoV device)
SEIS_5-2A2-1                          Initiation relays and panels (20-07)                   0.010 1.96 0.32 0.24   Functionality      CDFM SEIS_1C-              Wall 95 falls towards 480v RMOV 3A and 250vdc RMOV                   0.010 2.18 0.54 0.27    Block Wall          SoV 1W95MOV                                               BD 3A                                                           Failure SEIS_1B-1               480V BD 1A, 2A, 3A, 1B, 2B, 3B (Fragility Group 02-01)              0.008 1.7  0.32 0.24   Functionality      CDFM SEIS_1C-                      Wall 95 falls towards 480v SD BD 3A and 3B                   0.008 2.18 0.54 0.27    Block Wall          SoV 1W95SD                                                                                                               Failure SEIS_12-1P-1                   RHRSW pumps based on pipe frag (Pipe Calc)                    0.007 2.45 0.61 0.47   Soil (Buried         SoV Piping)
Failure SEIS_12-1b             RHRSW pumps and EECW Alternate (Fragility Group 06-                  0.006 1.45 0.32 0.24    Anchorage         CDFM 03)

See Section 5.7 for additional discussion on these representative fragilities.
Page 160 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 The EPRI SYSIMP software was used to calculate the importance measure of each fragility group, considering the combined F-V importance across all the seismic initiator bins.
LOOP represents the most significant contributor, which is consistent with the results of previous SPRA studies across the nuclear industry that have found that extended LOOP events are dominant for seismic risk. The most important fragility groups for Unit 1 are SEIS_5-2B (Initiation relays and panels) and SEIS_2-1-1 (Unit Batteries (Fragility Group 15-03)). SEIS_5-2B is important because its loss affects many accident initiation signal relays. SEIS_2-1-1 is important because it affects the availability of DC power from the 250V battery boards. For Unit 2 and Unit 3, the most important fragility groups are also SEIS_5-2B and SEIS_2-1-1.
5.5.4 SLERF Component Importance (Non-Seismic Failures)
Components were determined to be risk significant if the components RAW is greater than two or its F-V is greater than 0.005. Components are considered risk-significant if the component has a F-V value greater than 0.005 or a RAW greater than two for either the SCDF or SLERF importance measures.
One individual component-related basic event related to a main battery test and maintenance term (TM_0BATA2480001=0-BATA-248-0001 (MAIN BATTERY 1)
UNAVAILABLE DUE TO T&M) was risk-significant for Unit 1. Table 5.5-7 contains the risk-significant RAW importance measures for each individual component, common cause group, and test and maintenance basic event that appears in the seismic cutsets for Unit 1 SLERF.
One individual component-related basic event related to a main battery test and maintenance term (TM_0BATA2480002=0-BATA-248-0002 (MAIN BATTERY 2)
UNAVAILABLE DUE TO T&M) was risk-significant for Unit 2.
Table 5.5-8 contains the risk-significant RAW importance measures for each component, common cause group, and test and maintenance basic event that appears in the seismic cutsets for Unit 2 SLERF.
One individual component-related basic event related to a main battery test and maintenance term (TM_0BATA2480003=0-BATA-248-0003 (MAIN BATTERY 3)
UNAVAILABLE DUE TO T&M) was risk-significant for Unit 3. Table 5.5-9 contains the risk-significant RAW importance measures for each component, common cause group, and test and maintenance basic event that appears in the seismic cutsets for Unit 3 SLERF.
Page 161 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table 5.5-7: Unit 1 SLERF Risk-Significant Individual Component and Common-Cause Group Importance by RAW Basic Event                Description                                               RAW U1_CCFMECH                 CCF of mechanical equipment causes failure to scram        21.4 U0_24803BATFR_1_2_3        CCF of three components: BATFR0BATA2480001 &               13.6 BATFR0BATA2480002 & BATFR0BATA2480003 FANFR1FAN_0710602          SAI PANEL 9-82 FAN COOLER FAN-71-602 FAILED                 4.6 TM_1BDDD2810001A           250V RMOV BD 1A UNAVAILABLE DUE TO TEST AND                 4.5 MAINTENANCE TM_0BATA2480001            0-BATA-248-0001 (MAIN BATTERY 1) UNAVAILABLE DUE            4.5 TO T&M BUSFR1BDDD2810001A         250V RMOV BD 1A BUS FAILED                                  4.4 FUSSO0FU2_2801_111         FUSED SWITCH 111 FAILS                                      4.4 BATFR0BATA2480001          BATTERY 1 FAILS                                             4.3 BUSFR0BDDD2800001          BATTERY BD. 1 FAILS                                         4.3 CBKXO0BKR_2801_110         BREAKER 110 TRANSFERS OPEN                                  4.2 CBKXO0BKR_2801_202         FEEDER BRK 202 TRANSFERS OPEN DURING                        4.2 OPERATION.
CBKXO1BKR_2811A_2D         BREAKER 2D TRANSFERS OPEN                                   4.2 U0_08208DGGFR_ALL          CCF of all components in group 'U0_08208DGGFR'              3.4 U1_07102FANFR_1_2          CCF of two components: FANFR1FAN_0710601 &                  3.3 FANFR1FAN_0710602 U0_21104CBKFO_ALL          CCF of all components in group 'U0_21104CBKFO'              3.3 U0_03008FANFD_ALL          CCF of all components in group 'U0_03008FANFD'              3.3 U0_03002FANFR1_1_2         CCF of two components: FANFR0FAN_0300072 &                  3.3 FANFR0FAN_0300073 U0_08208DGGFD_ALL          CCF of all components in group 'U0_08208DGGFD'              3.3 U0_21104CBKFO_1_2_3        CCF of three components: CBKFO0BKR_211A_003 &               3.3 CBKFO0BKR_211B_002 & CBKFO0BKR_211C_022 U0_21104CBKFO_1_2_4        CCF of three components: CBKFO0BKR_211A_003 &               3.3 CBKFO0BKR_211B_002 & CBKFO0BKR_211D_022 U0_08208DGGFR_1_2_3        CCF of three components: DGGFR0EDG_082__DGA &               3.0 DGGFR0EDG_082__DGB & DGGFR0EDG_082__DGC U0_08208DGGFR_1_2_4        CCF of three components: DGGFR0EDG_082__DGA &               3.0 DGGFR0EDG_082__DGB & DGGFR0EDG_082__DGD U0_03008FANFR_ALL          CCF of all components in group 'U0_03008FANFR'              3.0 U0_08208DGGFD_1_2_3        CCF of three components: DGGFD0EDG_082__DGA &               2.3 DGGFD0EDG_082__DGB & DGGFD0EDG_082__DGC U0_08208DGGFD_1_2_4        CCF of three components: DGGFD0EDG_082__DGA &               2.3 DGGFD0EDG_082__DGB & DGGFD0EDG_082__DGD Page 162 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table 5.5-8: Unit 2 SLERF Risk-Significant Individual Component and Common-Cause Group Importance by RAW Basic Event                Description                                               RAW U2_CCFMECH                 CCF of mechanical equipment causes failure to scram        14.7 U0_24803BATFR_1_2_3        CCF of three components: BATFR0BATA2480001 &                7.4 BATFR0BATA2480002 & BATFR0BATA2480003 FANFR2FAN_0710602          SAI PANEL 9-82 FAN COOLER FAN-71-602 FAILED                 4.4 TM_0BATA2480002            0-BATA-248-0002 (MAIN BATTERY 2) UNAVAILABLE DUE            4.3 TO T&M TM_2BDDD2810002A           250V RMOV BD 2A UNAVAILABLE DUE TO TEST AND                 4.3 MAINTENANCE BUSFR2BDDD2810002A         250V RMOV BD 2A FAILS                                       3.9 FUSSO0FU2_2802_111         FUSED SWITCH 111 FAILS                                      3.9 BATFR0BATA2480002          BATTERY 2 FAILS INCLUDING COMMON CAUSE                      3.9 BUSFR0BDDD2800002          BATTERY BD. 2 FAILS.                                        3.9 U0_08208DGGFR_ALL          CCF of all components in group 'U0_08208DGGFR'              2.6 U0_03008FANFD_ALL          CCF of all components in group 'U0_03008FANFD'              2.5 U0_03002FANFR1_1_2         CCF of two components: FANFR0FAN_0300072 &                  2.5 FANFR0FAN_0300073 U0_08208DGGFD_ALL          CCF of all components in group 'U0_08208DGGFD'              2.4 U0_21104CBKFO_1_2_4        CCF of three components: CBKFO0BKR_211A_003 &               2.4 CBKFO0BKR_211B_002 & CBKFO0BKR_211D_022 U0_03008FANFR_ALL          CCF of all components in group 'U0_03008FANFR'              2.3 U2_07102FANFR_1_2          CCF of two components: FANFR2FAN_0710601 &                  2.2 FANFR2FAN_0710602 Page 163 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table 5.5-9: Unit 3 SLERF Risk-Significant Individual Component and Common-Cause Group Importance by RAW Basic Event                 Description                                               RAW U3_CCFMECH                  CCF of mechanical equipment causes failure to scram        13.8 U0_24803BATFR_1_2_3         CCF of three components: BATFR0BATA2480001 &                7.0 BATFR0BATA2480002 & BATFR0BATA2480003 TM_0BATA2480003             0-BATA-248-0003 (MAIN BATTERY 3) UNAVAILABLE DUE            4.3 TO T&M FANFR3FAN_0710602           SAI PANEL 9-82 FAN COOLER FAN-71-602 FAILED                 4.2 TM_3BDDD2810003A            250V RMOV BD 3A UNAVAILABLE DUE TO TEST AND                 4.2 MAINTENANCE FUSSO0FU2_2803_111          FUSED SWITCH 111 FAILS                                      3.8 BUSFR0BDDD2800003           BATTERY BD. 3 FAILS                                         3.8 BATFR0BATA2480003           BATTERY 3 FAILS DURING OPERATION INCLUDING                  3.8 COMMON CAUSE BUSFR3BDDD2810003A          250V RMOV BD 3A FAILS                                       3.7 U0_08208DGGFR_ALL           CCF of all components in group 'U0_08208DGGFR'              2.6 U3_07102FANFR_1_2           CCF of two components: FANFR3FAN_0710601 &                  2.3 FANFR3FAN_0710602 U0_08208DGGFD_ALL           CCF of all components in group 'U0_08208DGGFD'              2.3 CBKXO0BKR_2803_110          BREAKER 110 TRANSFERS OPEN                                  2.3 CBKXO3BKR_2803_203          FEEDER BREAKER 203 TRANSFERS OPEN DURING                    2.3 OPERATION CBKXO3BKR_2813A_2D          BREAKER 2D TRANSFERS OPEN DURING OPERATION                  2.3 5.5.5 Significant Human Failure Events Significant post-initiator operator actions are defined as those operator action basic events that have an F-V value greater than 0.005 or a RAW greater than two. Note that the common methods of calculating RAW for basic events will not yield useful results for the HRA events due to the processing of combination events. This is because the events are set to one during the quantification process, and a recovery event representing the combination or single event is appended to the cutset. Therefore, setting the event to one to determine the RAW value has no effect on SLERF. The F-V values of each operator action were determined in SYSIMP by defining groups where each particular operator action appearing in each seismic HRA bin (S1, S2, S3 and S4) were simultaneously set to FALSE in the combined cutset file to determine the combined importance across all seismic bins. Tables 5.5-10, 5.5-11, and 5.5-12 list the operator actions that were determined to be risk-significant in the seismic model for Unit 1, Unit 2 and Unit 3 SLERF, respectively.
Page 164 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                               December 2019 Table 5.5-10: Risk-Significant Operator Actions for Unit 1 SLERF Operator Action                   Description                                                              F-V HFA_0280ALNALTBBD                 Operator fails to align alternate feeder                              3.52E-02 HFA_0001HPRVD1                    Failure to initiate reactor-vessel depressurization (transient        3.44E-02 or ATWS)
HFA_0073MANLEVEL                  Operator fails to manually control level with HPCI                    3.40E-02 HFA_0001HPRVD2                    Operator Fails to Initiate Depressurization (SLERF)                   2.67E-02 HFA_0TD2_HPI                      Operator fails to manually initiate injection into drywell after      2.67E-02 core damage HFA_OPS_4KVSDBDRESET              Operator reset of 4kV Shutdown Board lockout relays                   9.30E-03 (seismic)
HFA_0IR2_LPI                      Operator fails to manually initiate injection for in-vessel           7.90E-03 recovery HFA_0TD2_LPI                      Operator fails to manually initiate injection into drywell after      5.10E-03 core damage Table 5.5-11: Risk-Significant Operator Actions for Unit 2 SLERF Operator Action                   Description                                                           F-V HFA_0073MANLEVEL                  Operator fails to manually control level with HPCI                 4.86E-02 HFA_0001HPRVD1                    Failure to initiate reactor-vessel depressurization (transient     4.34E-02 or ATWS)
HFA_0TD2_HPI                      Operator fails to manually initiate injection into drywell         3.27E-02 after core damage HFA_0001HPRVD2                    Operator Fails to Initiate Depressurization (SLERF)                3.23E-02 HFA_0280ALNALTBBD                 Operator fails to align alternate feeder                           3.22E-02 HFA_HINST*                        Seismic failure of Main Control Room instrumentation               1.65E-02 HFA_0IR2_LPI                      Operator fails to manually initiate injection for in-vessel        8.90E-03 recovery HFA_0TD2_LPI                      Operator fails to manually initiate injection into drywell         5.80E-03 after core damage HFA_OPS_4KVSDBDRESET              Operator reset of 4kV Shutdown Board lockout relays                5.00E-03 (seismic)
 *Although this is not an operator action but an equipment failure, it has an effect similar to that of multiple operator action failures.
Table 5.5-12: Risk-Significant Operator Actions for Unit 3 SLERF Operator Action                    Description                                                          F-V HFA_0073MANLEVEL                   Operator fails to manually control level with HPCI                3.93E-02 HFA_0001HPRVD1                     Failure to initiate reactor-vessel depressurization               3.46E-02 (transient or ATWS)
HFA_0280ALNALTBBD                  Operator fails to align alternate feeder                          3.24E-02 HFA_0TD2_HPI                       Operator fails to manually initiate injection into drywell        2.99E-02 after core damage HFA_0001HPRVD2                     Operator Fails to Initiate Depressurization (SLERF)               2.78E-02 HFA_OPS_4KVSDBDRESET               Operator reset of 4kV Shutdown Board lockout relays               2.52E-02 (seismic)
Page 165 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                          December 2019 Operator Action                Description                                                       F-V HFA_0231480SDBTIE              Failure to transfer 480V shutdown board to alternate           2.34E-02 source HFA_0IR2_LPI                   Operator fails to manually initiate injection for in-vessel    7.50E-03 recovery HFA_0TD2_LPI                   Operator fails to manually initiate injection into drywell     5.00E-03 after core damage 5.5.6 Significant SLERF Accident Sequences The SPRA evaluated the GTRAN, Direct Core Damage, LLOCA, MLOCA, IOORV, ISLOCA, and ATWS ETs, which were the same ET categories considered in the IEPRA, except for direct core damage, which is unique to the SPRA. The results were dominated by seismically induced LOOP scenarios, which are treated in the GTRAN ET.
An examination of the individual bin results in FRANX shows that the results are dominated by Level 1 accident sequences GTRAN-005, GTRAN-012, GTRAN-011, and GTRAN-001A. The sequence descriptions are given in Table 5.5-13.
Figures 5.5-1, 5.5-2, and 5.5-3 show the relative percentage contribution for each dominant accident sequence based on a consolidated CAFTA cutset file for Unit 1, Unit 2, and Unit 3, respectively. The differences in results between units are primarily the result of power dependency differences for the SPRA.
Page 166 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                     December 2019 Figure 5.5-1: Unit 1 SLERF Accident Sequence Contribution Figure 5.5-2: Unit 2 SLERF Accident Sequence Contribution Figure 5.5-3: Unit 3 SLERF Accident Sequence Contribution Page 167 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                            December 2019 Table 5.5-13: Significant SLERF Accident Sequences for Units 1, 2, and 3 Accident Sequence                             Description                                                       Discussion GTRAN-005/ CET1-  Sequence GTRAN-005 is the same as sequence GTRAN-002              The dominant cutset involves a GTRAN-005 sequence along 008        except all low-pressure injections fail. core damage occurs       with a containment isolation failure of 3 inches or greater (CET about 1.5 hours after scram with the RPV at low pressure.         1, release state 8 in the ET). This could be caused by both inboard and outboard Main Steam Isolation Valves (MSIVs)
In sequence GTRAN-002, the scram successfully occurs, the         failing to close in any of the main steamlines. In the case of the PCS fails, and there are no breaks outside containment or         dominant sequence, this is due to failure of fragility group stuck-open relief valves. HPCI or RCIC is successful for at least SEIS_5-2B (initiation relays and panels). It could also be 4 hours. Early suppression pool cooling is not successful or      caused by failure of high drywell trip signals coupled with an initiated in time to prevent exceeding HCTL and 190 deg F in      operator action failure to manually close the primary the suppression pool. Therefore, long-term HPCI or RCIC is not    containment isolation valves along with a low reactor trip signal successful. The CRD and HPMU systems fail, and manual             failure.
depressurization is challenged and initiated successfully with 2 SRVs at HCTL about 4 hours after the scram. After depressurization, low-pressure injection by RHR in the LPCI mode or CS is successful. ASDC is unsuccessful. Late suppression pool cooling fails allowing the primary containment to pressurize to PCPL in about 13 hours, but the hardened wetwell vent successfully establishes DHR. After the successful vent, the suppression pool no longer provides a successful suction source for RHR or CS, and late injection is not re-established when all the available injection sources fail. These sources are condensate injection, standby coolant injection, shutdown cooling and RHR or CS with suction on the CST.
Core damage is caused by loss of injection, and the RPV is at low pressure.
GTRAN-012/ CET1-  Sequence GTRAN-012 is the same as sequence GTRAN-007              The dominant cutset involves a GTRAN-012 sequence. The CILRT-16      except depressurization fails. Without depressurization, there    operator fails to depressurize the RPV because the SRVs fail are no other available injection sources, and core damage         as a result of loss of nitrogen supply to the X-50 drywell occurs. Core damage occurs in about 30 to 40 minutes with the     penetration. Also, in-vessel recovery fails due to loss of HPCI RPV at high pressure.                                             and RCIC because the suppression pool cooling path is unavailable. Finally, RPV/drywell injection after core damage In sequence GTRAN-007, the scram successfully occurs, the         fails because of loss of drywell sprays and loss of low-pressure PCS fails, and there are no breaks outside containment or         injection. The drywell spray flow path fails because of fragility stuck-open relief valves. Early HPCI or RCIC are unsuccessful     group SEIS_5-2A2-2 (initiation relays and panels (20-08)) or as an initial injection source. A cooldown is initiated but HPMU  SEIS_5-2B (initiation relays and panels). Low-pressure Page 168 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                           December 2019 is unsuccessful. CRD may be available but is not challenged       injection fails because core spray and LPCI are both lost. Core because it lacks sufficient capacity to be used as an initial     spray can be lost because the low reactor pressure permissive injection source. When RPV level drops to TAF, manual             signal fails (SEIS_5-2B) or the pumps could be lost due to depressurization is successful. Low-pressure injection by RHR     fragility group SEIS_5-2A2-2 (Initiation relays and panels (20-in the LPCI mode or CS is successful as an initial injection      08)). LPCI can be lost because of pump failures due to fragility source but must be initiated within 30 minutes. ASDC is           group SEIS_5-2A2-2 or the reactor low pressure permissive unsuccessful. Late suppression pool cooling and drywell spray     signal fails (SEIS_5-2B).
are unsuccessful. HWWV successfully accomplishes DHR.
Without SPC, PCPL is reached in approximately 10 hours.
Successful venting fails the suppression pool as a suction source, and RHR and CS with suction on the SP are lost. A post-vent injection source of either RHR with suction on the CST, CS with suction on the CST, shutdown cooling, standby coolant injection, condensate injection, or CRD (1 pump) is unsuccessful. CD occurs due to loss of injection, and the RPV fails at low pressure.
GTRAN-011/ CET1- Sequence GTRAN-011 is the same as sequence GTRAN-007              Depressurization of the RPV fails because of failure of operator CILRT-16      except all low-pressure injections fail. CD occurs in about 30 to action to initiate depressurization. Also, in-vessel recovery fails 40 minutes with the RPV at low pressure.                          due to loss of HPCI and RCIC, which both fail because the suppression pool cooling path is unavailable. The suppression pool cooling is not available because of pump failures due to loss of fragility group SEIS_5-2B (initiation relays and panels).
It could also be due to loss of fragility group SEIS_5-2A2-2 (initiation relays and panels (20-08)). There is a failure to inject into the RPV or drywell after core damage occurs because low-pressure injection and drywell sprays both fail. The two loops of drywell spray fail because of either fragility group SEIS_5-2B or SEIS_5-2A2-2. Core spray loop 1 and 2 also fail because of fragility group SEIS_5-2A2-2 or SEIS_5-2B. Both loops of LPCI also fail due to the loss of components in fragility groups SEIS_5-2B or SEIS_5-2A2-2.
GTRAN-001A/CET1  The scram successfully occurs, the PCS fails, and there are no    This sequence involves a GTRAN-001A sequence along with a breaks outside containment or stuck-open relief valves. HPCI      containment isolation failure of >=3 inches. Core damage or RCIC is successful for at least 4 hours. ASDC fails. CRD       occurs for the same reasons as those discussed in the SCDF fails. Low-pressure injection via core spray or LPCI fail.        descriptions for this sequence. The containment isolation failure occurs in the most dominant sequence because of MSIV failures. This is primarily due to initiation relay failures from fragility group SEIS_5-2B.
Page 169 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 5.6   SPRA Quantification Uncertainty Analysis The nature of a PRA is such that the results have inherent uncertainty; these uncertainties must be understood and appreciated when using PRA results. In addition, exploration of the models, inputs, and results promotes an improved understanding of the analysis, and aids in identifying areas for refinement to reduce uncertainty.
NRC RG 1.200 [45] states that an important aspect in understanding the PRA results is knowing the sources of uncertainty and assumptions and understanding their potential impact. They include: (1) parameter uncertainties; (2) model uncertainties and related assumptions; (3) completeness uncertainties; and (4) assumptions related to scope and level of detail.
The scope of the SPRA was limited to the base PRA results and sources of uncertainty for the at-power, Level 1 PRA plus SLERF for seismic events. The focus was also on epistemic uncertainty that results from incompleteness; it is noted that a PRA also includes aleatory uncertainty that results from randomness. The requirements of PRA applications will be evaluated separately for each application to determine whether sources of uncertainties and assumptions are acceptable. Uncertainties and sensitivities in the IEPRA base model are documented in the Quantification, Uncertainty and Sensitivity Analysis Notebook [44].
5.6.1 Parameter Uncertainty Parameter uncertainty relates to the uncertainty in the computation of the input parameter values used to quantify the model (i.e., initiating event frequencies, component failure probabilities, and HEPs). These uncertainties can be characterized by probability distributions that relate to the degree of belief in their values. A formal propagation of uncertainty is the best way to correctly account for this, and the PRA software UNCERT has the capability to propagate these uncertainties.
SCDF uncertainty analysis results are summarized in Tables 5.6-1 through 5.6-3 and Figures 5.6-1 through 5.6-3. The uncertainty analysis was performed with UNCERT 4.0, using Monte Carlo sampling with 20,000 samples and ACUBE processing of 200 cutsets. Reference [50] shows that the uncertainty bounds are not especially sensitive to the number of ACUBE cutsets processed.
The uncertainty mean for Unit 1 SCDF is 1.51E-05/ry, compared with the point estimate mean of 6.69E-06/ry. The uncertainty mean is typically larger than the point estimate mean for these types of analyses. The 5th percentile is 7.53E-06/ry, the median is 1.36E-05/ry, and the 95th percentile is 2.77E-05/ry.
The uncertainty mean for Unit 2 SCDF is 1.57E-05/ry, compared with the point estimate mean of 6.78E-06/ry. The uncertainty mean is typically larger than the point estimate mean for these types of analyses. The 5th percentile is 7.72E-6/ry, the median is 1.42E-05/ry, and the 95th percentile is 2.89E-05/ry.
The uncertainty mean for Unit 3 SCDF is 1.72E-05/ry, compared with the point estimate mean of 7.50E-06/ry. The uncertainty mean is typically larger than the point estimate mean for these types of analyses. The 5th percentile is 8.57E-6/ry, the median is 1.58E-05/ry, and the 95th percentile is 3.01E-05/ry.
Page 170 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                          December 2019 SLERF uncertainty analysis results are summarized in Tables 5.6-4 through 5.6-6 and Figures 5.6-4 through 5.6-6. The uncertainty analysis was performed with UNCERT 4.0, using Monte Carlo sampling with 20,000 samples and ACUBE processing of 200 cutsets.
The uncertainty mean for Unit 1 SLERF is 6.74E-06/ry, compared with the point estimate mean of 3.20E-06/ry. The uncertainty mean is typically larger than the point estimate mean for these types of analyses. The 5th percentile is 3.39E-06/ry, the median is 6.10E-06/ry, and the 95th percentile is 1.21E-05/ry.
The uncertainty mean for Unit 2 SLERF is 7.27E-06/ry, compared with the point estimate mean of 3.51E-06/ry. The uncertainty mean is typically larger than the point estimate mean for these types of analyses. The 5th percentile is 3.95E-06/ry, the median is 6.64E-06/ry, and the 95th percentile is 1.28E-05/ry.
The uncertainty mean for Unit 3 SLERF is 8.04E-06/ry, compared with the point estimate mean of 4.01E-06/ry. The uncertainty mean is typically larger than the point estimate mean for these types of analyses. The 5th percentile is 4.42E-06/ry, the median is 7.38E-06/ry, and the 95th percentile is 1.39E-05/ry.
The UNCERT analysis included distributions for seismic IEs, seismic fragility estimates, seismic HEPs and combinations, and IEPRA basic events. The seismic bin frequency distributions are presented in Table 5.6-7. Those distributions were generated by the FRANX code assuming lognormal distributions and estimating error factors (EFs) from the various seismic hazard curves input into the code (16th, median, and 84th).
Sampling of the individual seismic bin frequencies was performed using the correlated approach described in the FRANX manual. Seismic failure probability distributions are determined automatically by FRANX given the fragility parameter estimates (Am, R, and U). Distributions for HEPs and combination factors were calculated in the HRA Calculator Version 5.2. Distributions for IEPRA basic events were left unchanged from the IEPRA model.
Page 171 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                   December 2019 Table 5.6-1: Unit 1 SCDF Uncertainty Results Figure 5.6-1: Unit 1 SCDF Uncertainty Results Page 172 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                   December 2019 Table 5.6-2: Unit 2 SCDF Uncertainty Results Figure 5.6-2: Unit 2 SCDF Uncertainty Results Page 173 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                   December 2019 Table 5.6-3: Unit 3 SCDF Uncertainty Results Figure 5.6-3: Unit 3 SCDF Uncertainty Results Page 174 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                   December 2019 Table 5.6-4: Unit 1 SLERF Uncertainty Results Figure 5.6-4: Unit 1 SLERF Uncertainty Results Page 175 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                   December 2019 Table 5.6-5: Unit 2 SLERF Uncertainty Results Figure 5.6-5: Unit 2 SLERF Uncertainty Results Page 176 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                   December 2019 Table 5.6-6: Unit 3 SLERF Uncertainty Results Figure 5.6-6: Unit 3 SLERF Uncertainty Results Page 177 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                         December 2019 Table 5.6-7: Seismic Bin Frequency Distributions Seismic Bin                  Bin PGA               Mean Frequency           Error Factor (EF)
(g)                      (/ry)
          %G01                        0.14                   2.68E-4                       9.08
          %G02                        0.24                   5.56E-5                       5.60
          %G03                        0.42                   2.98E-5                       4.93
          %G04                        0.65                   2.17E-6                       4.77
          %G05                        0.79                   2.19E-6                       4.97
          %G06                        1.16                   1.82E-6                       5.71
          %G07                        1.60                   1.80E-7                       6.96
          %G08                        2.26                   3.32E-7                       9.13
          %G09                         3.3                   5.96E-8                       68.08 Note: Uncertainty in the bin PGA is assumed to be covered in the bin frequency distribution.
5.6.2 Model Uncertainty Model uncertainty arises because different approaches exist to represent plant response. A source of model uncertainty is one related to an issue in which no consensus approach or model exists, and where the choice of approach or model is known to have an effect on the SPRA. These uncertainties are typically addressed by making assumptions; e.g., the approach to address CCFs, or the approach to identify and quantify HFEs. In general, model uncertainties are addressed through sensitivity studies using different models or assumptions.
The guidance provided in EPRI 1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments [51], was used to address sources of model uncertainty and related assumptions. It provides a framework for the pragmatic treatment of uncertainty characterization to support risk-informed applications and decision-making. The process includes identification and characterization of sources of model uncertainty and related assumptions; the following sections summarize those in the Level 1 PRA.
5.6.3 Completeness Uncertainty Completeness uncertainty relates to risk contributors that are not in the SPRA model.
These include known types such as the scope of the PRA, which does not include some classes of IEs, hazards, and operating modes; and the level of analysis, which may have omitted phenomena, failure mechanisms, or other factors because their relative contribution is believed to be negligible. They also include types that are not known such as the effects on risk from aging or organizational changes; and omitted phenomena and failure mechanisms that are unknown. Both can have a significant impact on risk.
No completeness uncertainties were identified for the BFN SPRA, based on the PRA Standard.
5.6.4 Truncation Study A truncation study was performed on the Units 1, 2, and 3 SPRA models (SCDF and SLERF) to ensure that sufficient cutsets were generated to result in an accurate Page 178 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 estimate for SCDF and SLERF. The truncation study is more complex than typically performed for the IEPRA CDF because of several reasons:
: 1. Quantification of the SPRA SCDF and SLERF is performed separately by seismic bin, and the results are then combined to obtain a total SCDF and SLERF estimate.
: 2. ACUBE post-processing of bin cutsets is performed to obtain more accurate cutset summation estimates, and the number of cutsets that can be processed by ACUBE is limited.
: 3. The number of fragility events included in the model may be limited by software and hardware constraints.
Therefore, the truncation study is multi-dimensional. Results of the truncation study are presented in Reference [52], which shows that the BFN SPRA model uses a truncation such that SCDF and SLERF converge for Units 1, 2, and 3.
5.7    SPRA Quantification Sensitivity Analysis This section presents the results of several sensitivity studies chosen to determine the effects of changing various variables, parameters, or assumptions, most of which are particular to the seismic analysis. Most of these sensitivity studies were done for the Revision 0 (Peer Review) model and were not re-performed for the post-Peer Review model. Sensitivity study #2 is not applicable to the post-Peer Review model since late sequences above %G04 are not considered to be early in the current model.
One new sensitivity was run for the post-Peer Review model (sensitivity #6). In the pre-Peer Review model, SLERF sequences that are normally considered late were considered early for earthquakes in bin %G04 and above. In the post-Peer Review model, the late sequences were not considered early. Sensitivity #6 shows the results for Unit 1 SLERF if late sequences are considered early for seismic bins %G04 and above. The sensitivity cases presented below were not re-run subsequent to the Closure Review.
An additional sensitivity study (Sensitivity #7) is documented in Reference [53] that discusses the effects of increasing the Am values of certain fragility groups from their original representative non-detailed (i.e., conservative) fragility values since these were risk-significant fragility groups. The fragility groups changed in the sensitivity study analyzed in Reference [53] were SEIS_2-1-1, SEIS_4-5, SEIS_5-4, and SEIS_5-8.
These four were selected for improved fragility evaluations due to their importance in SLERF. The additional representative fragilities that show up on the top risk contributor list for U2 SLERF (SEIS_9-1, SEIS_1A-2, SEIS_5-9-1) were not included in this evaluation because these were on the lower end of importance and improvements were judged to not provide significant benefit to the results.
The uncertainties in the assessment of the seismic hazard curve, and of SSC fragilities, are captured in the parameters that define these intermediate results; i.e., by the family of seismic hazard exceedance curves and the parameters for each of the SSC fragilities (Am, r, and u). Section 5.6.1 presents the SCDF and SLERF results of the uncertainty analysis, which captured the variation in these risk metrics accounting for these uncertainties.
Page 179 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                     December 2019 Sensitivity studies described below are used to investigate other sources of uncertainty, which impact the modeling of seismic impacts and the quantification methods used.
The following areas were investigated:
* Modeling of Seismic Impacts
* Impact of Modeled FLEX HEPs
* Relay Chatter
* HEP Seismic Bin Divisions
* HEP Recovery Actions Added for Seismic The results for each of the seismic-related sensitivity cases are summarized in Table 5.7-1.
Table 5.7-1: Sensitivity Studies Case #                     Base      Sensitivity    Delta % Notes 1-No FLEX       U1CDF=     5.07E-06   7.23E-06        43%   This sensitivity assumed there was nitrogen        U1LERF=    3.53E-06   3.61E-06         2%   no credit for the backup nitrogen backup to                                                   supply to the hardened wetwell vent hardened                                                    and the FLEX nitrogen supply to wetwell vent                                                drywell control air by assuming the FLEX nitrogen storage bottles had a high likelihood of failure (SEIS_HLF) and the operator action to align them failed (HFA_OPS_FLEXN2ALIGN=1). It resulted in a relatively large increase in SCDF and a small increase in SLERF.
2- Late LERF    U1CDF=     5.07E-06      N/A            --- This sensitivity does not affect CDF.
not early       U1LERF=    3.53E-06   2.45E-06        -30%  It results in a large reduction in above bin                                                   SLERF. The gate U1S_LLR_TOP1
 %G04                                                        was removed from the model.
3-84% hazard    U1CDF=     5.07E-06   5.14E-06         1%   This sensitivity set the hazard curve curve           U1LERF=    3.53E-06   3.59E-06         2%   mean exceedance frequencies to exceedance                                                  their 84% percentile values rather values                                                      than their mean values. There was a small increase in both SCDF and SLERF.
4-Redefine      U1CDF=     5.07E-06   5.07E-06       -0.03% This sensitivity re-defined the bin 7 break point for U1LERF=    3.53E-06   3.48E-06         -2%  and 8 break points so that the EPRI EPRI HRA bin                                                HRA bin S3 break point with bin S4 S3                                                          could more closely match those used in the Sequoyah SPRA. This resulted in a very small reduction in SCDF and SLERF.
5-No Credit for U1CDF=     5.07E-06   8.30E-06        64%   The model was modified to take no relay chatter   U1LERF=    3.53E-06   5.26E-06        49%   credit for operator actions recovery                                                    HFA_OPS_4KVSDBDRESET_S* or actions                                                     HFA_OPS_EDGRESET_S* (they were set to 1 for all bins). This resulted in a relatively large increase in SCDF and SLERF.
Page 180 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                    December 2019 Case #                      Base     Sensitivity    Delta % Notes 6-Late          U1CDF=      6.72E-06     N/A            --  The flag file was modified such that sequences                                                   FLG_LATE_SEIS_EARLY was set considered                                                  to 1 rather than FALSE. Also, since early for bin   U1LERF=     3.29E-06  4.16E-06        27%   there was difficulty in quantifying bin
%G04 and                                                    %G07, all the basic events above. (post-                                               associated with fragility group Peer Review                                                 SEIS_LOOP were set to TRUE for Model)                                                      bin 7.
7-Sensitivity   U1CDF=      6.30E-06  6.30E-06       -2.65% The purpose of this evaluation is to Study for                                                   document the sensitivity study and Representative  U1LERF=     3.00E-06  2.58E-06      -14.24% to determine the impact of Fragilities                                                 increasing fragilities of the risk-(post-Closure                                               significant components that have Review Model)                                               representative fragilities. The increased fragilities are taken from Appendix H of the fragility report
[40]. The study showed that an increase in fragilities did not result in a significant increase in SCDF or SLERF.
Page 181 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 5.8   SPRA Logic Model and Quantification Technical Adequacy The BFN SPRA risk quantification and results interpretation methodology [44] were subjected to an independent peer review against the pertinent requirements in the PRA Standard. The risk quantification and results interpretation methodology were peer reviewed relative to Capability Category II for the full set of SRs in the PRA Standard.
After completion of the subsequent independent assessment, the full set of SRs was met, and the seismic hazard analysis was determined to be acceptable for use in the SPRA.
The peer review assessment and subsequent disposition of peer review findings through an independent assessment are further described in Appendix A, and references [6] and [16].
6.0 Conclusions A SPRA has been performed for BFN in accordance with the guidance in the PRA Standard and the SPID [2]. The point estimate results of the BFN SPRA are summarized below:
Unit 1         Unit 2      Unit 3 Core Damage Frequency              6.30E-06       6.40E-06     7.13E-06 Large Early Release Frequency          3.00E-06       3.10E-06     3.31E-06 Appendix A includes an assessment of plant changes not included in the model and a discussion of how the changes impact the model results.
No seismic hazard vulnerabilities were identified, and no plant actions have been taken or are planned given the insights from this study.
Page 182 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                           December 2019 7.0 References
: 1. NRC Letter to All Power Reactor Licensees et al., 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, March 12, 2012, ML12056A046.
: 2. EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, Electric Power Research Institute, Palo Alto, CA: February 2013.
: 3. TVA Letter to NRC, Tennessee Valley Authoritys Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 31, 2014, ML14098A478.
: 4. TVA Letter to NRC, Spent Fuel Pool Evaluation Supplemental Report for Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal regulations 50.54(f)
Regarding Recommendation 2.1 of the Near-Term, Task Force Review of Insights from the Fukushima Dai-ichi Accident, December 21, 2016, ML16356A596.
: 5. NRC Letter to TVA, Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Staff Review of Spent Fuel Pool Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1 (CAC Nos.
MF3764, MF3765 and MF3766), January 27, 2017, ML17024A164.
: 6. BWR Owners Group, Browns Ferry Nuclear Plant Seismic PRA Peer Review Report Using the PRA Standard Requirements, Revision 0, June 2019.
: 7. ASME/ANS RA-S CASE 1, Case for ASME/ANS RA-Sb-2013 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME and the American Nuclear Society, November 22, 2017.
: 8. ASME/ANS RA-Sb-2013, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME and the American Nuclear Society, June 2013.
: 9. NEI-12-13, External Hazards PRA Peer Review Process Guidelines, Revision 0, Nuclear Energy Institute, Washington, DC, August 2012.
: 10. NRC Letter to NEI, U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os), ML17079A427, May 3, 2017.
Page 183 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                            December 2019
: 11. Fugro PR No.160029-PR-01, Revision 4, Probabilistic Seismic Hazards Analysis for TVA Browns Ferry Nuclear Plant PSHA Results Report, October 2019.
: 12. NUREG-2115, EPRI Product ID 3002005288, Central and Eastern United States Seismic Source Characterization for Nuclear Facilities, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Washington DC 20555, Sep 29, 2015.
: 13. NRC Regulatory Guide 1.208, Revision 0, A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion, U.S. Nuclear Regulatory Commission, March 2007.
: 14. NUREG/CR-6728, Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard and Risk-Consistent Ground Motion Spectra Guidelines, October 2001.
: 15. EPRI Report 3002004396, High Frequency Program, Application Guidance for Functional Confirmation and Fragility Evaluation, Electric Power Research Institute, July 2015.
: 16. Jensen-Hughes Report 006069-RPT-01, Revision 0, Brown Ferry Seismic PRA Fact and Observation Closure Independent Assessment, November 2019.
: 17. Report CJC-BFN-C-001, Revision 1, Updated Soil Failure and Fragility Analysis for the Browns Ferry Nuclear Plant (BFN), September 2019.
: 18. EPRI 3002000709, Seismic PRA Implementation Guide, Electric Power Research Institute, Palo Alto, CA, December 2013.
: 19. TVA Calculation MDN0009992019000269, Revision 1, BFN Seismic PRA Seismic Equipment List, September 2019.
: 20. ENERCON Report TVAEBFN062-REPT-003, Revision 2, Browns Ferry Units 1-3 Seismic PRA Chatter Analysis Report, September 2019.
: 21. TVA Letter to NRC, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 -
Generic Letter (GL) 87-02, Supplement 1, Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue A- 46 and GL 88-20, Supplement 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - Submittal of Seismic Evaluation Reports (TAC Nos. M69431, M69432, M83596 and M83589), June 28, 1996.
: 22. TVA Letter to NRC Browns Ferry Nuclear Plant (BFN) - Unit 1 Response to NRC GL 88-20, Supplement 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - Submittal of Browns Ferry Nuclear Plant Unit 1 Seismic and Internal Fires IPEEE, January 14, 2005.
: 23. Tennessee Valley Authority, CDQ0999960096, Revision 9, Resolution of A-46/IPEEE Seismic Programs at Browns Ferry Nuclear Plant, February 2016.
Page 184 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                          December 2019
: 24. EPRI NP 6041-SL, Revision 1, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Electric Power Research Institute, Palo Alto, CA, August 1991.
: 25. TVA Letter to NRC, L44 121127 001, TVA Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f)
Regarding the Browns Ferry Nuclear Plant Seismic Walkdown Results of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, November 27, 2012.
