ML18079B140

Browns Ferry Nuclear Plant, Units 1, 2, and 3: Proposed Technical Specifications (TS) Change TS-510 - Request for License Amendments - Maximum Extended Load Line Limit Analysis Plus

ML18079B140
Person / Time
Site:	Browns Ferry
Issue date:	02/23/2018
From:	Shea J W Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML18086A089	List: ML18079B140
References
CNL-18-002
Download: ML18079B140 (1585)

Text

[[:#Wiki_filter:, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-18-002 February 23, 2018 10 CFR 50.90 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:

In accordance with the provisions of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Tennessee Valley Authority (TVA) is submitting a request for a Technical Specification (TS) amendment (TS-510) to Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 for Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3, respectively. The proposed amendment will allow operation in the expanded Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain and use of the Detect and Suppress Solution - Confirmation Density (DSS-CD) stability solution. The enclosure to this letter provides a summary of the proposed changes to the BFN Units 1, 2, and 3 renewed facility operating licenses and Technical Specifications to support operation in the MELLLA+ operating domain and the use of the DSS-CD stability solution, the no significant hazards considerations, and environmental considerations. It should be noted that AREVA nuclear reactor operations has been recently renamed "Framatome." For the purposes of this LAR, "AREVA" should be considered synonymous with "Framatome." Attachments 1 though 39 include a description of the proposed changes to the licensing basis and provide supporting documentation. This LAR is subdivided as follows. U. S. Nuclear Regulatory Commission CNL-18-002 Page 2 February 23, 2018 Attachment 1 Proposed Renewed Facility Operating License and Technical Specification Changes (Markups) Attachment 1 provides a markup of the affected renewed facility operating license and Technical Specifications pages indicating the proposed changes for MELLLA+. Attachment 2 Retyped Proposed Renewed Facility Operating License and Technical Specification Changes Attachment 2 provides a camera-ready retype of the affected renewed facility operating license and Technical Specification pages with changes incorporated. Attachment 3 Proposed Technical Specification Bases Changes (Markups) Attachment 3 provides a markup of the affected Technical Specification Bases pages indicating the proposed changes for MELLLA+. Attachment 4 Retyped Proposed Technical Specification Bases Changes Attachment 4 provides a camera-ready retype of the affected Technical Specification Bases pages with changes incorporated. Attachment 5 NEDC-33877P, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Maximum Extended Load Line Limit Plus (proprietary) Attachment 5 provides a proprietary version of the GE-Hitachi Nuclear Energy Americas LLC (GEH) Maximum Extended Load Line Limit Analysis Plus Safety Analysis Report (M+SAR) documented in NEDC-33877P, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Maximum Extended Load Line Limit Analysis Plus. This report summarizes the results of safety analyses and evaluations performed that justify the expansion of the core flow operating domain for BFN Units 1, 2, and 3. The proposed changes expand the operating domain, but do not increase the licensed power level. The MELLLA+ core flow operating domain expansion requires minor plant system modifications, such as changes to the operating power/core flow map, application of the DSS-CD stability solution, and changes to some instrument setpoints. Electric Power Research Institute (EPRI) and GEH separately consider portions of the information provided in Attachment 5 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. Affidavits for withholding information, executed by GEH and EPRI, are provided in Attachment 37 and Attachment 39, respectively. A non-proprietary version of the document is provided in Attachment 6. Therefore, on behalf of GEH and EPRI, TVA requests that Attachment 5 be withheld from public disclosure in accordance with the GEH and EPRI affidavits and the provisions of 10 CFR 2.390. Attachment 6 NEDO-33877, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Maximum Extended Load Line Limit Plus (non-proprietary) U. S. Nuclear Regulatory Commission CNL-18-002 Page 3 February 23, 2018 Attachment 7 ANP-3551P, AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 (proprietary) Attachment 7 provides a proprietary version of the AREVA Inc. (AREVA) MELLLA+ Safety Analysis Report (AMSAR) documented in ANP-3551P, "AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3." This report supplements the M+SAR and summarizes the results of fuel-related safety analyses and evaluations performed that justify expansion of the core flow operating domain and the use of the DSS-CD stability solution for AREVA fuel installed in BFN Units 1, 2, and 3. The AMSAR technical evaluations supporting the proposed changes are based on a series of NRC-approved AREVA methods. The safety evaluations documented in Attachment 7 are based on the continued use of AREVA's ATRIUM 10XM fuel design. Where appropriate, evaluations for the existing ATRIUM-10 fuel design have also been included because the potential exists that some fuel of this type may still be resident in a BFN reactor core upon implementation of MELLLA+. The general format of the technical evaluations in Attachment 7 is consistent with NEDC-33006P-A, Revision 3, "General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus," June 2009, as it applies to the licensing basis of BFN Units 1, 2, and 3. However, because only selected portions of NEDC-33006P-A are addressed, the numbering of individual sections is discontinuous. AREVA considers portions of the information provided in Attachment 7 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 8. Therefore, on behalf of AREVA, TVA requests that Attachment 7 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 8 ANP-3551NP, AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 (non-proprietary) U. S. Nuclear Regulatory Commission CNL-18-002 Page 4 February 23, 2018 Attachment 9 ANP-3544P, Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design (proprietary) Attachment 9 provides the design results for the equilibrium cycle reactor core loading, including projected control rod patterns and evaluations of thermal and reactivity margins. The equilibrium cycle results are summarized for operation of BFN Units 1, 2, and 3 with ATRIUM 10XM fuel at EPU conditions with MELLLA+. AREVA considers portions of the information provided in Attachment 9 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 10. Therefore, on behalf of AREVA, TVA requests that Attachment 9 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 10 ANP-3544NP, Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design (non-proprietary) Attachment 11 ANP-3546P, Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) (proprietary) Attachment 11 provides the results of a loss-of-coolant accident (LOCA) break spectrum analysis for BFN Units 1, 2, and 3. The break spectrum analysis is used to identify the parameters that result in the highest calculated peak cladding temperature (PCT) during a postulated LOCA. AREVA considers portions of the information provided in Attachment 11 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 12. Therefore, on behalf of AREVA, TVA requests that Attachment 11 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 12 ANP-3546NP, Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) (non-proprietary) Attachment 13 ANP-3547P, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10XM Fuel (EPU MELLLA+) (proprietary) Attachment 13 provides the results of the LOCA-ECCS analyses for BFN Units 1, 2 and 3. The LOCA-ECCS analysis specifies the maximum average planar linear heat generation rate (MAPLHGR) limit versus exposure for ATRIUM 10XM fuel and demonstrates that the MAPLHGR limit is adequate to ensure that the LOCA-ECCS criteria in 10 CFR 50.46 are satisfied for operation at or below the limit. Attachment 13 also documents the licensing basis PCT and corresponding local cladding oxidation from the metal water reaction (MWR) for ATRIUM 10XM fuel U. S. Nuclear Regulatory Commission CNL-18-002 Page 5 February 23, 2018 used at BFN Units 1, 2, and 3. AREVA considers portions of the information provided in Attachment 13 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 14. Therefore, on behalf of AREVA, TVA requests that Attachment 13 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 14 ANP-3547NP, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10XM Fuel (EPU MELLLA+) (non-proprietary) Attachment 15 ANP-3548P, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10 Fuel (EPU MELLLA+) (proprietary) Attachment 15 provides the results of the LOCA-ECCS analyses for BFN Units 1, 2 and 3. The LOCA-ECCS analysis specifies the MAPLHGR limit versus exposure for ATRIUM-10 fuel and demonstrates that the MAPLHGR limit is adequate to ensure that the LOCA-ECCS criteria in 10 CFR 50.46 are satisfied for operation at or below the limit. Attachment 15 also documents the licensing basis PCT and corresponding local cladding oxidation from the MWR for ATRIUM-10 fuel used at BFN Units 1, 2, and 3. AREVA considers portions of the information provided in Attachment 15 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 16. Therefore, on behalf of AREVA, TVA requests that Attachment 15 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 16 ANP-3548NP, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10 Fuel (EPU MELLLA+) (non-proprietary) U. S. Nuclear Regulatory Commission CNL-18-002 Page 6 February 23, 2018 Attachment 17 ANP-3552P, Browns Ferry Unit 3 Cycle 19 Representative Reload Analysis (EPU MELLLA+) (proprietary) Attachment 17 provides reload licensing analyses results in support of MELLLA+ operation at EPU conditions at BFN. The analyses reported in Attachment 17 were performed using NRC-approved methodologies for generic application to boiling water reactors. Reload licensing analyses were performed for potentially limiting events. The results are used to establish the Technical Specifications/Core Operating Limits Report limits and ensure design and licensing criteria are met. The results of the representative reload licensing analysis for BFN Unit 3 Cycle 19 operation are presented and support operation with equipment out-of-service. AREVA considers portions of the