: 26. TVA Letter to NRC, L44 150831 002, Supplemental Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding the Browns Ferry Nuclear Plant, Unit 3 Seismic Walkdown Results of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, September 1, 2015.
: 27. TVA Letter to NRC, L44 130628 001, Supplemental Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding the Browns Ferry Nuclear Plant, Unit 2 Seismic Walkdown Results of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, June 28, 2013.
: 28. TVA Letter to NRC, CNL-14-210, Tennessee Valley Authoritys Browns Ferry Nuclear Plant Expedited Seismic Evaluation Process Report (CEUS Sites)
Response to NRC Request for Information Pursuant to 10 CFR 50.54(f)
Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, December 22, 2014.
: 29. SQUG/GIP, Revision 3A, SQUG Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, December 2001.
: 30. ENERCON Report TVAEBFN062-REPT-001, Revision 1, Browns Ferry Nuclear Plant Seismic Walkdowns, October 2019.
: 31. SC Solutions Report BFN-17-001, Revision 2, Browns Ferry Nuclear Plant Seismic Probabilistic Risk Assessment: Structural Response Analysis, September 2019.
: 32. ASCE/SEI 4-98., Seismic Analysis of Safety-Related Nuclear Structures, American Society of Civil Engineers, 1998.
: 33. ASCE/SEI Standard 4-16, Seismic Analysis of Safety-Related Nuclear Structures, American Society of Civil Engineers, 2016.
: 34. ASCE/SEI 43-05, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, American Society of Civil Engineers and Structural Engineering Institute, 2005.
: 35. EPRI TR-103959, Methodology for Developing Seismic Fragilities, Electric Power Research Institute, Palo Alto, CA, June 1994.
Page 185 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                         December 2019
: 36. EPRI TR-1019200, Seismic Fragility Application Guide Update, Electric Power Research Institute, Palo Alto, CA, December 2009.
: 37. EPRI 3002012994, Seismic Fragility and Seismic Margin Guidance for Seismic Probabilistic Risk Assessment, Electric Power Research Institute, Palo Alto, CA, September 2018.
: 38. ENERCON Report TVAEBFN062-REPT-002, Revision 2, Browns Ferry Nuclear Components and Structures Fragility Evaluation, December 2019.
: 39. NUREG/CR-4840, Procedure for the External Event Core Damage Frequency Analyses for NUREG-1150, November 1990.
: 40. NUREG/CR-6544, A Methodology for Analyzing Precursors to Earthquake-Initiated and Fire-Initiated Accident Sequences, April 1998.
: 41. TVA Calculation MDN0009992019000266, Revision 1, BFN Seismic PRA Human Reliability Analysis, October 2019.
: 42. TVA Calculation MDN0009992019000258, Revision 1, BFN Seismic PRA Quantification, Sensitivity and Uncertainty Notebook, November 2019.
: 43. Regulatory Guide 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, U.S. Nuclear Regulatory Commission, March 2009.
: 44. PRA Evaluation BFN-0-18-116, Revision 1, BFN Seismic PRA Seismic-Fire Interaction, March 2019.
: 45. EPRI 3002012980, Methodology for Seismically Induced Internal Fire and Flood Probabilistic Risk Assessment, June 2018.
: 46. TVA Calculation NDN00099920070031, Revision 2, BFN Probabilistic Risk Assessment - Internal Flooding Analysis, October 2018.
: 47. EPRI 3002008093, An Approach to Human Reliability Analysis for External Events with a Focus on Seismic, December 2016.
: 48. PRA Evaluation BFN-0-19-065, Revision 0. BFN SPRA Uncertainty Analysis to Address F&O 25-3, May 2019.
: 49. EPRI 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, Electric Power Research Institute, Palo Alto, CA, December 2008.
: 50. PRA Evaluation BFN-0-19-022, Revision 1. Convergence of the BFN Seismic PRA Model, July 2019.
: 51. PRA Evaluation BFN-0-19-115, Revision 0. Updated Fragility Estimates for Selected Components, November 2019.
: 52. BWR Owners Group Report BROWNS FERRY UNITS 1,2,3 PRA Peer Review Report Using ASME PRA Standard Requirements, August 2009.
Page 186 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                         December 2019
: 53. NEI 05-04, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, Revision 2, Nuclear Energy Institute, Washington, DC, November 2008.
: 54. ABS Consulting, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, October 2009.
: 55. ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME and the American Nuclear Society, 2009.
: 56. P3146000102 Revision 0, Focused Scope Peer Review of the Browns Ferry Nuclear Power Plant (BFN) Internal Flood PRA Model Against the ASME PRA Standard Requirements, October 2018.
: 57. P3146100001, Revision 0, F&O Closeout by Independent Assessment Report for the Browns Ferry Nuclear Plant (BFN) Internal Events PRA Model Against the ASME PRA Standard Requirements and NEI 0504 Appendix X, November 2018.
: 58. P0132150002-5175, Revision 0, BFN PRA Focused Scope Peer Review Final Report, August 31, 2015.
: 59. NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, March 2017.
: 60. TVA Calculation NDN00099920100001, Revision 8, BFN Probabilistic Risk Assessment - Summary Document, February 2018.
Page 187 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                  December 2019 8.0 Acronyms and Abbreviations AFE           Annual Frequency of Exceedance Am            Median Acceleration Capacities ANS           American Nuclear Society AOV           Air-Operated Valve ASCE          American Society of Civil Engineers ASDC          Alternate Shutdown Cooling ASME          American Society of Mechanical Engineers ATWS          Anticipated Transient Without Scram BE            Best Estimate BFN           Browns Ferry Nuclear Plant BWROG         Boiling Water Reactor Owners Group CAD           Containment Air Dilution CB          Control Bay CCDP          Conditional Core Damage Probability CCF           Common Cause Failure CDF           Core Damage Frequency CDFM          Conservative Deterministic Failure Model CET           Containment Event Tree CEUS          Central and Eastern United States CEUS-SSC Central and Eastern United States Seismic Source Characterization CG            Center of Gravity CLERP         Conditional Large Early Release Probability CR            Center of Rigidity CRD           Control Rod Drive CS            Core Spray CST           Condensate Storage Tank DCA           Drywell Control Air DGB           Diesel Generator Building DHR           Decay Heat Removal DM            Direct Method DOE           Department of Energy Dp            Compression-wave Damping Page 188 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report            December 2019 Ds            Shear-wave Damping ECCS          Emergency Core Cooling System EDG           Emergency Diesel Generator EECW          Emergency Equipment Cooling Water EF            Error Factor EL            Elevation EPRI          Electric Power Research Institute ESEP          Expedited Seismic Evaluation Process ET            Event Tree FEM           Finite Element Model F&O           Facts and Observations FIRS          Foundation Input Response Spectra FLEX          Diverse and Flexible Coping Strategies FPIE          Full Power Internal Events FSAR          Final Safety Analysis Report FSPR          Focused-Scope Peer Review F-V           Fussell-Vesely GERS          Generic Equipment Ruggedness Spectra GIP           Generic Information Procedure GMC           Ground Motion Characterization GMPE          Ground Motion Prediction Equation GMRS          Ground Motion Response Spectra GTRAN         General Transient HEP           Human Error Probability HCLPF         High Confidence of Low Probability of Failure HCTL          Heat Capacity Temperature Limit HCV           Hardened Containment Vent HF            High Frequency HFE           Human Failure Event HLR           High-Level Requirements HPMU          High Pressure Make-Up HRA           Human Reliability Analysis HVAC          Heating, Ventilating, and Air Conditioning Page 189 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report               December 2019 IE             Initiating Event IEPRA          Internal Events Probabilistic Risk Assessment IOORV          Inadvertent Opening of One Relief Valve IPEEE          Individual Plant Examination for External Events IPS            Intake Pumping Station ISLOCA         Interfacing System LOCA ISRS           In-Structure Response Spectra JCNRM          Joint Committee on Nuclear Risk Management LAR            License Amendment Request LB             Lower Bound LERF           Large Early Release Frequency LF             Low Frequency LLOCA          Large LOCA LOCA           Loss-of-Coolant Accident LOOP           Loss of Offsite Power LPCI           Low Pressure Coolant Injection LMSM           Lumped Mass Stick Model MAFE           Mean Annual Frequency of Exceedance MCC            Motor Control Center MLOCA          Medium LOCA MOV            Motor-Operated Valve MSIV           Main Steam Isolation Valve MSL            Mean Sea Level MSM            Modified Subtraction Method NEI            Nuclear Energy Institute NRC            United States Nuclear Regulatory Commission NSSS           Nuclear Steam Supply System NTTF           Near Term Task Force OBE            Operating Basis Earthquake PCS            Power Conversion Systems PDS            Plant Damage States PGA            Peak Ground Acceleration PRA            Probabilistic Risk Assessment Page 190 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                     December 2019 PSHA          Probabilistic Seismic Hazard Analysis RAW           Risk Achievement Worth RB            Reactor Building RE            Reference Earthquake RG            Regulatory Guide RHR           Residual Heat Removal RHRSW         Residual Heat Removal Service Water RLE           Review Level Earthquake RLGM          Review Level Ground Motion RLME          Repeated Large Magnitude Earthquakes RMOV          Reactor Motor-Operated Valve SASSI         System for Analysis for Soil-Structure Interaction SAF           Site Amplification Factor SEL           Seismic Equipment List SEWS          Screening and Evaluation Walkdown Sheets SFR           Seismic Fragility Element Within ASME/ANS PRA Standard SHA           Seismic Hazard Analysis Element Within ASME/ANS PRA Standard SIET          Seismic Initiating Event Tree SLOCA         Small LOCA SoV           Separation of Variables SPID          Screening, Prioritization and Implementation Details SPR           Seismic PRA Modeling Element Within ASME/ANS PRA Standard SPRA          Seismic PRA SQUG          Seismic Qualification Utility Group SR            Supporting Requirement SRT           Seismic Review Team SRV           Safety Relief Valve SSC           Structures, Systems and Components SSE           Safe Shutdown Earthquake SSEL          Safe Shutdown Equipment List SSHAC         Senior Seismic Hazard Analysis Committee SSI           Soil-Structure Interaction TAF           Top of Active Fuel Page 191 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report       December 2019 TB            Turbine Building TH            Time History TVA           Tennessee Valley Authority UB            Upper Bound UHRS          Uniform Hazard Response Spectrum USI           Unresolved Safety Issue V/H           Vertical over Horizontal Vp            Compression-wave Velocity Vs            Shear-wave Velocity VSLOCA        Very Small LOCA Page 192 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                           December 2019 Appendix A Summary of Seismic PRA Peer Review and Assessment of PRA Technical Adequacy for Response to NTTF 2.1 Seismic 50.54(f) Letter A.1     Introduction This Appendix provides a summary of the SPRA peer review and F&O closure reviews and provides the bases for why the SPRA is technically adequate for the 50.54(f) response.
A.2     Peer Review of BFN SPRA The BFN SPRA was performed in three phases. A pre-visit review, a one-week onsite review, and a post-review phase. The pre-visit consisted of an independent review of SPRA documentation with and exchange of questions and answers between the BFN SPRA team and the peer reviewers. The onsite phase of the BFN SPRA Peer Review was performed May 6 through May 10, 2019, at the Browns Ferry Nuclear Plant (BFN) training center in Athens, AL. As part of the peer review, a walkdown of portions of BFN Units 1, 2, and 3 was performed on May 7, 2019, by several members of the peer review team who have the appropriate SQUG training. The post-review time period was used to resolve comments and perform technical editing on the final peer review report.
The information presented here establishes that the SPRA has been peer reviewed by a team with adequate credentials to perform the assessment, establishes that the peer review process followed meets the intent of the peer review characteristics and attributes in Table 16 of RG 1.200 Rev. 2 [45] and the requirements in Section 1-6 of the PRA Standard [8], and presents the significant results of the peer review.
A.2.1 Summary of the BFN Seismic PRA Peer Review Process The BFN SPRA Peer Review was performed using the NEI 12-13 process and the PRA Standard (ASME/ANS RA-Sb-2013, Code Case 1). The purpose of this review was to establish the technical adequacy of the BFN SPRA (Units 1, 2, and 3) for the spectrum of potential risk-informed plant applications for which the SPRA may be used. The 2019 BFN SPRA Peer Review was a full-scope review of all the technical elements of the BFN at-power SPRA (model files provided in April 2019) against all technical elements in ASME/ANS RA-Sb-2013, Code Case 1.
The peer review team consisted of eight team members with extensive qualifications in all areas of SPRA as required by NEI 12-13. The team members experience averaged more than 20 years in PRA, seismic hazard and fragility analyses, with extensive experience in SPRA and the SPRA Section of the PRA Standard. Team member experience is discussed further in Section A.2.2, with resumes provided in Appendix D of the peer review report [6].
The peer review was performed against the requirements in the PRA Standard (ASME/ANS RA-Sb-2013, Code Case 1), using the peer review process defined in NEI 12-13 [9]. The review was conducted over a four-day period, with a summary and exit meeting on the morning of the fifth day.
Page 193 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 The SPRA peer review process defined in NEI 12-13 involves an examination by each reviewer of their assigned PRA technical elements against the requirements in the PRA Standard to ensure the robustness of the model relative to all the requirements.
Implementing the review involves a combination of a broad scope examination of the PRA elements within the scope of the review and a deeper examination of portions of the PRA elements based on what is found during the initial review. The SRs, in combination with the peer reviewers PRA experience, provide the structure and basis for examining the various PRA technical elements. If a reviewer identifies a question or discrepancy, then that issue is further investigated until it is either resolved or an F&O is written describing the issue and its potential impacts with suggestions for possible resolutions.
For each technical element, i.e., SHA, SFR, SPR, at least two peer reviewers were assigned, with one having lead responsibility for a given area. For each supporting requirement (SR) reviewed, the responsible reviewers reached consensus regarding which of the Capability Categories defined in the PRA Standard the PRA meets for that SR, and the assignment of the Capability Category for each SR was ultimately based on the consensus of the full review team. The PRA Standard also specifies high-level requirements (HLR). Consistent with the guidance in the PRA Standard, Capability Categories were not assigned to the HLRs, but a qualitative assessment of the applicable HLRs in the context of the PRA technical element summary was made based on the associated SR Capability Categories.
As part of the review teams assessment of Capability Categories, F&Os were prepared.
There are three types of F&Os defined in [9]: Findings, which identify issues that must be addressed for an SR (or multiple SRs) to meet Capability Category II; Suggestions, which identify issues that the reviewers have noted as potentially important but not requiring resolution to meet the SRs; and Best Practices, which reflect the reviewers opinion that a particular aspect of the review exceeds normal industry practice. The focus in this Appendix is on Findings and their disposition relative to this submittal.
A.2.2 Peer Review Team Qualifications The review was conducted by Mr. Paul Amico of Jensen Hughes, Mr. Russell Childs of Duke Energy, Mr. Eddie Guerra of ARUP, Mr. Jeffrey Kimball of Rizzo International, Mr.
Lawrence Lee of Jensen Hughes, Dr. Glenn Rix of Geosyntec, Mr. Habib Shtaih of Energy Northwest, and Mr. Philip Tarpinian of Exelon. In addition, support was provided by Dr. Mayasandra Ravindra of MK Ravindra Consulting. The working observer was Ms. Coreen McDonald of American Electric Power (12 years of experience (7 years fragility)), who supported SFR.
Mr. Paul Amico, the team lead, is a nuclear engineer with over 40 years of experience in the performance and management of domestic and international programs involving risk and reliability technology and its application to the design and operation of nuclear plants. He has been involved in Seismic PRA since 1981 and is currently very active in Seismic PRA standards development and the performance of Seismic PRAs.
Dr. Glenn Rix was the lead for the Seismic Hazard Analysis (SHA) technical element.
He has 30 years of experience in geotechnical earthquake engineering and engineering Page 194 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 seismology (particularly for the eastern and central U.S.), and seismic hazard assessment and risk mitigation for civil infrastructure. Dr. Rix is an expert in site characterization; site response analysis; and secondary hazards such as liquefaction-induced ground deformation. He was assisted by Mr. Jeffrey Kimball. Mr. Kimball has 39 years of experience with the evaluation and characterization of natural phenomena hazards and the design of critical facilities to resist these hazards. He led the preparation of Department of Energy (DOE) standards and guides to define requirements and procedures to complete assessment of natural phenomena hazards and has full knowledge of a wide range of nuclear facility regulations, regulatory guides, standards, manuals, guides and review plans associated with nuclear facility design and evaluation. Mr. Kimball is a recognized expert in site characterization; ground motion modeling including site response and probabilistic seismic hazard analysis (PSHA) including guidance for completing PSHA.
Mr. Eddie Guerra was the lead for the seismic fragility analysis (SFR) technical element.
Mr. Guerra has nine years of experience in seismic engineering, including seismic risk assessments for nuclear plants. This includes Seismic Equipment List (SEL) development, building analysis, equipment fragility, and walkdowns. He was assisted by Mr. Rusty Childs. Mr. Childs is the Lead Nuclear Engineer for the Oconee Nuclear Station and has over 30 years of experience in civil engineering and seismic qualification projects in the nuclear industry. Mr. Childss experience in support of Seismic PRA and margin assessments includes the fragility and walkdown scopes for A46/IPEEE program, NTTF 2.3 Walkdowns, ESEP and 2.1 NTTF submittal for Duke Energy's Oconee Nuclear Station. Mr. Childs is the Chairman of the EPRI Seismic Qualification Utility Group (SQUG), where he oversees the application of the SQUG database and methodology for U.S. utilities as well as international vendors. Assistance was also provided Dr. Mayasandra Ravindra, who was originally scheduled to be the SFR lead. He completed the pre-visit review but was unable to attend.
Mr. Lawrence Lee was the lead for the Seismic Plant Response (SPR) technical element. Mr. Lee has a degree in Mechanical Engineering and over 27 years of experience in the risk assessment area. Mr. Lee has contributed to, led and reviewed numerous nuclear power risk assessments (Level 1, Level 2, all modes, internal and external events), as well as numerous other risk-related projects. He has experience in the performance and peer review of SPRAs. He is a Principal Engineer with Jensen Hughes. Mr. Lee was assisted by Mr. Habib Shtaih and Mr. Philip Tarpinian. Mr. Shtaih has 12 years of experience, including the performance of most major elements of Level 1 internal events PRA as well as various applications such as MSPI, SDP and MOV ranking. He is currently the technical lead for the SPRA for the Columbia Generating Station. Mr. Tarpinian has 36 years of experience in the nuclear field in engineering and PRA positions, all at the Limerick Station and Exelon corporate nuclear offices. His work since 2012 has been focused on Exelons post-Fukushima (10CFR50.54f) evaluations regarding seismic, external flooding, FLEX and containment venting.
The peer review team members met the peer reviewer independence criteria in NEI 12-13 [9].
Page 195 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 A.2.3 Summary of the Peer Review The review teams assessment of the SPRA elements is excerpted from the peer review report as follows. Where the review team identified issues, these are captured in peer review findings, the dispositions for which are summarized in the next section of this appendix.
A.2.3.1        Seismic Hazard Analysis (SHA)
As required by the PRA Standard, the frequency of occurrence of earthquake ground motions at the site was based on a probabilistic seismic hazard analysis (PSHA). The seismic source characterization inputs to the PSHA are based on the Central and Eastern U.S. (CEUS) regional seismic source characterization model published in NUREG-2115 (i.e., the CEUS-SSC model). The ground motion characterization (GMC) inputs to the PSHA are based on an updated CEUS ground motion model published by EPRI (Reference 8). The seismic hazard analysis for BFN also accounts for the effects of local site response for those structures, systems, and components that are not founded on CEUS reference rock (i.e., hard rock); site response analyses were performed to calculate Ground Motion Response Spectra (GMRS) and Foundation Input Response Spectra (FIRS) at several elevations (EL): EL 515 ft (GMRS/FIRS1, base of reactor building (RB)); EL 556 ft (FIRS2, base of the diesel generator building (DGB));
EL 515 ft (FIRS3, base of the intake pumping station (IPS)); and EL 565 ft (FIRS4, base of the Yard equipment).
The Senior Seismic Hazard Analysis Committee (SSHAC) methodology defines a process of structured expert interaction (elicitation) that is considered a minimum technical requirement for conduct of a PSHA. The SSHAC process (NUREG/CR-6372 and NUREG- 2117) of conducting a PSHA was used to develop both the SSC and GMC models used as inputs to the analysis. Use of the SSHAC methodology ensures that data, methods and models supporting the PSHA are fully identified and incorporated and that uncertainties are fully considered and quantified in the process at sufficient depth and detail necessary to satisfy scientific and regulatory needs. The SSHAC-related guidance documents define and describe four levels. The level of study is not mandated in the PRA Standard; however, both the SSC and GMC parts of the PSHA were developed as a result of SSHAC Level 3 analyses. In the case of the GMC, a SSHAC Level 2 analysis was carried out to update a prior Level 3 study. These Level 3 studies satisfy the requirements of the PRA Standard related to the method of conduct of the PSHA generally, as well as addressing several individual requirements related to data collection, data evaluation and model development, and quantification of uncertainties supporting HLR-A to HLR-D.
The PRA Standard requires compilation of an up-to-date database, including regional geological, seismological, geophysical data, local site topography, and information on surficial geologic and geotechnical site properties. These data include a catalog of relevant historical, instrumental, and paleoseismic information. The CEUS-SSC study involved an extensive data collection effort that satisfies the requirements of the PRA Standard as it relates to developing a regional-scale seismic source model. In the implementation of the CEUS-SSC model for the BFN site, all distributed (i.e.,
background) and repeated large-magnitude earthquake (RLME) seismic sources in the Page 196 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 CEUS-SSC model were included in the PSHA calculations. By including these seismic sources in the analysis, the contributions of all credible sources of potentially damaging earthquakes to ground motions at the BFN site were considered. Likewise, the EPRI (Reference 8) GMC study involved an extensive data collection effort that satisfies the requirements of the PRA Standard as it relates to developing a GMC model to estimate the range of vibratory ground motion that may occur at the BFN site.
The CEUS-SSC and EPRI regional models discussed above are existing models, and the seismicity database that underpins significant aspects of the CEUS-SSC only includes earthquakes through 2008. According to the PRA Standard, if an existing model is used, a data collection and evaluation effort should be conducted to determine (1) whether new information has become available since the data was compiled for the existing model and, if so, (2) whether any new information challenges the validity of the technical basis of the existing study. It is not the case that identification of new data automatically requires an update to the PSHA existing model. Rather, an evaluation of the new data determines whether the existing model is appropriate for its continued use in the intended application. In the case of the PSHA for the BFN site, the analysts developed an updated seismicity catalog that was quantitatively assessed to ensure that (1) assumptions regarding the distribution of the maximum magnitude are not violated and (2) no new data exists that undermines the earthquake recurrence parameters of sources in the CEUS-SSC model important to the seismic hazard at the BFN site. The analysts also performed a literature review to confirm that no new (since 2012) information is available that may impact the CEUS-SSC model.
The PRA Standard requires consideration of all sources that can potentially cause important vibratory ground motion at the BFN site, including non-tectonic, human-induced earthquakes. The CEUS-SSC model used to assess vibratory ground motions explicitly removes non-tectonic earthquakes, which is appropriate because the underlying causation is different from tectonic earthquakes and is non-stationary (i.e., it may change over relatively short time periods). However, human-induced and other non-tectonic seismicity (e.g., earthquakes from wastewater injection and reservoir-induced seismicity) can produce damaging ground motions in some cases. For the BFN site, a separate seismicity catalog of non-tectonic earthquakes was compiled and evaluated via a screening process. Because of the very low magnitude of non-tectonic earthquakes within 400 km of the site, it was concluded that no revisions to the CEUS-SSC model were required to account for nontectonic seismicity.
The PSHA results are provided over an appropriately wide range of spectral frequencies and annual frequencies of exceedances (AFEs). Uncertainties on the reference rock hazard are quantified, analyzed and reported, as required in the standard. The lower-bound magnitude chosen for the analysis is consistent with standard practice. The results include fractile and mean hazard curves, median and mean uniform hazard response spectra (UHRS), and deaggregation results by magnitude and distance and by seismic source.
The SHA for the BFN site included a site response analysis for structures, systems and components not founded on reference hard rock. As noted previously, GMRS and FIRS were developed for several elevations within the plant. As part of the characterization of the site, an extensive geophysical study was performed in 2016 to evaluate the shear-Page 197 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 wave velocity of the rock material below EL 515 ft at the site to inform the site response analysis. The study utilized state-of-the-art methods, including active and passive surface wave methods, vertical-over-horizontal (V/H) spectral ratios, first arrival information, and sonic logs to constrain the interpretation of the data via joint inversion, resulting in a Best Practice. The site response analysis included the effects of site topography, surficial geologic deposits, and site geotechnical properties on ground motions at the site. However, the data and information related to the various site materials above the sedimentary rock (EL 515 ft) were not compiled in an integrated effort, leading to inconsistencies between analyses for site response, soil-structure interaction (SSI), and soil failure.
Both the aleatory and epistemic uncertainties have been addressed in characterizing the seismic sources, ground motion models, and site response analyses. For the SSC and GMC, the characterization of uncertainties is included in NUREG-2115 (Reference
: 9) and EPRI (Reference 8), respectively. For the site response component, epistemic uncertainty is represented by three shear-wave velocities and two sets of modulus reduction and damping curves for firm rock. Aleatory variability is represented by 60 random realizations of each profile, including random variations in shear-wave velocity and modulus reduction and damping curves. In general, the parameters selected to model each type of uncertainty are consistent with values recommended in EPRI (Reference 8). Correlation between properties is modeled when appropriate.
The reference hard rock hazard calculations are based on the CEUS-SSC and EPRI GMC models. During the development of these models, uncertainties in the seismic sources and ground motion prediction equations were included, and appropriate sensitivity analyses were performed to demonstrate the sensitivity of the results to uncertainties in key model parameters. These sensitivity analyses are also documented in the associated reports. In addition, the PSHA analysts for BFN performed site-specific sensitivity analyses for the ground motion prediction equations (GMPEs),
magnitude completeness, earthquake recurrence rate, and maximum magnitude. A sensitivity analysis was performed to evaluate the potential impacts of implementing the Next Generation Attenuation-East ground motion model at BFN. Sensitivity analyses were also performed to evaluate the sensitivity of the calculated soil amplification factors and/or surface hazard curves to variations in site response parameters, including alternative profile randomization schemes and the presence/absence of a soft layer near the bottom of the Fort Payne formation.
The PRA Standard requires that spectral shapes be based on a site-specific evaluation that considers the contributions of deaggregated magnitude-distance results of the PSHA. The horizontal UHRS used in the SPRA is based on site-specific results and incorporates analysis results for spectral frequencies ranging from 0.5 to 100 Hz (i.e.,
PGA). A Finding under SHA-G1 was developed due to the lack of a sufficient technical explanation and justification for how the high-frequency portion (25 to 100 Hz) of the reference rock spectrum was calculated. Appropriate V/H ratios were used to calculate vertical response spectra.
SR SHA-I1 addresses the bases and methodology used for any screening of the seismic hazards other than vibratory ground motion. The purpose of SHAI1 is to ensure that a screening analysis is performed using a structured approach to ensure that all Page 198 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 possible secondary (or other seismic-related) hazards are identified and appropriately dispositioned. Many secondary hazards screened out for BFN, but soil failure modes related to liquefaction-induced settlement and lateral spreading are evaluated further as part of addressing SR SHA-I2, using an approach to estimate the magnitude of displacement and associated uncertainty that is based on local soil conditions. A Finding was developed regarding the need to more robustly model uncertainties with respect to liquefaction-induced settlement and lateral spreading to provide more confidence in the resulting fragilities.
The PRA Standard requires that documentation of the PSHA that supports the PRA applications, peer review and potential future upgrades of the SHA be provided. This requirement establishes a high standard for documentation of the PSHA that allows for examination of the PSHA methodology, its implementation, and the PSHA results to evaluate whether the approach is appropriate, the analyses were performed correctly, and the results are reasonable. Overall, the PSHA documentation is very good and fulfills these requirements. Items that must be addressed to improve the documentation on specific topics and items that should be addressed to improve the readability of the documentation are included in Findings linked to SHA-J1.
A.2.3.2         Seismic Fragility Analysis (SFR)
The SFR assessment of BFN SPRA covered three principal elements of the fragility analysis: building seismic response analysis, plant walkdowns, and fragility analysis calculations. These three elements are briefly summarized below.
The building seismic response analyses of the BFN structures that feed into the fragility evaluations are based on input response spectra corresponding to the control point (top of rock) UHRS with AFE of 10-5 anchored to a horizontal PGA of 0.52g. BFN developed soil-structure interaction (SSI) models for all the buildings and structures included in the PRA using the computer code SC-SASSI. The modeling was done following the recommendations of ASCE/SEI 4-16. The model configurations and properties were based on available drawings, seismic analysis reports, and other documents. The peer review team concurs that the structural models are realistic.
BFN performed deterministic SSI analyses using five (5) combinations of soil-structural properties varying from a median-centered (Best Estimate) model. Median and 84th percentile structural responses were obtained from multiple deterministic analyses. The analyses were performed for three (3) structures: RB, IPS and the DGB.
Approximations were made to model a single representative RB SSI and two separate DGB+RB models in lieu of a combined 2XDGB+3XRB model. The selection of the representative Unit 1 RB structure among all three RB units is well-justified based on seismic response sensitivity studies.
As part of the building response analysis task, the BFN team investigated the effect of potential impact between the Turbine Building (TB) and the RB structure. A coupled finite element model was developed to assess the impact effects between TB and RB structures, capturing the differing foundation conditions of both the RB (rock-founded) and TB (pile- supported in soft soil), and with nonlinear impact elements between the RB and TB. The analysis results demonstrated that the impact would not result in significant building structural damage, nor meaningful shock response across floors in Page 199 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 the RB that would propagate into cabinets housing vibration-sensitive devices such as relays. Nominal impact-induced vibration effects were included in subsequent fragility evaluations.
The walkdown effort included representatives from the fragility team and the systems modeling team. During the walkdowns, the team verified that SSCs met the high-capacity and inherently rugged capacity criteria.
* Inherently rugged items were judged by the Seismic Review Team (SRT) to have sufficiently high seismic capacity such that they should have a negligible contribution to seismic risk.
* High-capacity items were judged by the SRT to have a HCLPF capacity of at least 2g. A fragility curve with 2g HCLPF capacity convolved with the hazard curve yields a point estimate SCDF of 1.4E-8. This is less than 2 orders of magnitude below the BFN SCDF so it is judged to be an adequate level for screening of rugged components.
* SSCs that did not fall within inherently rugged or high-capacity were judged to require further evaluation.
The walkdown effort relied heavily on previous IPEEE and A-46 walkdown data. Where original A-46/IPEEE Screening and Evaluation Walkdown Sheets (SEWS) existed, they were credited and a walk-by was performed to evaluate interaction and potential changes to the area since the original walkdown. Where original A-46/IPEEE SEWS did not exist, a full walkdown was performed. The existing A- 46/IPEEE SEWS were reviewed and found to be well documented, although the original A- 46/IPEEE SEWS were performed several years prior to the SPRA evaluations, the BFN design change process implements Tennessee Valley Authority (TVA) seismic qualification standards that achieve equipment capacities at least as high as earthquake experience data. Any changes in the configuration of adjacent commodities to SEL items were specifically addressed during the SPRA walk-bys in the seismic interactions review.
The BFN walkdown team utilized training mock-ups when available to pre-walkdown components and gain insights into cabinet load paths, breaker functions and relay sensitivities. The peer review team identified this as a Best Practice for SPRAs and one that should be utilized when possible to ensure optimal efficiency and quality of inspections during plant walkdowns.
Plant-specific data was collected and applied in the fragility calculations. The Walkdown Team collected information to determine whether capacities from seismic experience data and Generic Equipment Ruggedness Spectra (GERS) and SQUG data were applicable to each SSC. The seismic fragility evaluations relied on the walkdown documentation to establish as-built conditions and identify likely seismic failure modes.
SSCs were evaluated for critical failure modes that could govern fragility of the components, including functional, load path, and anchorage failure modes. Potential failure modes were assessed during the walkdown and documented in the seismic walkdown report and corresponding SEWS. Operator pathways for access to SSCs outside the main control room were also walked down.
Page 200 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 Potential seismic-induced flooding and fire sources were considered in the walkdowns.
Credible sources were added to the SEL using a combination of input from the SPRA systems analysts and area based walkdowns to identify additional components that may be flood sources from past earthquake experience. Credible seismic-induced flood and fire interactions were identified during the walkdown for further evaluation. Other potential seismic interactions were also identified during the walkdowns and documented on the SEWS. Key issues included interactions between buildings, interactions between cabinets with chatter non-acceptable relays, and limited masonry wall interactions. The peer review team concurs that the walkdowns met the applicable SRs of the ASME/ANS RA-Sb Code Case.
Fragility parameters were calculated for the SSCs listed in the SEL and credited in the plant SPRA Model. Based on a sample review, fragilities for SSCs in the SEL were determined, whether by assigning high capacities or calculated fragilities, and discussions on the definition of failure modes took place throughout the project.
Seismically induced soil-related failure modes and embankment failures that affect the SSC included on the SEL were developed. Although the final SEL does not include fragilities for soil-related and embankment failures, these are credited in the final SPRA model quantification.
A capacity-based HCLPF screening level of 2g PGA was established for high-capacity SSCs based on a single point estimate risk contribution of 1x10-7. However, no items were screened out of the PRA model and fragility estimates for high-capacity and inherently rugged SSCs were provided to the PRA analyst.
Fragility calculations for all SSCs not categorized as high-capacity and inherently rugged were developed in a two-phase approach. First, fragilities were initially developed for large groups of SSCs that were treated as correlated. A lead component was chosen from the group, and the representative fragility based on scaling of existing IPEEE calculations was assigned. As quantification iterations identified a need for refinement, the fragility groups were divided into more realistic subsets, and separate fragilities were developed for the potentially important SSCs. For components determined to contribute to CDF/LERF, fragilities were calculated based on the Conservative Deterministic Failure Margin (CDFM) hybrid approach and the Separation of Variables (SOV) method as described in industry-accepted EPRI guidelines. SOV calculations were specifically developed for SSCs with high contribution to CDF/LERF.
A review of the final PRA Quantification Notebook showed that some SSCs with fragilities calculated via hybrid or conservative representative calculations were ranked as high contributors to CDF/LERF. However, the BFN SPRA provided documentation for sensitivity PRA (BFN-0-19-028, PRA Evaluation Response) runs showing the minimal impact of such conservatism in the overall PRA results.
Although a large number of SSCs remain in the SPRA model with conservative fragilities (more than 90% of fragilities), the BFN SPRA team justified this scenario based on the individual impact of the risk-significant components on the overall SPRA results (BFN- 0-19-028, PRA Evaluation Response). Overall, the peer review team finds the fragilities developed for the risk-significant SSCs to be acceptable for the BFN SPRA.
Page 201 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 In summary, the fragility analysis generally meets the applicable requirements of the ASME/ANS RA-Sb Code Case #1.
A.2.3.3         Seismic Plant Response Analysis (SPR)
The SPRA model was developed by modifying the Full Power Internal Events (FPIE)
PRA model to incorporate specific aspects of seismic analysis that are different from the FPIE. The logic model appropriately includes seismic-caused initiating events and other failures including seismic-induced SSC failures, non-seismic-induced unreliability and unavailability failure modes (based on the FPIE model), and human errors. Some refinements to the SPRA logic model were identified, including the need for a more comprehensive evaluation for crediting FLEX mitigation strategies.
The BFN SPRA used the EPRI External Hazards Human Reliability Analysis (HRA) approach to account for adjustments to the performance shaping factors to modify the internal events human error probabilities (HEPs), followed by a detailed SHRA of risk-significant human failure events (HFEs). The HRA was generally well-done, but additional refinements were identified to the HRA that could be performed to be more realistic (e.g., assess the impact of seismic effects on HFE time parameters).
The implementation of the EPRI 3002012980 [47] process for seismic-induced fire was comprehensive and complete. Every fire source was considered and clearly dispositioned. Qualitative screening was carefully applied and performed correctly and clearly. The evaluation of seismic-induced fire sources was documented in a clear, comprehensive, and traceable manner. The evaluation of seismic-induced fire impacts was identified as a Best Practice.
No SSCs in the BFN FPIE PRA were screened from the SPRA model. All fragilities associated with FPIE PRA SSCs were incorporated into the SPRA model. The comprehensive modeling of SPRA fragilities helped to ensure that potentially significant risk contributors were not inappropriately screened.
A number of sensitivities were performed to understand the impact of the various modeling and screening assumptions. In addition, a detailed evaluation was performed for the identification and disposition of sources of modeling uncertainty. In these aspects, the quantification of the BFN SPRA is judged to meet the PRA Standard.
The Level 2 SPRA LERF analysis addressed the uncertainty associated with potential extended evacuation times during a seismic event. However, some shortcomings were identified in the modifications incorporated into the Level 2 FPIE LERF analysis to support the development of the Level 2 SPRA LERF analysis.
In conclusion, the SPRA model integrates the seismic hazard, the seismic fragilities, and the systems-analysis aspects appropriately to quantify CDF and LERF, albeit with the above-noted deficiencies.
The seismic-PRA analysis was documented in a manner that facilitates applying and updating the SPRA model.
A.2.3.4         Peer Review Findings Based on the peer review, the BFN SPRA was judged to be consistent with the PRA Standard and can be used for risk-informed applications. If the areas identified for Page 202 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                        December 2019 enhancements in the SPRA impact a specific risk-informed application, then additional bounding analyses may be required to support that application.
In summary, the peer review team concludes that the technical adequacy of the BFN SPRA is very good and meets most of the requirements of the PRA Standard.
However, the peer review team identified specific areas for improving the technical adequacy of the SPRA. These areas are documented as F&Os. At the conclusion of the peer review, there were 33 open Finding-Level F&Os as shown in Table A-1. Four SRs (all in the SPR element) were identified as Not Met.
Table A-1 Summary of Facts & Observations for the Browns Ferry Unit 1, 2 and 3 SPRA Peer Review F&Os Element Findings       Suggestions             Best Practice        Total by Element SHA(1)              4                  2                      1                     7 SFR(1)             4                  0                      1                     5 SPR(1)             25                  3                      1                    29 TOTAL(2)             33                  5                      3                    41 Notes:
(1) F&Os by element refer to linked F&Os (i.e., a single F&O can be linked to more than one SR)
(2) Total refers to unique F&Os (i.e., not linked)
A.3      Revision of Model and Documentation Following the peer review, the BFN SPRA model and documentation were updated to address each of the 33 Finding-Level F&Os. In addition, TVA generated closure documentation for each of the F&Os from the peer review against the PRA Standard (ASME/ANS RA-Sb-2013, Code Case 1).
Subsequently, the updated BFN SPRA model and documentation were subjected to an independent closure review. This review is described in Section A.4.
A.4      Finding-Level F&O Independent Closure Review The BFN Seismic Probabilistic Risk Assessment (SPRA) Finding-Level F&O Independent Assessment & Focused-Scope Peer Review was performed at the TVA Corporate offices in Chattanooga, Tennessee, October 1-4, 2019. The purpose was to perform an independent assessment in accordance with Appendix X of NEI 05-04/12-13
[55, 9] to review TVAs proposed close out of Finding-Level F&Os of record from prior PRA peer reviews against the PRA Standard (ASME/ANS RA-Sb-2013, Code Case 1).
The process used for the independent technical review is outlined in the Appendix X of NEI 12-13, which has been accepted by NRC. The review focused on the closure of the 33 open Finding-Level F&Os.
Page 203 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                            December 2019 The review was based on results of a completed Boiling Water Reactor Owners Group (BWROG) review of the BFN SPRA (final report issued June 2019). The result of this independent assessment is intended to be used to support future License Amendment Request (LAR) submittals. Finding-Level F&O dispositions reviewed and determined to have been adequately addressed through the independent assessment are considered resolved and no longer relevant to the current PRA model. Therefore, these F&Os do not need to be carried forward or discussed in future LAR submittals.
The Independent Assessment Team consisted of 6 team members with extensive qualifications and extensive experience in all areas of SPRA. All reviewers met the criteria specified in NEI 05-04 [55], NEI 12-13 [9], ASME/ANS RA-Sb-2013 PRA Standard Section 1-6.2 [8], and in NRCs memoranda outlining expectations for a finding closure independent assessment. Detailed resumes for each of the team members are provided in the closure review report.
A.4.1 Summary of the Finding Level F&O Independent Technical Review Process Review team criteria (NEI 12-13 Section 2.2) and Review Schedule (NEI 12-13 Section 2-3) were addressed in recruiting and approving the closure review team members and defining the schedule for the review. Reviewer independence was established, approved, and documented in the closure review report. Reviewer experience meets the criteria specified in the NEI guidance documents and ASME/ANS RA-Sb-2013 PRA Standard Section 1-6.2. Overall review team experience is such that there were two qualified reviewers for each F&O.
TVA provided the PRA model files and PRA notebooks sufficiently in advance of the start of the onsite review to allow the reviewers to prepare and conduct a more efficient technical review. As input to the review, TVA provided a copy of the BFN peer review report, the list of peer review findings to be considered, and their suggested resolution of each finding.
In accordance with the guidance in NEI 12-13, Appendix X, a lead reviewer and supporting reviewer was assigned for each Technical Element. The reviewers reviewed the associated Finding-Level F&Os and made the initial determination regarding adequacy of resolution of each finding within their scope. A consensus process was followed during which the full team present on the day of the associated consensus session considered and reached consensus on the adequacy of resolution of each F&O using the appropriate SRs of the PRA Standard for the review criteria. The team performed additional consensus sessions via teleconference to disposition F&Os not fully resolved at the conclusion of the onsite review.
A.4.2 Independent Technical Review Team Qualifications Mr. Lawrence Lee is a Principal Engineer with 27 years of experience in the nuclear field specializing in Probabilistic Safety Assessment. Mr. Lee has experience in leading Level 1 and Level 2 PSA updates (internal and external events), shutdown safety assessment, On-line Maintenance, In-Service inspection of piping, MOV prioritization, air-operated valve (AOV) prioritization, and utility response to NRC compliance using PSA techniques. He has been involved in multiple projects ranging from FPIE updates, Page 204 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 SPRA model development, and shutdown PRA model development. He has also supported various risk-related industry projects through the EPRI and the BWROG.
Mr. Jeffrey Kimball is a Chief Seismologist with RIZZO International. Mr. Kimball has 38 years of experience with the evaluation and characterization of natural phenomena hazards and the design of critical facilities to resist these hazards. He led the preparation of DOE standards and guides to define requirements and procedures to complete assessment of natural phenomena hazards. Mr. Kimball has extensive knowledge of a wide range of nuclear facility regulations, regulatory guides, standards, manuals, and review plans associated with nuclear facility design and evaluation. He is also a recognized expert in site characterization; ground motion modeling including site response and PSHA, including guidance for completing PSHA.
Dr. Glenn Rix is a Senior Principal in Kennesaw, Georgia, with expertise in seismic hazard evaluation, geotechnical earthquake engineering, and performance-based and risk-based analyses. Dr. Rix joined Geosyntec in 2013 after a distinguished 24-year career as a faculty member in the School of Civil and Environmental Engineering at the Georgia Institute of Technology specializing in geotechnical and earthquake engineering.
Mr. Apostolos (Paul) Karavoussianis is a Lead Engineer with 31 years of experience in project management and structural engineering. He manages and leads analysis and design projects of nuclear and heavy industrial facilities, semiconductor fabrication plants, petrochemical process plants, and pulp and paper mills. Mr. Karavoussianis managed and was the technical lead for the SPRA projects of Callaway Energy Center, Wolf Creek Generating Station and the new AP1000 units, Virgil C. Summer Nuclear Generating Station Units 2 & 3 and Vogtle Electric Generating Plant Unit 3 & 4. He has also consulted on the Seismic Fragility Assessment project at Palo Verde Nuclear Generating Station.
Mr. Stuart Lewis is currently a Principal Engineer with 40 years of experience in the nuclear industry. In this role, he draws on this experience to support risk assessments and risk-informed applications for nuclear power plants and to identify opportunities for further development of risk methods. Mr. Lewis provides consulting services in the performance of PRAs and in the application of risk results and insights for improving the safety and operational flexibility of nuclear power plants. Projects have related to risk-informed decision-making; various specific risk-informed applications; assessment of the risk benefits of advanced nuclear fuel concepts; advancements in methods for SPRA; peer reviews, including reviews for closure of existing findings; and developing and providing training of utility PRA staff.
Dr. Todd Radford is a Lead Engineer with 13 years of experience including 9 years at Jensen Hughes. In addition to acting as a Structural Engineering Manager in the Wakefield, MA, office of Jensen Hughes, Dr. Radford has been involved in numerous seismic evaluation projects for Jensen Hughes as both project manager and lead analyst. He is an expert in building modeling and SSI and has led response analysis efforts for multiple SPRA projects for nuclear power plants. Dr. Radford has been responsible for post-Fukushima engineering support including R2.3, ESEP, FLEX, SFP evaluations, and R2.1 HF confirmations. Dr. Radford has also led development efforts Page 205 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 for Jensen Hughes internal engineering analysis software including Spectra, SULTAN, and ANCHOR.
A.4.3 Independent Technical Review Team Conclusions All the F&Os were assessed to be resolved during the closure review. All four SRs that that were previously assessment as Not Met were re-assessed as Capability Category II.
A.5     Summary of SPRA Capability Relative to SPID Tables 6-4 through 6-5 The BWROG performed a peer review of the SPRA in 2019. The SPRA was peer reviewed relative to Capability II for the full set of requirements in the PRA Standard (ASME/ANS RA-Sb-2013, Code Case 1). After completion of the subsequent independent assessment in 2019, which utilized the process given in Appendix X of NEI 12-13, the full set of SRs were met.
The final F&O dispositions are provided in the following pages in this Appendix.
Table A-1 provides the dispositions for the original peer review findings within the scope of the F&O independent assessment.
Table A-1 is sorted by Review Unit in the first column. The columns in the table provide the following information (numbers denote column number):
: 1. Review Unit (SHA, SFR, or SPR).
: 2. The SR number against which the peer review Finding was referenced.
: 3. The original peer review teams assessment of Capability Category for the referenced SR.
: 4. The Finding Number from the peer review report.
: 5. The Finding Description from the peer review report.
: 6. A summary of the Basis and Suggested Resolution for the Finding from the peer review report.
: 7. TVAs description of the resolution of the Finding.
: 8. References to appropriate portions of the BFN SPRA Model and documentation to support TVAs resolution.
: 9. The Independent Assessment Teams assessment of whether TVAs resolution of the Finding represents PRA Maintenance or Upgrade.
: 10. The Independent Assessment Teams basis for Maintenance or Upgrade determination.
: 11. The Independent Assessment Teams assessment of adequacy of the Finding resolution.
: 12. The Independent Assessment Teams assessment of the new Capability Category of the referenced SR given the Finding resolution.
Page 206 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                            December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                          Self-Assessment Reference          Maint (M) or Basis for Maint RU  SR                  Finding No. Description                      Prior Peer Review Assessment           Self-Assessment Closure Basis                                                                           Independent Review Team Disposition                                            Review Team Assessment                                                                                                                         Document(s)                        Upgrade?     (M) or Upgrade SR Assessment SHA C-SHA-I2 Met        20-4        SHA-I2 requires that secondary   As in the SSC, GMC, and SRA            Aleatory and epistemic              CJC-BFN-C-001, Rev 1               Maintenance  The resolution was  In Report CJC-BFN-C-001, Revision 1 (September 2019), sources of               Remains Met hazards that do not screen out   components of the SPRA, there are      uncertainties are modeled in                                                        judged to be PRA    epistemic and aleatory uncertainty in the estimates of liquefaction-be further analyzed to           epistemic and aleatory uncertainties   separate event and logic trees,                                                     Maintenance         induced settlement and lateral spreading have been identified and evaluate their frequency of      involved in estimating the             respectively with an adaptation                                                     because it is a     modeled via logic and event trees, respectively (see Figures 5.1 through occurrence and severity. The     magnitude of liquefaction-induced      to the Latin Hypercube Sampling                                                     refinement of an    5.4). Sources of epistemic uncertainty include: (i) liquefaction triggering method used to estimate the      settlements and lateral spreading. method used for variance                                                            existing            methods for sand-like soils, (ii) shear wave velocity profiles, (iii) residual median liquefaction-induced      The approach used in CJC-BFN-C-        reduction. Incorporation of logic                                                   calculation that    undrained strength models, and (iv) modulus reduction and damping settlement and lateral           001 R0 as shown in Figures 5.1 and     and event trees into the                                                            partitions          curves. Sources of aleatory variability include: (i) earthquake magnitude spreading and associated         5.2 commingles the two types of        analyses methodology results in                                                     uncertainty into    and distance, (ii) various soil properties (see p. 40), (iii), SPT blow count, uncertainty commingles           uncertainty, making it difficult to    changes to much of the text in                                                      two types. The      (iv) the depth to groundwater, (v) soil behavior (clay-like vs. sand-like),
epistemic and aleatory           evaluate whether the results are       the report in order to obtain                                                       method used to      (vi) probability of liquefaction or cyclic softening, and (vii) soil uncertainties and makes it       realistic.                             consistently between the                                                            combine the         amplification. The sources of epistemic and aleatory uncertainty are difficult to judge whether the                                          process, interim results, final                                                     results from each   combined in a manner that yields estimates of Am, Bu, and Br that are results are realistic.                                                  results, and conclusions. Thus,                                                     type to calculate   more accurate and realistic than previous analyses (in which all the the use of the logic and event                                                      Am, Bu, and Br is   above items were modeled as aleatory uncertainty). This F&O is assessed (This F&O originated from SR                                            trees is described in Section 5 of                                                  documented in       as CLOSED.
C-SHA-I2.)                                                              CJC-BFN-C-001 R1 while the                                                          Appendix B of EPRI resulting changes are                                                               (2013). No new incorporated as needed in the                                                       methods were appropriate sections and                                                            used.
Appendices. Specific changes associated with the updated methodology are reflected in the Record of Revision of CJC-BFN-C-001 R1.
SHA C-SHA-J1 Met        20-5        In some cases, the specific      1. Section 3 of the PSHA does not      1. Section 3.0 now refers to the    1 through 6; 160029-PR-01, Rev 3   Maintenance  The resolution was  Three reports were updated to address SHA F&O 20-5; (1) Fugro                  Remains Met steps included in the PSHA are   adequately describe the literature     detailed literature search and                                                      judged to be PRA    Consultants, Inc. (Fugro): Probabilistic Seismic Hazard Analysis (PSHA) for based on technical approaches    search and review of information       evaluation performed in             7; TVA-BFN-SPRA-001, Rev 2                      Maintenance         TVA Browns Ferry Nuclear Plant, PSHA Results Report PR No. 160029-PR-which require additional         and data to determine if the CEUS-     Appendix G of 160029-PR-01                                                          because it is a     01, Revision 3 (August 2019), (2) TVA Report TVA-BFN-SPRA-001, discussion or documentation      SSC model can be used without          Rev. 3.                             8 through 11; CJC-BFN-C-001, Rev 1              correction of       Revision 2, Secondary Seismic Hazards for Browns Ferry Nuclear Plan, to provide sufficient            modification. A list of references                                                                                         documentation       Seismic Probabilistic Risk Assessment (September 2019), and (3) Carl J.
information to ensure that the   consulted to evaluate potential                                                                                            deficiencies that   Costantino and Associates (CJC): Updated Soil Failure and Fragility PSHA results are valid. In these changes to the CEUS-SSC to be          2. Section 5.1 of 160029-PR-01                                                      provides            Analysis for Browns Ferry Nuclear Plant (BFN), Report CJC-BFN-C-001, cases, the PSHA                  considered in the BFN PSHA and a       Rev. 3 includes updated                                                             additional,         Revision 1 (September 2019). F&O 20-5 includes eleven (11) items. The documentation should be          brief summary of the disposition of    documentation on the steps                                                          important           resolution for each item is listed below.
improved to provide an           new information contained in each      taken to derive the HF and LF                                                       information that improved technical basis and     reference should be provided.          smoothed spectrum and the                                                           does not change     1. Appendix G, of Fugro PSHA PR No. 160029-PR-01 Rev. 3 provides documentation.                                                          procedure for smoothing the                                                         the basis for the   documentation of the references reviewed including a brief technical
: 2. Section 8.3 of the report does not  response spectra is included in                                                     inputs and results. assessment for each reference of the possible relevance and need to (This F&O originated from SR     adequately describe the steps taken    Section 8.3.                                                                                            modify the CEUS seismic source model. Appendix G is referenced in C-SHA-J1.)                       to develop the smoothed UHRS                                                                                                                   Section 3.0 of Fugro PSHA PR No. 160029-PR-01 Rev. 3.
(site-specific spectral shape factors  3. The UHRS used at 1E-5 is now                                                                         2. Section 5.1 of Fugro PSHA PR No. 160029-PR-01 Rev. 3 provides an that are scales to match the spectral  included in Section 8.3.                                                                                enhanced explanation of the steps taken to derive the HF and LF accelerations at the seven                                                                                                                     smoothed UHRS.
frequencies. The procedure used to develop smooth spectra should be       4. Text was added to Section 6.4                                                                        3. Section 8.3, Table 8-43, and Figure 8-79 of Fugro PSHA PR No. 160029-explained in more detail.              of 160029-PR-01 Rev. 3 to                                                                               PR-01 Rev. 3 provides the smoothed UHRS for FIRS2 at a MAFE of 1x10-5 acknowledge that there is a                                                                             consistent with the spectra used for the structural analysis. Section 7.4
: 3. Section 8.3 of the PSHA describes   minor portion in principle that                                                                         describes the derivation of V/H ratios at this MAFE, and Section 8.4 the development of the smoothed        could be attributed to epistemic                                                                        describes the development of strain-compatible soil properties at this horizontal GMRS and FIRS and the       uncertainty and that in principle,                                                                      MAFE.
corresponding smoothed vertical        we are not stating that epistemic uncertainty is zero,                                                                          4. Section 6.4 of Fugro PSHA PR No. 160029-PR-01 Rev. 3 provides an GMRS and FIRS. However, the but rather that it is very minor                                                                        enhanced technical explanation and basis for why the site response Structural Response Analysis and insignificant from a                                                                                analysis is adequately modeled using a single base-case shear wave described in SC Solution Report numerical standpoint.                                                                                   velocity (VS) model for the rock layers at the BFN site. Fugro BFN-17-001, Rev. 1, uses the acknowledges that a portion of the uncertainty resulting from Uniform Hazard Response Spectra developing sixty (60) randomized profiles (modeling aleatory variability Page 207 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                    December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                      Self-Assessment Reference Maint (M) or Basis for Maint RU SR            Finding No. Description Prior Peer Review Assessment             Self-Assessment Closure Basis                                                             Independent Review Team Disposition                                           Review Team Assessment                                                                                                     Document(s)               Upgrade?     (M) or Upgrade SR Assessment at a MAFE of 1x10-5 as input into                                                                                                  in VS, layer thickness, and total depth to hard rock) for the rock layers the structural analysis. The PSHA                                                                                                  can be attributed to epistemic uncertainty. The impact on the derivation report should provide sufficient         5. Section 6.2.1 and 6.4                                                                  of hazard curves and associated FIRS and GMRS is judged to be information and explanation to           (160029-PR-01 Rev. 3)                                                                     insignificant.
develop the smoothed horizontal          documents the steps used to and vertical response spectra used       develop the shear wave velocity                                                           5. Sections 6.2.1 and 6.4 of Fugro PSHA PR No. 160029-PR-01 Rev. 3 for the structural analysis.             profiles in the fill.                                                                     provide an enhanced explanation and technical basis of the steps taken
: 6. Section 6.4.1 (160029-PR-01                                                            to develop the VS profiles for the compacted earth fill and crushed rock
: 4. Section 6.4 of the PSHA indicates     Rev. 3) documents what low-                                                               fill. This includes explicit citation and references for the industry average that a single base-case profile (i.e.,   strain damping is used and what                                                           VS values used, and improved comparison of the available site data with no epistemic uncertainty) was found      kappa values were obtained and                                                            the base-case VS profiles used as input for site response modeling.
to be appropriate to model the rock      confirms its consistency.
: 6. Section 6.4.1 of Fugro PSHA PR No. 160029-PR-01 Rev. 3 provides layers at the site for the SRA. The                                                                                                clarification of the low-strain damping used for the rock layers below a decision to neglect epistemic            7. Report TVA-BFN-SPRA-001, Revision 2 includes a discussion                                                          depth of 500 ft. This section also provides a check of the relative kappa uncertainty in the shear wave                                                                                                      contribution for the rock layers as further support to the low-strain velocity profile below El. 515 ft-MSL    of the potential for seismically induced collapse of karst                                                                 damping for the deeper rock layers.
must be better justified as a modeling choice rather than the          features based on the                                                                     7. Section 3 of TVA Report TVA-BFN-SPRA-001 Rev. 2 provides discussion absence of uncertainty in principle. information presented in                                                                  of the steps taken to screen out and eliminate the potential impacts Section 3.3 of the PSHA report.                                                           resulting from possible karst features based on the discussion found in 5.Section 6.4 of the PSHA indicates                                                                                                Section 3 of Fugro PSHA PR No. 160029-PR-01 Rev. 3.
that for the compacted earth fill and    8, 9, 10, 11 CJC-BFN-C-001 is modified in Revision 1 to                                                                 8. Section 2.3 of CJC Report CJC-BFN-C-001, Rev, 1 now indicates that the crushed rock fill, industry average address Items 8, 9, 10, and 11.                                                           relative fragility of the piping and the cables/buried conduit is the shear wave velocity values were As suggested, Section 2.3 is                                                              primary reason for screening out the cables/buried conduit. This section used (listed in Tables 6-2 and 6-3).
modified to include the                                                                   was also revised to reference the TVA drawing that provides insight Insufficient information is provided discussion provided in the                                                                regarding the configuration of the cables.
to understand what specific steps were taken to derive the shear wave      response to GJR-03, Section 5.5
: 9. The test and section that were in Revision 0 of the CJC Report CJC-velocity values for the compacted        is deleted, the duplicate BFN-C-001 were deleted consistent with recommendation.
earth fill and crushed rock fill.        paragraph in Section 6.0 is deleted, and Table D1 is                                                                  10. CJC Report CJC-BFN-C-001, Rev, 1 eliminates the duplicate paragraph
: 6. Section 6.4.1 of the PSHA             updated to reflect the shear                                                              that existed in the previous version of the report.
indicates that the EPRI (rock)           wave velocities used in the analyses (consistent with the                                                             11. Table D1 of CJC Report CJC-BFN-C-001, Rev, 1 was revised to reflect dynamic properties are used to                                                                                                     the rock VS used in the site response analyses for secondary hazards, model the site response of the rock      PSHA).
and the revised values are now consistent with those used in the PSHA layers at the site. Tables 6-3 and 6-4                                                                                             as found in Fugro PSHA PR No. 160029-PR-01 Rev. 3.
do not clearly indicate what low-strain damping is used for the SRA for deeper (> 500 ft within the rock)
This F&O is assessed as CLOSED.
rock layers. The modeling of low-strain damping in the firm rock materials should be explained in more detail.
7.Given the prevalence of limestone in the vicinity of the site that is highly susceptible to the formation of karst features, the seismically induced collapse of karst features should be included as a potential secondary hazard at BFN.
: 8. Section 2.3 of CJC-BFN-C-001 R0 states that "based on the configuration of the cables within the duct bank and cable tunnels, it is judged that sufficient slack in the cables is available to accommodate the potential deformations of the Page 208 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                     December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                            Self-Assessment Reference Maint (M) or Basis for Maint RU  SR                  Finding No. Description                      Prior Peer Review Assessment            Self-Assessment Closure Basis                                                                   Independent Review Team Disposition                                          Review Team Assessment                                                                                                                           Document(s)               Upgrade?     (M) or Upgrade SR Assessment cable tunnel and duct bank caused by potential movement from soil deformation." Similar statements are made for both the power cables for the RHRSW pumps beneath the turbine building Unit 3 and buried conduit exiting Units 1-2 and 3 DGB.
As a result, this mode of failure is screened out, with further justification provided in the response to GJR-03. Although the relative fragility of the piping and the cables/buried conduit is noted (p. 18), the relative fragility is a more intuitive, compelling, and persuasive reason for screening out the cables than is the slack in the cables in the opinion of the PRT.
: 9. Section 5.5 of CJC-BFN-C-001 R0 is potentially misleading because the sensitivity study was performed using soil profiles from Sequoyah Nuclear Plant (SQN), and the application to BFN is questionable.
More importantly, Section1 of the BFN PSHA (PR No. 160029-PR-01, Rev. 2) presents a summary of results from BFN that demonstrate that an equivalent linear site response analysis is indeed valid.
: 10. In CJC-BFN-C-001 R0, the second and third paragraphs in Section 6.0 (p. 51) are duplicates.
: 11. T able D1 of CJC-BFN-C-001 R0 does not reflect the rock shear wave velocities used in the site response analyses for secondary hazards.
SHA C-SHA-B3 Met        20-7        SHA-B3 requires that 'SHA        The geologic and geotechnical           Section 6.2.1 and 6.4 (160029-       160029-PR-01, Rev 3       Maintenance  The resolution was  Existing geotechnical data was reviewed and synthesized jointly by the       Remains Met analysts ensure that the'        information and data described in       PR-01 Rev. 3) documents the                                                 judged to be PRA    two vendors responsible for performing the PSHA and evaluation of geologic and geotechnical data   the PSHA Report (Pr No. 160029-Pr-      clarifications as related to the     CJC-BFN-C-001, Rev 1                   Maintenance         liquefaction-induced settlement and lateral spreading to address and information are sufficient   01) and the report assessing soil       PSHA report.                                                                because it is       inconsistencies in previous, separate interpretations of the data. An to characterize local site       failure mechanisms and                                                                                              limited to a        assessment was completed and documented clarifying: (1) the nature of effects, including their         deformations (CJC-BFN.C-001, Rev.       The text of CJC-BFN-C-001 is                                                reinterpretation of geologic materials on both sides of the reactor building (turbine building associated uncertainties.' Data  0) result in the following              modified in Revision 1 of the                                               existing            side versus berm side), including their low-strain stiffness definition, (2) and information related to the   inconsistencies: (1) whether the        document to incorporate the                                                 geotechnical data. whether the geologic materials under the Yard area are in-situ soils or various site materials above     geologic materials on both sides of     responses to each of the Items                                              No new data were    compacted fill and the similarity of material properties, (3) the extent the bedrock are discussed in     the reactor building (turbine           listed in the F&O. These changes                                            obtained.           and applicability of the SPT blow-count data and how that data is used to the PSHA Report (Pr No.          building side versus berm side) are     are described in Section 3 and                                                                  aid in establishing low-strain stiffness properties, and (4) whether the 160029-PR-01, Rev. 2) and the    the same or different, including        Appendix D.2. It is noted that as                                                               material above El. 515 ft should be treated differently for site response report related to assessing soil their low-strain stiffness definition,  a result of these changes the                                                                   analysis, soil-structure interaction analysis (see F&O 23-4), and soil failure mechanisms and           (2) whether the geologic materials      berm properties used in the                                                                     deformations. The previous inconsistencies were resolved, and analyses deformation (CJC-BFN-C-001,      under the Yard area are in-situ soils   analyses are affected and                                                                       were updated to reflect the revised geotechnical profiles.
Rev. 0).? Review of these        or compacted fill, and (3) the extent   therefore liquefaction demands documents indicates the lack     and applicability of the SPT blow-      resulting in changes to the text                                                                This F&O is assessed as CLOSED.
of an integrated effort to       count data and how that data is         in results and conclusions as Page 209 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                     December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                             Self-Assessment Reference  Maint (M) or Basis for Maint RU  SR                  Finding No. Description                        Prior Peer Review Assessment            Self-Assessment Closure Basis                                                                   Independent Review Team Disposition                                        Review Team Assessment                                                                                                                            Document(s)                Upgrade?     (M) or Upgrade SR Assessment compile and integrate the          used to aid in establishing low-strain  well as in supporting sections available data on in-situ and      stiffness properties. While there is    describing the analyses.
fill soils at the site, leading to no impact on the GMRS/FIRS1/FIRS3       Additionally, as indicated in the inconsistencies between            and the assessment of ground            response to F&O 20-4, the analyses for site response, soil-  motion at the top of bedrock, the       approach used to include structure interaction, and soil    lack of an integrated assessment for    epistemic and aleatory failure.                           the overlying site materials makes it   uncertainty results in changes to difficult to determine if there are     the results. Specific changes to (This F&O originated from SR       any impacts on the assessment of        the report are indicated in the C-SHA-B3.)                         site response including the resulting   Record of Revision of CJC-BFN-C-strain-compatible properties, and       001 Rev. 1 the assessment of soil failure modes and the resulting soil deformations. Refer to F&O 23-4 for the treatment of different soil properties in the SSI.
SFR C-SFR-D2 Met        19-13       Isolation valve FCV-069-0012       This is an important isolation valve    Fragility Report TVAEBFN062-        TVAEBFN062-REPT-002, Rev 1 Maintenance  The closure team    Photographs developed by TVA Radiological Control personnel were           Remains Met was not accessible during          for CDF as it isolates the              REPT-002 BFN Fragility Report                                               assessed and        previously obtained and included in TVAEBFN062-REPT-001. These were walkdowns due to high rad          regenerative heat exchanger in case     Rev. 1 was updated to more                                                  agreed with the     used to screen the valve for spatial interactions and to judge that the lockout. The follow- up            of breakout. Need to provide            technically address the fragility                                           TVA                 valve was of very high seismic capacity (low eccentricity, light operator, resolution states that this        stronger argument for                   capability of FCV 69-0012. Data                                             determination       robust yoke) and was conservatively assigned the standard TVA BFN 3g-valve is assumed similar to        walkdown/fragility resolution.          was collected from TVA                                                      that the Browns     2g capacity from design criteria BFN-50-C-7106. Based on this, FCV-069-001 and 002. This is                                               engineering design basis                                                    Ferry SPRA          TVAEBFN062-REPT-002 Rev. 1 develops a fragility for valve FCV-069-not a strong basis for                                                     documents and valve data                                                    response to this    0012. As there is no longer an assumption of similarity, this F&O is similarity. Reliance on plant                                              information. Spatial interaction                                            F&O is a PRA        assessed as CLOSED.
design documents is preferred.                                             information around FCV69-0012                                               Maintenance was observed based on                                                       activity and not an (This F&O originated from SR                                               photographs developed by TVA                                                Upgrade. The C-SFR-D2.)                                                                 Radiological Control personnel.                                             activity includes The TVA photographs validate                                                documentation that there are no adverse spatial                                           update and interactions associated with FCV                                            fragility 69-0012. Based on review of                                                 calculation using TVA engineering data and                                                    previously used available photographs it can be                                             methods.
concluded that FCV 69-0012 meets all GIP requirements and GERS caveats. Consequently, a HCLPF capacity calculation was developed and inserted into Fragility Report TVAEBFN062-REPT-002 BFN Fragility Report Rev. 1. The result of the HCLPF capacity calculation demonstrated that the controlling isolation valve remained the HCLPF capacity established for the 69-001 valve.
SFR C-SFR-D7 Met        23-3        During the confirmatory            SFR-D7 specifies the identification of  The BFN Seismic Review Team         TVAEBFN062-REPT-001, Rev 1 Maintenance  The closure team    An extent of condition review was performed and a list of credible but     Remains Met walkdowns it was noted for         credible seismic interactions during    (SRT) agrees with the                                                       assessed and        non-consequential interactions was included in TVAEBFN062-REPT-001 cabinet 0-BDAA-211-0000B           the walkdown effort. While the          observation cited by the peer                                               agreed with the     with justification supporting the designation as non-consequential. As along with several other           Seismic Review Team (SRT) may           reviewer in the description of                                              TVA                 documentation has been updated to show that the subject interactions cabinets that fire extinguishers   ultimately conclude that the            F&O 23-3. The CO2 distribution                                              determination       from the F&O were considered, this F&O is assessed as CLOSED.
were hung from wall brackets       significance of these credible          nozzle that is in contact with the                                          that the Browns near SEL equipment. The fire       interaction hazards is negligible, the  top of cabinet skirt of cabinet 0-                                          Ferry SPRA extinguishers are not              identification of these interaction     PNLA-009-0017 does not alter                                                response to this addressed in either the            hazards should be incorporated into     the seismic behavior of the                                                 F&O is a PRA Page 210 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                     December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                             Self-Assessment Reference Maint (M) or Basis for Maint RU  SR                  Finding No. Description                       Prior Peer Review Assessment           Self-Assessment Closure Basis                                                                    Independent Review Team Disposition                                        Review Team Assessment                                                                                                                            Document(s)               Upgrade?     (M) or Upgrade SR Assessment original A-46 SEWS or the         the walkdown process and               cabinet. Further, per Drawing                                                Maintenance SPRA walk by SEWS. Although       documentation with an evaluation       BFN-0-47E610-39-1-CC                                                         activity and not an there are arguments which         of significance.                       Mechanical Control Diagram                                                  Upgrade. The can be made as to why these                                              CO2 Storage, Fire Protection &                                               activity includes fire extinguishers are                                                   Purging System, the CO2                                                     an extent of acceptable as is, it should be                                           system is passive and open                                                   condition review documented that they were                                                nozzle. The impact to the CO2                                                and considered.                                                              spray nozzle has no                                                          documentation consequence on CO2 system.                                                   update with no During the confirmatory                                                  Therefore, although the                                                      new methods or walkdowns it was noted that a                                            interaction is credible, it is non-                                          calculations.
CO2 distribution nozzle along                                            consequential for fragility with a light conduit were in                                             development purpose.
direct contact with the top of the cabinet skirt for cabinet 0-                                         Also, during a seismic event the PNLA-009-0017. Although                                                  fire extinguishers that are hung there are arguments which                                                from wall brackets or hooks can can be made as to why this                                               potentially fall from the hook potential interaction is                                                 and pose an interaction acceptable as is, it should be                                           concern. The SRT judged that documented that they were                                                such interactions even though considered.                                                              credible are non-consequential, as the fire extinguishers are (This F&O originated from SR                                             unlikely to become missiles or C-SFR-D7.)                                                               cause harm to the SEL equipment.
As an extent of condition review, the SRT revisited the field walkdown notes and photos to document the list of credible but non-consequential seismic interaction concerns.
The list is provided in Section 5.28.6 Non-Consequential Credible Interaction Concerns of Walkdown Report, TVAEBFN062-REPT-001, Rev 1.
SFR C-SFR-B5 Met        23-4        A review of the soil properties   This inconsistency in definition of    Revision 1 of the soil failure and    CJC-BFN-C-001, Rev 1      Maintenance  The closure team    Revision 1 of the soil failure and fragility report (document CJC-BFN-C-   Remains Met at the site showed that the soil  soil properties can lead to            fragility report (document CJC-                                              assessed and        001 Rev. 1) includes an assessment of potential differences in the soil column south of the RB            differences in in-structure response   BFN-C-001 Rev. 1) includes an         BFN-17-001, Rev 1                      agreed with the     properties on the north and south sides of the Rector Building (see F&O structure has different           and therefore changes in fragilities. assessment of potential                                                      TVA                 20-7). The results provided in that report show that the properties of the properties when compared                                                 differences in the soil properties                                           determination       soils surrounding the reactor buildings are similar in shear wave velocity with the soil on the north side                                          on the north and south sides of                                              that the Browns     (and thus in stiffness), and the differences are not large enough to of the RB structure. However,                                            the Rector Building. The results                                             Ferry SPRA          significantly affect the computed SSI response of the buildings. As the the RB SSI computer model                                                provided in that report show                                                 response to this    impact of the difference between soil profiles on each side of the RB has uses the soil properties                                                 that the properties of the soils                                             F&O is a PRA        been evaluated and the impact on fragilities determined to be non-corresponding to the north                                               surrounding the reactor                                                      Maintenance         significant, this F&O is assessed as CLOSED.
side at both sides of the RB                                             buildings are similar in shear                                               activity and not an structure. This inconsistency in                                         wave velocity (and thus in                                                   Upgrade. The definition of soil properties                                            stiffness), and the differences                                              activity includes can lead to differences in in-                                           are not large enough to                                                      documentation structure response and                                                   significantly affect the                                                     update for therefore changes in fragilities.                                        computed SSI response of the                                                 additional A justification should be                                                buildings. The applicable text                                               discussion with no provided on the impact of this                                           from document CJC-BFN-C-001                                                  new methods or inconsistency on the fragilities                                         Rev. 1. Section D2.2, is                                                     calculations.
and PRA results.                                                         excerpted below:
Page 211 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                 December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                            Self-Assessment Reference              Maint (M) or Basis for Maint RU  SR                  Finding No. Description                       Prior Peer Review Assessment            Self-Assessment Closure Basis                                                                               Independent Review Team Disposition                                    Review Team Assessment                                                                                                                           Document(s)                            Upgrade?     (M) or Upgrade SR Assessment (This F&O originated from SR                                              deep foundations are used C-SFR-B5.)                                                                to support the Turbine Building to the north of the Reactor Building, [and] limited information (qualitative description) of soil overlaying the bedrock is available in this area. From descriptions in the boring log taken to investigate the underlying rock prior to construction (CD-Q0000-884999, Browns Ferry Geology
                                                                                                              & Seismology Bore Hole Logs),
the soil materials are consistent with the overall site description of clayey alluvial and residual soils.
SFR C-SFR-F1 Met        23-5        The work encompassing the         --Documentation and Reference           1. The reference to the             1. TVAEBFN062-REPT-002, Rev 1.         Maintenance  The closure team    1 - TVAEBFN062-REPT-002 is updated appropriately                       Remains Met fragility analysis is provided in recommendations--                       sensitivity study BFN-0-19-028 is                                                       assessed and three separate notebooks -                                                added in TVAEBFN062-REPT-           2. TVAEBFN062-REPT-001, Rev 1.                      agreed with the     2 - TVAEBFN062-REPT-001 is updated appropriately walkdown report, building         1 - TVAEBFN062-REPT-002, Section        002, Rev 1, Section 13 and                                                              TVA                 3 - SEIS_12-1P-1 is in the Master Fragility Table associated with report and fragility report.      13 documents risk quantification        Section 16.                         3. Master Fragility Table; TVAEBFN062-              determination       component BFN-0-PIPE-023-XXXX and SEIS_12-1P-2 is in the Master These reports provide the         sensitivity studies to address                                              REPT-002 Rev. 1, Appendix D.                        that the Browns     Fragility Table associated with component BFN-0-PIPE-067-XXXX.
input, methodology and            sources of uncertainty in the SPRA      2. Reference for the A-46/IPEEE                                                         Ferry SPRA          TVAEBFN062-REPT-002 is updated appropriately.
results of the overall fragility  model due to potential fragility        SEWS that are documented in         4. Master Fragility Table; TVAEBFN062-              response to this work in relation to SFR. In       improvements and assumptions            TVA document W78 070428             REPT-002, Rev 1, Appendix D.                        F&O is a PRA        4 - Relays BFN-3-RLY-082-DRRA-1 and BFN-3-RLY-082-DRRC-1 have been general, the notebooks are        made during component fragility         003. Walkdown Package No.                                                               Maintenance         assigned fragility group SEIS_1C-4 and Relays BFN-3-RLY-082-DRRB-1 and well organized. However,          development. This sensitivity study     BFN1 CEB A46 IPEEE is added to      5. TVAEBFN062-REPT-002, Rev 1.                      activity and not an BFN-3-RLY-082-DRRD-1 have been assigned fragility group SEIS_1C-5.
there were some specific          is documented in BFN-0-19-028. For      TVAEBFN062-REPT-001. This                                                               Upgrade. The        These are consistent with the fragility groups of the host cabinets, instances where additional        completeness, this sensitivity study    reference is added as Ref. 3.63                                                         activity includes   therefore TVAEBFN062-REPT-002 is updated appropriately.
documentation and                 should be referenced in                 in Section 3, References of                                                           documentation 5 - TVAEBFN062-REPT-002 is updated appropriately corrections to existing           TVAEBFN062-REPT-002, Section 13.        Walkdown Report. Additionally,                                                          update only.
documentation needs to be                                                 the appropriate pointers to                                                                                 As all documentation has been updated appropriately, this F&O is made.                             2 - Add references for A-46             reference are added to the text                                                                             assessed as CLOSED.
walkdowns SEWS to TVAEBFN062-           in Section 1.5.1 USI A-46/IPEEE (This F&O originated from SR      REPT-001. The A-46/IPEEE SEWS are       Walkdowns.
C-SFR-F1.)                        in TVA document W78 070428 003.
Walkdown Package No. BFN1 CEB           3. The following fragility groups A46 IPEEE. This reference should be     are added to the Master added to TVAEBFN062-REPT-001.           Fragility Table
: a. SEIS_12-1P RHRSW Pumps
                                                                      -- Cut and paste and typo --            based on pipe frag (Pipe Calc)
: b. SEIS_12-1P EECW Pumps The following risk significant SSC's    based on pipe frag calc are not listed in Master Fragility table: SEIS_12-1P RHRSW Pumps       4. The relays identified in the based on pipe frag (Pipe Calc)          F&O are not chatter sensitive.
SEIS_12-1P EECW Pumps based         These relays are modeled as on pipe frag calc                       part of the host cabinet and are regrouped with the host cabinet
                                                                      -- Errors in DG fragility group --      Model Fragility Group. This Relays BFN-C5-RLY-082-DRRA-1,           update has been made in the BFN-3-RLY-082-DRRB-1, BFN-3-RLY-        Master Fragility Table.
082-DRRC-1 & BFN-3-RLY-082-DRRD-1 are identified C5in the          5. The excerpt in the finding wrong Model Fragility Group. They       belongs to Section 11.2 (not should be in the Model Fragility        Section 11.1) of Fragility Report.
Page 212 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                             December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                          Self-Assessment Reference         Maint (M) or Basis for Maint RU  SR                  Finding No. Description                    Prior Peer Review Assessment            Self-Assessment Closure Basis                                                                             Independent Review Team Disposition                                        Review Team Assessment                                                                                                                         Document(s)                       Upgrade?     (M) or Upgrade SR Assessment Group of their Host cabinet             The finding has been resolved as suggested.
Refer to Peer Review Report for Table of UNID, Component Type, Basis for SRT Review, Fragility Calculation Group, and Model Fragility Group.
Cut and paste errors in Section 11.1 of TVAEBFN062-REPT-002 (entire labeled as {} need revision)
From the Internal Flooding Analysis (Ref. 16.66), the conditional probability of failure of fire protection piping considering random pipe failure in U1 SE and U3 SW Corner Room is determined as 5.79E-04 and 4.23E-04, respectively.
The flooding frequency is obtained by convolving the block wall failure fragility with the RE hazard. A representative fragility for block wall around the elevator shaft is developed based on {review of the 80-11 design evaluation} for masonry block wall and is documented in Appendix A, Attachment 22 to this report.
The fragility parameters are summarized below:
Am = {1.70}g Beta-R = 0.24 Beta-U = {0.32}
HCLPF = {0.67}g SPR C-SPR-B2 Met        19-1        Based on document BFN-0    The basis for why these F&Os do not     Based on the discussion of the       Methodology, Input and Model; PRA Maintenance  The closure team      Discussion of F&O 6-50                                                     Remains Met 029, BFN has 10 open finding   affect the SPRA is not clear.           FPIE F&O 6-5 and SPRA F&O 19-        Evaluation BFN-0-19-062                        assessed and F&Os on the internal event                                             1, the credit of 74-52 and 76-66                                                    agreed with the       Based on the discussion of FPIE F&O 6-50 and SPRA F&O 19-1, the and 8 open findings F&Os on    FPIE F&O 6-50                           to mitigate the ISLOCA is                                                           TVA                   crediting of outboard isolation MOVs 74-52 and 74-66 to mitigate the the internal flooding PRA. The F&O 6-50 identifies that some of the    questionable because these                                                          determination         ISLOCA is questionable because these valves are not designed to close above open F&Os were           MOVs credited in the BFN ISLOCA         valves are not designed to close                                                    that the BFN SPRA     against the full RPV differential pressure. Even with successful RPV reviewed and it was            fault tree are not tested to close      against the full differential                                                       response to SPRA      depressurization, it is likely that these valves would not isolate the determined that F&O 6-50 and   against full DP. Therefore, there is    pressure. Depressurization or                                                       F&O 19-1 is a PRA     interface pipelines. Thus, the SPRA model has been revised to set MOVs F&O 2-31 may affect the        significant uncertainty if the MOVs     not by the operators, it is likely                                                  Maintenance           74-52 and 74-66 to failed (i.e., setting basic events for MOVs 74-52 and results of the seismic-PRA. could actually close when exposed       these valves would not isolate                                                      activity and not an   74-66 to TRUE in flag file BFN123_Flag_r8_SEIS.FLG) and there is no to full reactor pressure and be able    the interface pipelines. Thus, 74-                                                  Upgrade because       credit taken for operator actions to mitigate ISLOCA scenarios. The to mitigate an ISLOCA event. The        52 and 74-66 are set to failed in                                                   the activity          inboard isolation MOVs 74-53 and 74-67 are calculated to have F&O identifies an example where         the SPRA modeling and there is                                                      includes minor        Am=6.09g (MOV Group 08A-74-3). However, they are conservatively gate U1_ISLVSS_2 credits isolation      no credit taken for operator                                                        modeling              modeled with Am=2.94g (MOV Group 08A-74-4) using SPRA Fragility of MOVs FCV-74-52 and FCV-74-66,        actions to mitigate ISLOCA                                                          enhancements          Group ID SEIS_14-1 that represents a large group of MOVs.
where FCV-74-66 is not designed to      scenarios. The functional failure                                                   (i.e., setting basic  The PRA model fragility group SEIS_14-1 contains various sub-groups of close against full DP.                  modes are the consequential                                                         events to TRUE in     MOVs (Fragility Calculation Groups 08A-74-1, 08A-74-2, 08A-74-3, 08A-effect of the seismic impact;                                                       a flag file) using no 74-4 etc.). The lowest fragility calculated for any of these subgroups has The BFN SPRA review and                 therefore, the current modeling                                                     new methods or        an Am=2.94. This fragility was used to represent the failure of all MOVs disposition for F&O 6-50 states that    is acceptable to map the seismic                                                    calculations.         in SEIS_14-1. MOV FCV-74-67 could have been based on MOV Group all the MOVs included in the ISLOCA                                                                                                               08A-74-3 (Am=6.09g) instead of MOV Group 08A-74-4 (Am=2.94g). It is fault tree are in the same fragility                                                                                                              understood that using MOV Group 08A-74-4 with Am=2.94g as the Page 213 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                   December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                     Self-Assessment Reference Maint (M) or Basis for Maint RU SR            Finding No. Description Prior Peer Review Assessment            Self-Assessment Closure Basis                                                             Independent Review Team Disposition                                           Review Team Assessment                                                                                                    Document(s)               Upgrade?     (M) or Upgrade SR Assessment group. If one valve seismically fails,  group fragility to the group of                                                           fragility of seismic group SEIS_14-1 is conservative. However, the seismic they all fail. Therefore, all the MOVs  valves.                                                                                   group SEIS_14-1 did not show up as being risk significant and there were would fail to operate due to                                                                                                      no ISLOCA sequences that appear in the cutsets. This assumption for mapping from the same fragility         For FPIE F&O 2-31 and SPRA                                                                grouping will be acknowledged in a revision to BFN-0-19-062. The review group. However, this disposition is     F&O 19-1, the valve status                                                                that shows no ISLOCA sequences appear in the cutsets will also be not clear for the BFN ISLOCA            required for LPI and SPC was                                                              documented in the revision to BFN-0-19-062.
discussion.                             listed based on the internal events modeling. These valves                                                             Discussion of F&O 2-31 For example, in the ISLOCA              and their associated basic events representing different                                                             For FPIE F&O 2-31 and SPRA F&O 19-1, the valve status required for LPI evaluation for the RHR B injection                                                                                                and SPC was listed based on the internal events modeling. These valves path, MOV FCV-74-67 is normally         failure modes were all mapped to the same fragility group in the                                                        and their associated basic events representing different failure modes closed (inboard valve) and MOV                                                                                                    (i.e., valve fails to open, valve fails to close) were all mapped to the same FCV-74-66 is normally open              Fragility to Component table.
fragility group associated with a large group of MOVs in the Fragility to (outboard valve), where FCV-74-66                                                                                                 Component table using a conservative fragility value (i.e., Am=2.94g is not designed to close against full                                                                                             using SPRA Fragility Group ID SEIS_14-1).