information provided in Attachment 17 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 18. Therefore, on behalf of AREVA, TVA requests that Attachment 17 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 18 ANP-3552NP, Browns Ferry Unit 3 Cycle 19 Representative Reload Analysis (EPU MELLLA+) (non-proprietary) Attachment 19 ANP-3553P, Browns Ferry Unit 3 Cycle 19 EPU (120% OLTP) MELLLA+ LAR Reference Fuel Cycle Design (proprietary) Attachment 19 provides the design results for the BFN Unit 3 Cycle 19 reactor core loading, including projected control rod patterns and evaluations of thermal and reactivity margins. The Cycle 19 results are summarized for operation of BFN Unit 3 with ATRIUM 10XM fuel at EPU conditions with MELLLA+ based on projected Cycle 18 core operation. AREVA considers portions of the information provided in Attachment 19 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 20. Therefore, on behalf of AREVA, TVA requests that Attachment 19 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 20 ANP-3553NP, Browns Ferry Unit 3 Cycle 19 EPU (120% OLTP) MELLLA+ LAR Reference Fuel Cycle Design (non-proprietary) U. S. Nuclear Regulatory Commission CNL-18-002 Page 7 February 23, 2018 Attachment 21 ANP-3568P, Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Extended Power Uprate/MELLLA+ (proprietary) Attachment 21 provides the results of fuel rod thermal-mechanical analyses to demonstrate that the applicable design criteria are satisfied. The analyses are for ATRIUM 10XM fuel. These evaluations assess fuel rod performance at EPU conditions and operation in the MELLLA+ operating domain based on NRC-approved methodologies and design criteria. AREVA considers portions of the information provided in Attachment 21 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 22. Therefore, on behalf of AREVA, TVA requests that Attachment 21 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 22 ANP-3568NP, Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Extended Power Uprate/MELLLA+ (non-proprietary) Attachment 23 FS1-0029291, Browns Ferry Units 1, 2, and 3 EPU MELLLA+ MCPR Safety Limit Analysis With SAFLIM3D Methodology (proprietary) Attachment 23 provides the Safety Limit Minimum Critical Power Ratio (MCPR) results using the methodology in AREVA's licensing topical report, ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, dated June 2011, to support an extension of the licensed power/flow operating domain into the MELLLA+ regime at BFN. AREVA considers portions of the information provided in Attachment 23 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 24. Therefore, on behalf of AREVA, TVA requests that Attachment 23 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 24 FS1-0029292, Browns Ferry Units 1, 2, and 3 EPU MELLLA+ MCPR Safety Limit Analysis With SAFLIM3D Methodology (non-proprietary) Attachment 25 ANP-3550P, Evaluation of AREVA Fuel Thermal-Hydraulic Performance for Browns Ferry at EPU MELLLA+ (proprietary) Attachment 25 provides the results of thermal-hydraulic analyses that demonstrate that ATRIUM 10XM fuel is hydraulically compatible with ATRIUM-10 fuel for BFN at EPU MELLLA+ conditions. This report also provides the hydraulic characterization of the BFN ATRIUM 10XM and ATRIUM-10 fuel designs. AREVA considers portions of the information provided in Attachment 25 of U. S. Nuclear Regulatory Commission CNL-18-002 Page 8 February 23, 2018 this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 26. Therefore, on behalf of AREVA, TVA requests that Attachment 25 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 26 ANP-3550NP, Evaluation of AREVA Fuel Thermal-Hydraulic Performance for Browns Ferry at EPU MELLLA+ (non-proprietary) Attachment 27 ANP-2860P Revision 2, Supplement 3P, Revision 2, Browns Ferry Unit 1 - Summary of Responses to Request for Additional Information, Extension for Extended Power Flow Operating Domain (proprietary) Attachment 27 provides a review of certain NRC-approved licensing methodologies to demonstrate their applicability to the operation of BFN with AREVA's ATRIUM 10XM fuel in the MELLLA+ operating domain. General applicability of AREVA's licensing methods to BFN is provided in AREVA technical report, ANP-2860P, Revision 2, Browns Ferry Unit 1- Summary of Responses to Request for Additional Information. Applicability of these methods to the ATRIUM 10XM fuel design in the current operating domain is provided in ANP-2860P Revision 2, Supplement 2. AREVA considers portions of the information provided in Attachment 27 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 28. Therefore, on behalf of AREVA, TVA requests that Attachment 27 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 28 ANP-2860NP Revision 2, Supplement 3NP, Revision 2, Browns Ferry Unit 1 - Summary of Responses to Request for Additional Information, Extension for Extended Power Flow Operating Domain (non-proprietary) U. S. Nuclear Regulatory Commission CNL-18-002 Page 9 February 23, 2018 Attachment 29ANP-3572P, Browns Ferry EPU MELLLA+ 10 CFR 50.46c Evaluation (proprietary) Attachment 29 addresses the proposed new criteria for the postulated LOCA in 10 CFR 50.46c. At present these items are not regulatory requirements and therefore are not included in the current licensing basis for BFN. However, because the issue represents potential changes to regulatory requirements that may become effective during the NRC staff's review of this LAR, the potential effect of their adoption into regulatory requirements on the MELLLA+ LAR is being addressed. AREVA considers portions of the information provided in Attachment 29 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 30. Therefore, on behalf of AREVA, TVA requests that Attachment 29 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. Attachment 30 ANP-3572NP, Browns Ferry EPU MELLLA+ 10 CFR 50.46c Evaluation (non-proprietary) Attachment 31 ANP-3633P, Browns Ferry EPU MELLLA+ CRDA Assessment with DG-1327 Criteria (proprietary) Attachment 31 addresses the proposed changes to acceptance criteria for the Boiling Water Reactor Control Rod Drop Accident provided in Draft Regulatory Guide DG-1327, Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents, dated November 2016. At present these criteria are not regulatory requirements and therefore are not included in the current licensing basis for BFN. However, because the issue represents potential changes to regulatory requirements that may become effective during the NRC staff's review of this LAR, the potential effect of their adoption into regulatory requirements on the MELLLA+ LAR is being addressed. AREVA considers portions of the information provided in Attachment 31 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by AREVA, is provided in Attachment 38. A non-proprietary version of the document is provided in Attachment 32. Therefore, on behalf of AREVA, TVA requests that Attachment 31 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390. U. S. Nuclear Regulatory Commission CNL-18-002 Page 10 February 23, 2018 Attachment 32 ANP-3633NP, Browns Ferry EPU MELLLA+ CRDA Assessment with DG-1327 Criteria (non-proprietary) Attachment 33 Basis for Feedwater Temperature Reduction Input Parameter Values for Browns Ferry MELLLA+ ATWS-I Analysis (non-proprietary) Attachment 33 provides the technical basis for the feedwater temperature reduction input parameter values using BFN-specific information. The feedwater temperature reduction rates are used in the Anticipated Transient Without Scram with Instability (ATWS-I) analysis. Attachment 34Steady-state Core Simulator Comparison Using GEH/GNF and AREVA Methods (proprietary) Attachment 34 is a report providing a comparison of the steady state modeling and benchmarking results of the BFN ATRIUM 10XM equilibrium core from the AREVA computer code MICROBURN-B2 and from the GEH computer code PANAC11. GEH consider portions of the information provided in Attachment 34 of this LAR to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by GEH, is provided in Attachment 37. A non-proprietary version of the document is provided in Attachment 35. Therefore, on behalf of GEH, TVA requests that Attachment 34 be withheld from public disclosure in accordance with the GEH affidavit and the provisions of 10 CFR 2.390. Attachment 35 Steady-state Core Simulator Comparison Using GEH/GNF and AREVA Methods (non-proprietary) Attachment 36Differences Between Browns Ferry Nuclear Plant Units Potentially Impacting Maximum Extended Load Line Limit Analysis Plus (MELLLA+) Analyses (non-proprietary) Attachment 36 provides a description of differences between the BFN units that could impact MELLLA+ analyses and evaluations. This attachment also includes a discussion of how the differences were addressed in the BFN MELLLA+ analyses and evaluation to ensure the applicability of these analyses and evaluations to each of the units. Attachment 37 Affidavits - GE-Hitachi Nuclear Energy Americas LLC Attachment 38 Affidavits - AREVA Inc. Attachment 39 Affidavit - Electric Power Research Institute U. S. Nuclear Regulatory Commission CNL-18-002 Page 11 February 23, 2018 TVA has determined that there are no significant hazards considerations associated with the proposed changes and that the renewed facility operating license and TS changes qualify for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91 (b)(1 ), TVA is sending a copy of this letter and the enclosure to the Alabama State Department of Public Health. The BFN Plant Operations Review Committee and the TVA Nuclear Safety Review Board have reviewed this proposed change and determined that operation of BFN in accordance with the proposed change will not endanger the health and safety of the public. TVA requests approval of these proposed renewed facility operating license and TS changes within 26 months of the date of this letter. Once approved, the amendments will be fully implemented within 60 days. This implementation period will provide adequate time for the completion of testing of required modifications supporting the DSS-CD stability solution and revision of the affected station documentation in accordance with the appropriate change control mechanisms. There are no new regulatory commitments associated with this submittal. If there are any questions or if additional information is needed, please contact Mr. Edward D. Schrull at (423) 751-3850. I declare under penalty of perjury that the foregoing is true and correct. Executed on the 23rd day of February 2018. Respectf u I ly, J. W. Shea Vice President, Nuclear Regulatory Affairs and Support Services