DP. Given that inboard valve FCV-74-67 is normally closed it is unclear                                                                                            This F&O is assessed as CLOSED.
what seismic failure mode is being applied in the BFN SPRA with respect to contributing to a seismic ISLOCA (e.g., boundary rupture failure mode). In addition, under gate !MOCFC1FCV_0740067 for VALVE FCV-74-67 FAILS TO CLOSE ON LAST DEMAND, the failure mode for FCV-74-67 contributing to an ISLOCA event is modeled as a failure to close event, which appears to assume that the normally closed FCV- 74-67 is normally open.
Although this issue may be rooted in the modeling of the ISLOCA scenarios in the BFN FPIE PRA model, this issue appears to impact the SPRA model in how the MOV seismic fragilities are mapped into the SPRA model. If the MOV seismic failure is inappropriately mapped to model failure of an MOV to close to mitigate an ISLOCA, when the MOV is normally closed (e.g., inboard MOV FCV-74-67), then the assumption that the same MOV seismic fragility would apply to the normally open FCV-74-66 may not be valid. Typical MOV failure modes in an ISLOCA analysis include, but are not limited to: Failure to Hold on Demand, Internal Rupture, Mispositioned (very small probability).
FPIE F&O 2-31 The BFN SPRA review and disposition for F&O 2-31 does not clarify how when swapping from the LPCI mode to the SPC mode, the LPCI injection valves need to cycle closed to prevent potential flow Page 214 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                               December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                              Self-Assessment Reference           Maint (M) or Basis for Maint RU  SR                  Finding No. Description                        Prior Peer Review Assessment             Self-Assessment Closure Basis                                                                           Independent Review Team Disposition                                     Review Team Assessment                                                                                                                             Document(s)                         Upgrade?     (M) or Upgrade SR Assessment diversion if the valves had previously successfully opened for LPCI operation. The same is true if needed to swap from SPC mode to LPCI mode during the mission time.
Although likely true for the BFN SPRA, it is not clarified if the MOV seismic failure mode applies to both Failure to Open and Failure to Close failure modes. In addition, it is not clarified if the LPCI injection valve MOVs and the SPC return valve MOVs are in the same seismic group.
SPR C-SPR-C4 Met        19-3        The fire sources not credited      A significant portion of the credible    Fire ignition sources retained in   Seismic PRA Equipment List Notebook Maintenance  The review team    Review of Attachment C to the BFN Seismic PRA Seismic Equipment List    Remains Met in the SPRA model (assumed         seismic-fire ignition sources were       the BFN Fire Scenario Summary                                                        concurs with the   (Calculation MDN0009992019000269, Rev. 1) confirmed that a large failed or inherently rugged)       not included in the SEL, thus the SEL    list were checked against the                                                        assessment that    number of entries were added to the SEL to reflect equipment that could were not added to the SEL,         is not complete. This is a systematic    SEL. If an ignition source was                                                       the changes to the constitute a fire ignition source.
although ultimately they were      deficiency in the SEL.                   already in the SEL, no further                                                       SPRA constitute evaluated for fragility as if they                                          actions were taken. For ignition                                                     Maintenance. The   This F&O is assessed as CLOSED.
were on the SEL. The SEL is                                                 sources that were not in the                                                         change to the therefore incomplete,                                                       existing SEL, these UNIDs were                                                       SPRA was to although it is noted that there                                             added individually. For ignition                                                     document in the is no effect on the results since                                           sources that did not have UNIDs,                                                     seismic equipment they were included in the                                                   these sources were added to the                                                      list (SEL) the quantitative analysis.                                                      SEL individually by scenario                                                         equipment that name. The note field was                                                             had already been populated listing all additions as                                                   considered to fire ignition sources.                                                               constitute potential ignition sources.
SPR C-SPR-C5 Not Met    19-4        The only credible failures         While there were only two failures       Two basic events were created       Seismic PRA Equipment List Notebook Maintenance  The review team    Review of Attachment C to the BFN Seismic PRA Seismic Equipment List    Met induced by secondary hazards       induced by secondary hazards that        in the SEL to represent the                                                          concurs with the   (Calculation MDN0009992019000269, Rev. 1) confirmed that events coming from SHA-I2 were            were retained, neither of them was       buried piping secondary hazards                                                      assessment that    BFN0PIPE023XXXX and BFN0PIPE067XXXX were added to the SEL liquefaction and lateral           incorporated in the SEL, so a finding    for the RHRSW and EECW                                                               the changes to the to reflect failures of buried piping in the RHRSW and EECW systems, spreading. The credible            under this SR is required.               piping.                                                                              SPRA constitute    respectively.
induced events were identified                                                                                                                                   Maintenance.
and supplied for fragility                                                                                                                                       Additions to the   This F&O is assessed as CLOSED.
analysis, and were treated as if                                                                                                                                 SEL document the they were in the SEL, but they                                                                                                                                   impacts of were not actually incorporated                                                                                                                                   secondary failures into the SEL.                                                                                                                                                    already considered in the SPRA.
SPR C-SPR-C6 Met        19-5        The identification of the failure  While the process followed did           The listed requirements were        Seismic PRA Equipment List Database Maintenance  The review team    Review of the Seismic Equipment List (Access database Composite         Remains Met modes of interest was only         result in consideration of the PRA       added as columns to the SEL.                                                         concurs with the   SEL_draft_20190910 - Copy.accdb) confirmed that information has been performed for the dominant         failure modes for the dominant           For the initial state, components                                                    assessment that    added for each SSC to indicate normal status, desired status, and PRA SSCs. Initial fragilities were     SSCs, the fact that the information      like pumps and valves in the PRA                                                     the changes to the (basic event) failure mode(s).
provided before the SEL was        for all SSCs is missing from the SEL     were filled in based on the de-                                                      SPRA constitute developed, using lists of          and thus is not available for the        energized state shown in flow                                                        Maintenance.       This F&O is assessed as CLOSED.
equipment such as IPEEE,           fragility analysis is a deficiency in    diagrams or system notebooks.                                                        Additions to the NTTF 2.3, and ESEP. These          the SEL.                                 The desire state and failed                                                          SEL document were used for the initial runs,                                             support system were filled in                                                        information that and the dominant contributors                                               where applicable if information                                                      provides a clearer were discussed, including the                                               was readily available. The basic                                                     understanding of Page 215 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                          Self-Assessment Reference              Maint (M) or Basis for Maint RU  SR                  Finding No. Description                      Prior Peer Review Assessment           Self-Assessment Closure Basis                                                                               Independent Review Team Disposition                                       Review Team Assessment                                                                                                                         Document(s)                            Upgrade?     (M) or Upgrade SR Assessment specific PRA failure mode. The                                          events failure modes were taken                                                         the relevant status PRA failure mode was not                                                from the internal events analysis                                                       and failure modes provided for the non-                                                   and input into the database.                                                            of the SSCs dominant contributors and not                                                                                                                                   considered in the documented in the SEL. This                                                                                                                                     SPRA.
SR is considered MET because the correct information was exchanged for the dominant contributors and the fragilities were based on that, and for the non-dominant contributors it is not as important. However, the SEL is missing this information and is therefore incomplete.
SPR C-SPR-D5 Met        19-8        Detailed analysis was            HFA_0IR2_LPI has LERF F-V 0.0084       This HEP, along with all the        Seismic PRA Human Reliability Analysis Maintenance  The review team     Detailed assessments have been provided for event HFA_0IR2_LPI for        Remains Met performed on all HEPs that       for U1 and 0.009 for U2                other seismic HEPs, was subject     Notebook                                            concurs with the    use in the internal-events PRA and for seismic HRA bins 1 through 4.
had an F-V importance greater                                           to a detailed analysis. As part of                                                      assessment that     These assessments are documented in Appendix A of the Seismic PRA than 0.005, with one                                                    the resolution to F&O 19-9, the                                                         the changes to the  HRA Notebook (MDN0009992019000266, Rev. 1).
exception. HFA_0IR2_LPI does                                            screening approach was                                                                  SPRA constitute not have detailed HRA.                                                  abandoned and a detailed HRA                                                            Maintenance.        For bins 1 and 2, the assessments are identical to that for the internal-was done for all HEPs. This is                                                          Detailed analyses   events PRA, based on the relatively long system time window; the action documented in the BFN Seismic                                                           were completed      becomes relevant about 49 min after the earthquake would have HRA Notebook.                                                                           for one event that  occurred, and the total time window is over 2 hr. The assessment for bin had previously      3 accounts for a higher probability of failure due to reduced credit for been assessed       review that could affect the cognitive contribution. This treatment is using a screening   consistent with that for other HFEs as assessed for the seismic HRA.
approach. The       This F&O is assessed as CLOSED.
analyses were performed using the same methods as for other HFEs previously evaluated.
SPR C-SPR-D3 Met        19-9        The general approach to the      The screening approach from EPRI       Since in some cases the             Seismic PRA Human Reliability Analysis Maintenance  The review team     Table 9-2 of the Seismic PRA HRA Notebook (MDN0009992019000266,           Remains Met quantification is sound. The     3002008093 was used. These             screening method used               Notebook                                            concurs with the    Rev. 1) summarizes the post-initiator HFEs, including the relevant approach is clearly to adjust    screening values are expected to be    produced HEPs that were lower                                                           assessment that     probabilities from the assessment for internal events and for each of the the PSFs to reflect the impact   conservative relative to the detailed  than the HEPs used in the                                                               the changes to the  four seismic bins. All the events that were determined to be feasible of the seismic events.           values, otherwise the purpose of       detailed analysis, screening HEPs                                                       SPRA constitute     were assessed using a detailed approach. Screening probabilities are no However, there are some          using the screening values is not      were not used in the HRA                                                                Maintenance.        longer employed for these events. All HFEs determined to be infeasible deficiencies noted. One such     achieved. For the affected HFEs, the   analysis. Instead each HEP was                                                          Additional          or assessed for seismic HRA bin 4 are appropriately assigned failure deficiency is as follows:        screening values were                  subject to a detailed analysis                                                          detailed analyses   probabilities of 1.0.
underestimated by as little of 10%     where timing factors and stress                                                         have been
                                     - A table was provided that     up to as much as a factor of 20. This  levels were adjusted according                                                          performed for       This F&O is assessed as CLOSED.
compared the screening HEPs      illustrates a systematic problem       to the seismic bin being                                                                certain HFEs using that were used for the HFEs      with the application of the            analyzed for each HRA. This is                                                          the same methods that were eventually analyzed    screening approach, and it is clearly  discussed in the BFN Seismic                                                            as previously using detailed analysis. In a    possible that HFEs that retained       HRA Notebook. Additionally, the                                                         employed in the number of cases, the             their screening values could also be   boundaries of each EPRI seismic                                                         SPRA.
screening values were lower      a lot higher, and thus could be        bin were redefined such that than the detailed values.        significant contributors.              they were more closely tied to plant-specific factors, rather than the previous criteria which tied the bin boundaries to the percentage of components failed at each bin boundary.
Page 216 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                  December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                           Self-Assessment Reference              Maint (M) or Basis for Maint RU  SR                  Finding No. Description                     Prior Peer Review Assessment             Self-Assessment Closure Basis                                                                               Independent Review Team Disposition                                          Review Team Assessment                                                                                                                          Document(s)                            Upgrade?     (M) or Upgrade SR Assessment SPR C-SPR-D3 Met        19-10       The general approach to the     The adjusted cognitive HFEs are only     Each seismic HRA was modified       Seismic Probabilistic Risk Assessment  Maintenance  The review team     As documented in Section 6 and Appendix A of the Seismic PRA HRA             Remains Met quantification is sound. The    evaluated using CBDTM. EPRI TR-          such that the CBDTM/HCR-ORE         Human Reliability Analysis Notebook                 concurs with the    Notebook (MDN0009992019000266, Rev. 1) and in the calculations in approach is clearly to adjust   100259 states that CBDT was not          combination (max) method was                                                            assessment that     the HRA Calculator file (BFN SPRA 8-21-19.hra), both the cause-based the PSFs to reflect the impact  created as a method of its own, but      used in order to more effectively                                                       the changes to the  decision-tree method (CBDTM) and the HCR/ORE method were used in of the seismic events.          it was created as a supplement to        account for HEPs that were                                                              SPRA constitute     assessing post-initiator HFEs. The higher of the two results was However, there are some         HCR/ORE to take care of the cases        sensitive to timing issues. This is                                                     Maintenance.        employed for the cognitive contribution to the probability of each HFE.
deficiencies noted. One such    where extrapolation of the               documented in the seismic HRA                                                           Although the use deficiency is:                  HCR/ORE curves may not be                Notebook and reflected in the                                                           of the HCR/ORE      Timing impacts followed a convention with respect to assessing each HFE
                                    - Although the timing effects   appropriate. The concern with using      HEPs calculated in the HRA                                                              method was new      for different levels of seismic demand. In general, this entailed increasing were assessed to some extent,   CBDT alone is that it is completely      database, which calculates the                                                          to the SPRA for     the delay time (Tdelay) and, for actions taken outside the main control they were not incorporated      time-insensitive, and so by itself       cognitive portion of each HEP                                                           Browns Ferry, it    room, increasing the travel time contributing to the time to execute the into the quantification         does not account for the effects of      using both the HCR/ORE method                                                           was based on an     action (Texe).
because there was no time-      time, as required by the SR. This is     and the CBDTM method and                                                                earlier set of      A report was generated from the HRA Calculator file to permit reliability correlation used in emphasized in EPRI 3002008093,           uses the maximum of the two                                                             calculations        comparing the times applied in the HCR/ORE calculations for events the quantification.             which states 'Similar to internal        values for the final HEP.                                                               performed for the   across the spectrum of the seismic bins. The conventions for adjusting events HRA, both the CBDTM and                                                                                                   internal-events     timing were applied in a generally consistent manner. Exceptions were the HCR/ORE are to be considered                                                                                                 PRA and             found for HFEs that accounted for immediate, memorized actions, and for external events HRA. Both                                                                                                    documented in       for some HFEs that were ultimately not used in the SPRA.
methods address detection,                                                                                                       the BFN PRA -
diagnosis, and decision makingthe                                                                                               Human Reliability   This F&O is assessed as CLOSED.
HCR/ORE implicitly and the CBDTM                                                                                                 Analysis (NDN-explicitly. The CBDTM was                                                                                                        000-999-2007-developed to provide a lower limit                                                                                               0032, Rev. 0). That on the probability because the                                                                                                   PRA was subjected HCR/ORE calculates very low                                                                                                      to a peer review in probabilities for HFEs for which the                                                                                             2009.
time available is long relative to the time required.'
SPR C-SPR-D3 Met        19-11       The general approach to the     Section 6.3.1 of the EPRI                The EPRI bin definitions were       Seismic PRA Human Reliability Analysis Maintenance  The review team     Table 7-1 of the Seismic PRA HRA Notebook (MDN0009992019000266,              Remains Met quantification is sound. The    3002008093 methodology provides          redefined such that they more       Notebook                                            concurs with the    Rev. 1) summarizes the adjustments made to the assessments of post-approach is clearly to adjust   guidance on adjusting the PSFs for       closely align with specific plant                                                       assessment that     initiator HFEs to account for the impacts of earthquakes of increasing the PSFs to reflect the impact  CBDTM. Additional guidance is            related phenomena during a                                                              the changes to the  magnitude. Included in the table is a description of the level of plant of the seismic events.          provided on table 6-15. While this       seismic event. EPRI bin 1 is                                                            SPRA constitute     damage associated with each of the four HRA seismic bins.
However, there are some         guidance was followed to some            defined the same as in the peer                                                         Maintenance. The deficiencies noted. One such    extent, the adjustments appear to        review model where the upper                                                            HRA seismic bins    The impacts on timing and other PSFs are noted. These include deficiency is as follows.       be minimal considering the               bound is the safe-shutdown                                                              were re-aligned to
* Increases in the delay time (1 min for bin 1, up to 5 min for bin 3), with
                                    - While PSF adjustments were    guidance. The Pc assessments are         earthquake. In bin S1 there is no                                                       be more             no changes to Tcog or Texe, consistent with the recommendations of made in some cases, there       based on minimal operator input,         damage to the plant safety-                                                             consistent with     EPRI 3002008093, and increased travel time for actions taken outside were also many cases where      usually only at a high level as          related SSCs or non-safety SSCs                                                         the seismic hazard  the main control room; no changes were made. This is   opposed to specifically addressing       required for operation. There is                                                        for BFN, and some counter to the experience in    the context of the HRA bins. This        limited damage to non-safety,                                                           of the factors
* Increase in assessed stress levels except where the stress was already most other SHRAs that have      appears to be in large measure           non-seismic designed SSCs like                                                          affecting           assessed to be high; been seen for recent PRAs.      because the plant specific context of    residences and office buildings.                                                        quantification
* Increases the potential for attention to be diverted from important each bin (i.e., what specifically has    Bin S2 has been defined such                                                            were adjusted, but cues, due, for example, to workload; failed in each bin that would affect     that at the upper bound the                                                             no new methods the HFE context) has not been            turbine building has a 25%                                                              or significant
* Expectations regarding availability of crew for review of actions and clearly defined.                         chance of failure and a loss of                                                         change in           omissions in the main control room.
offsite power is likely                                                                 capability were (approximately an 89% chance                                                            introduced.         These are judged to be reasonable steps in adapting the internal-events of a LOOP occurring). In bin S2,                                                                            HRA to apply to seismic scenarios, and they are appropriate in the there is no expected damage to                                                                              context of the EPRI guidance.
the plant safety-related SSCs or                                                                            This F&O is assessed as CLOSED.
to rugged industrial type non-safety SSCs required for operation. Damage may be expected to non-safety-related SSCs not important to plant Page 217 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                    December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                           Self-Assessment Reference              Maint (M) or Basis for Maint RU  SR                  Finding No. Description                    Prior Peer Review Assessment           Self-Assessment Closure Basis                                                                                 Independent Review Team Disposition                                      Review Team Assessment                                                                                                                          Document(s)                            Upgrade?     (M) or Upgrade SR Assessment operations and to the switchyard (e.g., LOOP expected). Some falling of suspended ceiling panels. EPRI bin S3 is defined such that there is significant damage to category 2 structures. Widespread damage to non-safety related SSCs and/or some damage expected to safety related SSCs.
In bin S3 no credit for alarmed versus non-alarmed cues is taken due to the fact that there are multiple competing alarms and/or vibration trips and alarms occurring. The Turbine building failure probability at the upper bound of bin S3 is 96%. Also in bin S3, the Intake Pumping Station (the most fragile Category I structure) has a 25% chance of failing. In bin S4
(>1.5g) there is substantial damage to both safety-related and non-safety related SSCs.
SPR C-SPR-D4 Not Met    19-12       Section 7.1 of                 The impact of seismic effects on HFE   Timing factors were adjusted for       Seismic PRA Human Reliability Analysis Maintenance  The review team    As summarized in Table 7-1 of the Seismic PRA HRA Notebook               Met MDN0009992019000266            time parameters were not assessed      each EPRI seismic bin. This            Notebook                                            concurs with the   (MDN0009992019000266, Rev. 1), certain time parameters were states that HFE timing was     in accordance with the statements      generally consisted of adjusting                                                           assessment that    adjusted to account for impacts of seismic events. For example, the adjusted to account for        made in MDN0009992019000266.           the Tdelay of each HEP                                                                     the changes to the value from the internal-events PRA for Tdelay was increased by 1, 2 and seismic effects. Specifically: Those statements are in accordance     according to the given seismic                                                             SPRA constitute    5 min for HRA seismic bins 1, 2 and 3, respectively. For ex-control room
                                    - Main Control Room Actions -  with the requirements of the SR.       bin in order to account for such                                                           Maintenance. The   actions, travel time was increased as well.
Timing is adjusted depending   Guidance that would comply with        factors as increased briefing                                                              timing impacts on the seismic bin to account  the SR is provided in Section 6.3.2 of times and other distractions                                                               were incorporated  A new timing report was generated from the HRA Calculator file (BFN for additional time spent on   the EPRI 3002008093 methodology.       caused by the seismic event. In                                                            into the           SPRA 8-21-19.hra). Comparisons of the timing information for various crew briefs, competing         Additional guidance is provided on     addition, stress factors were                                                              assessments of     events confirmed that the intended timing impacts were appropriately actions, or other distractions table 6-15.                            adjusted to high for all seismic                                                           individual HFEs in accounted for.
in the control room.                                                  HEPs, and higher seismic bins                                                              a consistent       This F&O is assessed as CLOSED.
                                    - Local Actions - Timing is                                           assumed cues were not alarmed                                                              manner.
adjusted depending on the                                             to account for the distraction of seismic bin to account for                                            multiple alarms going off additional time spent on crew                                         simultaneously. For operator briefs and increased travel                                           actions external to the control time to the execution location                                        room the Texe was increased by due to seismic impacts on the                                         increasing the travel times to operator pathway.                                                     perform the operator action.
However, a review was                                                 The sigma values were also conducted on a timing output                                          adjusted higher for the HCR-ORE spreadsheet from the HRA                                              method in the higher seismic calculator and it was                                                 bins. HRAs occurring in the determined that there were                                            highest seismic EPRI bin were virtually no changes in time                                          assumed to fail.
parameters from the FPIE parameters or between the bins.
SPR C-SPR-B3 Met        25-1        Seismically induced failure    Table E-1 of the BFN SPRA              Based on the Master Fragility          PRA Evaluation BFN-0-19-062            Maintenance  The review team    As summarized in the response to Finding 25-1 in the PRA Evaluation      Remains Met modes and fragility values     Methodology Inputs and Model           Table Excel file, all fragilities                                                          concurs with the   Response (BFN-0-19-029, Rev. 0), a review was made to determine the Page 218 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                         December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                              Self-Assessment Reference   Maint (M) or Basis for Maint RU  SR                  Finding No. Description                      Prior Peer Review Assessment              Self-Assessment Closure Basis                                                                     Independent Review Team Disposition                                        Review Team Assessment                                                                                                                             Document(s)                 Upgrade?     (M) or Upgrade SR Assessment have been linked to all the      Notebook identifies that SPRA             were reviewed individually and                                                assessment that     extent of condition for this inconsistency in mapping to fragility groups.
components modeled in the        model fragility group SEIS_11-3           updated or re-mapped in some                                                  the changes to the  As a result, a significant number of SSCs were reassigned to new or IEPRA. Section 5.1               HPCI RCIC pumps (Fragility Group         cases where necessary. All                                                    SPRA constitute     different fragility groups. These reassignments are tabulated in the
                                    'Assumptions' of the Seismic     05-01) is modeled with Am=4.48g.         changes are documented in PRA                                                 Maintenance. The    description of the response.
Methods Notebook identifies      The Master Fragility Table (MFT)          Evaluation BFN-0-19-062. The                                                  changes to the that seismic SSC failures are    (i.e., Column O) identifies that the      'Fragility' table was updated in                                              mapping to          A spot check was performed for selected entries in this table relative to treated as a unique kind of      majority of SSCs included in fragility    the FRANX software to reflect                                                 fragility groups    the FRANX fragility mapping (BFN_Seismic_Rev1_U1CDF.franx) and the failure mode and are assumed     group SEIS_11-3 are calculated to         updated and newly added                                                       more                tabulation of fragilities in Appendix D of the Fragility Evaluation Report to be complete failures, in that have Am=4.48g. However, fragility         fragility groups. Since some                                                  appropriately       (TVAEBFN062-REPT-002, Rev. 1). No instances of apparently the SSC fails to perform its     group SEIS_11-3 also includes two         components were re-mapped to                                                  capture some SSC    inappropriate assignment to fragility groups were identified.
function. Seismic failure        (2) SSCs (i.e., CHILLED WATER             different fragility groups, the                                               capacities, but no  This F&O is assessed as CLOSED.
modes are identified by the      CIRCULATING CHW PUMP A and B)             'Fragility to Comp' table also                                                new methods fragility team for anchorage,    with a lower Am=2.52g                     required updating in the FRANX                                                were employed functional, and block wall                                                 software. These changes were                                                  and no significant failures. Additional failure     Approximately 10 other examples           populated to the CAFTA model                                                  changes to the modes for other types of SSCs    were identified similar to the above      database.                                                                     insights from the (e.g., structures) are evaluated comment.                                                                                                                PRA resulted.
(e.g., shear and bending) as needed.                          This combines fragility groups that are not actually seismically correlated, which results in collapsing of cutsets. As a result, it can be said that certain specific failures that should be specifically in the model are not.
This is an extent of condition issue.
The above is only an example.
SPR C-SPR-B6 Met        25-2        The component chatter            Contact chatter fragility calculation     Table 3 of the Enercon               PRA Evaluation BFN-0-19-062 Maintenance  The closure team    Based on the recommendation in F&O 25-2, fragility group SEIS_11-1R1       Remains Met analysis developed in a          SEIS_11-1R1 'Relay group 1 for            refinement document,                                                          assessed and        is sub-divided into groups SEIS_11-1R1-1 (EECW Pp B3 UV device) and detailed and systematic          group SEIS_11-1 (EECW Pp B3&D3            attachment 32 of Report                                                       agreed with the     SEIS_11-1R1-2 (EECW Pp D3 UV device. Using the FRANX software, in the process for the Browns Ferry     UV device)' includes relays               TVAEBFN062-REPT-002 was                                                       TVA                 'Fragility' table, group SEIS_11-1R1 was removed and groups SEIS_11-Units 1-3 Seismic PRA. The       associated with host cabinets on U2       reviewed to determine the                                                     determination       1R1-1 and SEIS_11-1R1-2 were created. In addition, the 'FireInitiatorHRA' identification and initial       RB Elevation 593 and 621. SPRA            relays of interest in the                                                     that the BFN SPRA   table was also updated with basic events tied to these relay groups and screening process for the        Fragility Group SEIS_11-1R1 is            refinement calculation. Based                                                 response to this    populated within the CAFTA database. The cutsets were reviewed to component chatter analysis is    shown in the results as risk              on the review, RHRSW pumps                                                    F&O is a PRA        verify that new fragility groups SEIS_11-1R1-1 and SEIS_11-1R1-2 documented in the Seismic        significant (e.g., U1 Seismic CDF FV =    B3 and D3 should be divided                                                   Maintenance         appeared appropriately.
PRA Chatter Analysis Report      2.3E-2). Relays in host cabinets that     further. Following the                                                        activity and not an (TVAEBFN062-REPT-003). The       are on different elevations should        recommendation, fragility group                                               Upgrade because     The BFN SPRA team discussed how the FRANX fragility table and the Chatter analysis report          not he same calculation group.            SEIS_11-1R1 is sub-divided into                                               the activity        'FireInitiatorHRA' table were used in conjunction to apply seismic determined which                                                           groups SEIS_11-1R1-1 and                                                      includes minor      induced fragility events to multiple PRA basic event failure modes (e.g.,
components could not be                                                    SEIS_11-1R1-2. Using the FRANX                                                modeling            pump fails to start, pump fails to run).
screened from the SEL for                                                  software, in the 'Fragility' table,                                           enhancements        This F&O is assessed as CLOSED.
chatter concerns. Fragilities                                              group SEIS_11-1R1 was removed                                                 (i.e., revisions to were then developed for the                                                and SEIS_11-1R1-1 and SEIS_11-                                                FRANX input file) unscreened relays in the                                                   1R1-2 was created. The                                                        using no new Fragility Analysis report                                                  'FireInitiatorHRA' table was also                                             methods or (Section 10 and Appendix E)                                                updated with basic events tied                                                calculations.
(TVAEBFN062-REPT-002).                                                     to these relay groups and populated within the CAFTA database.
SPR C-SPR-E5 Met        25-3        Per Section 9.1 of the BFN       The Monte Carlo evaluation was            PRA evaluation BFN-0-19-065          PRA Evaluation BFN-0-19-065 Maintenance  The closure team    Additional parametric uncertainty calculations were performed by TVA       Remains Met SPRA Quantification,             performed with 20,000 samples,            completed using the one-top                                                   assessed and        for processing various levels of cutsets (i.e., from 0 up to 500 cutsets)
Sensitivity and Uncertainty      using 1000 cutsets out of                 model shows that the                                                          agreed with the     through UNCERT and ACUBE. Figures 1 and 2 in BFN-0-19-065, BFN Notebook, 'The uncertainty       approximately 7000 total Seismic          uncertainty bands of CDF and                                                  TVA                 SPRA Uncertainty Analysis indicate that the propagated mean and analysis was performed with      CDF cutsets (and processed without        LERF are not significantly                                                    determination Page 219 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                         December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                            Self-Assessment Reference    Maint (M) or Basis for Maint RU  SR                  Finding No. Description                     Prior Peer Review Assessment              Self-Assessment Closure Basis                                                                      Independent Review Team Disposition                                     Review Team Assessment                                                                                                                           Document(s)                  Upgrade?     (M) or Upgrade SR Assessment UNCERT 4.0, using Monte         ACUBE). Typical industry parametric       affected by the number of                                                     that the BFN SPRA    uncertainty bands are relatively insensitive to the number of ACUBE Carlo sampling with 20,000      uncertainty analyses have used a          cutsets processed by ACUBE.                                                   response to this     cutsets processed.
samples and ACUBE processing    larger number of cutsets. Using                                                                                         F&O is a PRA of 1,000 cutsets.' TVA          more than 1000 of cutsets may                                                                                           Maintenance          The Unit 1 UNCERT sensitivity case results show that increasing the identified that, in reality,    provide different risk insights for the                                                                                 activity and not an  number of cutsets processed through ACUBE from 200 to 500 decreases ACUBE was not used (i.e., 0     parametric uncertainty results.                                                                                         Upgrade because      the UNCERT mean value for CDF and LERF by approximately 3% and 6%,
cutsets processed with ACUBE)   Using only 1000 cutsets with                                                                                            the activity         respectively. Therefore, further increasing the number of cutsets in the UNCERT runs because it   UNCERT (and 0 cutsets processed                                                                                         includes running     processed through ACUBE for the UNCERT evaluation likely would not caused the computer to crash. with ACUBE) may explain why the                                                                                         additional           result in a significant decrease in the UNCERT mean value.
BFN Unit 1, 2, and 3 propagated                                                                                         parametric           This F&O is assessed as CLOSED.
(This F&O originated from SR    uncertainty mean CDFs are                                                                                               uncertainty C-SPR-E5.)                      approximately a factor of two (2)                                                                                       calculations using higher than the respective point                                                                                        UNCERT and estimate CDFs. Processing more                                                                                          ACUBE using no than 1000 cutsets with ACUBE may                                                                                        new methods or result in calculating significantly                                                                                     types of lower propagated uncertainty mean                                                                                       calculations.
CDF values. Similarly, processing more than 1000 cutsets with ACUBE may also result in calculating significantly lower propagated uncertainty mean LERF values.
SPR C-SPR-E6 Met        25-4        The seismic impacts (e.g.,      To support the BFN ILRT LAR               The flag basic event                PRA Evaluation BFN-0-19-062. Maintenance  The closure team     Based on the recommendation in F&O 25-4, the sensitivity case for not   Remains Met fragilities, operator action,   submittal, selected sequences in the      FLG_LATE_SEIS_EARLY is used to                                                assessed and         converting Large-Late sequences to Large-Early sequences (i.e.,
containment bypass scenarios)   BFN FPIE Level 2 CETs were re-            reverse the base case and the                                                 agreed with the      Sensitivity Case 2 in Table 10-1 of the BFN SPRA Quantification were appropriately              defined from the No LERF end            sensitivity case 2 of BFN SPRA                                                TVA                  Notebook) is incorporated into the updated Base Case SPRA model.
incorporated into the Level 2   state to the Containment Intact         Rev0 model. In this Rev1 model,                                               determination        Typical industry SPRA models have assumed no significant changes to PRA structure taken from the    end state for the BFN SPRA. The           in the flag file                                                              that the BFN SPRA    the definition of "early" for the evaluation of LERF.
FPIE Level 2 PRA model. The     sequences in question include the         BFN123_Flag_r8_SEIS.FLG of                                                  response to this SPRA model assumed that any     following issues:                         the base case, the flag                                                       F&O is a PRA         New Sensitivity Case 6 in Table 10-1 of the BFN SPRA Quantification Large Late Releases from the
* For CET 1 (for Classes 1, 3A, 3B,       FLG_LATE_SEIS_EARLY is set to                                                 Maintenance          Notebook shows that assuming that Large-Late sequences lead to the FPIE PRA modeling insights      and 3C), the Containment Intact         FALSE such that the large late                                                activity and not an  Large-Early end state would increase the LERF from 3.29E-6/yr to 4.16E-would be treated as Large       sequence does not appear to               release (gate U1S_LLR_TOP1 for                                                Upgrade because      6/yr (i.e., +27%). For the new sensitivity case, the Level 2 model also Early Releases for seismic      consider the status of containment        earthquake >0.5g, %G4 and                                                     the activity         incorporates the conservative assumption that the Containment Intact ground motion >0.5g.            heat removal in the CET. Without          above) will not be quantified as                                              includes minor       sequences lead to the Large-Early end state.
considering containment heat              large early release. For                                                      modeling             This F&O is assessed as CLOSED.
However, the Level 2 CET        removal, an intact containment            sensitivity case, the flag                                                    enhancements structures used as input to the cannot be guaranteed.                     FLG_LATE_SEIS_EARLY is set to                                                 (i.e., setting basic Level 2 SPRA model identified
* For CET 2 (for Class 3D), the           1.0 by commenting the setting                                                 events to FALSE in selected sequences as           containment is already assumed            out to quantify the large late                                                a flag file and Containment Intact with no    failed in the Level 1 because of the      release as the large early                                                    inserting new apparent basis. The             LOCA with loss of vapor                   release.                                                                      Level 2 basic Containment Intact            suppression. By default, the                                                                                            events and gates) sequences were excluded         containment is modeled as failed in       To model the containment                                                      using no new from the SPRA Level 2 model. the Level 2.                              intact sequences that potentially                                             methods or Based on this, the base SPRA
* For CET 2 (for Class 4), the            could be LERF cases, the fault                                                calculations.
model may under-predict the     containment has already failed in         tree is updated to include all the LERF because some of the        the Level 1 because of the                containment intact groups:
scenarios included in the       unmitigated ATWS event (e.g., high        Sequences 1and 10 from CET1, Containment Intact            suppression pool temperature and          Sequence 1 from CET2 Class 3D sequences do not have an        hydrodynamic loads). By default,          and Sequence 1 from CET2 Class intact containment (e.g., Level the containment is assumed failed         4 according to BFN-0-16-040 for 1 ATWS scenarios with           in the Level 2.                           Units 1, 2 and 3 integrated containment modeled as                                                    leakage rate test risk evaluation.
already failed in the Level 2). In addition, the Level 1 Class II (i.e.,  The flag basic event Loss of Containment Heat Removal)         FLG_CONT_INT is set to 1.0 to core damage sequences are not             quantify all the containment transferred and evaluated in the          intact sequences as LERF cases.
Page 220 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                      December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                        Self-Assessment Reference   Maint (M) or Basis for Maint RU  SR                  Finding No. Description                   Prior Peer Review Assessment            Self-Assessment Closure Basis                                                                    Independent Review Team Disposition                                         Review Team Assessment                                                                                                                       Document(s)                 Upgrade?     (M) or Upgrade SR Assessment BFN Level 2 SPRA.                       This assumption is very conservative, but there is no The BFN Full Power Internal Events      basis to determine the split (FPIE) Level 2 PRA, Level 1 Class II    fraction of the LERF sequences (i.e., Loss of Containment Heat         out of the intact sequences, in Removal) core damage sequences          the sensitive case, do not contribute to LERF               FLG_CONT_INT can be set to sequences because the BFN               different values like 0.25, 0.5, Emergency Action Levels (EALs)          0.75 in order to account for the support that a General Emergency        uncertainty of the split fraction.
will be declared sufficiently early     New Basic events and gates are such that any offsite releases from     added into the fault tree. It Class II events will be non-LERF.       should be noted that the unit 1 Therefore, a Level 2 Class II           case is discussed specifically, but Containment Event Tree (CET) was        similar logic was also built for not developed for the BFN FPIE          Units 2 and 3. The gates Level 2 PRA.                            U1_CONT_INTACT_G10, A Level 2 Class II CET would have a     U1_CONT_INTACT_G11 and similar accident sequence structure     U1_CONT_INTACT_G20 are as the Class IIID and Class IV CETs. added to the larger late release to the large early release gate U1S_LLR_TOP for CET1, CDT2_3D and CET2_4 respectively.
SPR C-SPR-E3 Met        25-5        QU-D6: Met as significant     Fragility group SEIS_12-1a EECW        The mapping in the                  PRA Evaluation BFN-0-19-062 Maintenance  The closure team    The mapping in the "FireInitiatorHRA" table was updated with the            Remains Met contributors were identified. Pumps (Fragility Group 06-03-01)       "FireInitiatorHRA" table was                                                 assessed and        'dummy' basic events (i.e., basic events with the _dum on the end of However, it appears that two  with Am of 1.34g has a U1 Seismic       updated with the 'dummy' basic                                               agreed with the     the basic event name). The 'dummy' basic event gets created from the significant contributors were CDF FV=0.0 while fragility group        events. The 'dummy' basic event                                              TVA                 "FireInitiatorHRA" table and is injected into the fault tree along with the missed.                       SEIS_12-1P-2 EECW Pumps based          gets created from the                                                        determination       internal event basic event. A test run was conducted to show the Seismic on pipe frag calc is modeled with a    "FireInitiatorHRA" table and is                                              that the BFN SPRA   CDF AND Seismic LERF. Both SEIS_12a and SEIS_12b appeared in the higher Am=1.86g and is a top risk       injected into the fault tree along                                           response to this    cutsets as well as SEIS_1P-1 and SEIS_1P-2. The purpose of the test run contributor (e.g., U1 Seismic CDF FV    with the internal event basic                                                F&O is a PRA        was to show these fragility groups indeed exist in the cutsets and the
                                                                  = 6.0E-2 in Table 8-4 of the BFN        event. A test run was conducted                                              Maintenance         SYSIMP risk importance results. The results showed that these groups SPRA Quantification Notebook            to show the Seismic CDF AND                                                  activity and not an are represented correctly.
(MDN0009992019000268)). Both            Seismic LERF. Both SEIS_12a and                                              Upgrade because SEIS_12-1a and SEIS_12-1P-2 should      SEIS_12b appeared in the                                                     the activity        This F&O is assessed as CLOSED.
have the same consequences in the       cutsets as well as SEIS_1P-1 and                                             includes minor SPRA model (i.e., loss of all EECS). SEIS_1P-2. The purpose of the                                                modeling Therefore, SEIS_12-1a should have a     test run was to show these                                                   enhancements higher FV.                              fragility groups indeed exist in                                             (i.e., revisions to the cutsets and the SYSIMP risk                                              FRANX input file)
Similarly, fragility group SEIS_12-1b   importance results. The results                                              using no new RHR[SW] Pumps (Fragility Group         showed that these groups are                                                 methods or 06-03) with Am of 1.45g has a U1       represented correctly. See PRA                                               calculations.
Seismic CDF FV=0.0 while fragility      Evaluation BFN-0-19-062.
group SEIS_12-1P-1 RHRSW Pumps based on pipe frag (Pipe calc) is modeled with a higher Am=1.82g and is a top risk contributor (e.g., U1 Seismic CDF FV = 6.6E-2 in Table 8-4 of the BFN SPRA Quantification Notebook (MDN0009992019000268)). Both SEIS_12-1b and SEIS_12-1P-1 should have the same consequences in the SPRA model (i.e., loss of all RHRSW).
Therefore, SEIS_12-1b should have a higher FV.
Page 221 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                               December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                        Self-Assessment Reference            Maint (M) or Basis for Maint RU  SR                  Finding No. Description                   Prior Peer Review Assessment            Self-Assessment Closure Basis                                                                             Independent Review Team Disposition                                      Review Team Assessment                                                                                                                       Document(s)                          Upgrade?     (M) or Upgrade SR Assessment In the Groups to Components window in FRANX file BFN_Seismic_Rev0_U1CDF.franx, it appears that fragility groups SEIS_12-1P-1 and SEIS_12-1P-2 (i.e.,
for the pipe fragilities) are not mapped to any SPRA basic events.
However, fragility groups SEIS_12-1a and SEIS_12-1b (i.e., for the pump fragilities) are mapped to SPRA basic events (e.g., RHRSW pump fails to start or run).
Based on the above, it would be expected that fragility groups SEIS_12-1P-1 and SEIS_12-1P-2 (i.e.,
for the pipe fragilities) do not appear in the cutsets, while fragility groups SEIS_12-1a and SEIS_12-1b (i.e., for the pump fragilities) do appear in the cutsets.
However, the opposite appears to be true. For example, in file U1_CDF-consolidated cutset file.cut, fragility groups SEIS_12-1P-1 and SEIS_12-1P-2 (i.e., for the pipe fragilities) do appear in the cutsets, while fragility groups SEIS_12-1a and SEIS_12- 1b (i.e., for the pump fragilities) do not appear in the cutsets.
SPR C-SPR-F1 Met        25-6        The seismic plant-response    Discuss the process for accounting      An update to section 2.0 of the     TVABFN062-REPT-003, Seismic PRA      Maintenance  The closure team    Section 2.0 of the seismic PRA Chatter Analysis Report (TVABFN062-       Remains Met analysis and quantification   for all of the unscreened relays        seismic PRA Chatter Analysis        Chatter Analysis Report, Section 2.0              assessed and        REPT-003) was updated to include the clarification for the response to was documented in 'SPR'       identified in the Seismic PRA Chatter   Report (TVABFN062-REPT-003).                                                          agreed with the     F&O 25-6. This was a documentation update to identify the process for Notebooks (Seismic Methods,   Analysis Report (TVAEBFN062-REPT-       The changes include the                                                               TVA                 accounting for all the unscreened relays compared to the contact chatter SQU) in a manner that         003) compared to the contact            clarification for the response to                                                     determination       groups ultimately included in the SPRA model.
facilitates PRA applications, chatter groups ultimately included      F&O 25-6, correction for the                                                          that the BFN SPRA upgrades, and peer review. in the SPRA model.                      Errata and additional editorial                                                       response to this    This F&O is assessed as CLOSED.
changes.                                                                              F&O is a PRA The BFN SPRA team identified the                                                                                              Maintenance following:                                                                                                                    activity and not an Upgrade because The Seismic PRA Chatter Analysis                                                                                              the activity Report (TVAEBFN062-REPT-003)                                                                                                  involves Appendix B Components of Chatter                                                                                             documentation Concerns Requiring Functional                                                                                                 changes only using Fragility Analysis has 615 line items                                                                                        no new methods but does not actually have 615                                                                                                or calculations.
unscreened relays. The chatter analysis was performed based on component mis-operation and appendix B of the report listed the affected component (columns 2 and 3 of the table) and then the relay (column 4 and 5) that causes Page 222 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                     December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                         Self-Assessment Reference   Maint (M) or Basis for Maint RU  SR                  Finding No. Description                      Prior Peer Review Assessment            Self-Assessment Closure Basis                                                                  Independent Review Team Disposition                                       Review Team Assessment                                                                                                                        Document(s)                 Upgrade?     (M) or Upgrade SR Assessment the mis-operation. This results in duplicate relay entries due the same components on each of the units and multiple functions of individual relays. For example, the first line of table identifies the component as BFN-1-FCV-071-0002 and the controlling relay as 13A-K33. This relay appears 6 times in the chatter report. This relay affects an additional valve (BFN-1-FCV-071-0003) and occurs on all three units.
This was recognized, and the duplicate items were eliminated during the fragility development.
Also, during this review several relays chatter events were noted to be acceptable due to operating conditions such as diesel generator lockouts when the diesel is in a test configuration. These were left in the table but annotated as Chatter Acceptable.
SPR C-SPR-B4 Met        26-2        Fragility group definitions      SPRA model fragility group SEIS_12-     The Unit 2 and Unit 3 pumps are   PRA Evaluation BFN-0-19-062 Maintenance  The review team     The response to Finding 26-2 in the PRA Evaluation Response (BFN-0    Remains Met were provided by the fragility   2 (RHR CS pump (Fragility Group        mapped differently from the                                                concurs with the    029, Rev. 0), provides a thorough discussion of the steps taken to change team. The correlations           06-01 AND 06-02) is                    Unit 1 RCIC pumps in the BFN                                               assessment that     the mapping of the RCIC pumps to the proper fragility group. Review of between different                inappropriately mapped to fail U2       SPRA Rev 0 model. This mapping                                             the changes to the  the FRANX fragility mapping (BFN_Seismic_Rev1_U1CDF.franx) components were established      and U3 RCIC pumps in addition to        was corrected by mapping the                                               SPRA constitute     confirmed that the re-assignments have been properly implemented.
based on the failure mode,       U1, U2, and U3 RHR and Core Spray       unit 2 and unit 3 RCIC pumps to                                            Maintenance. The similarity, orientation,         pumps. The U2 and U3 RCIC SSCs          SEIS_11-3 within the FRANX                                                 changes to the      This F&O is assessed as CLOSED.
location, fragility, and other   mapped to Model fragility group         software in the                                                            assignment to seismic concerns. Section 6.7    SEIS_12-2 should be changed to          'Fragility_To_Comp' table. The                                             fragility groups of TVAEBFN062-REPT-002           Model fragility group SEIS_11-3.        new mapping was then                                                       has a small impact (Browns Ferry Nuclear Power      Model Fragility group SEIS_11-3         populated to the CAFTA                                                     on the PRA results, Plant Components and             contains the corresponding U1 RCIC      database. A test run was                                                   generally leading Structures Fragility Evaluation) pump.                                   conducted to show Seismic CDF                                              to a very small summarizes technical basis for                                           and Seismic LERF for one of the                                            decrease in CDF establishing correlations and                                            updated units. See PRA                                                     and LERF. No new grouping. The Fragility group                                            Evaluation BFN-0-19-062.                                                   methods were definitions provided by the                                                                                                                         introduced, and fragility team are summarized                                                                                                                       no significant new in Column "Fragility                                                                                                                                insights resulted Calculation Group" of                                                                                                                               from this change.
Appendix D "Master Fragility File" of the TVAEBFN062-REPT-002 BFN Fragility Report.
The fragility groups actually included in the SPRA model are shown in Column "Model Fragility Group" of Appendix D.
SPR C-SPR-B8 Not Met    26-3        The only new logic related to    1. Clarify if the FLEX Nitrogen         In the 'Components' table, a      PRA Evaluation BFN-0-29-062 Maintenance  The closure team    In the 'Components' table, a UNID was created to represent the four (4)   Met system modeling added to the     bottles (basic event                    pseudo UNID was created to                                                 assessed and        nitrogen carts functioning as an alternate supply to the SRVs with the BFN SPRA was regarding           TKURP0TNK_N2FLEX) and                   represent the 4 nitrogen carts                                             agreed with the     assumption of complete correlation. A new UNID name was created for crediting the FLEX nitrogen      associated operator action              functioning as an alternate                                                TVA                 the HCVS N2 bottles and a new component failure basic event bottles for hardened wetwell     HFA_OPS_FLEXN2ALIGN (Seismic -          supply to the SRVs with the                                                determination       TKURP0TNK_N2HCVS and a newly created operator action Page 223 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                               December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                         Self-Assessment Reference             Maint (M) or Basis for Maint RU  SR                  Finding No. Description                     Prior Peer Review Assessment           Self-Assessment Closure Basis                                                                              Independent Review Team Disposition                                        Review Team Assessment                                                                                                                        Document(s)                           Upgrade?     (M) or Upgrade SR Assessment venting. An operator action to  Operator fails to align FLEX N2        assumption of complete                                                                 that the BFN SPRA   HFA_OPS_HWWVN2ALIGN are associated with primary containment connect and operate the         backup to DCA) discussed in Section    correlation. A new pseudo UNID                                                         response to this    vent. HEPs for both of the operator actions were developed equipment was added to the      6.5.2 support backup pneumatic         name was created for the HCVS                                                          F&O is a PRA        appropriately.
model. No other new             supply for both the Hardened           N2 bottles and a new                                                                   Maintenance equipment was added to the      Wetwell Vent valves and the SRVs       component failure basic event                                                          activity and not an Fragility grouping was evaluated separately for both sets of nitrogen model. This is discussed in     for Reactor Depressurization. The      TKURP0TNK_N2HCVS and a                                                                 Upgrade because     bottles, although both were determined to be rugged and assigned to Section 6.5.2 of the            fault tree logic (e.g., gate           newly created operator action                                                          the activity        fragility group ID SEIS_0-20 Plant Ruggedness Fragility. Since the HCVS Methodology, Inputs, and        U1_VENT_N2FLEX) models that the        HFA_OPS_HWWVN2ALIGN are                                                                includes minor      N2 bottles are stored in the diesel building, the structure fragility Model Notebook.                 same FLEX Nitrogen bottles and         associated with primary                                                                modeling            corresponding to SEIS_BLD-DGB was modeled as a limiting fragility for The FLEX Nitrogen bottles       operator action supports both the      containment vent. Fragility                                                            enhancements        basic event TKURP0TNK_N2HCVS. Similarly, SEIS_BLD-RB was the (basic event                    Hardened Wetwell Vent valves and       grouping was evaluated                                                                 (i.e., add basic    structure fragility modeled as a limiting fragility for the SRV N2 bottles TKURP0TNK_N2FLEX) and           the SRVs for Reactor                   separately for both sets on                                                            events and gates    since they reside in the reactor building. N2 bottle fault tree logic was associated operator action      Depressurization.                      nitrogen bottles, although both                                                        to fault tree logic updated to ensure that the basic events for FLEX N2 bottle alignment for HFA_OPS_FLEXN2ALIGN             2.Based on information provided by     were determined to be rugged                                                           and revisions to    RPV depressurization correctly feeds into inputs for basic events (Seismic - Operator fails to    BFN, the (permanent) Nitrogen          and assigned to SEIS_0-20. Since                                                       FRANX input file)   TKURP0TNK_N2FLEX, and logic for backing up hardened wet well vent align FLEX N2 backup to DCA)    bottles providing backup supply to     the HCVS N2 bottles are stored                                                         using no new        correctly feeds into inputs from basic event TKURP0TNK_N2HCVS. The discussed in Section 6.5.2      the Hardened Wetwell Vent valves       in the diesel building, the                                                            methods or          associated mapping was updated in the Fragility_To_Comp table and support backup pneumatic        appear to be different than the        structure fragility corresponds                                                        calculations.       the human actions in the 'FireInitiatorHRA' table. All updates were supply for both the Hardened    (portable, i.e., on carts) Nitrogen    to SEIS_BLD-DGB for basic event                                                                            populated to the CAFTA database.
Wetwell Vent valves and the     bottles providing backup supply to     TKURP0TNK_N2HCVS. SEIS_BLD-                                                                                This F&O is assessed as CLOSED.
SRVs for Reactor                the SRVs for Reactor                   RB was the structure fragility of Depressurization. The fault     Depressurization.                      the SRV N2 bottles since they tree logic (e.g., gate          3.The fragility analysis should        reside in the reactor building.
U1_VENT_N2FLEX) models          distinguish between the Nitrogen       N2 bottle fault tree logic was that the same FLEX Nitrogen     bottles providing backup supply to     updated to ensure that the basic bottles and operator action     the Hardened Wetwell Vent valves       events for flex n2 bottle supports both the Hardened      and the Nitrogen bottles providing     alignment for RPV Wetwell Vent valves and the     backup supply to the SRVs for          depressurization correctly feed SRVs for Reactor                Reactor Depressurization. The same     in inputs for basic events Depressurization. However,      fragility group was applied to both    TKURP0TNK_N2FLEX, and logic based on information provided   N2 bottle sets - (1) for portable N2   to backing up hardened wet well by BFN, the Nitrogen bottles    bottles (carts) used for backup        vent correctly feed in inputs providing backup supply to the  supply to ADS SRVs, and (2) for        from basic event Hardened Wetwell Vent valves    permanently installed N2 bottles for   TKURP0TNK_N2HCVS. The are different than the Nitrogen backup supply to HCVS AOVs.            associated mapping was bottles providing backup        4.The human reliability analysis       updated in the supply to the SRVs for Reactor  (HRA) should distinguish between       Fragility_To_Comp table and Depressurization. In addition,  the operator action to align the       the human actions in the the operator actions should be  (permanent) Nitrogen bottles           'FireInitiatorHRA' table. All different.                      providing backup supply to the         updates were populated to the Crediting the FLEX Nitrogen     Hardened Wetwell Vent valves and       CAFTA database.
bottles was the only change     the operator action to align incorporated for the SPRA       (portable) Nitrogen bottles logic and the fault tree model  providing backup supply to the SRVs was not updated appropriately   for Reactor Depressurization. The (i.e., assumed the same         current HRA for operator action fragility and the same operator HFA_OPS_FLEXN2ALIGN references action).                        procedure 1-EOI Appendix-20H (Alternate N2 Supply to SRVs) 5.Provide the MAAP run (identified as from 'EVB') that is used as the basis for the time available (i.e.,
Tsw) of 6.27 hours for operator action HFA_OPS_FLEXN2ALIGN.
SPR C-SPR-A3 Not Met    26-4        There is no discussion of the   SPRA notebooks provide no              A table was assembled listing a     Methodology, Inputs and Model         Maintenance  The review team     Table 6-8 of the (MDN-000-999-2019-000267, Rev. 1) provides a              Met available seismic risk          evidence that this assessment of       review of external events for       Notebook; PRA Evaluation BFN-0-19-062              concurs with the    summary of responses to earthquakes for several plants and identifies applicability to seismic events.                                                       assessment that     the relevance of those responses to the BFN SPRA.
This list is derived from                                                              the changes to the Page 224 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                         December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                             Self-Assessment Reference     Maint (M) or Basis for Maint RU  SR                  Finding No. Description                       Prior Peer Review Assessment            Self-Assessment Closure Basis                                                                      Independent Review Team Disposition                                          Review Team Assessment                                                                                                                            Document(s)                   Upgrade?     (M) or Upgrade SR Assessment evaluations for other nuclear     other nuclear plants or industry        reference 3002000709, Seismic                                                   SPRA constitute    A comparison of the seismically induced initiating events from three plants or industry experience. experience was performed                Probabilistic Risk Assessment                                                   Maintenance. This  other BWRs to those relevant to BFN was made. This comparison is Implementation Guide, EPRI,                                                     comparison was     summarized in a table in the PRA Evaluation Response (BFN-0-19-029, Palo, Alto, CA.                                                                 made to            Rev. 0), which describes in detail the manner in which each of the document the       findings from the peer review was addressed.
check made of the adequacy of the    This F&O is assessed as CLOSED.
set of initiating events considered in the SPRA for BFN. No changes were made to the SPRA based on this comparison.
SPR C-SPR-A4 Met        26-5        Table 4-3 in the Seismic          For Break Outside of Containment        The seismic impact of these pipe     PRA Evaluation BFN-0-19-062   Maintenance  The review team    The rationale for screening of specific lines whose failure might            Remains Met Methodology, Inputs, and          (BOC), the BFN SPRA identifies          lines on the system functions                                                   concurs with the   constitute a break outside containment is provided in Table 6-8 of the Model Notebook includes a         Feedwater and Main Steam piping         has been modeled in the SPRA                                                    assessment that    Seismic Methodology, Inputs and Model Notebook (MDN-000-999-2019-listing of the seismic initiating failures in fragility group SEIS_23-    model; Given the rugged piping                                                  the changes to the 000267, Rev. 1). In the summary of actions taken to address Finding 26-5 events that were included in      2_BOC. However, the following           and relative high fragilities of                                                SPRA constitute    in the PRA Evaluation Response (BFN-0-19-029, Rev. 0), further details the model. It includes a          additional BOC initiators have not      the valves, given an earthquake,                                                Maintenance.       supporting these rationales are provided. These summaries include discussion of where the           been identified or assessed.            the risk contribution from these                                                Additional         reference to specific penetrations, facilitating review of each of the lines initiating events were inserted   No fragility value was provided for     lines as an initiating event is                                                 documentation of   cited above based on the information in Table 6-8.
into the model.                   BOC for the following system in         negligible and can be capped by                                                 the potential for table 6-8.(ISLOCA) or in the SPRA       the reactor building failure. The                                               breaks outside     No issues were identified with respect to the treatment of these lines.
model, whereas this info was            failure of the reactor building is                                              containment was    The amplified rationales provided in BFN-0-19-029 provide additional provided for FW /MS (SEIS_23-           modeled as leading directly to                                                  provided, but not  perspective on the ruggedness of the respective lines.
2_BOC                                   core damage. Due to the rugged                                                  changes were       This F&O is assessed as CLOSED.
piping line and high fragilities of                                             made to the PRA the associated valves, the risk                                                 models to address
: 1. RCIC - for BOC, not modeled and      contribution from these piping                                                  this Finding.
no basis provided for screening         lines is very small and negligible.
: 2. HPCI - for BOC, not modeled and      No further modeling change is no basis provided for screening         necessary.
: 3. SDV - for BOC, not modeled and no basis provided for screening
: 4. MSL Drains - for BOC, not modeled and no basis provided for screening
: 5. RWCU - for BOC, not modeled and no basis provided for screening.
(Note: ISLOCA is different than BOC (in BWRs).
These issues made it difficult for PRT SPR reviewers to assess BOC.)
SPR C-SPR-F1 Met        26-6        The seismic plant-response        1. Ref. BFN SPRA Methods                Methodology, model and inputs        Methodology, Inputs and Model Maintenance  The review team    Review of available documentation confirmed that appropriate changes         Remains Met analysis and quantification       Notebook do not explicitly state        notebook was updated in              Notebook                                   concurs with the   have been made. At the end of Section 6.2.5 of the Seismic was documented in 'SPR'           that 'no other secondary hazards        section 6.2.5 to include the                                                    assessment that    Methodology, Inputs and Model Notebook (MDN-000-999-2019-000267, Notebooks (Seismic Methods,       (from SHA-I2) were explicitly           statement "No other secondary                                                   the changes to the Rev. 1), a summary is provided of the secondary hazards considered and SQU) in a manner that             retained in the SPRA.'                  hazards were explicitly retained                                                SPRA constitute    screened for BFN. This summary includes the statement that no other facilitates PRA applications,     2. Ref. BFN SPRA Methods                in the SPRA." The description for                                               Maintenance. The   secondary hazards were explicitly addressed in the SPRA.
upgrades, and peer review.        Notebook, table E-1, page 364,          SEIS_23-2 was updated to                                                        changes affect fragility group SEIS_23-2 _BOC          include main steam in Table E-1                                                 only               The entry for fragility group SEIS_23-2_BOC in Table E-1 of the same (Feedwater piping (BOC)) also apply     of the methodology, inputs and                                                  documentation of   document has been revised to note that it included feedwater and main to main steam piping.                   model notebook.                                                                 aspects of the     steam piping.
SPRA.              This F&O is assessed as CLOSED.
Page 225 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                         Self-Assessment Reference             Maint (M) or Basis for Maint RU  SR                  Finding No. Description                    Prior Peer Review Assessment            Self-Assessment Closure Basis                                                                              Independent Review Team Disposition                                      Review Team Assessment                                                                                                                        Document(s)                           Upgrade?     (M) or Upgrade SR Assessment SPR C-SPR-F2 Met        26-8        The process to perform the     1.In section 4.4 of                     Operator action HFA_OP-             Methodology, Inputs and Model         Maintenance  The review team     Review of available documentation confirmed that appropriate changes     Remains Met BFN SPRA seismic plant-        MDN0009992019000266, HFA_OP-            LOCKOUT_4kVSDBD_S in section        Notebook; Quantification, Sensitivity              concurs with the    have been made:
response analysis and          LOCKOUT_4kVSDBD_S is incorrect.         4.4 of the HRA Notebook was         and Uncertainty Notebook; Seismic PRA              assessment that quantification was             It should be                            corrected to read                   Human Reliability Analysis Notebook                the changes to the
* HFE HFA_OP-LOCKOUT_4kVSDBD_S is now HFE documented in the SEL,         HFA_OPS_4KVSDBDRESET.                   HFA_OPS_4kVSDBDRESET.                                                                  SPRA constitute     HFA_OPS_4KVSDBDRESET in the Seismic PRA HRA Notebook (MDN-000-Seismic Methods, Model         2. In the SEISMIC METHODOLOGY,          In the Seismic Methodology,                                                            Maintenance. The    999-2019-000266, Rev. 1) and in the HRA Calculator file (BFN SPRA 8                                     Development _ Q1,Q2,Q3         INPUTS, AND MODEL                       Inputs and Model Notebook, in                                                          changes correct     19.hra). It has also been incorporated into the PRA model via the use of (027), SHRA, SQU, Truncation   6.3.1.1 SIET Top Event S-DCD (SEIS-     section 6.3.1.1, the line that                                                         minor               FRANX.
(028) and Convergence (022)    08)                                     reads, "The SIET sequence SEIS-                                                        documentation
* In Section 6.3.1.1 of the Seismic Methodology, Inputs and Model Notebooks, and associated      The SIET sequence SEIS-08 covers        08 covers seismic failures that                                                        errors in various   Notebook (MDN-000-999-2019-000267, Rev. 1), the figure number has PRA Model documentation.       seismic failures that lead directly to  lead directly to core damage.                                                          parts of the SPRA. been corrected to Figure 6-5.
core damage. Figure 6- shows.           Figure 6- shows.." is now
: 3. SEISMIC METHODOLOGY, INPUTS,         updated to read Figure 6-5.
* The events in Figure 6-3 and the entries in Table 6-9 of the Seismic AND MODEL                               The mismatch in Table 6-9 and                                                                              Methodology, Inputs and Model Notebook have been made consistent.
The SIET top events in page 74 and      Figure 6-3 was corrected in the
* In Section 6.5.16 of the Seismic Methodology, Inputs and Model 75 doesnt match table 6.9              Methodology, Inputs and Model Notebook, unitized has been corrected to utilized.
: 4. SEISMIC METHODOLOGY, INPUTS,         Notebook.
AND MODEL                               Section 6.5.16 was corrected to
* In Table 8-2 of the Seismic PRA Quantification, Sensitivity and shutdown boards have been               "should be utilized." in the                                                                               Uncertainty Notebook (MDN-000-999-2019-000268, Rev. 1), the unitized.it should be utilized.         Methodology, Inputs and Model                                                                              contributions by percent (the last column in the table) has been
: 5. Quantification, Sensitivity and      Notebook.                                                                                                  corrected.
Uncertainty Notebook                    Table 8.2 of the Quantification, Table 8-2 last column is wrong.         Sensitivity and Uncertainty
* In Section 5.3.2 of the Seismic Methodology, Inputs and Model
: 6. SEISMIC METHODOLOGY, INPUTS,         Notebook was corrected in the                                                                              Notebook, switch gear has been corrected to switchgear.
AND MODEL                               last column.                                                                                               This F&O is assessed as CLOSED.
occurrence of the failures from the     Section 5.3.2 of the switch gear to. It should be            methodology, inputs and model switchgear                              Notebook was updated from "occurrence of the failures from the switch gear" to read switchgear.
SPR C-SPR-B2 Met        26-9        Under gate                     Failure of equipment due to seismic     Three pseudo components             Methodology, Inputs and Model         Maintenance  The closure team    Three components (i.e., one for each unit) representing the backup       Remains Met
                                    'U1S_REMOTE_SHUTDN', the       event especially panel inside the       representing the backup control     Notebook; PRA Evaluation BFN-0-19-062              assessed and        control panel were created in the components table within the FRANX failure of the remote          remote shutdown room should be          panel were created in the                                                              agreed with the     software. The three components were modeled with SPRA fragility ID shutdown is model as a failure modeled. Without adding this to the     components table within the                                                            TVA                 SEIS_5-3, RCIC Local Control, BFN-1/2/3-LPNL-925- 0031 and 0032. The of human action                model it is not possible to             FRANX software. The                                                                    determination       components were then tied to the human action event
                                    'U1_CREVACSTDNFAILS_S'.        determine the importance of the         components were then tied to                                                           that the BFN SPRA   Ux_CREVACSTDNFAILS_S in the Fragility_to_Components table so that There is no failure of the     equipment failure or its contribution   the human action event                                                                 response to this    the failure of the panel fails the operator action. The new mapping was remote shutdown associated     to risk:                                Ux_CREVACSTDNFAILS_S in the                                                            F&O is a PRA        populated in the CAFTA database.
to equipment failure.                                                  Fragility_to_Components                                                              Maintenance table so that the failure of the                                                       activity and not an This F&O is assessed as CLOSED.
panel fails the operator action.                                                       Upgrade because The new mapping was                                                                    the activity populated in the CAFTA                                                                 includes minor database.                                                                              modeling enhancements (i.e., revisions to FRANX input file) using no new methods or calculations.
SPR C-SPR-F2 Met        26-10       The process to perform the     1.CST Am fragility value is             Section 6.5.13 of the               Methodology, Inputs and Model         Maintenance  The review team     The median capacity for the CST has been corrected in Table E-1 of the   Remains Met BFN SPRA seismic plant-        incorrectly reported as 0.1 g in        Methodology, Inputs and Model       Notebook                                           concurs with the    Seismic Methodology, Inputs and Model Notebook (MDN-000-999-2019-response analysis and          FRANX Fragility Editor (Mapped to       Notebook was updated to                                                                assessment that     000267, Rev. 1) to be the updated value of 0.42. The updated value (for quantification was             SEIS_HLF, High Likelihood of Failure    accurately reflect the new CST                                                         the changes to the documented in the SEL,         Fragility) and does not agree with      fragility value of 0.42g. The                                                          SPRA constitute Page 226 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                              December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                            Self-Assessment Reference          Maint (M) or Basis for Maint RU  SR                  Finding No. Description                       Prior Peer Review Assessment            Self-Assessment Closure Basis                                                                            Independent Review Team Disposition                                         Review Team Assessment                                                                                                                           Document(s)                        Upgrade?     (M) or Upgrade SR Assessment Seismic Methods, Model            value of 0.3 g reported in Methods      FRANX table was changed to                                                          Maintenance. The     group SEIS_19-3) is also entered into the fragility table in the FRANX file Development _ Q1,Q2,Q3            Notebook, Section 6.5.13.               reflect this value as well.                                                         changes entail       (e.g., in BFN_Seismic_Rev1_U1CDF.franx).
(027), SHRA, SQU, Truncation                                                                                                                                  updating a fragility (028) and Convergence (022)                                                                                                                                   value and ensuring   This F&O is assessed as CLOSED.
notebooks, and associated                                                                                                                                     that it is correctly PRA Model documentation.                                                                                                                                      reflected in the documentation.
The update did not entail the use of new methods or require significant changes to the PRA model or results.
SPR C-SPR-C2 Met        27-1        Additional SSCs, e.g.,            Permanent N2 bottles in U1/U2 and       The UNIDs representing the          Seismic PRA Seismic Equipment List Maintenance  The review team      A review of the Seismic Equipment List (MDN-000-999-2019-000269,            Remains Met structures, passive               U3 DG corridor that were added as       HCVS nitrogen bottles, storage      Notebook                                        concurs with the     Rev. 1) and a search of the Access database containing the SEL components, panels, and           a FLEX strategy for HCVS B/U N2         racks and associated tubing,                                                        assessment that      (Composite SEL_draft_20190910 - Copy.accdb) confirmed that the cabinets / relays that house      supply are not included in SEL but      valves and instrumentation                                                          the changes to the   following two entries had been added:
PRA components not included       are credited in the SPRA via fragility  were added to the SEL. The                                                          SPRA constitute in IEPRA were added to SPRA.      basic events (TKURP0TNK_N2FLEX,         permanent equipment installed                                                       Maintenance. The
* BFN-0-LPNL-925-6100 - U2/U3 Nitrogen gas bottle rack These are listed in SEL           SSC ID C0_TNK_N2FLEX-U- Gxxx)           by the DCN related to the N2                                                        changes entailed
* BFN-1-LPNL-925-6100 - Nitrogen Gas Bottle Rack Notebook Appendix D-2, Table      and HEP basic events                    storage only included the N2                                                        updates to of New Seismic Basic Events,      (HFA_OPS_FLEXN2ALIGN) in BFN            bottle storage racks. The bottles                                                   documentation        This F&O is assessed as CLOSED.
and include but are not limited   SPRA for each unit.                     themselves do not have UNIDs                                                        only, and no to the following.                                                         since they are not permanent                                                        changes to the
: 1. Structures (Rx, DG and IPS     And, the N2 bottles have no UN ID       and can be recharged and                                                            SPRA model itself.
Buildings)                        cited in the SPRA notebooks.            replaced as needed.
: 2. Block wall failures
: 3. Rule of the Box failures
: 4. Relay chatter
: 5. Various LOCA types, including ISLOCA
: 6. IPEEE / A-46 SSEL SSCs.
Distribution systems (e.g.,
cable tray, HVAC ducts) were assessed via fragility walkdown and screened out from being added to SEL.
Distribution piping failures (ex.
Fire piping) and associated impacts were identified in other SR SPR-C5.
SPR C-SPR-E7 Met        27-2        The sources of uncertainty are    FLEX PPs / GENs were not                A sensitivity study has been        PRA Evaluation BFN-0-19-074        Maintenance  The closure team     A sensitivity study has been performed to evaluate the risk significance    Remains Met discussed in the Quantification   adequately considered for inclusion     performed to evaluate the risk                                                      assessed and         of crediting additional FLEX strategies. A recovery file was created to Notebook (Section 9.0) and        in the SPRA for ELAP or non-ELAP        significance of FLEX. A recovery                                                    agreed with the      credit FLEX in a simplified manner in sequences where FLEX would be assessed sources of model         scenarios, as a Rx injection source,    file was created only allowing in                                                   TVA                  implemented (e.g., ELAP sequences with initial success of HPCI or RCIC to uncertainty and their probable    for Seismic CDF / Seismic LERF          sequences where FLEX would be                                                       determination        allow sufficient time to align FLEX equipment). FLEX was not credited for effects on the model are          reduction, based on feasibility to      implemented. Although FLEX                                                          that the BFN SPRA    scenarios with judged inadequate time to align FLEX (e.g., ELAP discussed in Appendix G). In      deploy within time needed for           may reduce the risk of certain                                                      response to this     sequences without HPCI and RCIC, ATWS).
addition, sensitivity studies     success given expected battery life     accident sequences, BFN risk                                                        F&O is a PRA were performed in the             for HPCI/RCIC (4 hrs), MAAP run,        evaluation BFN-0-19-074                                                             Maintenance          The assumptions for the FLEX sensitivity case were reasonable and Quantification Notebook           time to TAF (Rx core damage) and        concluded that the inclusion of                                                     activity and not an  appropriate. BFN risk evaluation BFN-0-19-074 concluded that the (Section 10.0) to address the     FLEX deployment time.                   FLEX recovery does not have a                                                       Upgrade because      inclusion of FLEX recovery does not have a significant overall effect on potential effects of changing                                             significant effect on CDF and                                                       the activity various variables, parameters,    TVA BFN PRA engineers responded                                                                                             includes various Page 227 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                 December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                                                 Self-Assessment Reference Maint (M) or Basis for Maint RU SR            Finding No. Description               Prior Peer Review Assessment            Self-Assessment Closure Basis                                                               Independent Review Team Disposition                                         Review Team Assessment                                                                                                                Document(s)               Upgrade?     (M) or Upgrade SR Assessment or assumption in the SPRA to Peer Review Team Q's PJT-06,         LERF and does not need to be                                            modeling            CDF, LERF, or the risk insights and does not need to be explicitly added to model.                    PJT-07 and PJT-21, and reported         added to the model.                                                     enhancements        the Base Case SPRA model.
that (1) FLEX Strategy Validation                                                                               (i.e., revisions to Report for BFN Units 1, 2, and 3                                                                                flag and recovery   This F&O is assessed as CLOSED.
(03/23/2018) allows up to 8 hrs to                                                                              files) using no new deploy FLEX PPs and GENs.                                                                                       methods or The report states that during FLEX                                                                              calculations.
validation, it was done in 6 hours and 10 minutes (6.17 hrs).
A MAAP case associated with a different HEP, HFA_OPS_FLEXN2ALIGN (N2 bottles, showed a Tsw of 6.27 hours are available prior to core damage based on initial RPV injection from HPCI or RCIC for 4 hours.
BFN PRA engineers reported since only 6 minutes of time margin exists between 6.27 hrs available and 6.17 hrs to deploy, the time margin was too small to warrant crediting FLEX PPs / GENs in SPRA. However, the follow 2 factors warrant more consideration of FLEX for the SPRA, especially since FLEX has been included in the SPRAs developed and peer reviewed IAW the PRA Standard (Add. B or Code Case) in the past 2 years thus making it a state of practice.
: 1. The SPRA dominant accident sequences are GTRAN_005 & 006 (total of 43%) where HPCI & RCIC are available for 4 hrs prior to battery depletion at 4 hours for S-R 250 VDC, whereas less dominant seq. include GTRAN_011 & 012 (total of 13%) where HPCI & RCIC are initially unavailable and use of FLEX PPs / GENs does not avert early core damage (< 4 hours).
: 2. Per BFN 0-FSI-1, page 2, safety-related battery life for HPCI/RCIC can be extended from 4 hrs to 11 or 12 hours if S-R DC load shed is accomplished. BFN 0-FSI-1, page 2 states "shed loads from BB 1, 2, 3 in accordance with 0-FSI-3F within 1 hour to extend battery coping time to 12 hours, or within 2 hours to extend coping time to 11 hours."
Thus, this can extend a MAAP run Tsw from 6.27 hours to > 12 hours when accounting for Rx level Page 228 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                              December 2019 Table A-2 BFN SPRA F&O Closure Review Consensus Table Independent PRT CC II                                                                                               Self-Assessment Reference Maint (M) or Basis for Maint RU SR            Finding No. Description Prior Peer Review Assessment        Self-Assessment Closure Basis                                                           Independent Review Team Disposition Review Team Assessment                                                                                              Document(s)               Upgrade?     (M) or Upgrade SR Assessment decrease to TAF and resulting in core damage.
Page 229 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                            December 2019 A.6    Summary of Technical Adequacy of the Seismic PRA The set of SRs from the PRA Standard [8] that are identified in Tables 6-4 through 6-6 of the SPID [2] define the technical attributes of a PRA model required for a SPRA used to respond to implement the 50.54(f) letter. The conclusions of the peer review discussed above and summarized in this submittal demonstrates that the BFN SPRA model meets the expectations for PRA scope and technical adequacy as presented in NRC RG 1.200, Rev. 2 [45] as clarified in the SPID.
The main body of this report provides a description of the SPRA methodology, including:
* Summary of the seismic hazard analysis (Section 3)
* Summary of the structures and fragilities analysis (Section 4)
* Summary of the seismic walkdowns performed (Section 4)
* Summary of the internal events at-power PRA model on which the SPRA is based, for SCDF and SLERF (Section 5)
* Summary of adaptations made in the internal events PRA model to produce the SPRA model and bases for the adaptations (Section 5)
Detailed archival information for the SPRA consistent with the listing in Section 4.1 of NRC RG 1.200, Rev. 2 is available if required to facilitate the NRC staffs review of this submittal.
The BFN SPRA reflects the as-built and as-operated plant as of the cutoff date for the SPRA, January 2016.
A.7    Summary of Technical Adequacy of the BFN Internal Events PRA The BFN SPRA was built on the BFN Internal Events and Internal Flooding PRA model.
The following sections describe the peer reviews performed on these models.
The BFN Internal Events (excluding Internal Flooding) PRA was subjected to a full scope peer review in May 2009 [54], in accordance with the requirements of NEI 05-04
[55]. The review covered all technical elements in Part 2 plus the configuration control element from the PRA Standard RA-Sa-2009 that was endorsed by RG 1.200, Rev. 2.
Table A-3 presents the results of this peer review.
Table A-3 Internal Events PRA Model 2009 Peer Review SR Capability Category Distribution Capability Number          Percent Category Not Met               53            20 I                10            4 I/II              10            4 II               26            10 II/III             20            8 Page 230 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                           December 2019 Capability Number         Percent Category III               3             1 Met (All)           140            53 Not Applicable           2             1 TOTAL                    264           100%
Internal Flooding was not included in the scope of this review. Of these 264 PRA Standard SRs reviewed, approximately 76% are supportive of Capability Category II or greater. A total of 189 unique F&Os were generated by the peer review team, from which 95 were Findings, 92 were Suggestions, and 2 were Best Practices.
A separate Internal Flooding Focused-Scope Peer Review (FSPR) was performed on the BFN PRA in September 2009 [56]. The review covered all technical elements from the PRA Standard Part RA-Sa-2009 [57]. Table A-4 presents the results of this FSPR.
Table A-4 IF PRA Model 2009 Focused-Scope Peer Review SR Capability Category Distribution Capability Number       Percent Category Not Met              26          42 I               3           5 II or better           30          48 Not Applicable            2           3 Not Reviewed              1           2 TOTAL                     62         100%
The BFN internal flood PRA met Capability Category II or higher for about 48% of the applicable SRs. The BFN internal flood PRA met Capability Category I level for an additional 5% of the applicable SRs. A total of 50 F&Os were generated during this focused-scope peer review, including 29 Findings and 21 Suggestions.
The key problem areas for the internal flood PRA were documentation and flood scenario development. All 15 documentation SRs were rated as not meeting the standard requirements. The primary problem associated with documentation was lack of details, numerous inconsistencies, and incomplete information in the input data, process, and results. The internal flood PRA was not prepared in a manner that can facilitate PRA applications, upgrades and peer review. To be consistent with the applicable SRs, more effort was needed to enhance the documentation. The major problem associated with the flood scenario development was that the development of flood scenarios was not rigorously performed. Many flood areas, flood sources, and flood scenarios were dismissed without adequate considerations of all the possible flooding effects that may cause damage to SSCs credited in the PRA. As a result, the total number of flood scenarios that were quantitatively evaluated was far less than Page 231 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                          December 2019 expected, and the results from some top internal flood-induced risk contributors were not completely realistic.
The internal flooding model was updated, and an additional Internal Flooding model FSPR was completed in September 2018 [58], again covering all technical elements from the PRA Standard Part RA-Sa-2009 Part 3 [57]. This internal flood FSPR was conducted concurrently with an Internal Events F&O closure review activity [59]. The internal flood FSPR and associated conclusions supersedes the internal flood PRA Peer Review and associated findings from 2009. As a result of this FSPR, all existing F&Os were considered to be no longer applicable, and 11 new F&Os were generated. A total of 7 Findings, 3 Suggestions and 1 Best Practice were reported by the peer review team. The results of this assessment are reported in Table A-5.
Table A-5 Internal Flood PRA Model 2018 Focused-Scope Peer Review SR Capability Category Distribution Capability Number Percent Category Not Met or CC I                 7   11 II or better               55   89 Not Applicable                 0    0 Not Reviewed                   0    0 TOTAL                             62 100%
The peer review team concluded that, from a technical perspective, the internal flood analyses appeared to address the appropriate inputs and outputs, and the modeling approaches appeared sound. In addition, the team concluded that the changes made appeared to meet most of the requirements in the PRA Standard at or above Capability Category II, with the caveats provided with the F&Os.
Internal Events PRA F&O Closure Review The 95 Internal Events Finding F&Os identified in the peer review in May 2009 were subjected to a F&O Resolution FSPR in 2015 [60], which followed the guidance from NEI 05-04 [55]. The review was conducted over a three-day period by a team of four independent PRA experts, and included a consensus process to determine the adequacy of the resolution to each reviewed Finding. Following that review, there were 48 Findings that remained open, including 9 that were not assessed due to time constraints.
A subsequent F&O Closeout Assessment was completed in September 2018 at the TVA Chattanooga offices [59] for the 48 Internal Events Findings that remained open.