Enclosure:

Discussion of Changes, No Significant Hazards Considerations, and Environmental Considerations Attachments: 1. Proposed Renewed Facility Operating License and Technical Specification Changes (Markups) 2. Retyped Proposed Renewed Facility Operating License and Technical Specification Changes 3. Proposed Technical Specification Bases Changes (Markups) 4. Retyped Proposed Technical Specification Bases Changes U. S. Nuclear Regulatory Commission CNL-18-002 Page 12 February 23, 2018 5. NEDC-33877P, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Maximum Extended Load Line Limit Plus (proprietary) 6. NEDO-33877, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Maximum Extended Load Line Limit Plus (non-proprietary) 7. ANP-3551P, AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 (proprietary) 8. ANP-3551NP, AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 (non-proprietary) 9. ANP-3544P, Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design (proprietary) 10. ANP-3544NP, Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design (non-proprietary) 11. ANP-3546P, Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) (proprietary) 12. ANP-3546NP, Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) (non-proprietary) 13. ANP-3547P, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10XM Fuel (EPU MELLLA+) (proprietary) 14. ANP-3547NP, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10XM Fuel (EPU MELLLA+) (non-proprietary) 15. ANP-3548P, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM-10 Fuel (EPU MELLLA+) (proprietary) 16. ANP-3548NP, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM-10 Fuel (EPU MELLLA+) (non-proprietary) 17. ANP-3552P, Browns Ferry Unit 3 Cycle 19 Representative Reload Analysis EPU MELLLA+ (proprietary) 18. ANP-3552NP, Browns Ferry Unit 3 Cycle 19 Representative Reload Analysis EPU MELLLA+ (non-proprietary) 19. ANP-3553P, Browns Ferry Unit 3 Cycle 19 EPU (120% OLTP) MELLLA+ LAR Reference Fuel Cycle Design (proprietary) 20. ANP-3553NP, Browns Ferry Unit 3 Cycle 19 EPU (120% OLTP) MELLLA+ LAR Reference Fuel Cycle Design (non-proprietary) 21. ANP-3568P, Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Extended Power Uprate/MELLLA+ (proprietary) 22. ANP-3568NP, Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Extended Power Uprate/MELLLA+ (non-proprietary) 23. FS1-0029291, Browns Ferry Units 1, 2, and 3 EPU MELLLA+ MCPR Safety Limit Analysis With SAFLIM3D Methodology (proprietary) 24. FS1-0029292, Browns Ferry Units 1, 2, and 3 EPU MELLLA+ MCPR Safety Limit Analysis With SAFLIM3D Methodology (non-proprietary) U. S. Nuclear Regulatory Commission CNL-18-002 Page 13 February 23, 2018 25. ANP-3550P, Evaluation of AREVA Fuel Thermal-Hydraulic Performance for Browns Ferry at EPU MELLLA+(proprietary) 26. ANP-3550NP, Evaluation of AREVA Fuel Thermal-Hydraulic Performance for Browns Ferry at EPU MELLLA+ (non-proprietary) 27. ANP-2860P Revision 2, Supplement 3P, Revision 2, Browns Ferry Unit 1 - Summary of Responses to Request for Additional Information, Extension for Extended Power Flow Operating Domain (proprietary) 28. ANP-2860NP Revision 2, Supplement 3NP, Revision 2, Browns Ferry Unit 1 - Summary of Responses to Request for Additional Information, Extension for Extended Power Flow Operating Domain (non-proprietary) 29. ANP-3572P, Browns Ferry EPU MELLLA+ 10 CFR 50.46c Evaluation (proprietary) 30. ANP-3572NP, Browns Ferry EPU MELLLA+ 10 CFR 50.46c Evaluation (non-proprietary) 31. ANP-3633P, Browns Ferry EPU MELLLA+ CRDA Assessment with DG-1327 Criteria (proprietary) 32. ANP-3633NP, Browns Ferry EPU MELLLA+ CRDA Assessment with DG-1327 Criteria (non-proprietary) 33. Basis for Feedwater Temperature Reduction Input Parameter Values for Browns Ferry MELLLA+ ATWS-I Analysis (non-proprietary) 34. Steady-state Core Simulator Comparison Using GEH/GNF and AREVA Methods (proprietary) 35. Steady-state Core Simulator Comparison Using GEH/GNF and AREVA Methods (non-proprietary) 36. Differences Between Browns Ferry Nuclear Plant Units Potentially Impacting Maximum Extended Load Line Limit Analysis Plus (MELLLA+) Analyses (non-proprietary) 37. Affidavits - GE-Hitachi Nuclear Energy Americas LLC 38. Affidavits - AREVA Inc. 39. Affidavit - Electric Power Research Institute cc (Enclosure): NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant State Health Officer, Alabama State Department of Public Health (w/o Attachments 5, 7, 9, 11, 13, 15, 17, 19, 21, 23, 25, 27, 29, 31, and 34)

Evaluation of Proposed Change

SUBJECT:

Maximum Extended Load Line Limit Analysis Plus 2.1 Background.............................................................................................................. 4 2.2 Renewed Facility Operating License and Technical Specification Changes ............. 5 2.3 Stability Solution ...................................................................................................... 5 2.4 Technical Specification Task Force (TSTF)-493 Applicability ................................... 6 2.5 Emergency Core Cooling System Net Positive Suction Head Evaluation ................. 6 3.1 Renewed Facility Operating License and Technical Specification Changes ............. 6 4.1 Applicable Regulatory Requirements/Criteria ......................................................... 14 4.2 Precedent .............................................................................................................. 15 4.3 Significant Hazards Consideration ......................................................................... 16 4.4 Conclusion ............................................................................................................. 17 Enclosure Page 2 of 20 Pursuant to 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," the Tennessee Valley Authority (TVA) proposes to revise the Browns Ferry Nuclear Plant (BFN), Units 1, 2 and 3 Renewed Facility Operating License (RFOL) and Technical Specifications (TS) to allow plant operation in the expanded Maximum Extended Load Line Limit Analysis Plus (MELLLA+) domain with the Detect and Suppress Solution - Confirmation Density (DSS-CD) long-term reactor core thermal-hydraulic stability solution. Proposed RFOL and TS changes include: Prohibiting operation in the MELLLA+ domain when operating in the following plant configurations: Reactor Recirculation System Single Loop Operation (SLO). A feedwater heater out of service resulting in more than a 10°F reduction in feedwater temperature below the design feedwater temperature. Replacing TS 3.3.1.1 Conditions/Required Actions/Completion Times for Conditions I and J with new Conditions/Required Actions/Completion Times I, J, and K, due to implementation of DSS-CD stability solution. Eliminating Surveillance Requirement (SR) 3.3.1.1.17, which is no longer required by the proposed DSS-CD stability solution. Revising TS Table 3.3.1.1-1 Function 2.b to change the Allowable Value for APRM Flow Biased Simulated Thermal Power - High trip function and to add a new note to implement Automated Backup Stability Region setpoints when Function 2.f is inoperable. Revising TS Table 3.3.1.1-1 Function 2.f to set the new operability power level for OPRM Upscale and to add a new note due to implementation of the DSS-CD stability solution. Adding new Condition B and corresponding Required Action B.1 and a Completion Time to TS 3.4.1 to reflect the fact that SLO is prohibited in the MELLLA+ region and to require immediate action required when the MELLLA+ domain is entered with one recirculation loop in operation. The previous Condition B regarding no recirculation loops in operation is re-designated as Condition C. Revising Administrative Controls TS 5.6.5.a to require certain content in the Core Operating Limits Report (COLR) and updating the applicable references in Subsection 5.6.5.b. due to implementation of DSS-CD stability solution. Adding new Administrative TS 5.6.7, which specifies the contents of a new report required by new TS 3.3.1.1 Required Action I.3. This MELLLA+ License Amendment Request (LAR) is based on the following General Electric - Hitachi (GEH) Licensing Topical Reports (LTRs): Enclosure Page 3 of 20 NEDC-33006P-A, Revision 3, Maximum Extended Load Line Limit Analysis Plus (M+LTR) (Reference 1). NEDC-33075P-A, Revision 8, General Electric Boiling Water Reactor Detect and Suppress Solution - Confirmation Density (DSS-CD LTR) (Reference 2). NEDC-33173P-A, Revision 4, Applicability of GE Methods to Expanded Operating Domains (Methods LTR) (Reference 3). The marked-up RFOL and TS pages for the proposed changes are provided in Attachment 1. Retyped RFOL and TS pages with the proposed changes incorporated are provided in Attachment 2. Associated proposed changes to the TS Bases are provided for information only in Attachment 3 (markups) and Attachment 4 (retyped pages). Attachment 5 contains the proprietary version of GEH Report NEDC-33877P, Revision 0, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 - Maximum Extended Load Line Limit Analysis Plus (M+SAR) and Attachment 6 provides the non-proprietary version of this report. The M+SAR provides the technical bases for this request and contains an integrated summary of the results of the underlying safety analyses and evaluations performed specifically for BFN to justify the expansion of the core flow operating domain and use of the DSS-CD stability solution. The M+SAR follows the guidelines contained in the M+LTR (Reference 1). The NRC has specified limitations and conditions for the Methods LTR (Reference 3), M+LTR (Reference 1), the DSS-CD LTR (Reference 2). These are addressed in Appendices A, B, and C of the M+SAR, respectively. BFN currently operates with AREVA fuel. As such, Attachment 7 contains the proprietary version of AREVA Report ANP-3551P, AREVA MELLLA+ Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 (AMSAR) and Attachment 8 provides the non-proprietary version of this report. This reports supplements the GEH M+SAR and summarizes the results of fuel-related safety analysis and evaluations performed specifically for BFN to justify the expansion of the core flow operating domain and use of the DSS-CD stability solution for AREVA fuel. The AMSAR is also supported by the following fuel-related reports, proprietary and non-proprietary versions, where applicable, that are included in Attachments 9 through 32: ANP-3546, Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) ANP-3547, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10XM Fuel (EPU MELLLA+) ANP-3548, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10 Fuel (EPU MELLLA+) ANP-3544 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3553, Browns Ferry Unit 3 Cycle 19 EPU (120% OLTP) MELLLA+LAR Reference Fuel Cycle Design Enclosure Page 4 of 20 ANP-3552, Browns Ferry Unit 3 Cycle 19 Representative Reload Analysis at EPU MELLLA+ ANP-3568, Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Extended Power Uprate/MELLLA+ ANP-3550, Evaluation of AREVA Fuel Thermal-Hydraulic Performance for Browns Ferry at EPU MELLLA+ FS1-0029291/2, Browns Ferry Units 1, 2, and 3 EPU MELLLA+ MCPR Safety Limit Analysis With SAFLIM3D Methodology ANP-2860 Revision 2, Supplement 3, Browns Ferry Unit 1 - Summary of Responses to Request for Additional Information, Extension for Extended Power Flow Operating Domain ANP-3572, Browns Ferry EPU MELLLA+ 10 CFR 50.46c Evaluation ANP-3633, Browns Ferry EPU MELLLA+ CRDA Assessment with DG-1327 Criteria The specific proposed RFOL and TS changes (described in Section 3.1, below) are consistent with the M+LTR and DSS-CD LTR. 2.1 Background BFN Units 1, 2 and 3 were each originally licensed to operate at a maximum power level of 3293 MWt. A rerate implemented in 1998 for BFN Units 2 and 3 and 2007 for BFN Unit 1 increased the licensed thermal power by approximately 5% to 3458 MWt. An Extended Power Uprate (EPU) approved on August 14, 2017, (Reference 4) increased the maximum power level to the Current Licensed Thermal Power (CLTP) of 3952 MWt. Operation of BWRs requires that reactivity balance be maintained to accommodate fuel burnup. BWR operators have typically two methods to maintain this reactivity balance that include (1) control rod movements, and (2) reactor recirculation core flow adjustments. Because of strong void reactivity feedback and its distributed effect through the reactor core, recirculation flow adjustments are the preferred reactivity control method. Operating at low core flow conditions at rated power level also increases the fuel capacity factor through spectral shift. In addition, an increased flow region compensates for reactivity reduction due to fuel depletion during the operating cycle. EPUs are implemented by extending the MELLLA operating domain up to EPU rated thermal power (RTP) levels. However, this reduces the available core flow window at these levels. In addition, the increased core pressure drop limits recirculation flow capability. Consequently, EPU plants generally operate with a greatly reduced core flow window and compensate for reactivity loss with control rod movement. Enclosure Page 5 of 20 MELLLA+ increases the operating boundary to permit BFN operation at a CLTP of 3952 MWt with a core flow as low as 85% thus adding a 14% flow-control window. The entire MELLLA+ domain is shown in Figure 1-1 of the M+SAR (Attachment 5). By operating in the MELLLA+ domain, a significantly lesser number of control rod manipulations are required than is currently required in the present operating domain. Lessening the number of rod manipulations represents a significant improvement in operating flexibility as well as providing safer plant operation. Specifically, this minimizes the likelihood of fuel failures and reduces the likelihood of events initiated by reactor maneuvers required to achieve an operating condition where control rods can be withdrawn. The MELLLA+ core operating domain expansion requires minor plant system modifications. It involves changes to the operating power/core flow map, changes to a small number of instrument setpoints included in the proposed TS changes, and application of the DSS-CD stability solution. 2.2 Renewed Facility Operating License and Technical Specification Changes The following Renewed Facility Operating License (RFOL) and Technical Specifications (TS) are affected by this change: RFOL Section 2.C added new Item (21) for BFN Units 1 and 2 and new item (17) for BFN Unit 3 to limit feedwater temperature reductions as a result of a feedwater heater being out of service. TS 3.3.1.1, Reactor Protection System (RPS) Instrumentation. TS 3.4.1, Recirculation Loops Operating. TS 5.6.5, Core Operating Limits Report (COLR). TS 5.6.7, Oscillation Power Range Monitor (OPRM) Report (new section). Section 3.1 of this Enclosure provides the details of the above changes along with the associated technical justification. 2.3 Stability Solution BFN currently operates with the Option III stability solution. The NRC Safety Evaluation for the M+LTR (Reference 1) requires that plants implementing MELLLA+ change from the Option III solution to the DSS-CD stability solution. The DSS-CD solution being implemented at BFN is designed to identify the power oscillation upon inception and initiate control rod insertion (scram) to terminate the oscillations prior to any significant amplitude growth. DSS-CD is based on the same hardware design as Option Ill. However, it introduces an enhanced detection algorithm that detects the inception of power oscillations and generates an early power suppression trip signal based on successive period confirmation recognition and an amplitude component. The existing Option III algorithms are retained to provide defense-in-depth protection for unanticipated reactor instability events. Enclosure Page 6 of 20 The DSS-CD solution algorithm, licensing basis, and application procedures are generically described in NEDC-33075P-A (Reference 2) and are applicable to BFN, including any limitations and conditions associated with their use and approval. To support MELLLA+ operation at BFN, the Power Range Neutron Monitoring System will be upgraded with the DSS-CD solution. The associated modifications will be installed prior to implementing MELLLA+ on the respective BFN unit. 2.4 Technical Specification Task Force (TSTF)-493 Applicability This LAR proposes changes to allowable values (AVs) for two TS Reactor Protection System (RPS) functions: (1) the Oscillation Power Range Monitor (OPRM) Upscale; and (2) the Average Power Range Monitor (APRM) Flow Biased Simulated Thermal Power - High. TSTF-493 clarifies the application of setpoint methodology for Limiting Safety System Setting (LSSS) functions in TS. The OPRM setpoints are unique to a particular core design for a particular fuel cycle. As described in the proposed TS Bases, the OPRM function setpoints do not have traditional TS AVs. The OPRM Upscale Function is not LSSS Safety Limit (SL)-related (NEDC-33075P-A). OPRM setpoints are discussed in more detail in the M+SAR Section 2.4.1. As such, application of TSTF-493 is not required for OPRM function. The APRM Flow Biased Simulated Thermal Power - High function is not credited in the accident or transient analysis for BFN. The proposed calculated Allowable Value follows the methodology that the NRC approved by Amendment Nos. 257, 296, and 254 for BFN Units 1, 2, and 3, respectively, dated September 14, 2006. (Reference 5) (Refer to Attachment 5, M+SAR Section 5.3.1.) This setpoint methodology is unchanged from that used in the Extended Power Uprate LAR approved by the NRC in Reference 4. As such, application of the TSTF-493 is not required for the APRM Flow Biased Simulated Thermal Power - High function. 