This assessment was completed in accordance with the process documented in Appendix X to NEI 05-04/07-12/12-13 [9], as well as the requirements published in the PRA Standard (RA-Sa-2009) and RG 1.200 Rev. 2 [45], including NRC expectations. A team of three independent PRA experts performed the F&O reviews along with the consensus sessions. The review met the Appendix X requirement that each F&O review include two qualified reviewers. Furthermore, the team examined the changes made to Page 232 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                          December 2019 the BFN PRA model, data, and documentation to address the findings to determine if the Capability Category II (or better) requirements of the PRA Standard, including clarifications imposed by RG 1.200, Rev. 2 were met.
The closure peer review team had significant PRA experience, and each team member confirmed they were not TVA employees, had no involvement in development of the BFN PRA or performance of risk applications for BFN, and no conflicts of interests, incentives, or disincentives.
The closure review team concluded that all but 10 of the 48 F&Os reviewed met the criteria for closure. In addition, an assessment was performed to determine if the F&O resolution resulted in an upgrade to the PRA or used new PRA methods. The peer review team concluded that those F&Os that were closed did not fall in the upgrade category and did not use new PRA methods (those F&Os remaining open were not assessed). Table A-5 presents the BFN Internal Events PRA F&O Closure Review Consensus, which includes the 10 Internal Events Finding F&Os that remain open.
Finally, the 7 Finding F&Os identified in the 2018 internal flood FSPR remain open, as no formal closure review has been performed to date. These Internal Flooding Finding F&Os that remain open are listed in Table A-6.
The open Internal Events and Internal Flooding F&Os were reviewed by the BFN SPRA peer review team. The SPRA peer review team created one F&O (19-1) that stated two of the open Internal Events may impact the SPRA. That F&O has been closed by the closure team as discussed in Table A-1. No Internal Flooding F&Os were identified as having an impact on the SPRA.
Page 233 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                              December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                                   Self-Assessment Reference         Maint (M) or Basis for Maint                                                                         Independent Review RU SR                         Finding No. Description                            Prior Peer Review Assessment           Self-Assessment Closure Basis                                                                         Independent Review Team Disposition Assessment                                                                                                                                        Document(s)                       Upgrade?     (M) or Upgrade                                                                          Team SR Assessment AS SY-B14 AS-B3 is Not Met    1-6         The sequence descriptions              The SR calls for identification of the Revision 3 of the Accident          NDN-000-999-2007-0036, Rev.5;     Maintenance  This resolution is The documentation provides explanation of the failures defined and   AS-B3 was previously AS-B3  SY-B14 is Met.                  generally include a description of     phenomenological conditions for        Sequence Notebook Section           BFN-0-16-031                                   limited to         the basis for assessment related to the postulated failure mode      Not Met and is now the sequences, but the                 each sequence.                         6.3.4.5.7 has been updated to                                                      enhancing          associated with LERF.                                                Met.
phenomenological conditions                                                   include the following:                                                             documentation to                                                                        SY-B14 was created are not specifically                                                  "The suppression pool suction                                                      provide a more     This F&O is assessed as Closed.                                      previously Met and identified. Some references to                                                strainers are modeled under                                                        detailed                                                                                remains Met.
phenomenology are provided but                                                gate Ux_LPC_G11 for MLOCA                                                          discussion of the not consistently (e.g., ATWS                                                  where x is the Unit number.                                                        quantification sequence descriptions conclude                                                Suppression pool suction                                                           results.
with the statement "There are no                                              strainer plugging would fail the                                                   (1) No new phenomenological conditions                                                   suppression pool cooling                                                           methods were identified.")                                                                 pathway."                                                                          used.
(2) There was no Revision 3 of the Accident                                                         change in PRA Sequence Notebook Section                                                          scope.
6.3.5.5.7 has been updated to                                                      (3) There was no include the following:                                                             change in PRA "The suppression pool suction                                                      capability.
strainers are modeled under gate Ux_LPC_G10 for LLOCA where x is the Unit number.
Suppression pool suction strainer plugging would fail the suppression pool cooling pathway. During an excessive LOCA SPC is not credited."
AS AS-A7  AS-A7 is met at Cat 5-5         Section 6.3.2.4.1 of the Accident      The omission of this sequence          The AS Notebook was updated         NDN-000-999-2007-0036, Rev.5;     Maintenance  This resolution is The assessment utilizes a screening approach based on low frequency  AS-A7 was previously I/II                            Sequence Analysis states that if       could result in an incorrectly low     to explain that not only would      NDN-000-999-2007-0041, Rev.8                   limited to         of occurrence as is allowed by the Standard. The discussion provides met at Cat I/II and Alternate Rod Insertion succeeds       CDF or cause the analyst to miss       the ATWS Induced LOCA                                                              enhancing          the supporting information to conclude that the sequence can be      remains met at Cat and either the recirculation pumps     important insight about the event. probability be below the ASME                                                      documentation to   excluded and is bounded by other similar scenarios.                  I/II.
fail to trip of the SRVs fail to open,                                        initiator frequency cutoff                                                         provide a more then a non-ATWS LOCA occurs                                                   recommended by IE-C4 of 1e-7                                                       detailed           This F&O is assessed as Closed.
which is not modeled in the PRA.                                              but would also be bounded by                                                       discussion of the While this new LOCA might be                                                  other LOCA IEs.                                                                    quantification quantitatively insignificant, no                                                                                                                                 results.
qualitative argument is made to                                                                                                                                  (1) No new justify its omission.                                                                                                                                            methods were used.
(2) There was no change in PRA scope.
(3) There was no change in PRA capability.
DA DA-C6  DA-C6 is Not Met. 1-17        Reviewed DA.01. The source of          Post-maintenance testing must be       As mentioned in the DA              NDN-000-999-2007-0033, Rev. 9     Open         none               BFN uses an automatic demand counter to populate the data. As such   DA-C6 remains Not demands is not discussed. Based        excluded from the exposure data        Notebook, the only demands          (DA.01);                                                          this would include all related surveillance, maintenance and         Met.
upon discussions with the PRA          per the SR.                            that are included in the data       BFN-0-15-079;                                                     operational demands. Because the system may count additional staff, exposure is collected directly                                         analysis update of failure rates    BFN-0-18-006;                                                     demands for PMTs BFN has estimated these additional demands and from plant data systems and is                                                are those that come directly        SY.21 (Safety Relief Valve System                                 performed sensitivities to support the impact on the failure rates.
therefore actual component                                                    from PEDs, from the IST             Notebook)                                                         Although the sensitivities may justify a minimal impact, it does not exposure. However, post-                                                      database or from the system                                                                           meet the SR (DA-C6).
maintenance testing demands are                                               engineer directly. The IST also included in these numbers and                                            database gives just those                                                                             This F&O is assessed as Open.
are not removed.                                                              successful demands that occur for each test (i.e., no post Page 234 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                 December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                                Self-Assessment Reference            Maint (M) or Basis for Maint                                                                              Independent Review RU SR                        Finding No. Description                          Prior Peer Review Assessment          Self-Assessment Closure Basis                                                                             Independent Review Team Disposition Assessment                                                                                                                                     Document(s)                          Upgrade?     (M) or Upgrade                                                                               Team SR Assessment maintenance demands included). PEDs/ the system engineer gives the actual number of demands the component observes which could potentially include post maintenance demands, however a sensitivity was performed (BFN-0-15-079) which shows that the model is not sensitive PMTs.
DA DA-C10 DA-C10 is not Met. 1-22        There is no discussion of the        All levels of capability in this SR   DataWare takes the actual            NDN-000-999-2007-0033, Rev. 10       Maintenance  The change         BFN uses an automatic demand counter to populate the data. As such        DA-C10 is now Met.
process to be applied in the use of  indicate that the process for use of  component demands and hours          (DA.01),                                          provides           this would include all related surveillance tests. Because the system surveillance test data. The use of   surveillance data needs to possess    as documented in PEDS. When          SY.07                                             documentation      may count additional demands from incomplete tests or unplanned this data is required for situations specific attributes. There is no      PEDS does not track the                                                                describing how     operation as a success, the data should be reviewed adjusted as in which there is no MR data         process defined.                      component being looked at,                                                             the data is        necessary to account for these demands. BFN reviewed all data for available (for example), so a                                              either the IST database was                                                            collected and      inappropriate inclusion of data and made adjustments, if necessary.
process for its use should be in                                           used, or the system engineer                                                           reviewed.          This additional step was added to the Data Notebook (DA.01) by place.                                                                     was contacted to figure out the                                                        (1) No new         Revision 10 and is described in Section 7.1. In addition, Section 7.3 and number of demands/hours that                                                           methods were       Assumption 7 of DA.01 describe the process to be followed when the occurred for that particular                                                           used.              electronic data system does not include the PRA component failure component. This information is                                                         (2) There was no   mode of interest, which includes the requirement to review test described in the DA Notebook                                                           change in scope of procedures. A sample of system notebooks (SY.07 and SY.21) were and Table 13 shows specifically                                                        the PRA.           reviewed confirming that a listing of applicable test procedures and a where the success information                                                          (3) There was no   discussion for how demands are determined was used for these cases.
comes from. The process                                                                change in the described in the DA notebook                                                           capability of the  This F&O is assessed as Closed.
was updated to clarify how the                                                         PRA.
data collection is performed.
DA DA-B2  DA-B2 is not Met. 5-3         The data analysis does not appear    The inclusion of outlier              During the data collection the       NDN-000-999-2007-0033, Rev.9 (DA.01) Maintenance  The change is      The process documented in Attachment E of DA.01 ensures that the          DA-B2 is now Met.
to consider outlier components.      components can incorrectly impact     current grouping of the                                                                documentation      data groupings are compared to generic data to identify outlier the failure rate assigned to a        component is used, however                                                             only.              behaviors. Where plant data is outside the generic 5% or 95% bounds, component group. Such outlier         the data analyst looks at the                                                          (1) No new         plant-specific data is used. The bases that outliers components are not components should be placed into      data and any components that                                                           methods were       inappropriately grouped is documented in DA-01. No outliers were a separate suitable component         are never tested would have                                                            used.              identified, and these results are unchanged.
group.                                little or no data to update the                                                        (2) There was no failure rate of the typecode in                                                        change in scope of This F&O is assessed as Closed.
the model with. These are                                                              the PRA.
looked at and determined                                                               (3) There was no whether it is more appropriate                                                         change in the to keep them within the same                                                           capability of the grouping as they are the same                                                          PRA.
type of component, experience the same type of environmental conditions, and have about the same type of failure rates or whether they should be put into a separate grouping. This was the intent of bullet 3 of Section 5.0. As shown in the DA Notebook Appendix E, the prior and posterior distributions were reviewed, and it was determined whether generic data was a suitable representation for BFN.
Page 235 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                          December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                                  Self-Assessment Reference    Maint (M) or Basis for Maint                                                                          Independent Review RU SR                      Finding No. Description                            Prior Peer Review Assessment         Self-Assessment Closure Basis                                                                       Independent Review Team Disposition Assessment                                                                                                                                       Document(s)                  Upgrade?     (M) or Upgrade                                                                           Team SR Assessment HR HR-C3 HR-C3 is not Met. 2-14        HFL_1003_CCFT0056 is Common            The pair CCFs will have a higher     HFL_1003_CCFT0056 is                   BFN IE PRA Model, Rev 8      Maintenance  The change         The Calibration CCF of 2 of 4 instruments has been added to the       HR-C3 is now Met.
HR-D5 HR-D5 is not Met.             cause miscalibration of all 4 level    value than the 4 of 4 event thus     Common cause miscalibration                                                      involved adding    model.                                                                HR-D5 is now Met.
transmitters, inspection of the        impact the results.                  of all 4 level transmitters;                                                     the basic events fault tree shows that specific pairs                                        Events for the critical 2 of 4                                                   for 2 of 4 CCF     This F&O is assessed as Closed.
of failures (AC, BD) would also                                             failure groups have not been                                                     events.
cause a failure to initiate the logic.                                      added to the model and the                                                       (1) No new These CCF pairs should be added                                             original F&O issue has not been                                                  methods were to the                                                                      addressed.                                                                       used. The model model. This will apply to other
* ADDED pre-initiator HEs for                                                    change was made miscalibration CCFs also.                                                   miscalibration of 2 of 4                                                         using processes instruments to the PRA model.                                                    and tools.
See gates U1_LM_G01,                                                             (2) The scope U2_LM_G01, U3_LM_G01                                                             included several new basic events.
There was no significant change to the risk insights.
(3) There was no change in PRA capability.
HR HR-G2 HR-G2 is Met. 4-18        Some operator actions assume           Execution failure is a required part
* Execution error has not been         NDN-000-999-2007-0032, Rev 6 Maintenance  The update         Execution failure probability has been added to some HFEs but not     HR-G2 remains Met.
that the execution failure             of the HEP calculation, and the      included for ADS inhibit                                                         involves updating  others. HFA_0024RCWINTAKE involves physically cleaning the intake probability (Pe) is including:         argument for ignoring execution      (HFA_0_ADSINHIBIT). This is                                                      HEP execution      screens within time to prevent a plant trip or equipment overheating.
HFA_0_ADSINHIBIT,                      failure is not necessarily           modeled only for ATWS in the                                                     failure            Assuming the execution failure probability is zero is inappropriate.
HFA_0_ATWSLEVEL,                       compelling, especially for           PRA. There is a single step to                                                   probabilities.
HFA_0024RCWINTAKE,                     maintaining level                    implement this action, errors of                                                 (1) No new         This F&O is assessed as Partially Closed.
HFA_0027INTAKE, HFA_0IR2_LPI,          (HFA_0_ATWSLEVEL). Some of the       omission are integral to the                                                     methods were HFA_1063SLCINJECT,                     actions for which Pe is not          cognitive error to omit the                                                      used. The HEP HFA_0024IFISOL                         considered are important to the      action. Errors of commission are                                                 update uses the Example 1: Several operator            overall results.                     neglected because the action to                                                  HRA Calculator actions for ATWS scenarios (e.g.,                                           inhibit ADS is unique (no                                                        consistent with HFA_1063SLCINJECT: Failure to SLC      Note 1: The explanation given for    transition to any EOI Appendix                                                   current practice.
in response to an ATWS event)          no execution failure for             is required, and there are                                                       (2) The scope was assume the execution failure           HFA_0_ATWSLEVEL describes the        several places in the EOI that                                                   limited to a probability (Pe) is 0.0.               actions required for starting SLC    call for inhibiting ADS), and                                                    limited set of Example 2: Operator action             (HFA_1063SLCINJECT).                 because it is routinely                                                          HFEs.
HFA_0024RCWINTAKE (Failure to          Note 2: Cleaning debris from         performed for every reactor                                                      (3) There was no clear debris at intake before          traveling screens is not a simple    scram, graphically distinct and                                                  change in PRA reactor scram) assumes an              action, an assumption, that if the   performed after SLC.                                                             capability.
execution error of 0.0 based on the    actions are started, they are
* Execution error was added for following: 'Cleaning traveling         guaranteed to be completed in 1      SLC. This is a time critical screens does not relate to a series    hour, is not justified.              operator action, and the EOI of manual actions, but to an effort                                         specifies the appropriate steps among several operators. It is                                              required in EOI-Appendix 3A.
assumed that, if the action is                                              While the actions are simple, initiated within 1 hr, it will be                                           these require transition successful.' The same rationale is                                          between procedures for the provided for no execution error in                                          execution, so it is appropriate to HFA_0027INTAKE.                                                             include execution errors.
* HFA_0_ATWSLEVEL -Execution errors are included for this event. NO CHANGE.
* HFA_0024RCWINTAKE -
Execution error set to zero and it deemed not necessary to add detail for this activity. Clearing traveling screens does not Page 236 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                           December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                                  Self-Assessment Reference    Maint (M) or Basis for Maint                                                                               Independent Review RU SR                     Finding No. Description                            Prior Peer Review Assessment           Self-Assessment Closure Basis                                                                      Independent Review Team Disposition Assessment                                                                                                                                       Document(s)                  Upgrade?     (M) or Upgrade                                                                                Team SR Assessment relate to a series of manual actions, but to an effort among several operators, so errors of execution are in parallel and considered unlikely. It is assumed that, if the action is initiated within 1 hour, it will be successful (i.e., only the cognitive error is included). The RCW system is supplied river water from the CCW conduits of each unit through fine mesh strainers that include a dP alarm. Pumps are run periodically to avoid fouling.
* HFA_0027INTAKE - Basic event is not in the model. NO CHANGE
* HFA_0IR2_LPI -Execution errors are included for this event. NO CHANGE.
* HFA_0024IFISOL - This event is not used in the PRA model. NO CHANGE.
HR HR-F2 HR-F2 is Not Met 4-25        There are many operator actions        Without any real timing                Clarification on the basis for the    NDN-000-999-2007-0032, Rev.6 Open         none               BFN-0-16-031 list several HFEs with clarification of the timing            HR-F2 remains Not HR-G4 HR-G4 is Not Met             that use screening values; see         information, it is not possible to     timing has been added to the          BFN-0-16-031                                                 information. These are not the HFEs listed in Table 8 as referenced in     Met (F&O 4-25)
HR-G5 HR-G5 is Not Met             Table 8 of the HRA. None of these      estimate, even at a screening level,   HRA Notebook.                                                                                      the F&O, nor is there any discussion why these events were selected.       HR-G4 remains Not actions appear to use any              the probability of operator failure                                                                                                       NDN-000-999-2007-0032 Assumption 10 assumes that screened HFEs             Met (F&O 4-25) information to base the time           or success.                                                                                                                               all have a delay time of 24h. This is not consistent with several of the   HR-G5 remains Not available and the times to operator                                                                                                                                              event descriptions, which imply the timing would need to be less than      Met (F&O 4-25) cues and perform the actions are                                                                                                                                                 24h for success (some screened events list times of 15m or less in the not documented.                                                                                                                                                                  description).
This F&O is assessed as Open.
HR HR-C1 HR-C1 is Met. 4-28        Non-screened miscalibration            The requirements of HR-C1 cannot       A table has been added to             NDN-000-999-2007-0032, Rev.6 Maintenance  The resolution     HRA Notebook Section 6.2.3 discusses screening values for pre-             HR-C1 remains Met.
events are not provided with           be verified due to lack of             Appendix A of the HRA                 BFN IE PRA Model, Rev 8.                  enhanced the       initiators and references section 6.2.2.3 for detailed analysis. A table designators in Appendix A of the       traceability from HRA Appendix A       Notebook that list all the pre-                                                 documentation      has been added to NDN-000-999-2007-0032 Appendix A that HRA. Therefore, HFEs associated        table to the rest of the preinitiator  initiator CCF HFEs                                                              for screened HFEs. specifically lists the pre-initiator CCF events. The detailed results from with these miscalibration events       analysis.                                                                                                              (1) No new         the HRA Calculator are included as Appendix B.
cannot be readily determined.                                                                                                                                 methods were used.              This F&O is assessed as Closed.
(2) There was no change in PRA scope.
(3) There was no change in PRA capability.
HR HR-A1 HR-A1 is Met. 4-29        The list of activities reviewed in the The review of procedures should        As mentioned on Section 4.3 of        NDN-000-999-2007-0032, Rev.7 Maintenance  The resolution     The HRA Notebook discusses the required procedure reviews in               HR-A1 remains Met.
HR-A2 HR-A2 is Met.                HRA Appendix A table is primarily      not be limited to one unit.            the HRA Notebook, the                                                           involves a         section 4.3. Procedures are reviewed for all the units.                    HR-A2 remains Met.
focused on Unit 2 or Unit 0 SRs and    Differences between units may          operating practices, staffing and                                               documentation SIs. There are a few Unit 1            present additional pre-initiator       training for all three units are                                                change only to     This F&O is assessed as Closed.
procedures listed, but it is not clear actions. Although the one example      identical. Differences that could                                               revise procedure why certain procedures from Unit       found would not likely result in a     relate to the HRA are reflected                                                 references.
1 are reviewed but not others.         pre-initiator, the point is that there in the system fault trees. The                                                  (1) No new More importantly, there do not                                                procedures were reviewed and                                                    methods were appear to be any Unit 3                                                       only one unit was referenced                                                    used.
Page 237 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                       December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                             Self-Assessment Reference    Maint (M) or Basis for Maint                                                                           Independent Review RU SR                     Finding No. Description                           Prior Peer Review Assessment       Self-Assessment Closure Basis                                                                      Independent Review Team Disposition Assessment                                                                                                                                  Document(s)                  Upgrade?     (M) or Upgrade                                                                            Team SR Assessment procedures reviewed. A sample         are differences between the units' for each different procedure in                                                 (2) There was no review of one procedure between       procedures.                        the HRA Notebook. This is                                                       change in PRA all three units (3.5.1.5(CS I)) found                                    because each unit has the same                                                  scope.
that the Units 1/2 tests affected                                        steps within the procedure and                                                  (3) There was no two relays that are not tested in                                        the only thing that would be                                                    change in PRA the Unit 3 procedure.                                                    different would be specific                                                     capability.
UNIDs, but the overall result would be the same.
HR HR-H1 HR-H1 is Met. 6-25        Event HFA_3003P_START_A does          Significance is unknown, since     The modeling for                      BFN IE PRA Model, Rev 8      Maintenance  This item was a    HFA_0003P_START_A is no longer credited but is still in the model with HR-H1 remains Met.
not appear to be applied correctly    model modification is required in  HFA_0003P_START_A was                                                           modeling fix       probability of 1.0.
in the model. A question was asked    order to determine the impact. reviewed by plant personnel                                                     involving removal of the analysts on the logic, and                                        and it was determined that                                                      of credit for an   This F&O is assessed as Closed.
the response referred to gate                                            there are still some issues                                                     HFE.
U3_FWH_INIT for events were FW                                           related to this event that need                                                 (1) No new recovery is not credited.                                                to be resolved.                                                                 methods were However, the logic under gate
* It was determined that this                                                   used. The model U3_FWH_G50 limits the operator                                           event should be set to TRUE due                                                 was updated using failure event to only excessive FW                                       to inadequate time to restart a                                                 existing processes events; resulting in no failures                                         pump during at ATWS prior to                                                    and tools.
coming through for other events                                          MSIV isolation (After MSIV                                                      (2) The PRA scope were FW is credited.                                                     isolation steam is not available                                                was limited to a to drive the RFW pump). In                                                      single event.
addition, the ATWS procedure                                                    (3) There was no calls for terminating all injection                                             change in PRA except for RCIC, CRD and SLC.                                                   capability.
HR HR-G5 HR-G5 is Not Met 6-28        Basis for operator action time (30    Event provides over 5% of CDF. F&O 6-28 states that for              NDN-000-999-2007-0032, Rev.6 Maintenance  The change         Per NDN-000-999-2007-0032, Rev.6, HFA_0085ALIGNCST has been re-        HR-G5 remains Not min) for HFA_0085ALIGNCST                                                HFA_0085ALIGNCST, rough               0-OI-2B                                   enhanced the       evaluated. A basis has been included for Texe of 30 minutes based on   Met (due to F&O 4-appears to be roughly estimated,                                         estimates are used for Tm (30                                                   justification for  operator insights.                                                     25) as is the time available (7 hours).                                      minutes) and that Tsw=10 hours                                                  HFE timing is not consistent with info from                                                assumptions and    This F&O is assessed as Closed.
plant personnel. The basis for                                                  revised a HEP.
both of these are shown in the                                                  (1) No new HRA Notebook Revision 5. The                                                    methods were basis for the Tm value is given in                                              used. The HEP the Operator Interview Insights                                                 analysis uses the section. According to 1-ARP                                                  HRA Calculator 6B an alarm actuates in MCR                                                     consistent with when the CST reaches the 12-                                                    accepted foot level and there is a 10-                                                   guidance.
minute decision time (Tcog)                                                     (2) The scope was associated with whether to                                                      limited to a crosstie or not. OSC takes 20                                                   specific HFE.
minutes to staff and stage in                                                   (3) There was no field. Then the valves that need                                                change in PRA to be manipulated would take                                                    capability.
10 minutes to get to and manipulate (estimate from operator that has recently performed actions).
(Tm=Texe=20+10=30 min)
The Tsw issue was that we used a different inventory level of 180,000 gallons instead of 135,000 gallons (standpipe) as done for all other MAAP estimates. An assumption is specified in the HRA Notebook Page 238 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                               Self-Assessment Reference            Maint (M) or Basis for Maint                                                                            Independent Review RU SR                     Finding No. Description                            Prior Peer Review Assessment        Self-Assessment Closure Basis                                                                              Independent Review Team Disposition Assessment                                                                                                                                    Document(s)                          Upgrade?     (M) or Upgrade                                                                             Team SR Assessment Revision 5, that explains the 180,000 gallons corresponds to the available volume in the standpipe plus a CST height of 15 feet. This is used based on 0-OI-2B which cautions that a CST level below 15 feet may cause a loss of condenser vacuum.
See BFN-0-16-031.
HR QU-C2 QU-C2 is Not Met 6-30        Dependencies between operator          Systematic error affecting around
* The basis for ZD between            NDN-000-999-2007-0032, Rev.6,        Open         none               The stated resolution addresses only some specific HFEs, however        QU-C2 remains Not HR-G7 HR-G7 is Not Met             actions appear to be non-              1/2 of the combo events, including  early depressurization                NDN-000-999-2007-0041, Rev 8                                         during discussion it was identified that the dependency analyses were   Met conservatively applied. Mainly, the    combo 18.                           HFA_0001HPRVD1, and failure                                                                                completely redone. The actual process used to identify and process      HR-G7 remains Not Zero Dependence (ZD) between                                               to align suppression pool                                                                                  dependencies in general is not described, only that the "EPRI           Met actions is commonly applied,                                               cooling is significant differences,                                                                        recommended" method is used. More detail is needed. HRA NB simply when one of the actions                                             cues and timing. Early                                                                                     Section 6.3.3 points to the Quantification and Quantification NB points takes longer than 60 minutes.                                              depressurization is associated                                                                             back to HRA NB. The use of automated tools is mentioned but the What appears to be the mistake is                                          with failure to maintain RPV                                                                               actual tools and how they are used is not discussed.
applying the last event tree node                                          level, while failure to align SPC                                                                          There is and assumption (in HRA and Quant) that HFEs with screening in the Dependency Event Tree. In                                           (non-ATWS/IORV) is associated                                                                              HEPs of 0.1 or greater are treated as independent. Discussions with the this tree, if the stress of either HFE                                     with SP temperature. MAAP                                                                                  analyst indicated this is not how they are treated.
is moderate or high, the upper leg                                         analysis demonstrates that                                                                                 In the Quantification NB it states that the base quantification use a of the event tree is used. So for                                          operators have 3 hours to start                                                                            seed value of 0.15 for all HEPs. In section 6.3.1.9 its states that a combo 2, the HRA assumes ZD,                                               suppression pool cooling to                                                                                sensitivity is performed using 1.0 as the seed value and references the while the event tree would                                                 avoid exceeding 190F and thus                                                                              HRA calc. It is not clear how the dependent HFEs are identified.
designate Low Dependency.                                                  eventually impacting HPI systems taking suction from the                                                                            This F&O is assessed as Open.
SP. Since HPCI and RCIC take suction from the CST initially, it would take several hours to deplete the CST prior to any swapping suction to the SP.
Early SPC failure was included in the model under late failure for HPI since early failure would result in high SP temperature that may preclude late swap over of suctions for HPI.
* The basis for the User Defined dependency levels has been added to the HRA calculation in Appendix E.
IE IE-A7 IE-A7 is Met. 3-7         Scheduled manual shutdowns             CRNM ASME Standard                  Manual Scrams have been               NDN-000-999-2007-0030, Rev 2 (IE.01) Maintenance  The change         The Initiating Events Notebook, IE.01 was updated to include tabular    IE-A7 remains Met.
(especially for refueling outages)     Interpretation #5 (for FAQ 06-      broken up into its own initiator                                                        involves removal   information for each of the SCRAMs in the defined period. These tables should not be included in the          1060) states that normal controlled in the MOR R7. Further                                                                  of planned         assign an initiator bin to each SCRAM. While a statement that planned statistical basis for the scram        shutdowns should not be included    refinement of the manual scram                                                          shutdowns (e.g., a shutdowns are excluded is not included, a review of the calculation initiator. This can lead to an overly  when counting initiating events. initiator is currently being                                                            refueling outage   indicates that the number of events has excluded the planned manual conservative scram initiator           The current practice at Browns      investigated and will be                                                                shutdown with a    SCRAMs.
frequency.                             Ferry regarding this item,          included once finished. Since                                                           planned manual Note that CNRM interpretation for      therefore, does not meet the        the results are conservative                                                            SCRAM) from the    This F&O is assessed as Closed.
FAQ 06-1060 (should non-forced         requirements of the standard.       without the refinement, the                                                             initiating event manual trips which are part of the                                         model will continue to use the                                                          frequency.
normal shutdown procedure be                                               unrefined manual scram                                                                  (1) No new counted) states that 'a normal                                             initiator.                                                                              methods used.
controlled shutdown would not                                                                                                                                      This removal is present the same challenges as a                                                                                                                                   consistent with trip from full power if the manual                                                                                                                                 current practice Page 239 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                               December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                            Self-Assessment Reference            Maint (M) or Basis for Maint                                                                               Independent Review RU SR                       Finding No. Description                         Prior Peer Review Assessment         Self-Assessment Closure Basis                                                                            Independent Review Team Disposition Assessment                                                                                                                                 Document(s)                          Upgrade?     (M) or Upgrade                                                                                Team SR Assessment trip was prompted by conditions                                                                                                                                and uses the other than the normal shutdown                                                                                                                                 existing tools.
procedure which could occur at full                                                                                                                            (2) The scope power, it should be counted.                                                                                                                                   includes a change in the initiating event frequency with expected impact on the results. Risk insights are not significantly impacted.
(3) There is no change to the PRA capability.
IE IE-C8  IE-C8 is Not Met  6-10        CCF for Battery Chargers is not     Can affect the loss of DC initiating The IE Notebook lists an            IE.01 = NDN-000-999-2007-0030, Rev 2 Open         none               An assumption in IE.01 states that inclusion of common cause failures      IE-C8 remains Not included in the Initiating Event    events by a factor of 10, depending  Assumption about why inclusion      EPRI TR1016741, Support System                                       in the initiating event tree would yield inappropriate/conservatively      Met (see F&O 6-10).
Fault Tree for loss of 2 DC buses,  on how CCF is calculated.            of common cause is not              Initiating Events                                                    high frequencies. This is counter to current guidance in EPRI other than for the standby                                               included for support system                                                                              TR1016741. An update to IE.01 should be prepared following the EPRI chargers (not in the yearly failure                                      initiators. Inclusion of common                                                                          process which allows for appropriate screening of events and other rate logic).                                                             cause into the support system                                                                            adjustments.
initiator development would produce overly conservative                                                                              This F&O is assessed as Open.
initiator frequencies as mentioned in the previous response. In order to obtain a more realistic model TVA decided to leave out the common cause events for initiator development. Inclusion of the common cause for support system initiator development will be reevaluated and incorporated as required following completion of the evaluation.
IE IE-C14 IE-C14 is Not Met 6-13        The impact of Surveillance          Unknown impact on the ISLOCA         There is an open permissive         IE.01 - NDN-000-999-2007-0030, Rev 2 Maintenance  The change is      SY.04, Core Spray System Notebook and IE.02, ISLOCA Initiating Events      IE-C14 is now Met Procedures is not included in the   Frequency, without analyzing the     interlock between the inboard       SY.04 Core Spray System Notebook                  documentation      document the interlock between the inboard and outboard isolation ISLOCA Calculation.                 specifics of the site procedure. If  and outboard injection valves       IE.02 - ISLOCA Initiating Events                  only. The          valves will not allow the condition of an undetected rupture of the For example, for Core Spray,        the procedure has the operator       that allows both valves to be                                                         additional         testable check valve and the opening of the MOV because the pressure Surveillances in the CS Notebook    check downstream pressure (etc.)     open only when reactor                                                                documentation      interlock is set at 450 psig. On this basis, the exclusion of the MOV test indicate an MOV opening every 92    prior to opening the MOV, likely     pressure is below the low                                                             justifies the      is justified.
days. The likelihood of an ISLOCA   there is minimal impact. However,    reactor pressure setpoint. The                                                        continued during this MOV test is not         given the ISLOCA has a large         CS inboard and outboard                                                               exclusion of a     This F&O is assessed as Closed.
calculated in the ISLOCA IE Fault   impact on LERF, the impact could     injection valves have in-line                                                         potential ISLOCA Tree, including the sequence        be significant.                      valve interlocks to prevent both                                                      initiator.
where the check valve would have                                         valves from being opened with                                                         (1) No new previously failed prior to the                                           RPV pressure at or above 450                                                          methods were surveillance.                                                            psig. Both receive auto open                                                          used.
signals when there is a CS                                                            (2) There was no initiation signal and RPV                                                             change in scope of pressure is below 450 psig. The                                                       the PRA.
inboard valve may be throttled                                                        (3) There was no immediately after initiation.                                                         change in the Therefore, failure of the                                                             capability of the operator to check downstream                                                          PRA.
pressure prior to opening the Page 240 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                  December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                                 Self-Assessment Reference            Maint (M) or Basis for Maint                                                                                Independent Review RU SR                        Finding No. Description                         Prior Peer Review Assessment           Self-Assessment Closure Basis                                                                              Independent Review Team Disposition Assessment                                                                                                                                      Document(s)                          Upgrade?     (M) or Upgrade                                                                                 Team SR Assessment MOV for testing would not occur due to the low-pressure permissive interlocks.
IE IE-A5 IE-A5 meets Cat II. 6-2         Loss of HVAC as an initiating event Modeling changes have resulted in      Revision 2 of the Initiating Event   IE.01 = NDN-000-999-2007-0030, Rev 2 Maintenance  Additional          The HVAC System Notebook SY.08 was updated to identify                     IE-A5 remains met at IE-C6 IE-C6 is Not Met.               is screened, based on the 1995      HVAC becoming one of the top 5         Notebook Section 6.2.3.3 now         Initiating Events                                 documentation       components requiring HVAC for success, including areas with                Cat II.
PRA of the event. It appears the    systems in the present PRA. Based      states:                              SY.08, HVAC System Notebook                       was provided for    temperature detectors that if not cooled would cause MSIV closure.         IE-C6 is now Met.
model and the assumptions for       on this, a loss of HVAC initiating     "The initial plant faults that       SY.09, Main Steam System Notebook                 the exclusion of    The Main Steam System Notebook, SY.09 provides that for loss of HVAC have changed, and      event is likely to be significant as a result in reactor pressurization                                                       the CRD cooling as  temperatures greater than 189F in unspecified areas the MSIVs are loss of HVAC as an initiating event contributor to core damage and         are given in Table 10. The                                                             an initiator, no    failed closed. Based on SY.08, this area is assumed to be the Main should not be screened.             should not be screened.                reactor pressurization initial                                                         change in           Steam Tunnel. The fault tree model includes BEs (e.g.,
plant faults include RCS high                                                          methods was         FANFR1FAN_0640135) that cause closure of the MSIVs in the power pressure, loss of condenser heat                                                       introduced. The     conversion system tree; however, this BE is not an input to the sink and turbine trip. The loss of                                                     documentation       inadvertent MSIV closure initiator fault tree, rather a point estimate is condenser heat sink is made up                                                         was also updated    provided for IMSIV. There is no discussion about timing for area heatup of Inadvertent Closure of all                                                          to provide the      or operator indication for areas with temperature detectors causing MSIVs (including from loss of                                                          basis for the       MSIV closure. Adding this information could provide a basis that this HVAC), Loss of Condenser                                                               inclusion of        initiator could potentially be screened. Other initiators, such as loss of Vacuum, Turbine Bypass                                                                 another initiator,  power to the shutdown board have a similar affect as the IMSIV Unavailable, and Loss of Plant                                                         loss of ventilation imitator and would be caused by some of the same issues causing Air. None of these SCRAMs are                                                          to the steam        failure of the fan. Therefore, the addition of the ventilation fan failure grouped. For the loss of                                                               tunnel causing      to the IMSIV initiator is not expected to cause a significant change in condenser heat sink, the MSIV                                                          inadvertent MSIV    results. The Initiating Events Notebook IE.01 includes a systematic closure, turbine bypass                                                                closure. The        Failure Modes and Effects Analysis (FMEA) of key support systems unavailable, and loss of                                                               additional changes  failures that could cause a reactor SCRAM, including this item.
condenser vacuum have                                                                  described in the    Separately, other information was incorporated into IE.01 to exclude different effects on the pressure                                                      disposition are not loss of cooling to CRD pumps as an initiator. The basis for this exclusion spike. The loss of plant control                                                       expected to have    is provided.
air has a different timing and                                                         a significant affects a number of                                                                    impact on results. This F&O is assessed as Closed.
components."
Revision 2 of the Initiating Event Notebook Section 6.2.3.8 now states:
                                                                                                                    "The loss of important HVAC systems is well annunciated, and heatup calculations show that there is ample time for the operators to restore HVAC or take procedurally guided steps to prevent unnecessary isolation or SCRAM.
Additionally, many of the systems cool areas that do not have high heat loads during normal power operations or do not have equipment necessary for normal operation. For additional discussion see the BFN PRA HVAC system notebook. An exception to this is the CRD system which is only available for 4 hours after a loss of HVAC. The difference in the CRD system is that it does not cause a SCRAM or preclude a SCRAM to occur from another accident signal. The Technical Specifications do not require Page 241 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                             Self-Assessment Reference            Maint (M) or Basis for Maint                                                                              Independent Review RU SR                      Finding No. Description                         Prior Peer Review Assessment        Self-Assessment Closure Basis                                                                                 Independent Review Team Disposition Assessment                                                                                                                                  Document(s)                          Upgrade?     (M) or Upgrade                                                                               Team SR Assessment the CRDs to be functional, just that the control rods be able to insert if the need arises, but it is stated that if both CRD pumps are not available to start, a manual SCRAM would occur.
Therefore, HVAC induced loss of CRD is not a direct initiator in the PRA but is modeled implicitly in the manual SCRAM initiator."
See BFN-0-16-031.
IE IE-C8 IE-C8 remains Not 6-36        The ISLOCA Conditional Pipe Break   ISLOCA is a significant contributor The indicated parameters were         NDN-000-999-2007-0039, Rev 2 (ID.02) Maintenance  The change            The model file was confirmed to be consistent with the intended IE.02  IE-C8 remains Not Met                           Frequencies calculated for the      to LERF.                            updated in MOR 7 to indicate          BFN IE PRA Model, Rev 8                           involves both         documented values for ISLV21 and ISLV23. This corrected the noted      Met (see F&O 6-10).
analysis appear to be too low, in                                       the documented values. The                                                              correction of a BE    deficiency. The pdf "broken link" issues were also corrected in the comparison with other pants. From                                       reference links were broken                                                             value in the model    updated document.
NUREG/CR-5102, Appendix F,                                              when the document was                                                                   and a Table 2, the RHR and CS piping                                          converted into a PDF for record                                                         documentation         This F&O is assessed as Closed.
would generally get a failure                                           submittal and this was missed                                                           correction.
probability of 2.65E-02 and 2.54E-                                      prior to issuance into the vault.                                                       (1) No new 03 respectively. Other reference                                        The references were fixed for all                                                       methods were documents used should get similar                                       applicable instances and the                                                            used. The updated results. The BFN analysis is                                            specific references for the                                                             used existing supported by and Excel                                                  tables for section 6 are either                                                       methods and Spreadsheet for the overpressure                                        labeled directly or taken from                                                          tools.
estimate, and this analysis is not                                      references 26 and 27 of the                                                             (2) There was no included in the system notebook.                                        ISLOCA Notebook. Section 6.3.3                                                          change in scope of In the excel spreadsheet it appears                                     specifies that these are the                                                            the PRA. No the temperature assumed for the                                         references for the overpressure                                                         significant impact CS and RHR analysis assumes room                                        analysis.                                                                               in results or in risk temperature, whereas full RCS                                                                                                                                   insights resulted temperature is more appropriate.                                                                                                                                from this level of change.
(3) There was no change in the capability of the PRA. The corrected values resulted in the expected change in results with the
                                                                                                                                                                                                       % contribution for these ISLOCA initiators remaining very low after the change, approximately 0.5% of the total contribution from all initiators.
IE IE-B4 IE-A5 is Met      6-5         The calculation of HPCI Steam       Pipe break in the HPCI line can     The pipe rupture frequency            NDN-000-999-2030 Rev 2 (IE.01);      Maintenance  This response         The IE.01 Notebook has been updated by using the latest pipe break     IE-A5 remains met at IE-A5 IE-B4 is Met.                 Lines breaks (IE Section 6.2.3.8)   affect RCIC and many other          numbers in Section 6.2.3.8 were       EPRI Report 3002000079.                           applies the most      frequencies to calculate the HPCI unisolable break frequency.          Cat II.
IE-C6 IE-C6 is Not Met.             does not appear to be reasonable,   components, due to the HPCI         updated to the reference the                                                            recent data for       However, there is no discussion of why the specific values were        IE-B4 remains Met.
using older EPRI data and Wash-     pump being open to other areas. current pipe failure rates for                                                          calculating the       selected (For example, the HPCI line break uses the table for NPS >10  IE-C6 is now Met.