2.5 Emergency Core Cooling System Net Positive Suction Head Evaluation As approved in the BFN Extended Power Uprate License Amendments (Reference 4), BFN does not rely on containment accident pressure (CAP) credit to demonstrate adequate net positive suction head (NPSH) for the Emergency Core Cooling System (ECCS) pumps. For MELLLA+, the ECCS NPSH evaluations continue to demonstrate that CAP credit is not required. Refer to Attachment 5, M+SAR Section 4.2.6 for additional information regarding the ECCS NPSH evaluations performed for MELLLA+. 3.1 Renewed Facility Operating License and Technical Specification Changes The technical analyses and justifications for the proposed changes are provided in the M+SAR (Attachment 5) and AMSAR (Attachment 7). The M+SAR and AMSAR summarize the results of the significant safety evaluations performed that justify: (1) Implementing the MELLLA+ expanded operating domain; Enclosure Page 7 of 20 (2) Changing the BFN stability solution from Option III to DSS-CD; and (3) Applying the GEH TRACG04 analysis code to DSS-CD and the ATWS analysis. The evaluations contained in the M+SAR and AMSAR demonstrate that BFN can safely operate in the MELLLA+ expanded operating domain using the DSS-CD stability solution. The DSS-CD stability solution is required by the SER for the M+LTR (Reference 1) and is being implemented at BFN using the guidelines contained in the DSS-CD LTR (Reference 2). The use of TRACG04 is being implemented using the guidelines contained in the TRACG LTR (Reference 6). The results of the DSS-CD evaluation and the use of TRACG04 are provided in M+SAR Section 2.4. Table 1, below, identifies the RFOL and TS sections being changed, a description of the change, the Attachment 5, M+SAR, section that justifies the change, and any pertinent comments. RFOL/TS Section Description of Change M+SAR Section Comments RFOL Section 2.C 2.C (21) for Units 1 and 2 (new) 2.C (17) for Unit 3 (new) Implement the following new license condition that prohibits operation of the facility within the MELLLA+ operating domain with a feedwater heater out of service resulting in more than a 10°F reduction in feedwater temperature below the design feedwater temperature. Maximum Extended Load Line Limit Analysis Plus (MELLLA+) Special Consideration The licensee shall not operate the facility within the MELLLA+ operating domain with a feedwater heater out of service resulting in more than a 10°F reduction in feedwater temperature below the design feedwater temperature of 394.5°F. 1.2.4 2.4.4 The NRC has stated that any plant-specific application intending to operate with FWHOOS must provide the bases in the plant-specific application (Reference 1, page B-39). The bases for the proposed license condition are provided in M+SAR Sections 1.2.4 and 2.4.4 which satisfies the M+LTR Safety Evaluation Report Limitation and Condition 12.5.b). Prohibiting a FW temperature reduction of greater than 10°F includes prohibiting use of the Enclosure Page 8 of 20 RFOL/TS Section Description of Change M+SAR Section Comments FWHOOS and FFWTR operating flexibility options. TS 3.3.1.1, Reactor Protection System (RPS) Instrumentation TS 3.3.1.1 Condition I Replace Required Action I.1 (i.e., Initiate alternate method to detect and suppress thermal hydraulic instability oscillations (Completion Time: Immediately)) with the following new Required Actions I.1, I.2, and I.3 to implement DSS-CD for the Oscillation Power Range Monitor (OPRM). REQUIRED ACTION I.1 Initiate action to implement the Manual Backup Stability Protection (BSP) Regions defined in the COLR. (Completion Time: Immediately) AND I.2 Implement the Automated BSP Scram Region using the modified APRM Flow Biased Simulated Thermal Power - High scram setpoints defined in the COLR. (Completion Time: 12 hours) AND I.3 Initiate action to submit an OPRM report in accordance with Specification 5.6.7. (Completion Time: Immediately) 2.4.3 This change is discussed in Sections 7.2 and 7.4, and Table 8-1 of the DSS-CD LTR (Reference 2). TS 3.3.1.1 Condition J Replace Required Action J.1 (i.e., Be in Mode 2 (Completion Time: 4 hours)) with the following new Required Actions J.1, J.2, and J.3 to implement DSS-CD for the OPRM. 2.4.3 This change is discussed in Sections 7.2 and 7.3, and Table 8-1 of the DSS-CD LTR (Reference 2). Enclosure Page 9 of 20 RFOL/TS Section Description of Change M+SAR Section Comments REQUIRED ACTION J.1 Initiate action to implement the Manual BSP Regions defined in the COLR. (Completion Time: Immediately) AND J.2 Reduce operation to below the BSP Boundary defined in the COLR. (Completion Time: 12 hours) AND J.3 ------------NOTE------------------ LCO 3.0.4 is not applicable. -------------------------------------- Restore required channel to OPERABLE status. (Completion Time: 120 days) TS 3.3.1.1 Condition K (new) Insert the following new Condition K and Required Action K.1 to implement DSS-CD for the OPRM. CONDITION K. Required Action and associated Completion Time of Condition J not met. REQUIRED ACTION K.1 Reduce THERMAL POWER to < 18% RTP. (Completion Time: 4 hours) 2.4.2 Section 3.5 of the DSS-CD LTR (Reference 2) requires DSS-CD to be operable above a power level set at 5% below the lower boundary of the Armed Region defined by the MCPR monitoring threshold level, which is 23% for BFN (see M+SAR Section 2.4.2). Therefore, DSS-CD must be operable at 18% (23% - 5%). This change is also specified in Table 8-1 of the DSS-CD LTR. Enclosure Page 10 of 20 RFOL/TS Section Description of Change M+SAR Section Comments SR 3.3.1.1.17 Delete SR 3.1.1.1.17 (i.e., Verify OPRM is not bypassed when APRM Simulated Thermal Power is 23% and recirculation drive flow is < 60% of rated recirculation flow (Frequency: 24 months)). The DSS-CD automatically arms, therefore, this surveillance requirement (which verifies the OPRM is not bypassed) is no longer necessary and is being deleted. 2.4.1 Deleting this SR is specified in Table 8-1 of the DSS-CD LTR. Table 3.3.1.1-1, Reactor Protection System Instrumentation Function 2.b Revise the Allowable Value (AV) of the Average Power Range Monitors (APRM) Flow Biased Simulated Thermal Power - High + 65.5% RTP" to ".3% RTP." 5.3.1 The change is made to maintain the margin between the operating domain and the current trip during two loop operation. The calculated value follows the methodology that the NRC approved by Amendment Nos. 257, 296, and 254 for BFN Units 1, 2, and 3, respectively, dated September 14, 2006. (Reference 5) This setpoint methodology is unchanged from that used in the Extended Power Uprate LAR approved by the NRC in Reference 4. Function 2.b Add the following new Note (e) in the "Allowable Value" column due to implementation of DSS-CD. (e) With OPRM Upscale 2.4.3 Section 7.4 of the DSS-CD LTR discusses this action. See also Table 8-1 of the DSS-CD LTR. Enclosure Page 11 of 20 RFOL/TS Section Description of Change M+SAR Section Comments (Function 2.f) inoperable, the Automated BSP Scram Region setpoints are implemented in accordance with Action I of this Specification. Function 2.f Change the entry in the Applicable Modes or Other Specified Conditions column from Mode "1 18%." 2.4.2 Section 3.5 of the DSS-CD LTR (Reference 2) requires DSS-CD to be operable above a power level set at 5% below the lower boundary of the Armed Region defined by the MCPR threshold power level, which is 23% for BFN (see M+SAR Section 2.4.2). Therefore, DSS-CD must be operable at 18% (23% - 5%). This change is also specified in Table 8-1 of the DSS-CD LTR. Function 2.f Add the following new Note (f) in the Applicable Modes or Other Specified Conditions column due to implementation of DSS-CD. (f) Following Detect and Suppress Solution - Confirmation Density (DSS-CD) implementation, DSS-CD is not required to be armed while in the DSS-CD Armed Region during the first reactor startup and during the first controlled shutdown that passes completely through the DSS-CD Armed Region. However, DSS-CD is considered 2.4 This change is discussed in Sections 2.2 and 7.4, and Table 8-1 of the DSS-CD LTR. Enclosure Page 12 of 20 RFOL/TS Section Description of Change M+SAR Section Comments OPERABLE and shall be maintained OPERABLE and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region. Function 2.f Delete reference to SR 3.3.1.1.17 from the Surveillance Requirements column due to implementation of DSS-CD. 2.4 This change is discussed in Table 8-1 of the DSS-CD LTR. Function 2.f Delete Note (e) (i.e., Refer to COLR for OPRM period based detection algorithm (PBDA) setpoint limits) from the Allowable Value column due to implementation of DSS-CD. 2.4 This change is discussed in Table 8-1 of the DSS-CD LTR. TS 3.4.1, Recirculation Loops Operating LCO 3.4.1 Note (new) Add the following note to the LCO to prohibit operation in the MELLLA+ operating domain with a single recirculation loop in operation. -------------------NOTE---------------------- Single recirculation loop operation is prohibited in the MELLLA+ operating domain. ------------------------------------------------- 1.2.4 SLO is prohibited while operating in the MELLLA+ operating domain in accordance with Limitation and Condition 12.5.a of the M+LTR safety evaluation (Reference 1). TS 3.4.1 Condition B Add the following new Condition B with corresponding Required Action B.1 and Completion Time to reflect the immediate action required to exit the MELLLA+ operating domain if a single recirculation loop is in operation. The previous Condition B is also re-designated as Condition C and the reference to "Condition A" is changed to "Condition A or B." CONDITION 1.2.4 SLO is prohibited while operating in the MELLLA+ operating domain in accordance with Limitation and Condition 12.5.a of the M+LTR safety evaluation (Reference 1). Enclosure Page 13 of 20 RFOL/TS Section Description of Change M+SAR Section Comments B. Operation in the MELLLA+ operating domain with a single recirculation loop in operation. REQUIRED ACTION B.1 Initiate action to exit the MELLLA+ operating domain. (Completion Time: Immediately) TS 5.6.5, Core Operating Limits Report TS 5.6.5.a To reflect implementation of DSS-CD, replace existing Subsection a.(4) (i.e., The period based detection algorithm (PBDA) setpoint for Function 2.f, Oscillation Power Range Monitor (OPRM) Upscale, for Specification 3.3.1.1) with the following new Subsection a.(4). (4) The Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Flow Biased Simulated Thermal Power-High scram setpoints used in the Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1 2.4.3 Section 7.5 of the DSS-CD LTR discusses and justifies this addition. See also Table 8-1 of the DSS-CD LTR. TS 5.6.5.b Add reference, in Subsection b, to NEDC-33075P-A, Revision 8, due to implementation of DSS-CD. NEDC-33075P-A, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, Revision 8, November 2013. 2.4.1.1 NEDC-33075P-A Revision 8 provides the generic licensing basis for DSS-CD applications. Enclosure Page 14 of 20 RFOL/TS Section Description of Change M+SAR Section Comments TS 5.6.7, Oscillation Power Range Monitoring (OPRM) Report (new) TS 5.6.7 Add the following new Section 5.6.7. 