1400 data. The resulting steam line The modeling as documented does     HPCI, RCIC, and RWCU and the                                                            subject event         and EBS of 10 in). The calculation is also using updated failure rates Page 242 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                                  December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                               Self-Assessment Reference             Maint (M) or Basis for Maint                                                                             Independent Review RU SR                     Finding No. Description                          Prior Peer Review Assessment          Self-Assessment Closure Basis                                                                               Independent Review Team Disposition Assessment                                                                                                                                    Document(s)                           Upgrade?     (M) or Upgrade                                                                              Team SR Assessment break calculated is 4.55E-10/year,   not provide basis for screening,      rationale shown below (in bold)                                                          frequencies.       and CCF factors. This updated frequencies for the subject initiators are which does not compare with          and if re-performed, the analysis     was added to the last paragraph                                                          (1) No new         below the IE-C6 screening criteria and additional discussion is provided results from other plants. Using     will likely result in orders of       to explain why each break                                                                methods were       to demonstrate that at least two trains of mitigating systems remain newer data, the pipe break           magnitude increases here.             outside containment was                                                                  used. Only data    available for the events being screened. As an example, for unisolable frequencies would likely be 2-                                             screened.                                                                                was updated.       HPCI line breaks, RCIC and LPCI are stated as being available. This orders of magnitude higher.                                                The 2009 Standard describes a                                                            (2) There was no   discussion does not provide supporting detail that the nearby RCIC Additionally, although the isolation                                       significant cutset as the                                                                change in scope of pump would not be affected by the steam and water released from the valves may be available to                                                 summed percentage is 95% and                                                             the PRA.           HPCI line break.
eventually isolate the break, the                                          the individual percentage is 1%                                                          (3) There was no impact of the break may have                                               of the applicable hazard group .                                                         change in the      This F&O is assessed as Closed.
already occurred prior to isolation.                                       The calculated probabilities of                                                          capability of the Also, the generic MOV FTC value                                            each of the lines shown above                                                            PRA.
(from NUREG/CR-6928) in Data                                               for an individual cutset is less Table 4 is 1.07E-03/demand.                                                than 1% of hazard group Finally, the CCF probability used                                          contribution. Summing the should be changed to the HPCI                                              contribution for RWCU, HPCI MOV FTC, with Alpha = 1.41E-02.                                            and RCIC (~1.5E-8) the sum is less than 95% of the hazard group. Therefore, each of these is considered to be an insignificant contributor to CDF and LERF, so the HPCI, RCIC, and RWCU BOC initiators are not included in the BFN model.
In addition, if the HPCI line fails then RCIC would provide a backup source of inventory into the reactor along with LPCI once the operators depressurize. If the RCIC line fails, then HPCI would provide a backup source of inventory into the reactor along with LPCI once the operators depressurize. If RWCU fails, then HPCI and RCIC can provide inventory into the reactor.
See BFN-0-16-031.
IE IE-C8 IE-C8 is Not Met 6-8         RCW initiating event appears to be   Loss of RCW initiating event          This was a legacy item that was       IE.01 = NDN-000-999-2007-0030, Rev 2, Maintenance  This issue was     Review of the BFN recovery rule file confirms that the invalid           IE-C8 remains Not incorrectly reduced by factor        appears to be reduced by a factor     never removed but should have         Initiating Events Notebook Recovery                addressed by       correction factor has been removed.                                      Met (see F&O 6-10).
RCWMTCF for combinations where       of 1E-02 from the actual.             been. This has been confirmed         Rule file = BFN123_MREC_r8.RECV                    removing the reduction factor does not                                              to have been removed from the                                                            incorrect factor   This F&O is assessed as Closed appear to be valid. In particular,                                         latest MOR.                                                                              and updating the event is applied to cutsets                                                                                                                                     documentation to containing common transformer                                                                                                                                       reflect this. No events. Also, reduction factor                                                                                                                                      new methodology appears to be calculated                                                                                                                                            was involved in incorrectly (1/365)**2.                                                                                                                                             making the change.
(1) No new methods were used. The updated used existing methods and tools.
(2) There was no change in scope of the PRA. The Page 243 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                            December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                              Self-Assessment Reference        Maint (M) or Basis for Maint                                                                               Independent Review RU SR                         Finding No. Description                        Prior Peer Review Assessment         Self-Assessment Closure Basis                                                                             Independent Review Team Disposition Assessment                                                                                                                                   Document(s)                      Upgrade?     (M) or Upgrade                                                                                Team SR Assessment percent contribution of the loss of Raw Water initiator to all initiators is very low (0.2% for MOR Rev0 and currently below truncation).
(3) There was no change in the capability of the PRA. This accident sequence and risk insights are not significantly changed.
LE LE-F2  SR LE-F2 is Not Met 1-33        There is no discussion of the      A review of the reasonableness of    The review of the CDF and LERF       LE.01 - LERF Main Report, Rev 5; Open         none                   The current documentation provides a listing of addressed phenomena    SR LE-F2 was review of the LERF contributors    the results of the analysis of the   cutsets was performed and            NDN-000-999-2007-0041;                                               and failures postulated to lead to LERF in Table A.1-2. How the BFN    previously Not Met (ASME/ANS RA-Sa2009 Table 2-       contributors to LERF is required per documented in Attachment D           NDN-000-000-2010-0001 Rev, 008                                       model maps to these postulated events is provided in Table 11. The     and remains Not 2.8-9) for reasonableness per the  the SR.                              and E of the Quantification                                                                               model mapping is again provided in the QU notebook in Table 6.3-11. Met.
review of the QU Notebook and                                           Notebook. Section 6.3.2.3 of the                                                                          The frequency results are tabular in the QU notebook and there is a LE.01.                                                                  Quantification Notebook                                                                                   comparison of absolute frequency to similar designs. However, there is specifies the types of things that                                                                        no documented review of the results to determine if the LERF results were looked at when reviewing                                                                             are reasonable and that the identified contributors (categories) are the cutsets. The Top 100                                                                                  consistent with expectations. A pointer to the summary document was cutsets, a sample of 100 cutsets                                                                          provided but the requested information was not found at that from the middle and the last                                                                              location.
100 cutsets were all reviewed and showed no signs of                                                                                    This F&O is assessed as Open.
inconsistencies in logic.
LE LE-D1  LE-D1 is met        2-35        The containment structural         All three unit containments must     The Containment Ultimate             LE.01 - LERF Main Report, Rev 5; Maintenance  No change              Supporting documentation for the LERF report identifies in Section     LE-D1 was previously analysis does not address the Unit be addressed                         Capacity that is currently           NDN-000-999-2007-0038                         required since the     6.1.2 Primary and Secondary Containment Building Walk-Through that     met at Cat II and 3 primary containment ultimate                                          addressed in the LERF Notebook                                                     documentation          "Based on a review of these photographs it is apparent that the BFN    remains Met at Cat II.
capacity in section 6.3.                                                is applicable for all 3 units.                                                     can be obtained in     Unit 1 Primary Containment is sufficiently similar to BFN Units 2 & 3 There were no identified                                                           the current            that a physical walkdown is not warranted." The report also states in differences between the three                                                      references. The        section 6.3 that "The Browns Ferry Unit 1 containment structural is units with respect to                                                              change is              identical in construction to the Unit 2 containment." Therefore, the containment parameters, so                                                         documentation          existing information seems appropriate and documented.
none were specified in the LERF                                                    only. (1) No new Notebook.                                                                          methods were           This F&O is assessed as Closed.
used.
(2) There was no change in scope of the PRA.
(3) There was no change in the capability of the PRA.
LE LE-C11 LE-C11 meets Cat I  4-48        No credit is taken for equipment   LE-C11 implies credit be taken for   The LERF Notebook Attachment         LE.01 - LERF Main Report, Rev 5; Maintenance  The resolution         The LERF model discusses SA/EQ in section 3.1.3 Equipment              LE-C11 was LE-C12 LE-C12 meets Cat I              survivability or human actions     equipment survivability following    A discusses the                      NDN-000-075-2007-0010, Rev, 005               involved               Survivability in a Severe Accident Environment. The existing model     previously met at Cat following containment failure. containment failure, for Cat II/III. phenomenological conditions                                                        correction of          does incorporate in the system modeling response options designed to   I and is now met at that go into CZ1. Figure A.6-2                                                     existing               improve performance. These are documented in the system modeled        Cat I/II, and Table A.6-2 show the fault                                                     information and        development. For example, the CS system can be aligned to the CST to   LE-C12 was was                    allow continued injection (NDN-000-075-2007-0010 Rev: 005, 3.2.3       previously met at Cat documentation          Alternate System Alignments).
Page 244 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                            December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                               Self-Assessment Reference        Maint (M) or Basis for Maint                                                                             Independent Review RU SR                        Finding No. Description                          Prior Peer Review Assessment          Self-Assessment Closure Basis                                                                        Independent Review Team Disposition Assessment                                                                                                                                    Document(s)                      Upgrade?     (M) or Upgrade                                                                              Team SR Assessment tree modeling and split fraction                                                  only.                                                                                       I and is now met at designations of CZ1.                                                              (1) No new         This F&O is assessed as Closed.                                          Cat II methods were used.
(2) There was no change in scope of the PRA.
(3) There was no change in the capability of the PRA.
LE LE-C10 LE-C10 meets Cat I 4-50        Although equipment survivability     LE-C10 Cat II/III requirements are    The significant accident            LE.01 - LERF Main Report, Rev 5; Maintenance  The resolution     The LERF model discusses SA/EQ in section 3.1.3 Equipment                LE-C10 was beyond equipment qualification       to REVIEW significant sequences to    sequences were reviewed as          NDN-000-075-2007-0010, Rev: 005               involved           Survivability in a Severe Accident Environment. The existing model       previously met at Cat limits is credited, there is no      determine if engineering analyses     described in the QU Notebook                                                      correction of      does incorporate in the system modeling response options designed to     I and is now met at indication that significant accident can be used to take credit for        Section 6.3.2.3. Equipment                                                        existing           improve performance. These are documented in the system modeled          Cat II.
progression sequences were           additional equipment operation        survivability was looked at and                                                   information and is development. For example, the CS system can be aligned to the CST to reviewed to determine if             beyond normal qualification limits    used where achievable as the                                                      documentation      allow continued injection (NDN-000-075-2007-0010 Rev: 005, 3.2.3 continued equipment operation        to reduce LERF.                       discussion in the ISLOCA                                                          only.              Alternate System Alignments).
could be credited to                                                       Notebook Section 6.3.4.5                                                          (1) No new REDUCE LERF.                                                               attests. The two cutsets that                                                     methods were       This F&O is assessed as Closed.
were brought up were not                                                          used.
refined for equipment reliability                                                 (2) There was no as they would both involve                                                        change in scope of addition of HRA events while we                                                   the PRA.
are already at the floor value                                                    (3) There was no with respect to HRA. Addition of                                                  change in the the HRA events would increase                                                     capability of the the number of combination                                                         PRA.
events which increases the number of HRA recovery rules, and that would in turn increase the time to quantify. There would not be much gain (if any) of LERF either, as the dependency analysis would limit the amount of credit the human action would give.
LE LE-C1  LE-C1 meets Cat-I  4-51        Class3A (B,C)-006 LERF sequences     Class3A(B,C)-006 LERF sequences       The LERF Notebook Attachment        LE.01 - LERF Main Report, Rev 5; Maintenance  The resolution     The model was reviewed to the original F&O related to Top TD and the     LE-C1 remains Met at LE-C8  LE-C8 is Met.                  are non-sensical. In these           are non-sensical. In these            A Section A.7.6.1 under the         CET1_r8.eta;                                  was to explain the resolution put forth by TVA was found to be accurate and the F&O is      Cat-II.
sequences, TD2 succeeds (i.e., DW    sequences, TD2 succeeds (i.e., DW     heading Nodes Affected by           BFN IE PRA Model. Rev 8                      modeling using     closed. A second F&O from the 2015 review as also examined related       LE-C8 remains Met.
Spray hardware is available and      Spray hardware is available and       Success/Failure of (FD/FC)                                                       the existing       to containment flooding related to the ability to have a flooded drywell operator initiates injection         operator initiates injection per      states: The upward path will be                                                   information. The   and still have an impaired containment (Top RME). CET top RME per Table A.5.7-1) but DWS fails     Table A.5.7-1) but DWS fails later in used to represent flooding of                                                     change is          addresses the effectiveness of the reactor building. For the case in later in the CET (DWS_ALL_SUP        the CET (DWS_ALL_SUP branch is        containment resulting in a                                                        documentation      question the flooding in the drywell is assumed to lift dampers and branch is questioned).               questioned).                          release through the drywell                                                       only.              result in bypass of the drywell leading to an early release path. The vent. This will not be a                                                          (1) No new         assessment is conservative, and the model is consistent. No modeling contributor to LERF if coolant                                                    methods were       changes are necessary.
injection is available to the                                                     used.
debris, i.e., TD/TR = S. This                                                     (2) There was a    This F&O is assessed as Closed.
question is treated in the                                                        minor change in RME/RBE node. From the                                                            scope of the PRA RME/RBE node section Table                                                        by the addition of A.10-3, RME6 results in LERF                                                      the events.
due to the DW vent and the                                                        (3) There was no operator action failure to                                                        change in the initiate DWS, while RME4 does                                                     capability of the not contribute to LERF because                                                    PRA.
Page 245 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                            December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                            Self-Assessment Reference           Maint (M) or Basis for Maint                                                                                 Independent Review RU SR                     Finding No. Description                           Prior Peer Review Assessment        Self-Assessment Closure Basis                                                                             Independent Review Team Disposition Assessment                                                                                                                                 Document(s)                         Upgrade?     (M) or Upgrade                                                                                  Team SR Assessment of the delayed containment failure.
LE LE-G2 LE-G2 is Met. 4-54        The method used to quantify split     Split fraction values could not be  As mentioned in the LERF            LE.01 - LERF Main Report, Rev 5;    Maintenance  The update added     A review of the documentation identified simplified fault trees that       LE-G2 remains Met.
fractions was very difficult to       determined by the reviewer, and     Notebook Attachment A Section       Browns Ferry Nuclear -8/25/2009- RG              fault tree models    represent the logic for the CET tops. Tables in Appendix A provide the review and appears to be based on     descriptions for many split         A.6.4, the detailed                 1.200 PRA Peer Review;                           that are now a       status for success and failure, failure probability and basis. The current an old LERF model that is not         fractions do not appear to be valid phenomenology fault tree was        P0132150002-5175                                 part of the base     CAFTA model also has the fault trees for the tops. The CET endstates consistent with the current Level 1   anymore.                            developed and quantified using                                                       CAFTA model and      are quantified through the combination of sequence events (example, model. The split fraction fault trees                                     RISKMAN and has not changed                                                          represented a        U1_CET1_003P). This allows for CD insights to be propagated were not provided. Further, many                                          for the current PRA. The nodes                                                       model change.        appropriately to the CET.
of the split fraction descriptions                                        are calculated using the old                                                         Since the same provided in Appendix A of LE.01 do                                        RISKMAN Program and are                                                              data is utilized and This F&O is assessed as Closed.
not appear to be current or are no                                        shown in Table A.6-4. The                                                            the model is only longer used in the LERF model.                                            phenomenological basic events                                                        used to represent do not need to have any                                                              the same logic importances assessed as they                                                         there is no change involve no equipment failures                                                        in method. Use of and the split fractions are                                                          the same model generated using current                                                              data and same industry practices.                                                                  model structure should result in the same results such that there is no chance in capacity. Overall the change should not significantly alter results.
(1) Putting the logic in a small fault tree is a minor change in methods.
(2) No new events were added but the small CAFTA models were added and represent a minor scope change for the PRA.
(3) There was no change in the capability of the PRA.
LE LE-B1 LE-B1 is Not Met 7-6         Section 7.1 of LE.01 directly         The SR requires the consideration   The LERF Notebook Section 7.1       LE.01 - LERF Main Report, Rev 5     Maintenance  The resolution is a  The LERF contributors are identified in the text defined in Section 7.1. LE-B1 was previously addresses those contributors from     of unique plant issues.             was revised to indicate that                                                         change to                                                                                       Not Met and is now the table, but plant specific issues                                      MSBOC and FWBOC are both                                                             documentation.       This F&O is assessed as Closed.                                            Met at Cat II.
do not appear to be addressed.                                            LERF contributors. The common                                                        (1) No new cause failure of the battery was                                                     methods were also included as a plant specific                                                    used.
contributor to LERF.                                                                 (2) There was no change in scope of the PRA.
(3) There was no change in the capability of the PRA.
Page 246 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                            December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                             Self-Assessment Reference        Maint (M) or Basis for Maint                                                                            Independent Review RU SR                      Finding No. Description                        Prior Peer Review Assessment          Self-Assessment Closure Basis                                                                         Independent Review Team Disposition Assessment                                                                                                                                  Document(s)                      Upgrade?     (M) or Upgrade                                                                             Team SR Assessment LE LE-B1 LE-B1 is Not Met  7-7         The definition of Early appears to Definition of the timing of accident  The referenced EALs do specify       LE.01 - LERF Main Report, Rev 6; Maintenance  This resolution is The documentation was reviewed to determine the basis for LERF          LE-B1 was previously be inconsistent and may eliminate  sequences determines whether a        a GE at containment pressure of      NDN-000-999-2008-0006, Rev 006                limited to         timing. The development in the SC notebook is well documented and       Not Met and is now some scenarios from consideration  sequence can contribute to LERF.      55 psig, however the referenced                                                    enhancing          identifies a period of 6 hours for the breakpoint between LERF and      met at Cat II for LERF.                          Timing based from accident            EALs also specify a GE at a                                                        documentation to   non-LERF cases. Representative values have been between 4 and 8 initiation will be different than     reactor water level not being                                                      provide a more     hours so this is considered reasonable for the BFN site and population timing from declaration of General    maintained (i.e., low). During a                                                   detailed           density. The time is based on the evacuation of the population such Emergency.                            loss of decay heat removal, the                                                    discussion of the  that any releases would not involve exposed population. Therefore the water level would decrease and                                                     quantification     6 hours is a fixed time that is tied to the initiation of a general HCTL limits would be exceeded                                                      results.           emergency (GE) such that evacuation is initiated. The timing for the GE which would signify impending                                                      (1) No new         is based on plant-specific MAAP assessments and is again, an loss of fission product barriers                                                   methods were       appropriate metric.
which would lead to the                                                            used.
declaration of the General                                                         (2) There was no   This F&O is assessed as Closed.
Emergency. It is TVAs belief                                                      change in PRA that the LERF Notebook                                                             scope.
adequately states the definition                                                   (3) There was no of LERF based on the above                                                         change in PRA discussion.                                                                        capability.
QU QU-D2 QU-D2 is not Met. 1-34        Additional attention should be     The top accident sequence cutset      The review of the CDF and LERF       NDN-000-999-2007-0041, Rev 8     Maintenance  This resolution is The quantification results are reviewed. The review identifies the most QU-D2 is now Met.
applied to significant cutsets to  for both CDF and LERF deals with      cutsets was performed and                                                          limited to         significant contributors. Section 6.3.2.3 includes an assessment of determine that the bases for the   clogging of the intake and includes   documented in Attachment D                                                         enhancing          modeling consistency.
cutsets are consistent with        events that are very uncertain. The   and E of the Quantification                                                        documentation to modeling and operating             attention given this cutset to        Notebook. Section 6.3.2.3 of the                                                   provide a more     QU-D2 is now Met. This F&O is assessed as Closed.
philosophies.                      minimize the uncertainty              Quantification Notebook                                                            detailed associated with the contributing      specifies the types of things that                                                 discussion of the basic events has not been             were looked at when reviewing                                                      quantification sufficient. The approach to dealing   the cutsets. The Top 100                                                           results.
with such important cutsets should    cutsets, a sample of 100 cutsets                                                   (1) No new assure that the contributors are      from the middle and the last                                                       methods were understood and are supported by       100 cutsets were all reviewed                                                      used.
appropriate rigorous analyses         and showed no signs of                                                             (2) There was no and/or assessment.                    inconsistencies in logic. In                                                       change in PRA addition, the top accident                                                         scope.
sequences were also reviewed                                                       (3) There was no as documented in Section                                                           change in PRA 6.3.2.2 of the QU Notebook.                                                        capability.
Each of these were reviewed to determine whether they were appropriate. In regards to the questions asked, the mechanisms by which 480V AC bus failures become initiating events would be documented in the system notebooks or initiating event notebooks, the basis for the 2 CRD pump requirement is in the Success Criteria Notebook based on the MAAP runs in the Thermal Hydraulics calculation, the conservatisms related to modeling transients with stuck open MSIVs directly as LERF events would be in the Accident Sequence Notebook.
Page 247 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                       December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                              Self-Assessment Reference     Maint (M) or Basis for Maint                                                                         Independent Review RU SR                      Finding No. Description                            Prior Peer Review Assessment        Self-Assessment Closure Basis                                                                     Independent Review Team Disposition Assessment                                                                                                                                   Document(s)                   Upgrade?     (M) or Upgrade                                                                          Team SR Assessment QU HR-I3 HR-I3 is not Met. 3-10        Modeling uncertainty comes from        Sources of uncertainty must be      EPRI Report 1016737 gives          NDN-000-999-2007-0041, Rev 8,  Maintenance  The is primarily a Attachment H to the Quantification Notebook includes a discussion of HR-I3 is now Met.
IE-D3 IE-D3 is not Met.             two general types of issues, plant     identified and documented.          guidance on how to perform an      BFN-0-16-031                                documentation      assumptions and modeling uncertainties and includes a table of the   IE-D3 is now Met.
LE-F3 LE-F3 is not Met.             specific and generic. Plant specific                                       uncertainty analysis. The report                                               update to provide  various items identified in the BFN PRA. The table includes a        LE-F3 is now Met.
SY-C3 SY-C3 is not Met.             uncertainties and assumptions                                              goes into parameter                                                            enhanced           description and basis to characterization the items and potential    SY-C3 is now Met.
SC-C3 SC-C3 is not Met.             should be identified and                                                   uncertainty, model uncertainty                                                 discussion of      impacts to the results.                                              SC-C3 is now Met.
QU-E2 QU-E1 is not Met.             documented during the model                                                and completion uncertainty.                                                    uncertainties.                                                                          QU-E1 is now Met.
QU-E4 QU-E2 is not Met.             development. The generic sources                                           BFN takes into account                                                         (1) No new         This F&O is assessed as Closed.                                      QU-E2 is now Met.
QU-F4 QU-E4 is not Met.             of uncertainty are listed in EPRI                                          parameter uncertainty by                                                       methods were                                                                            QU-E4 is now Met.
QU-E1 QU-F4 is not Met.             Report 1016737 Table A-1. Both                                             having an uncertainty value tied                                               used.                                                                                   QU-F4 is now Met.
DA-E3 DA-E3 is not Met.             types of uncertainties must be                                             to each basic event and                                                        (2) There was no                                                                        DA-E3 is now Met.
addressed for the base model.                                              initiating event. For those basic                                              change in scope of Examples of plant specific                                                 events that were updated using                                                 the PRA.
uncertainties include:                                                     plant specific data, WinBUGs                                                   (3) There was no (1) ISLOCA valve failing to close                                          was run to see how well the                                                    change in the after testing is not listed in the                                         data fit the Poisson Distribution                                              capability of the sources of uncertainty, nor is the                                         and this is documented in the                                                  PRA.
conditional probability that the                                           Data Notebook. A probability break is greater than 93 or 600                                            density function was also gpm.                                                                       created for all of the basic (2) For Initiating Events, the factors                                     events and initiating events that affecting INTAKE initiating event is                                       have been updated with plant not included in the assumptions                                            specific data and this was used section, nor are any of the other                                          to see how well the posterior assumptions in the analysis.                                               data compares to the prior data.
(3) Specific assumptions for the                                           BFN describes the model detailed HFEs is not discussed,                                            uncertainties in Attachment H including assumptions made for                                             of the Quantification Notebook.
timing of operator responses                                               All the model uncertainties that (versus analyzed or those observed                                         were identified are within this on a simulator)                                                            attachment with a reference as to what notebook the uncertainty is taken from. Any uncertainty related to completeness is specified in the specific notebook in which the completeness is not taken into account (i.e., the system notebooks has the excluded components section that list why components are not taken into account and any assumptions made from that are listed in the assumptions section).
See BFN-0-16-031.
QU QU-F2 QU-F1 is not Met. 3-28        A detailed discussion of the           This is an important part of the    The QU Notebook Section 4.3         NDN-000-999-2007-0041, Rev 9, Maintenance  This change is     Discussion of modeling asymmetries and the impact on results         QU-F1 is now Met.
QU-F1 QU-F2 is not Met.             quantification asymmetries (with       quantification documentation        was expanded to include             SY.08 - HVAC Rev 5                         limited to         between the units is provided in Section 4.3 of the Quantification   QU-F2 is now Met.
respect to different units, system     process.                            specific differences that impact                                               documentation      Notebook.
alignments, etc.) is not presented.                                        the results between units. The                                                 updates.
numbering was also revised in                                                  (1) No new         This F&O is assessed as Closed.
the QU Notebook.                                                               methods were used.
(2) There was no change in scope of the PRA.
(3) The was no Page 248 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                        December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                               Self-Assessment Reference     Maint (M) or Basis for Maint                                                                            Independent Review RU SR                        Finding No. Description                          Prior Peer Review Assessment        Self-Assessment Closure Basis                                                                       Independent Review Team Disposition Assessment                                                                                                                                    Document(s)                   Upgrade?     (M) or Upgrade                                                                             Team SR Assessment change in the PRA capability.
QU QU-F3 LE-G6 is not Met. 3-31        The definitions for significant when This issue causes the supporting    The QU Notebook Section              NDN-000-999-2007-0041, Rev 8  Maintenance  The change is       The Quantification Notebook documents the definition of significant    LE-G6 is now Met.
LE-G6 QU-F6 is not Met.               presenting lists of important        requirement QU-F6 not to be met. 6.3.2.2 was revised to indicate                                                 documentation       that is used consistent with the ASME Standard.                        QU-F6 is now Met.
QU-D6 QU-D6 is Met                    equipment, operator actions, etc.                                        significant accident sequences                                                  only. The                                                                                  QU-D6 was QU-F6 QU-F3 is Met                    do not always conform to the strict                                      were those that contributed at                                                  definition of       This F&O is assessed as Closed.                                        previously Met at CC-ASME standard definition of                                              least 1% to CDF or at least 1% to                                               significant                                                                                II/III and continues to significant. Justifications for the                                      LERF. The definition of                                                         accident                                                                                   be Met at CC-II/III.
alternatives used are not                                                significant accident sequences                                                  sequences used is                                                                          QU-F3 was previously presented.                                                               now appropriately reflects that                                                 documented and                                                                             Met at CC-II/III and of the ASME standard.                                                           the results are                                                                            continues to be Met presented                                                                                  at CC-II/III.
consistently.
(1) No new methods were used.
(2) There was no change in scope of the PRA.
(3) There was no change in the capability of the PRA.
QU QU-C2 QU-C2 is Not Met. ( 4-21        The joint HEP for several combined   If the joint HEP for combined       Basis for JHEP floor value (The      NDN-000-999-2007-0041, Rev 8, Maintenance  The change          A floor value has been implemented to limit HEP credit in the BFN PRA  QU-C2 remains Not HR-I3 HR-I3 is Not Met.               operator actions are too low and     events is too low, sequence and     floor value applied in the           NDN-000-999-2007-0032, Rev 5,              involves adding a   as described in Section 4.1 of the HRA Notebook and Section 6.3.3 of   Met. (see F&O 6-30)
HR-G7 HR-G7 is Not Met.               cannot be justified. Specifically,   overall results may be artificially dependency analysis lacks a          U1_CDF-1E-12.CUT                           floor value for HEP the Quantification Notebook Sensitivity Case 3. The basis of the floor HR-I3 is now Met.
three combined actions have joint    lowered, and the importance of      justification for divergence from                                               credit.             values is provided and the values are consistent with other plants. HR-G7 remains Not HEPs of less than 1E-7, and eight    the operator actions may be         industry standards and it has a                                                 (1) No new                                                                                 Met. (see F&O 6-30) are less than 1E-6. Note that the    understated.                        significant impact on BFN                                                       methods used.       This F&O is assessed as Closed.
HRA acknowledges these low                                               results. In addition, the                                                       The method uses combined HEPs but does not                                               automated HRAC dependency                                                       fault tree enforce any lower bound. Further,                                        process does not account for                                                    modeling and it states that a sensitivity will be                                     intervening successes in the                                                    recovery rules. It performed in the Quantification                                          accident sequences, which is an                                                 is consistent with Notebook, but none is performed.                                         element of this SR).                                                            current practice
* The HRA floor value                                                           and uses the recommended by is 1E-5.                                                         existing tools.
However, this arbitrary value                                                   (2) The scope tends to skew the PRA results.                                                  includes all HFE The HRA industry group has                                                      combinations with been working on developing                                                      HEPs less that 1e-guidance for the minimum Joint                                                  5 or 1e-6. The PRA Human Error Probability to be                                                   results are used in PRA, but this guidance is                                               impacted as not available at this time.                                                     demonstrated However, as recommended by                                                      with a sensitivity EPRI HRAUG, sensitivity #3 is                                                   study showing an included in the Quantification                                                  expected increase Notebook to determine the                                                       in CDF. Risk impact of the selected floor                                                    insights are not value.                                                                          significantly impacted.
(3) There is no change to the PRA capability.
Page 249 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                         December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                              Self-Assessment Reference     Maint (M) or Basis for Maint                                                                               Independent Review RU SR                      Finding No. Description                          Prior Peer Review Assessment         Self-Assessment Closure Basis                                                                        Independent Review Team Disposition Assessment                                                                                                                                   Document(s)                   Upgrade?     (M) or Upgrade                                                                                Team SR Assessment QU QU-D2 QU-D2 is Not Met. 4-36        The assumption that A HVAC is        The assumption that one train is     The run and standby flags have       NDN-000-999-2007-0041, Rev 8, Maintenance  The resolution       Based on inspection of the PRA fault tree and flag file, the normal/stby QU-D2 is now Met.
QU-F2 QU-D7 is Not Met.             normally running and B HVAC is in    always normally running (the HVAC    been reviewed and the model is       SY.08 - HVAC, Rev 5,                       involves a           flags for HVAC have been replaced with appropriate split fractions.      QU-D7 is now Met.
QU-D7 QU-F2 is Met.                 standby leads to skewed basic        is only an example) does not reflect reflective of the normal             BFN IE PRA Model Rev 8                     probability value    (Note, this is contrary to the stated resolution, which indicates the    QU-F2 was previously event importance's and non-          the plant operation and can result   operating configuration. A                                                      update to the PRA    change was not made. Also, the related discussion in the SY Notebook     Met and continues to sensical cutsets.                    in skewed importance results or      sensitivity was run on the PCS                                                  model.               is incorrect-see F&O 3-28)                                               be Met.
For example, with A HVAC always      missing cutsets/sequences (i.e.,     pump configuration which is                                                     (1) No new running:                             how would the results be different   documented in the QU                                                            methods used.        This F&O is assessed as Closed.
(1) The Loss of RMOV Board A         if the other train were assumed to   Notebook. This sensitivity                                                      The change importance is much higher than       be running?).                        showed no change to CDF/LERF                                                    involves fault tree RMOV Board B (10% vs. 2.5%)                                               for any unit.                                                                   and data updates (2) Non-sensical cutsets exist, such                                                                                                                      consistent with as where RMOV Board A is in                                                                                                                               current practice maintenance and B HVAC fails to                                                                                                                           and uses the start (due to operator or hardware                                                                                                                        existing tools.
failure).                                                                                                                                                 (2) The scope includes multiple equipment alignment flags/logic. The PRA results now distribute the risk appropriately between trains.
Risk insights are not significantly impacted.
(3) There is no change to the PRA capability.
QU HR-H3 QU-D5 is not Met. 4-40        A review of non-significant cutsets  This is an example of non-           Revision 7 of the Quantification     NDN-000-999-2007-0041, Rev 8, Maintenance  The changes for      The HFE dependency analysis was re-performed using the HRA               QU-D5 is now Met.
QU-D5 HR-H3 is not Met.             found many LOOP cutsets that         significant cutsets that, had they   Notebook Section 6.3.2.10 now        NDN-000-999-2007-0032, Rev 5,              this F&O are         Calculator. The non-significant cutsets are reviewed as discussed in     HR-H3 was previously have combinations of two             been reviewed, would have            states:                              U1_CDF-1E-12.CUT                           primarily            Section 6.3.2.10 of the Quantification Notebook. The cutsets were        Not Met and is now independent HFEs which should        uncovered the need to perform        "All accident sequences in each                                                 documentation.       reviewed and no independent HFEs were identified as described in the     Met.
have some level of dependency:       additional operator dependency       event tree model were                                                           The cutset reviews   F&O.
HFA_02114KVCRSTIE (Failure to        analyses.                            quantified and reviewed early in                                                are performed cross-tie 4kV SD Board) AND                                               the quantification process to                                                   and documented.      This F&O is assessed as Closed.
HFA_0231480SDBTIE (Failure to                                             check, debug, and finalize the                                                  The actual HRA provide alternate power to 480V                                           model. This was found to be                                                     dependency SD Board).                                                                helpful in identifying modeling                                                 analysis is enhancements and to ensure                                                      addressed by F&O that the event tree modeling                                                    6-30.
logic is correct. The bottom 100                                                (1) No new CDF and LERF cutsets for each                                                   methods were unit are provided in Attachment                                                 used.
D and Attachment E,                                                             (2) There was no respectively. These cutsets were                                                change in scope of reviewed in the same manner as                                                  the PRA.
the significant cutsets were                                                    (3) There was no reviewed, and there were no                                                     change in the identified inconsistencies in the                                               capability of the logic."                                                                         PRA.
See BFN-0-16-031.
QU QU-D3 QU-D3 is Met      4-41        Offsite power recovery is applied    Recoveries should only be applied    The offsite power recoveries are      BFN IE PRA Model Rev 8,      Maintenance  Resolution of this   Discussions with BFN staff demonstrated that offsite power could still   QU-D3 was in cutsets where it might not be     to scenarios or cutsets where the    applied to cutsets that involve a    U1_CDF-1E-12.CUT,                          F&O was based on     be applied for the cutsets described in the F&O. The breaker failures    previously Met and possible. See U1 CDF cutset at                                            loss of offsite power event or a                                                justification of the block use of the DGs but still allow off-site power to be restored.      Continues to be Met.
1.151E-08: LOOP with common                                               loss of a diesel generator.                                                     existing PRA Page 250 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                               December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                               Self-Assessment Reference         Maint (M) or Basis for Maint                                                                          Independent Review RU SR                      Finding No. Description                         Prior Peer Review Assessment         Self-Assessment Closure Basis                                                                            Independent Review Team Disposition Assessment                                                                                                                                    Document(s)                       Upgrade?     (M) or Upgrade                                                                           Team SR Assessment cause failure of shutdown board     recovery can be expected to be       Therefore, recovering power to         BFN-0-15E500-1-CC_004337876, Rev               model.
normal feeder breakers to open. successful                           the shutdown board would still         44                                             (1) No new         This F&O is assessed as Closed.
be a viable pathway as at least                                                       methods were one shutdown board would still                                                        used.
be available. In the LOOP                                                             (2) There was no recovery rules, there are some                                                        change in scope of instances where battery                                                               the PRA.
depletion or HVAC might be lost                                                       (3) There was no along with a LOOP, but this is                                                        change in the still a recoverable event as the                                                      capability of the operator would still have at                                                          PRA.
least one shutdown board available to recover power to.
In addition, a review of the CDF and LERF cutsets was performed and documented in Attachment D and E of the Quantification Notebook. The Top 100 cutsets, a sample of 100 cutsets from the middle and the last 100 cutsets were all reviewed and there were no identified instances where recoveries were applied to non-recoverable failures.
SC SC-B1 SC-B1 is Met      1-12        Several examples found for lack of  The SR expects that engineering      The keep-fill system is not            NDN-000-999-2007 (SC.01);         Maintenance  The change is a    Most of the success criteria are based on realistic (MAAP) analysis. SC-B1 remains met at SY-B6 SY-B6 is not Met.             engineering analyses regarding      analyses will be performed to        modeled as a CS support                SY.01 Condensate System Notebook;              documentation      Some of the supporting analysis for systems are based on generic or   Cat II.
SY-B7 SY-B7 is Met                  HVAC that could be justified by     determine whether these              system. Failures of this system        SY Core Spray;                            update only.       conservative assumptions for room heatup; however, these              SY-B6 is now Met.
calcs.                              statements are correct.              would be detected and                  SY-07 HPCI                                     (1) No new         assumptions do not affect the determination of which systems are      SY-B7 remains met at Condensate System Notebook                                               corrected during normal                                                               methods were       required to respond to imitating events. Additional documentation was Cat II.
(SY.01) assumes active ventilation                                       operation. Daily instrument                                                           used.              included in the system notebooks providing the basis for the room is not required due to plant                                             checks are performed through                                                          (2) There was no   heatup analysis. Where available, calculations were used for the room experience Core Spray System                                             procedure x-SR-2, attachment 2.                                                       change in scope of heatup.
Notebook (SY.04) assumes keepfill                                        (where x is the unit)                                                                 the PRA.
system is not required. HPCI                                                                                                                                   (3) There was no   This F&O is assessed as Closed.
System Notebook (SY.07) assumes                                                                                                                                change in the dependence on quad cooling for                                                                                                                                 capability of the the remaining 20 hours of post-                                                                                                                                PRA.
accident operation.
SC SC-A5 SC-A5 is not met  3-12        There is no evidence of an analysis A CC II/III for SC-A5 requires that  Basis for Safe and Stable for        SC.1 (NDN-000-999-2007-0035 R3)   Open         none               Additional discussion of the bases for "safe and stable" has been     SC-A5 remains met at for sequences that go beyond the    options other than assuming          HFA_0085ALIGNCST - During a            Success Criteria                                                  added. However, there is no discussion whether any sequences were     Cat I.
24-hour period to evaluate the      sequences in which a stable state    single unit accident, refill of the                                                                      identified that require a mission time beyond 24 hours to reach safe appropriate treatment relative to   has not been reached in 24 hours     CST inventory is credited in the                                                                         and stable. Note that Table 6-1 of SC.1 contains several statements the CC II/III requirements for SC-  goes to core damage.                 model                                                                                                    implies that sequences may not safe be and stable at 24 hours and a A5.                                                                      (HFA_0085ALIGNCST) by                                                                                    bounding PDS may be assigned. This instruction in Table 6-1 is refilling from the non-accident                                                                          consistent with SC-A5 Cat I.
units CST. During a multi-unit accident, it is assumed that the                                                                         This F&O is assessed as Open.
TSC would direct the operators to provide additional inventory to the CSTs from an outside source given the CST depletion would not occur for 10 hours.
This assumption is not documented in the current model.
* It is already considered in the Page 251 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                           December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                              Self-Assessment Reference       Maint (M) or Basis for Maint                                                                              Independent Review RU SR                     Finding No. Description                            Prior Peer Review Assessment         Self-Assessment Closure Basis                                                                        Independent Review Team Disposition Assessment                                                                                                                                   Document(s)                     Upgrade?     (M) or Upgrade                                                                               Team SR Assessment cognitive analysis for HFA_0085ALIGNCST and the assumption that the TSC would direct operators to provide additional inventory to the CSTs is documented in the HRA Notebook. The alarm response procedures 1(2,3)ARP-9-6B provides a list of alternative sources including: 1) Hotwell or Radwaste transfer to CST, 2)
Demin or another CST transfer to the affected CST, and 3) CST Crosstie. The TSC and OSC would determine and perform the appropriate actions based on conditions at the plant and the choices identified in ARP.
SY SY-A3  SY-A3 is Met. 2-23        In section 3.2.6.1 of the HVAC         A breaker failing to provide tripped Failure of the operating unit to     BFN IE PRA Model. Rev 8         Maintenance  No change is       The standby ACU will not be required until the normal operating unit      SY-A3 remains Met.
system Notebook, it states that the    indication for a start permissive    trip has been added to the                                                        required to close  fails. This will, in itself, meet the trip condition based on discussions running ACU for unit 3 electric        can happen and this failure mode     model as a failure mode of the                                                    this F&O.          with knowledgeable TVA staff. Therefore, no change is required to boards must be tripped before the      should be included.                  standby unit.                                                                     (1) No new         address this F&O. The model was examined (gate U1_U1_HVAC_EBR standby unit can be started. Failure                                                                                                                          methods were       and U1_U1_HVEBR_G005) and found be appropriate.
of this trip to occur is not reflected                                                                                                                        used.
in the fault tree.                                                                                                                                            (2) There was no   This F&O is assessed as Closed.
change in scope of the PRA.
(3) There was no change in the capability of the PRA.
SY SY-A5  SY-A5 is Met. 2-31        For SPC and LPCI, the LPCI injection   All active components should be      The injection valves do need to      NDN-000-074-2007-0025, Rev 005; Open         none               The model includes other valve realignments and common cause. It is       SY-A5 remains Met.