5.6.7 OPRM Report When an OPRM report is required by Condition I of LCO 3.3.1.1, "RPS Instrumentation," the report shall be submitted within the following 90 days. The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to OPERABLE status. 2.4.3 This change is discussed in Table 8-1 of the DSS-CD LTR. 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c)(2)(ii), Criterion 2, requires TS Limiting Conditions for Operations (LCO) include process variables, design features, and operating restrictions that are initial conditions of design basis accident analysis. Compliance with TS ensures that system performance parameters are maintained within the values assumed in the safety analyses. The proposed RFOL and TS changes are supported by the safety analyses and continue to provide a level of protection comparable to the current TS. Applicable regulatory requirements and significant safety evaluations performed in support of the proposed changes are described in the M+SAR (Attachment 5), the AMSAR (Attachment 7), and supporting fuel-related reports (Attachments 9 through 32). Implementation of MELLLA+ does not require (1) an increase in the current maximum normal operating reactor dome pressure, (2) an increase in core power, (3) an increase in the maximum licensed core flow, (4) a change to source term methodology, (5) a new fuel product line, or (6) a change in fuel cycle length. As such, the impact on plant operation is minimal and, as demonstrated in the M+SAR, the AMSAR, and supporting fuel-related reports, the MELLLA+ operating domain expansion can be accomplished without exceeding any existing regulatory limits or design allowable limits applicable to BFN. Enclosure Page 15 of 20 As part of MELLLA+ implementation for BFN, TVA will also implement the DSS-CD approach to automatically detect and suppress thermal hydraulic instability. Since the DSS-CD approach will continue to provide reliable, automatic detection and suppression of stability related power oscillations and provide protection against violation of the Safety Limit Minimum Critical Power Ratio for anticipated oscillations, compliance with GDC 10 and 12 of 10 CFR 50, Appendix A is maintained. This LAR is not being submitted as a risk informed licensing action, as defined by Regulatory Guide (RG) 1.174 (Reference 7). However, it was evaluated from a risk perspective using the RG 1.174 criteria and, as demonstrated in Section 10.5, Individual Plant Evaluation, of the M+SAR (Attachment 5), there is no increase in core damage frequency (CDF) or large early release frequency (LERF). 4.2 Precedent This LAR is being submitted using the following NRC-approved General Electric-Hitachi (GE-H) Licensing Topical Reports (LTR): NEDC-33006P-A, (M+ LTR), Revision 3 and its associated safety evaluation report (i.e., Reference 1) NEDC-33075P-A, (DSS-CD LTR), Revision 8 and its associated safety evaluation report (i.e., Reference 2) NEDC-33173P-A, (Methods LTR), Revision 4 and its associated safety evaluation report (i.e., Reference 3) The applicability of AREVA methods to the MELLLA+ operating domain is addressed in AREVA NP Report ANP-2860P, Revision 2, Supplement 3, Browns Ferry Unit 1 - Summary of Responses to Request for Additional Information, Extension for Extended Power Flow Operating Domain (MELLLA+ LAR Attachment 27). The AREVA computer codes, including AREVA's NRC-approved LTRs, are specified in Table 1-1 of the AMSAR (MELLLA+ LAR Attachment 7). March 28, 2014 - The NRC issued Amendment 180 to the Renewed Facility Operating License for the Monticello Nuclear Generating Plant allowing operation in the expanded MELLLA+ operating domain and changing the stability solution from Option III to DSS-CD (ADAMS Accession No. ML 14035A248). August 31, 2015 - The NRC issued Amendment 205 to the Facility Operating License for the Grand Gulf Nuclear Station allowing operation in the expanded MELLLA+ operating domain and changing the stability solution from Option III to DSS-CD (ADAMS Accession No. ML 15229A219). September 2, 2015 - The NRC issued Amendment 151 to the Facility Operating License for the Nine Mile Point Nuclear Station, Unit No. 2 allowing operation in the expanded MELLLA+ operating domain, changing the stability solution from Option III to DSS-CD, and to increase the isotopic enrichment of boron-10 in the sodium pentaborate solution utilized in the Standby Liquid Control System (ADAMS Accession No. ML 15096A076). Enclosure Page 16 of 20 March 21, 2016 - The NRC issued Amendments 305 and 309 to the Renewed Facility Operating Licenses for the Peach Bottom Atomic Power Station, Units 2 and 3, allowing operation in the expanded MELLLA+ operating domain and changing the stability solution from Option III to DSS-CD (ADAMS Accession No. ML 16034A372). In addition, while not yet approved by the NRC, the Brunswick Steam Electric Plant (BSEP) MELLLA+ LAR(ADAMS Accession No. ML16257A410) has been submitted and is currently under review by the NRC. TVA has been closely monitoring the NRC review of the BSEP MELLLA+ LAR to ensure issues identified during the review are addressed in the BFN MELLLA+ LAR. 4.3 Significant Hazards Consideration The Tennessee Valley Authority (TVA) is submitting an amendment request to Renewed Facility Operating Licenses DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3, respectively. This license amendment request proposes revisions to the BFN Renewed Facility Operating Licenses and Technical Specifications to allow operating in the expanded MELLLA+ operating domain. TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: Response: No. The proposed operation in the MELLLA+ operating domain does not significantly increase the probability or consequences of an accident previously evaluated. The probability (frequency of occurrence) of Design Basis Accidents (DBAs) occurring is not affected by the MELLLA+ operating domain because BFN continues to comply with the regulatory and design basis criteria established for plant equipment. There is no change in consequences of postulated accidents when operating in the MELLLA+ operating domain compared to the operating domain previously evaluated. The results of accident evaluations remain within the NRC approved acceptance limits. The spectrum of postulated transients has been investigated and is shown to meet the plant's currently licensed regulatory criteria. Continued compliance with the Safety Limit Minimum Critical Power Ratio (SLMCPR) will be confirmed on a cycle-specific basis consistent with the criteria accepted by the NRC. Challenges to the reactor coolant pressure boundary were evaluated for the MELLLA+ operating domain conditions (pressure, temperature, flow, and radiation) and were found to meet their acceptance criteria for allowable stresses and overpressure margin. Challenges to the containment were evaluated and the containment and its associated cooling systems continue to meet the current licensing basis. The Enclosure Page 17 of 20 calculated post-Loss-of-Coolant Accident (LOCA) suppression pool temperature remains acceptable. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. Response: No. The proposed operation in the MELLLA+ operating domain does not create the possibility of a new or different kind of accident from any previously evaluated. Equipment that could be affected by the MELLLA+ operating domain has been evaluated. No new operating mode, safety-related equipment lineup, accident scenario, or equipment failure mode was identified. The full spectrum of accident considerations has been evaluated and no new or different kind of accident has been identified. The MELLLA+ operating domain uses developed technology, and applies it within the capabilities of existing plant safety-related equipment in accordance with the regulatory criteria (including NRC-approved codes, standards and methods). No new accident or event precursor has been identified. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Response: No. The proposed operation in the MELLLA+ operating domain does not involve a significant reduction in the margin of safety. The MELLLA+ operating domain affects only design and operational margins. Challenges to the fuel, reactor coolant pressure boundary, and containment were evaluated for the MELLLA+ operating domain conditions. Fuel integrity is maintained by meeting existing design and regulatory limits. The calculated loads on affected structures, systems, and components, including the reactor coolant pressure boundary, will remain within their design allowables for design basis event categories. No NRC acceptance criterion is exceeded. The BFN configuration and responses to transients and postulated accidents do not result in exceeding the presently approved NRC acceptance limits. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, TVA concludes that the proposed amendments do not involve a significant hazard consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified. 4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is a reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Enclosure Page 18 of 20 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. The environmental effects of MELLLA+ operating domain expansion are controlled at the same limits as the current analyses. None of the present limits for plant environmental releases are increased as a result of MELLLA+ operating domain expansion. MELLLA+ has no effect on the non-radiological elements of concern, and the plant will be operated in accordance with applicable environmental requirements as documented by the Environmental Assessment for BFN's current licensed operating domain. Existing Federal, State and local regulatory permits presently in effect accommodate the MELLLA+ operating domain expansion without modification. The evaluation of the effects of MELLLA+ operating domain expansion on normal radiological effluents is included in Section 8.0 of the M+SAR (Attachment 5). There will be no change in the radiological effluents released to the environment due to the MELLLA+ operating domain expansion. The normal effluents and doses remain well within the 10 CFR 20 limits and the 10 CFR 50, Appendix I guidance. There is no change to the predicted doses from postulated accidents and the 10 CFR 50.67 dose criteria continue to be met. In addition, the quantity of spent fuel does not increase as a result of MELLLA+ operating domain expansion. A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, and would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration (as discussed in Section 4.3 of this BFN MELLLA+ LAR Enclosure), (ii) a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite (as discussed in Sections 8.6 and 9.2 of BFN MELLLA+ LAR Attachments 5 and 6), or (iii) a significant increase in individual or cumulative occupational radiation exposure (as discussed in Section 8.5 of BFN MELLLA+ LAR Attachments 5 and 6). Accordingly, the proposed license amendment meets the eligibility criterion for categorical exclusion as set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed license amendment. 1. GE Nuclear Energy Report NEDC-33006P-A, Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report, Revision 3, dated June 2009. 2. GE-Hitachi Nuclear Energy Report NEDC-33075P-A, Detect And Suppress Solution - Confirmation Density Licensing Topical Report, Revision 8, dated November 2013. 3. GE-Hitachi Nuclear Energy Report NEDC-33173P-A, Applicability of GE Methods to Expanded Operating Domains Licensing Topical Report, Revision 4, dated November 2012. Enclosure Page 19 of 20 4. NRC Letter to TVA, "Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Issuance of Amendments Regarding Extended Power Uprate (CAC Nos. MF6741, MF6742, and MF6743)," dated August 14, 2017; (ADAMS Accession No. ML17032A120). 5. NRC Letter to TVA, "Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Issuance of Amendments Regarding The Instrument Setpoint Program (TAC Nos. MC9518, MC9519, and MC9520) (TS-453)," dated September 14, 2006; (ADAMS Accession No. ML061680008). 6. GE-Hitachi Nuclear Energy Report NEDE-33147P-A, DSS-CD TRACG Application, (TRACG LTR), Revision 4, August 2013. 7. Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2, May 2011.