SY-A13 SY-A13 is Met.              valves and SPC return valves are       included in the failure modes of a   change position for split            BFN IE PRA Model, Rev 8                                         unclear why this specific change would warrant a unique modeling          SY-A13 remains Met.
required to reposition when            system.                              LPCI/SPC flow; two valves would                                                                      approach. The absence of this failure mode could alter the importance swapping RHR modes, but this is                                             have to fail to modulate or close                                                                    calculations for the identified components and impact the ability to not included in the model. The RHR                                          in either path to fail either                                                                        determine MSPI characteristics. It would be expected that these valves system Notebook indicates that                                              system. An operator interview                                                                        would need to be included since it does involve a physical change in these valves need to close for the                                          was conducted to address this                                                                        state.
opposite function. However, in one                                          issue. The common cause location in the notebook it is                                              failure probability of two MOVs                                                                     This F&O is assessed as Open.
indicated that flow can be split                                            to close is less than 1E5.
between LPCI and SPC.                                                       The RHR pump start failure probability is approximately 1.4E3. The failure of two MOVs to close is less than 2 orders of magnitude lower than another failure that would fail the system in a similar manner.
Therefore, failure to close (or modulate) either the LPCI or SPC injection path can be neglected. The RHR System Notebook was modified to reflect this and the operator interview was added.
Page 252 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                           December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                             Self-Assessment Reference      Maint (M) or Basis for Maint                                                                          Independent Review RU SR                     Finding No. Description                          Prior Peer Review Assessment          Self-Assessment Closure Basis                                                                        Independent Review Team Disposition Assessment                                                                                                                                  Document(s)                    Upgrade?     (M) or Upgrade                                                                           Team SR Assessment SY SY-A8  SY-A8 is Met. 4-32        Several electrical system boards     Component boundaries for              The EDG logic to start and load      NDN-000-082-2007-0012, Rev 4;  Maintenance  The resolution      The legacy resolution is not consistent with the current model. The  SY-A8 remains Met.
SY-B9  SY-B9 is Met.               are modeled to receive power         breakers do not include such          (close output breaker) are           NDN-000-999-2007-0007, Rev 5;               requires no         current model does contain sequencer logic events (e.g.,             SY-B9 remains Met.
from multiple sources (e.g., normal  circuitry, based on NUREG/CR-         currently modeled. The               BFN IE PRA Model, Rev 8                     changes to the      SEQFD0SEQ_082__DGA). This closes the F&O.
and alternate buses, and/or EDGs)    6928. Note that local circuitry and   component description for the                                                    model or without considering the need for     protection devices are included.      circuit breaker component in                                                     documentation.      This F&O is assessed as Closed.
undervoltage detection and                                                 Appendix A of NUREG/CR6928                                                      (1) No new operation circuitry for breakers                                           states:                                                                          methods were and EDGs.                                                                  The circuit breaker (CBK) is                                                     used.
defined as the breaker itself and                                                (2) There was a local instrumentation and                                                        minor change in control circuitry. External                                                      scope of the PRA equipment used to monitor                                                        by the addition of under voltage, ground faults,                                                    the events.
differential faults, and other                                                   (3) There was no protection schemes for                                                           change in the individual breakers are                                                          capability of the considered part of the breaker.                                                  PRA."
External equipment used to monitor under voltage is considered part of the breaker.
The modeling of automatic bus transfer in the BFN model contains both the normal supply breaker failure to open (FTO),
and the alternate supply breaker failure to close (FTC).
Since both failure modes are included, and the data from NUREG/CR6928 includes under voltage detection in the breaker boundary, the current modeling methodology is appropriate.
SY SY-A19 SY-A19 is Met. 4-33        The unavailability or failure of a   Unavailability or failure of the      The failure of the bus has been      BFN IE PRA Model, Rev 8;       Maintenance  The model           Model reviewed and events found to be properly located. For example, SY-A19 remains Met.
bus is not considered in the logic   alternate power supply would          included in the BFN PRA model.       NDN-000-999-2007-0007, Rev 005              revisions are not a event TM_3BDAA211003EA was found to be appropriately included in used to provide alternate electrical prevent being able to credit it as an The applicable 4kV shutdown                                                     change in the       the model and documentation.
power supplies to other buses and    alternate source. Although the        board failure has been added to                                                  modeling boards. Example:                     failure probability of a bus is much  gates U1_SDREC_A,                                                                approach. Twelve    This F&O is assessed as Closed.
U1_SDREC_A is used to re-energize    less than the failure probability of  U2_SDREC_A, U3_SDREC_A,                                                          new events were 4kV SD Board A from 4kV SD Board     other equipment that could affect     U1_SDREC_B, U2_SDREC_B,                                                          added to the 3A. However, the unavailability or   the power transfer (e.g., breaker     U3_SDREC_B, U1_SDREC_C,                                                          model with low failure of 4kV SD Board 3A does      demand failure), the unavailability   U2_SDREC_C, U3_SDREC_C,                                                          probability of not fail the function (it should). could be substantial, especially      1_SDREC_D, U2_SDREC_D, and                                                       failure and are during an outage of the other unit. U3_SDREC_D.                                                                      anticipated to have little impact on results.
(1) No new methods were used.
(2) There was no change in scope of the PRA.
(3) There was no change in the capability of the PRA.
Page 253 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                            December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                            Self-Assessment Reference        Maint (M) or Basis for Maint                                                                             Independent Review RU SR                      Finding No. Description                         Prior Peer Review Assessment         Self-Assessment Closure Basis                                                                          Independent Review Team Disposition Assessment                                                                                                                                 Document(s)                      Upgrade?     (M) or Upgrade                                                                              Team SR Assessment SY SY-A8  SY-A8 is Met. 4-42        Table 3 of the data Notebook says   Apparent inconsistency in data and   The EDG output breakers 1818,        NDN-000-082-2007-0012, Rev, 004; Open         none                The system notebook did indicate that the failure of output circuit     SY-A8 remains Met.
that EDG boundaries included the    component boundary definitions.      1822, 1812, 1816, 1838, 1842,        BFN IE PRA Model Rev 8                                            breakers was included within the EDG boundary. However, the CAFTA output breakers, but the EDG                                             1832, and 1836 have been                                                                               model still had separate events for breaker failure with probability system Notebook and the model                                            included within the boundary of                                                                        included (CBKFC0BKR_211A_022).
have them as separate events.                                            the EDG. The output breakers NUREG/CR-6928 lists breakers as                                          are no longer explicitly                                                                               This F&O is assessed as Open.
WITHIN the boundary of the EDG.                                          modeled. The EDG system Notebook and table 4 have been updated to reflect this change.
SY SY-A11 SY-A11 is Met. 5-7         Control power for the RHRSW and     Apparent inconsistency in data and   Control power was placed             NDN-000-023-2007-0026, Rev, 005; Maintenance  The update          A review of the system model identified control power associated to     SY-A11 remains Met.
SY-B6  SY-B6 is Not Met             RCW pumps is currently modeled      component boundary definitions.      under pump start gates for all       MOR R8, NDN-000-032-2007-0016                 represented a       standby pumps including HPCI (U3_SUPP_HPI116), RCIC (U3_RCI_G01)        SY-B6 remains Not SY-B9  SY-B9 is Met.                such that failure of control power                                       pumps and air compressors                                                          documentation       as is appropriate. Air compressors A-D did not have any separate        Met (see F&O 1-12).
will result in failure of the pumps                                      where it was determined that                                                       change.             power supply, so it appears to be appropriately modeled. The            SY-B9 remains Met.
to continue running. Typically,                                          control power was not                                                              (1) No new          modeling for RHRSW was also correctly developed since the pumps are control power is only needed for                                         necessary to maintain a running                                                    methods were        normally in standby. The modeling appears to be correct. The starting the pump                                                        pump.                                                                              used.               modeling for RCW includes flags to truncate or include start faults and (2) There was no    dc control power if the pumps are assumed to be in standby. The F&O change in scope of  is considered closed.
the PRA.
(3) There was no    This F&O is assessed as Closed.
change in the capability of the PRA.
SY SY-A2  SY-A2 is Met. 6-17        System models do not appear to      Review of experience from BFN        The writeup in the system           NDN-000-023-2007-0026, Rev, 005; Maintenance  Resolution          A review of several system notebooks was performed. In the Core         SY-A2 is now Met.
incorporate operating experience    and other plants does not appear     notebooks discussing the level       NDN-000-075-2007-0010, Rev, 005               required a          Spray Notebook section 3.2.6 included operating experience. Similar in developing the fault tree logic. to be used in developing the fault   of SER, OER and LER reviews has                                                    documentation       information was observed for Main Steam and Emergency Equipment RHR Service Water operating         tree system logic or data. In some   been enhanced. There is no                                                         change.             Cooling Water System. The intent of the SR is considered to be met experience does not appear to be    cases, review of BFN OE is not       requirement in the ASME                                                            (1) No new          and the F&O closed.
complete or reviewed. HVAC          included in the notebooks.           standard that requires a                                                           methods were Notebook says LERs and OER was                                           detailed listing or discussion of                                                  used.               This F&O is assessed as Closed.
reviewed, but none are listed (no                                        the generic or plant specific                                                      (2) There was no evidence of the review). Similarly,                                      experience reviewed.                                                               change in scope of for 120 VAC and others. CRD                                                                                                                                 the PRA.
Notebook includes only a                                                                                                                                    (3) There was no discussion of the BFN Fire, but no                                                                                                                          change in the review of OE is presented.                                                                                                                                  capability of the PRA.
SY SY-A14 SY-A14 is Met. 6-20        Event STRPL1STN_0750664, CS         Affects multiple Initiating Events. The strainer plugging event was      BFN IE PRA Model, Rev 8          Maintenance  Plugging was        The model was examined and the strainer modeling for one unit was       SY-A14 remains Met.
Suction Strainer Plugging, is only  Pre-existing material in the Torus   added for MLOCA. All SRVs                                                          defined previously  examined. The gate U3_SP_STN_01 was found to contain strainer assumed for Large LOCA in the       can also affect the strainer         discharge directly to the                                                          and only            plugging and it was appropriately linked to RHR and CS suctions for Model. The phenomenon causing       plugging likelihood.                 suppression pool, so a stuck                                                       expanded to         large and medium LOCA. The omission of SRV LOCA and the basis plugging is not limited to large                                         open SRV could not dislodge                                                        address medium      presented by TVA is appropriate. The F&O is considered to be closed.
LOCA only, and is possible on                                            material from the drywell.                                                         LOCA.
Medium LOCA, SRV opening, etc. A                                                                                                                            (1) No new          This F&O is assessed as Closed.
question was asked to the analyst                                                                                                                           methods were on this, and the reference to the                                                                                                                           utilized.
absence of permanently installed                                                                                                                            (2) The addition of air filters or other sources in the                                                                                                                         the sump plugging drywell. However, the debris, if                                                                                                                            for the medium present, would be swept into the                                                                                                                            LOCA represents a suction strainer by any LOCA.                                                                                                                               small change in the scope of the PRA.
(3) There was no Page 254 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                                        December 2019 Table A-6 BFN Internal Events PRA F&O Closure Review Consensus Table Prior PRT CC II                                                                                                                          Self-Assessment Reference       Maint (M) or Basis for Maint                                                                            Independent Review RU SR                     Finding No. Description                           Prior Peer Review Assessment       Self-Assessment Closure Basis                                                                       Independent Review Team Disposition Assessment                                                                                                                               Document(s)                     Upgrade?     (M) or Upgrade                                                                             Team SR Assessment change in the capability of the PRA.
SY SY-B11 SY-B11 is Met. 6-41        Fuel oil transfer pumps to refill the Issue with EDG Component           NUREG/CR6928 states that the       NDN-000-082-2007-0012, Rev 004; Maintenance  The current model  The stated text from the NUREG/CR-6928 is correct. The fuel oil         SY-B11 remains Met.
day tank are not part of the EDG      Boundary.                          EDG boundary is the following:       BFN IE PRA Model, Rev 8                     was correct as     transfer pump should be considered outside the component boundary boundary in NUREG/CR-6928.                                               The EDG boundary includes the                                                   presented.         of the EDG. The current modeling and documentation (Figure 10) are diesel engine with all                                                           Closure            correct and the F&O is resolved.
components in the exhaust                                                        documentation is path, electrical generator,                                                      the only change. This F&O is assessed as Closed.
generator exciter, output                                                        (1) No new breaker, combustion air, lube oil                                                methods were systems, fuel oil system, and                                                    used.
starting compressed air system,                                                  (2) There was no and local instrumentation and                                                    change in scope of control circuitry. However, the                                                  the PRA.
sequencer is not included. For                                                   (3) There was no the service water system                                                         change in the providing cooling to the EDGs,                                                   capability of the only the devices providing                                                       PRA.
control of cooling flow to the EDG heat exchangers are included. Room heating and ventilating is not included.
The fuel oil system is interpreted as up to the fuel oil day tank including the fuel oil transfer pumps. Each EDG at BFN has a 550gallon day tank that provides enough fuel to operate for 21/2 hours at full load. Fuel is then transferred from the 40,000gallon 7day diesel storage tank with the diesel fuel oil transfer pump to continue operation. There is one 40,000gallon 7day diesel storage tank for each diesel generator, and it is included in the diesel generator boundary.
The pumps that transfer fuel from the yard storage tank are outside the boundary and are not considered in the model.
SY IE-C11 SY-A22 is Met. 6-50        Some of the MOVs credited in the      MOVs closing for ISLOCA are risk   Assumption was added to the         vsloca_r8;                      Open         none               A review of the ET representation identifies operator mitigation        SY-A22 remains met SY-A22 IE-C11 is Met.              ISLOCA Fault Tree are not tested to   significant, with a RAW of greater ISLOCA Notebook.                    BFN IE PRA Model, Rev 8                                         actions are included in the ET. This was also found to be the case when at Cat II.
close against full DP. These MOVs     than 2.                            Depressurization is not modeled                                                                     the ISLOCA modeling in the CAFTA model was reviewed (for example,       IE-C11 remains Met.
are not originally included in the                                       in the ISLOCA initiator before                                                                      gate U1_VRLOCA_002 includes gate U1_ISLV55_2 dealing with design as RCS isolation valves.                                          valve closure. The probability of                                                                   isolation).
Examples include 74-55 and 74-66                                         this failing to occur is only (note: this is not a complete list,                                      5.077E-02. The fact that all                                                                        This F&O is assessed as Open.
but 2 of 4 valves reviewed were                                          ISLOCA events go directly to not in the MOVATs 89-10                                                  core damage without any program).                                                                mitigation actions is more than adequate to make up for not modeling the low probability of SRV failure.
Page 255 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table A-7 Disposition of Open BFN Internal Flooding F&Os It was stated that no credit was taken for the removal of water via the drain system, with the exception of spray events (100 gpm). No scenarios were modeled that included backflow F&O IFSN-A8-01           through drain lines. Although this is reasonable based on the layout of the large open areas in the Reactor Building and Turbine Building, no discussion of the elimination of backflow was provided in the documentation.
Expand discussion in the Internal Flooding Notebook that explains how drain backflow was treated in the internal flood model. Include enough detail to justify screening.
: 1) What screening criteria was used?
: 2) How is the drain system configured?
: a. Are there separate drain systems in each building? (i.e.,
RB, TB, CB, etc.)
: b. Can a drain line become blocked downstream?
Possible                     c. Where does the water end up? (Sump on lower level?,
Resolution                      Holding tank?, Outside?)
: 3) Include general references that can be validated by the reviewer, such as the system description and/or drawings used to support the assumptions for screening.
: 4) Is screening conservative? Why?
This does not need to be a large effort but a statement that any of the rooms within a building already show water propagating to the bottom elevation of that building does not provide enough detail to demonstrate that the drain impacts were sufficiently assessed for screening.
Associated SRs           IFSN-A8 In the BFN Internal Flooding Analysis, it was determined that the only place that drain backflow could occur and potentially cause any issues would be in the lowest elevations of each building.
The affect from this occurrence is already accounted for in each of the flooding scenarios as they all propagate to the lowest Plant Response           elevation. The drain lines are not connected for each building so water could not propagate from one building to another. The upper elevation drainage systems were not analyzed as a potential backflow situation as the drains are relatively small compared to the open hatches and stairwells that would cause the water to propagate to the lowest elevations. In addition, the areas in which the water would be susceptible to drainage are Page 256 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table A-7 Disposition of Open BFN Internal Flooding F&Os large rooms where the water would have to significantly fill in order to even reach a drain.
Section 6.1.3 of the Internal Flooding Notebook explains that we screened drainage backflow from the analysis and provides the justification.
This is a documentation issue. Therefore, there is no impact on Impact on SPRA the Seismic PRA.
No specific flow rate calculations were performed. Flow rates were modeled to be the maximum flow rate for a given break category. For example, all flood events were assumed to result F&O IFSN-A9-01 in a break flow of 2,000 gpm. This results in very conservative times to component failure. It could result in incorrect ranking of the risk importance of the flooding scenarios.
Possible                 As a minimum, perform calculations to estimate the actual flow Resolution               rates of modeled breaks for the most risk-significant scenarios.
Associated SRs           IFSN-A9 The BFN Internal Flooding Analysis conservatively assumed that the flows out of the pipe breaks were at the top end of each of the generic flow rate values. This was done to assure that we properly addressed the importance of each scenario. The pipe-break frequencies are given for the range of flows and the frequency does not change whether the top end flow rate or a lower flow rate is used unless it changes which range of flows you are using. The only time you would be concerned with the flow rate would be when you are performing an operator action Plant Response to prevent water accumulation within a room. The BFN Internal Flooding analysis did not credit any of these types of operator actions except for in the RB at EL 519. The flow rates that could cause this elevation to flood could be from any water source in the RB, so the highest flow rate possible for both the flood scenario and the major flood scenario was used in calculating timing for the HRA action. This gives the smallest possible timeframe with which to perform the action and ensures that the results are conservative and risk insights are reasonable.
Even with conservative flowrates assumed for pipe breaks, Impact on SPRA           seismically induced flooding is not a significant contributor to seismic risk.
Page 257 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table A-7 Disposition of Open BFN Internal Flooding F&Os Spray events in the RB general areas (multiple elevations) are assumed to result in a manual trip and are analyzed. Larger flooding events are not considered an initiating event unless operators fail to isolate the flood prior to reaching the level of F&O IFSN-A10-01 equipment damage (5') at EL 519'. This appears to be an inconsistency between the spray and flood events. Although less frequent than spray events, flood events in these areas could in total be a significant contributor to CDF.
Develop some initiating event that models floods in the general areas of the RB, along with successful isolation of the flood prior Possible to equipment damage on EL 519' of the RB. Based on the Resolution results, determine whether the entire group of these scenarios should be included in the internal flood model.
IFSNA10 Associated SRs IFEVA1 When analyzing the spray events, it was assumed that for every spray scenario the operators would manually scram the reactor.
This is a conservative assumption as the operators may not need to shut down the plant. By analyzing every spray scenario with a manual scram, we were able to see what the impact from a spray scenario would be to the plant. The flooding scenarios, on the other hand, were not analyzed as during an RB flood scenario all the water would propagate down to EL 519' of the RB. If the operators are successful in isolating the pipe rupture prior to reaching 5' in EL 519', the plant would not necessarily be tripped. While it is true that some equipment might be lost, Plant Response           which is similar to that seen for the spray events, the flooding analysis viewed the equipment impact separately from the flooding scenario as the flood has been terminated. Therefore, the impact from the equipment being lost would be characterized by the internal events PRA model.
Each of the RB flooding scenarios that are successfully mitigated by the HRA action for EL 519' submergence will be reviewed to determine whether a potential scenario would exist or not. In addition, the spray scenarios will be reviewed to determine whether those are potential scenarios, and the results will be documented within the Internal Flooding Notebook.
Page 258 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 Table A-7 Disposition of Open BFN Internal Flooding F&Os This F&O is not applicable to the SPRA since the seismic event is the initiating event. Human actions for high-magnitude seismic Impact on SPRA           events that could potential damage piping are assumed to be failed.
Only a 2000 gpm flood initiating event was modeled in the Unit 1 Shutdown Board Room A. Spray events were not modeled.
F&O IFSN-A10-02          Given that there are no drains nor indication in that room (and an informal analysis), there is a possibility that a spray event of 100 gpm could also result in similar consequences.
Perform a calculation at 100 gpm to determine whether a spray Possible scenario is, in fact, a valid initiating event in this area. If so, Resolution include spray events in that area in the model.
Associated SRs           IFSNA10 Each room was looked at for potential spray effects, including the 4KV Shutdown Board Room A. This spray scenario is in the model as U1-621-R02_025_S with a contribution of 1.54E-10 to CDF, which constitutes 0.002% of the Internal Flooding CDF for Plant Response Unit 1. This spray scenario will be reviewed to ensure that it is treated appropriately within the model, and any changes will be documented in the next revision of the Internal Flooding Notebook.
For a given area, the equipment damaged by seismically induced flooding is always assumed to be from the largest Impact on SPRA           possible flood in that area. Even with conservative flowrates assumed for pipe breaks, seismically induced flooding is not a significant contributor to seismic risk.
For spray events in the general areas of the RB, all the possible spray frequencies were added to obtain on combined frequency F&O IFEVA101 for one event. The impact of this spray event was the combined impact of all the possible spray events on that elevation.
Separate out spray events in these areas to provide a better Possible picture of which spray sources and which impacted equipment Resolution are the more significant contributor.
IFEVA1 Associated SRs IFEVA2 Page 259 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table A-7 Disposition of Open BFN Internal Flooding F&Os For the general areas of the RB, all spray scenarios were determined to occur at the same time, and all equipment affected by a certain system piping were all failed. Because this is such a big room, this modeling approach was too Plant Response           conservative. Each of the spray scenarios within the general area of the RB will be reviewed to determine which components can be failed by what portions of piping, and new scenarios will be developed to ensure that only the pipe ruptures that affect a component are used to fail a particular component.
This F&O is not applicable to the SPRA since the seismic event is the initiating event. For a given area, the equipment damaged by seismically induced flooding is always assumed to be from Impact on SPRA the largest possible flood in that area. Even with conservative flowrates assumed for pipe breaks, seismically induced flooding is not a significant contributor to seismic risk.
The HRA assessment needs to incorporate several items:
a) Cues and indicators need to be documented in the first mitigation HRA (HFA_0_519FLOOD) b) With a), indicators should be assessed for flood damage c) PSFs need to be altered for general worst case in F&O IFQUA601             environment (e.g., radiation, etc.). This is because the flood mitigation actions are general and are not specific in place or time.
d) Why is the belief in the adequacy of instruction set to "No"?
For nonmitigation post-initiator HRAs:
a) Needs to discuss blocked path for each scenario Incorporate the missing pieces to the mitigation HRAs.
a) Cues and indicators for the first mitigation HRA (HFA_0_519FLOOD) b) With a), indicators should be assessed for flood damage Possible                   c) Alter PSFs for general worst case in environment (e.g.,
Resolution                 radiation, etc.)
d) Alter or add some discussion on why the Belief in Adequacy is set to "No" For nonmitigation post-initiator HRAs:
a) Discuss or incorporate blocked path for each scenario Associated SRs           IFQUA6 Page 260 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table A-7 Disposition of Open BFN Internal Flooding F&Os The HRA Assessment on HFA_0_519FLOOD was done generically as the only indication would be the Alarm coming in saying that there is water building up in EL 519'. The operator would be sent out to see if the alarm was valid and then try and isolate the pipe rupture. The Cues and Indicators will be updated to reflect the Alarm Indication. The flooding detectors are designed to get wet and would not be damaged by a flood. In addition, there are multiple flooding detectors within EL 519, so if any of the detectors work, the operators would still be able to mitigate the flood. The belief in adequacy of instruction was set to No as the operators would most likely question whether there is an actual flood within the RB. The operators would still comply with the procedure and perform the action as stated.
There is a timing aspect included that is to assess whether the flood actually occurred.
Plant Response           The PSFs were reviewed to assess whether an operator would experience any adverse situation outside of what would be experienced through everyday work. Because the flood and associated mitigation accident would occur prior to reactor trip, the shaping factors were consistent with a normal workload within the RB. Lighting would not be affected by the flood; heat/humidity would be normal for the areas that would be traversed. All of the areas within BFN are radiation areas, so there is no increased stress from radiation; isolating the pipe would be a simple action, and the stress was expected to be low as there is plenty of time to perform the action. It is expected that the action to close a couple of valves would not increase the stress on the operator.
Each of the HRAs will be reviewed to determine what the impact would be from a blocked path, and this will be documented within the Internal Flooding Notebook.
This F&O is not applicable to the SPRA since the seismic event is the initiating event. Human actions for high-magnitude seismic Impact on SPRA           events that could potential damage piping are assumed to be failed.
Page 261 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 Table A-7 Disposition of Open BFN Internal Flooding F&Os No modeling of direct effects due to a flooding event were identified. The rational was that for large flooding events in the RB, only those floods that resulted in flood levels reaching 5' in EL 519' were modeled. For those events, the required SSCs have failed due the indirect effects of the flooding.
F&O IFQUA901           Therefore, the direct effects of the flooding need not be considered. It is our contention that floods in the RB that are successfully isolated before damage occurs to components on EL 519' should be included as initiators. These events will still result in damage to SSCs and direct failure to part of the breached system.
Include floods on the RB at EL 565' and above, even with Possible successful isolation prior to equipment damage on EL 519'. For Resolution those events, model the direct failure of the breached system.
Associated SRs           IFQUA9 This F&O is similar to F&O IFSN-A10-01. As mentioned in the response for that F&O, an operator may not need to scram the reactor for a loss of a component affected by a flooding event.
Plant Response           Each of the RB flooding scenarios that are successfully mitigated by the HRA action for EL 519' submergence will be reviewed to determine whether a potential scenario would exist or not.
This F&O is not applicable to the SPRA since the seismic event is the initiating event. Human actions for high-magnitude seismic Impact on SPRA events that could potential damage piping are assumed to be failed.
A.8    Identification of Key Assumptions and Uncertainties The PRA Standard [8] includes several requirements related to identification and evaluation of the impact of assumptions and sources of uncertainty on the PRA results.
NUREG-1855 [61] and EPRI 1016737 [51] provide guidance on assessment of uncertainty for applications of a PRA. As described in NUREG-1855, sources of uncertainty include parametric uncertainties, modeling uncertainties, and completeness (or scope and level of detail) uncertainties.
* Parametric uncertainty was addressed as part of the BFN Seismic PRA model quantification (see Section 5 of this submittal).
* Modeling uncertainties are considered in both the base internal events PRA and the Seismic PRA. Assumptions are made during the PRA development to Page 262 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 address a particular modeling uncertainty because there is not a single definitive approach. Plant-specific assumptions made for each of the BFN Seismic PRA technical elements are noted in the Seismic PRA documentation that was subject to peer review, and a summary of important modeling assumptions is included in Section 5.
* Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the PRA but are only considered for their impact on a specific application. No specific issues of PRA completeness were identified in the Seismic PRA peer review.
A summary of potentially important sources of uncertainty in the BFN Seismic PRA is listed in Table A-8.
Table A-8 Summary of Potentially Important Sources of Uncertainty PRA        Summary of Treatment of Sources            Potential Impact on Element        of Uncertainty per Peer Review           Seismic PRA Results Seismic     The BFN SPRA peer review team             The seismic hazard Hazard      noted that both the aleatory and          reasonably reflects epistemic uncertainties have been         sources of uncertainty.
addressed in characterizing the seismic sources. In addition, uncertainties in each step of the hazard analysis were propagated and displayed in the final quantification of hazard estimates for the BFN site.
A.9     Identification of Plant Changes Not Reflected in the Seismic PRA The BFN Seismic PRA reflects the plant as of the cutoff date for the Seismic PRA, which was January 2016. All plant changes have been reviewed since the 2016 cutoff date, and there are no significant plant changes subsequent to this date.
Page 263 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 Appendix B NRC Generic Concerns on Responses to NTTF 2.1 Seismic 50.54(f) Letter The purpose of this Appendix is to provide a response for each of the generic observations associated with the staffs review of SPRA reports provided in response to the March 12, 2012, 50.54(f) letter associated with reevaluated seismic hazards.
: 1. Resolution of finding level Facts and Observations (F&Os) for internal events probabilistic risk assessment (PRA)
NRC Observation 1: The internal events PRA forms the base for the SPRA. To date the staff has ensured that internal event F&Os are resolved/closed AND that the SPRA reflects those resolutions/closures through the audit process. The staff is looking for a more efficient way of addressing this issue. For those plants that have already dispositioned the internal events F&Os and the disposition have been fed into the SPRAs, the staff believes (1) adding a statement in the cover letter transmitting the SPRA submittal that this was done, and (2) adding a statement that the findings were closed through an NRC-accepted process or pointing to docketed information providing the dispositions would obviate the need for the staff trying to determine this through the audit process.
BFN Response The cover letter transmitting this submittal has the following statement: The BFN internal events PRA has had all finding-level peer review F&Os being dispositioned as discussed in Appendix A and the updated internal events model has been used as the basis for the BFN SPRA.
Section A.4 of Appendix A describes how each of the Finding-Level BFN SPRA F&Os were closed using an NRC-accepted process. Section A.6 of Appendix A describes how most of the Finding-Level BFN Internal Events F&Os were closed using an NRC-accepted process. Table A-5 shows the basis for closing these F&Os.
Table A-6 gives the disposition of each open Internal Events and Internal Flooding F&Os. The BFN Seismic PRA Peer Review and Closure Review teams have reviewed these dispositions and have agreed that there is no impact on the BFN Seismic PRA. There are no open BFN Seismic PRA F&Os associated with the open Internal Events or Internal Flooding F&Os.
: 2. Consideration of Staff Comments on Industry Documents NRC Observation 2 (i): The staff had several comments on the industry guidance for crediting Diverse and Flexible Coping Strategies (FLEX) equipment and actions in PRAs (Nuclear Energy Institute [NEI] 16-06), which were documented in a publicly available memorandum dated May 30, 2017 (ADAMS Accession No.ML17031A269). To date the staff has used the audit process to review the credit for FLEX equipment and actions with the intent of ensuring that the credit considers those comments. The staff is looking for a more efficient way of addressing this issue and focusing its review. A potential path that can gain efficiency would be a discussion in the SPRA submittal about the specific credit Page 264 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                              December 2019 for FLEX equipment and actions included in the SPRA, how the staffs comments on NEI 16-06 were appropriately considered, and the basis of as well as results from relevant sensitivity studies.
BFN Response There are 13 conclusions reached in the NRC memorandum dated May 30, 2017 (ADAMS Accession No.ML17031A269). Each of these conclusions is addressed below:
NRC Conclusion 1: NEI 16-06 has not provided accepted human reliability analysis methods for inclusion of offsite portable equipment to take quantitative risk credits in risk-informed applications that should meet the guidance of RG 1.200; therefore, claiming quantitative credits for offsite equipment is not appropriate until evaluations consistent with the guidance of RG 1.200 or improvements in the NEI guidance or state-of-art methods address the technical gaps BFN Response to NRC Conclusion 1 No credit was taken for any portable offsite FLEX equipment in the BFN SPRA. The FLEX operator actions to align backup nitrogen to drywell control air for the SRVs were included in the model. No other FLEX systems were credited in the model.
NRC Conclusion 2: For any new risk-informed application that has incorporated mitigating strategies and should meet the guidance of RG 1.200, the licensee should either perform a focused-scope peer review of the PRA model or demonstrate 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.
BFN Response to NRC Conclusion 2 This is not applicable to this submittal; however, a peer review has been performed on the BFN SPRA model that includes the FLEX operator actions to align backup nitrogen to drywell control air for the SRVs that were included in the model. No other FLEX systems were credited in the model. This is documented in Section A.2 of Appendix A.
NRC Conclusion 3: Licensees may incorporate mitigating strategies in PRA models after the issuance of amendments for applications that use PRA models to exercise self-approval for a plant change. For such applications, the licensee should, in addition to conforming with specific license condition(s) associated with those applications, either perform a focused scope peer review and resolve the focused scope peer-review findings before using the new models to support any risk-informed decision-making or document an evaluation demonstrating that none of the upgrade criteria is satisfied. NRC will monitor those evaluations and their documentation, along with evaluations and documents related to Page 265 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 other items identified in this assessment, through appropriate regulatory processes (e.g., inspections).
BFN Response to NRC Conclusion 3 This is not applicable to this submittal. It is not a risk-informed application.
NRC Conclusion 4: The use of expert judgment consistent with the PRA Standard as endorsed by RG 1.200 is acceptable for estimating parameter values under certain conditions and the rationale for estimated values should be documented. In reviewing future risk-informed applications, the staff may request additional information to understand the rationale for parameter values. Using the appropriate regulatory processes, the NRC will review the rationale for parameter values added to PRA models after issuance of applications that use PRA models to exercise self-approval for a plant change.
BFN Response to NRC Conclusion 4 This is not applicable to this submittal. It is not a risk-informed application.
NRC Conclusion 5: The NRC staff does not agree with crediting spare portable equipment not modeled in the PRA in lieu of using appropriate failure rates because this approach is not consistent with the PRA Standard and RG 1.200. Furthermore, the potential impact of underestimating failure rates could be larger than the unquantified risk benefits of spare equipment not modeled in PRAs.
BFN Response to NRC Conclusion 6 This is not applicable to this submittal. It is not a risk-informed application.
NRC Conclusion 6: The failure rates of permanently installed equipment cannot be used for portable equipment even if sensitivity analyses are performed. Licensees should use plant-specific generic data collected and analyzed using acceptable approaches to estimate the failure rates for portable equipment.
BFN Response to NRC Conclusion 6 The FLEX operator actions to align backup nitrogen to drywell control air for the SRVs were included in the model. No other portable FLEX systems were credited in the model. The valves and piping used to align the backup nitrogen tanks are permanently installed. The nitrogen tanks have been evaluated by the fragility team and were determined to be seismically rugged. The random failure of the tanks was determined to be insignificant when compared to the failure rates of the modeled operator actions required for aligning the nitrogen supply.
Page 266 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                                December 2019 NRC Conclusion 7: NEI 16-06 and risk-informed applications should address whether and how the analysis described in Supporting Requirement DA-D8 is performed.
BFN Response to NRC Conclusion 7 This is not applicable to this submittal. It is not a risk-informed application.
NRC Conclusion 8: The uncertainty associated with failure rates of portable equipment should be considered in the PRA models consistent with the PRA Standard as endorsed by RG 1.200. Risk-informed applications should address whether and how these uncertainties are evaluated.
BFN Response to NRC Conclusion 8 The FLEX operator actions to align backup nitrogen to drywell control air for the SRVs were included in the model. No other portable FLEX systems were credited in the model. The valves and piping used to align the backup nitrogen tanks are permanently installed. The nitrogen tanks have been evaluated by the fragility team and were determined to be seismically rugged. The random failure of the tanks was determined to be insignificant when compared to the failure rates of the modeled operator actions required for aligning the nitrogen supply. Therefore, the uncertainty associated with failure rates of portable equipment is insignificant.
NRC Conclusion 9: The NRC staff does not have access to and has not reviewed PWROG-14003. At this time, the NRC staff treats approaches proposed by that PWROG document as unreviewed methods.
BFN Response to NRC Conclusion 9 PWROG-14003 was not used in the development of the BFN Seismic PRA.
NRC Conclusion 10: Without any additional data or evaluations, the currently available common cause failure (CCF) parameter values should be used which should appropriately reflect the higher CCF failure rates of the portable equipment when applied to the higher independent failure rates.
BFN Response to NRC Conclusion 10 The FLEX operator actions to align backup nitrogen to drywell control air for the SRVs were included in the model. No other portable FLEX systems were credited in the model. The valves and piping used to align the backup nitrogen tanks are permanently installed. The nitrogen tanks have been evaluated by the fragility team and were determined to be seismically rugged. The random failure of the tanks was determined to be insignificant when compared to the failure rates of the modeled operator actions required for aligning the nitrogen supply.
Page 267 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                               December 2019 NRC Conclusion 11: The staff finds that using surrogates for specific actions or engineering judgement to estimate the failure probability do not adequately address the elements needed for a technically acceptable human reliability analysis described in the PRA Standard (e.g., the impact of the environment under which the operators work). Until gaps in the human reliability analysis methodologies are addressed by improved industry guidance, HEPs associated with actions for which the existing approaches are not explicitly applicable, such as actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06, along with assumptions and assessments, should be submitted to NRC for review.
BFN Response to NRC Conclusion 11 No credit is taken for debris removal, transportation of portable equipment, installation of equipment at a staging location, or routing of cables and hoses. The FLEX operator actions to align backup nitrogen to drywell control air for the SRVs were included in the model. No other portable FLEX systems were credited in the model. The valves and piping used to align the backup nitrogen tanks are permanently installed.
The operator actions associated with aligning the backup nitrogen are not significantly different than other ex-control room operator actions associated with other permanently installed equipment. These actions are properly evaluated using the existing HRA tools (HRA Calculator).
NRC Conclusion 12: If procedures for initiating mitigating strategies are not explicit and the associated failure probabilities are not directly analyzed by accepted approaches, technical bases for probability of failure to initiate mitigating strategies should be submitted to NRC for review.
BFN Response to NRC Conclusion 12 The procedures associated with aligning the backup nitrogen are explicit, and the associated failure probabilities are directly analyzed by accepted approaches.
NRC Conclusion 13: Until acceptable guidance is provided for identifying and assessing unique aspects of pre-initiator human failure events for mitigating strategies, the staff may request additional information regarding assessment of those human failure events.
BFN Response to NRC Conclusion 13 There are no unique aspects of pre-initiator human failure events associated with aligning the backup nitrogen when compared to other ex-control room operator actions credited in the model.
Page 268 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 NRC Observation 2(ii): The staff issued a formal acceptance letter for NEI 12-13 dated March 7, 2018 (ADAMS Accession No. ML18025C022), which included specific comments. The letter stated that the use of NEI 12-13 was acceptable when supplemented by the staffs comments. To date the staff has used the audit process to ensure that the implementation of NEI 12-13 was appropriately supplemented by the staffs comments. A potential path for efficiency in this area would be a discussion in the SPRA submittal about the consideration of the staffs comments in the aforementioned acceptance letter provided such confirmation exists in the peer-review report (an excerpt from the peer-review report that states as much would also be beneficial).
BFN Response This peer review was performed by an experienced, independent team, using the process defined in NEI guidelines NEI-12-13 as amended by the NRC on March of 2018 (ADAMS access ML18025C024 and ML18025C025).
: 3. Combining potential improvements during detailed screening:
NRC Observation 3: In alignment with the discussion in the Enclosure to letter dated September 21, 2016 (ADAMS Accession Nos. ML16237A108 [letter] and ML16237A114 [enclosure]) the staffs evaluation of each licensees SPRA submittal includes a determination whether additional regulatory actions are necessary (e.g., updating the design basis and structures, systems, and components (SSCs) important to safety. The staff uses guidance documents that have been developed to facilitate consistent and objective decision-making (ADAMS Accession No. ML17146A200). To date, in accordance with the cited guidance, the staff has engaged with the licensee to request information and insights, as necessary, as part of the audit process. A potential path for efficiency in this area would be the consideration of the enclosure and guidance document mentioned above and communication of the results therefrom in the submittal.
BFN Response Prior to the completion of the SPRA, select plant modifications were completed to reduce seismic risk based on IPEEE results and early SPRA quantifications. The modifications include 1) replacing the diesel auxiliary transformers and 2) installing fasteners on removable grating above the diesel generator intake and exhaust dampers.
Additionally, studies have been completed to evaluate a potential modification of SEIS_5-2B Initiation Panels and Relays to improve SLERF. Based on risk sensitivity studies, SLERF appeared to be driven by SEIS_5-2B. An increase in the fragility value up to functional failure made a difference in SLERF but did not seem a realistic target since it would be over double the current fragility level.
Therefore, an increase of 50% was evaluated and found to not significantly improve SLERF. The failure mode of the panels is anchorage. Completing a detailed SoV fragility calculation on these panels is not expected to provide more than approximately 20% increase in fragility. Based on these risk sensitivity studies, possible improvements to the anchorage were assessed to determine if it was feasible to perform a modification to obtain a substantial increase in the fragility value. The controlling failure is the concrete anchors. The welds between Page 269 of 270

BFN 50.54(f) NTTF 2.1 Seismic PRA Summary Report                             December 2019 the panel and the sill channel is the next controlling failure mode. A slight improvement (~ 5%) to the concrete anchors would shift the controlling failure mode to the weld. Thus, two modifications would be required; first, a replacement of the concrete anchors and second, additional weld of panel to the sill channel.
Interferences within the panel make replacement of the concrete anchors impractical. Additional possible modifications were considered such as top bracing or bracing from the floor. These panels are located in the auxiliary instrument room which contains many panels. Bracing to the floor would cause accessibility issues. Top bracing was also considered but deemed impractical due to the number of panels in the room and block walls as the only potential support point. Furthermore, these panels house relays that are operational during all modes of operation; while it may seem feasible to work modifications in during outages when some relay functions are not necessary, this would not be recommended due to those relays in the panel that are required for operability during the outage. Detailed studies on the amount of improvement to the fragility through modification was not evaluated further due to the impracticality of the modification.
Page 270 of 270}}