-6e-(h) The results of the visual inspections of the steam dryer shall be submitted to the NRC staff in a report in accordance with 10 CFR 50.4. The report shall be submitted to the NRC within 90 days following startup from each of the first two respective refueling outages. (i) Within 6 months following completion of the second refueling outage, after the implementation of the EPU, the licensee shall submit a long-term steam dryer inspection plan based on industry operating experience along with the baseline inspection results. The license condition described above shall expire: (1) upon satisfaction of the requirements in items (g) and (h), provided that a visual inspection of the steam dryer does not reveal any new unacceptable flaw(s) or unacceptable flaw growth that is due to fatigue, and; (2) upon satisfaction of the requirements specified in item (i). (19) Neutron Absorber Monitoring Program The licensee shall, at least once every ten years, withdraw a neutron absorber coupon from the spent fuel pool and perform Boron-10 (B-10) areal density measurement on the coupon. Based on the results of the B-1 O areal density measurement, the licensee shall perform any technical evaluations that may be necessary and take appropriate actions using relevant regulatory and licensing processes. (20) Radiological Consequences Analyses Using Alternative Source Terms TVA shall perform facility and licensing basis modifications to resolve the non-conforming/degraded condition associated with the Alternate Leakage Treatment pathway such that the current licensing basis dose calculations (approved in License Amendment Nos. 251/282 (Unit 1), 290/308 (Unit 2) and 249/267 (Unit 3)) would remain valid. These facility and licensing basis modifications shall be complete prior to initial power ascension above 3458 MWt. D. The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21 (d), shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71(e)(4) following the issuance of this renewed operating license. Until that update is complete, TVA may make changes to the programs and activities described in the supplement without prior Commission approval, provided that TVA evaluates such changes pursuant to the criteria set forth in 1 O CFR 50.59 and otherwise complies with the requirements in that section. E. The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. TVA shall complete these activities no later than December 20, 2013, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. BFN-UNIT 1 Renewed License No. DPR-33 Amendment No. 299 Insert LC for BFN Unit 1 (21) Maximum Extended Load Line Limit Analysis Plus (MELLLA+) Special Consideration The licensee shall not operate the facility within the MELLLA+ operating domain with a feedwater heater out of service resulting in more than a 10°F reduction in feedwater temperature below the design feedwater temperature. RPS Instrumentation 3.3.1.1 BFN-UNIT 1 3.3-2a Amendment No.266 December 29, 2006 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME I. As required by Required Action D.1 and referenced in Table 3.3.1.1-1. I.1 Initiate alternate method to detect and suppress thermal hydraulic instability oscillations. 12 hours J. Required Action and associated Completion Time of Condition I not met. J.1 Be in Mode 2. 4 hours TS 3.3.1.1 Insert A I. As required by Required Action D.1 and referenced in Table 3.3.1.1-1 I.1 Initiate action to implement the Manual Backup Stability Protection (BSP Regions defined in the COLR. AND I.2 Implement the Automated BSP Scram Region using the modified APRM Flow Biased Simulated Thermal Power-High scram setpoints defined in the COLR. AND I.3 Initiate action to submit an OPRM report in accordance with Specification 5.6.7. Immediately 12 hours Immediately J. Required Action and associated Completion Time of Condition I not met. J.1 Initiate action to implement the Manual BSP Regions defined in the COLR. AND J.2 Reduce operation to below the BSP Boundary defined in the COLR. AND J.3 ---------------NOTE---------------- LCO 3.0.4 is not applicable. ---------------------------------------- Restore required channel to OPERABLE status. Immediately 12 hours 120 days K. Required Actions and associated Completion Time of Condition J not met. K.1 Reduce THERMAL POWER to < 18% RTP. 4 hours SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE SR 3.3.1.1.10 Perform* CHANNEL CALIBRATION. SR 3.3.1.1.11 (Deleted) SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. SR 3.3.1.1.13 --------------------------NOTE-------------------------Neutron detectors are excluded. ------------------------------------------------------------Perform CHANNEL CALIBRATION. SR 3.3.1.1.14 Perform LOGIC SYSTEM FUNCTIONAL TEST. SR 3.3.1.1.15 Verify Turbine Stop Valve -Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions are not bypassed when THERMAL POWER 26% RTP. SR 3.3.1.1.16 --------------------------NOTE-------------------------For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. ------------------------------------------------------------Perform CHANNEL FUNCTIONAL TEST. SR 3.3.1.1.17 Verify OPRM is not bypassed when APRM Simulated Thermal Power is 23% and recirculation drive flow is < 60% of rated recirculation drive flow. RPS Instrumentation 3.3.1.1 FREQUENCY 184 days 24 months 24 months 24 months 24 months 184 days 24 months BFN-UNIT 1 3.3-5 Amendment No. 234, 262, 263, 266, 299 RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3) Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1 1. Intermediate Range Monitors a. Neutron Flux -High 2 3 G SR 3.3.1.1.1 <:. 120/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.1 <:. 120/125 SR* 3.3.1.1.4 divisions of full SR 3.3.U.9 scale SR 3.3.1.1.14 b. lnop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14 2. Average Power Range Monitors 2 3(b) G SR 3.3.1.1.1 <:.13% RTP a. Neutron Flux High, SR 3.3.1.1.6 Set down SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 b. Flow Biased Simulated 3(b) F SR 3.3.1.1.1 <:.0.55W Thermal Power -High SR 3.3.1.1.2 + 65.5% RTP SR 3.3.1.1.7 and<:. 120% SR 3.3.1.1.13 RTP(c) SR 3.3.1.1.16 c. Neutron Flux -High 3(b) F SR 3.3.1.1.1 120% RTP SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. (c) [0.55 W + 65.5% -0.55 8 W] RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating." BFN-UNIT 1 3.3-6 Amendment No. 236, 262, 269,299 RPS Instrumentation 3.3.1.1 BFN-UNIT 1 3.3-7 Amendment No. 281 234, 262, 259, 257,258, 266, 269 April 27, 2012 Table 3.3.1.1-1 (page 2 of 3) Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 2. Average Power Range Monitors (continued) d. Inop 1,2

3(b)

G

SR 3.3.1.1.16

NA e. 2-Out-Of-4 Voter 1,2 2 G SR 3.3.1.1.1 SR 3.3.1.1.14 SR 3.3.1.1.16 NA f. OPRM Upscale 1 3(b) I SR 3.3.1.1.1 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.17 NA(e) 3. Reactor Vessel Steam Dome Pressure - High(d) 1,2 2 G SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 1090 psig 4. Reactor Vessel Water Level - Low, Level 3(d) 1,2 2 G SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 528 inches above vessel zero 5. Main Steam Isolation Valve - Closure 1 8 F SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 10% closed 6. High Drywell Pressure 1,2 2 G SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 2.5 psig 7. Scram Discharge Volume Water Level - High a. Resistance Temperature Detector 1,2 2 G SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 50 gallons 5(a) 2 H SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 50 gallons (continued) (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. (d) During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable. Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable. The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report. (e) Refer to COLR for OPRM period based detection algorithm (PBDA) setpoint limits. TS 3.3.1.1 Insert B (f) Following Detect and Suppress Solution - Confirmation Density (DSS-CD) implementation, DSS-CD is not required to be armed while in the DSS-CD Armed Region during the first reactor startup and during the first controlled shutdown that passes completely through the DSS-CD Armed Region. However, DSS-CD is considered OPERABLE and shall be maintained OPERABLE and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region. Recirculation Loops Operating 3.4.1 BFN-UNIT 1 3.4-1 Amendment No. 236, 266 December 29, 2006 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating

LCO 3.4.1 Two recirculation loops with matched flows shall be in operation. OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable: a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR;

c. LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power - High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY: MODES 1 and 2. Recirculation Loops Operating 3.4.1 BFN-UNIT 1 3.4-2 Amendment No. 236, 266 December 29, 2006 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO not met. A.1 Satisfy the requirements of the LCO. 24 hours B. Required Action and associated Completion Time of Condition A not met.

OR No recirculation loops in operation. B.1 Be in MODE 3. 12 hours TS 3.4.1 Insert A B. Operation in the MELLLA+ operating domain with a single recirculation loop in operation. B.1 Initiate action to exit the MELLLA+ operating domain.

Immediately

Reporting Requirements 5.6 (continued) BFN-UNIT 1 5.0-24 Amendment No. 234, 239, 252, 281 April 27, 2012 5.6 Reporting Requirements (continued) 5.6.4 (Deleted). 5.6.5 CORE OPERATING LIMITS REPORT (COLR) a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: (1) The APLHGRs for Specification 3.2.1; (2) The LHGR for Specification 3.2.3; (3) The MCPR Operating Limits for Specification 3.2.2; (4) The period based detection algorithm (PBDA) setpoint for Function 2.f, Oscillation Power Range Monitor (OPRM) Upscale, for Specification 3.3.1.1; and (5) The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1. b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: 1. NEDE-24011-P-A, Revision 16, General Electric Standard Application for Reactor Fuel, October 2007. 2. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984. Reporting Requirements 5.6 (continued) BFN-UNIT 1 5.0-24c Amendment No. 234, 239, 252, 281, 285 July 31, 2014 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 18. EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000. 19. BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008. 20. BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008. 21. ANP-10298PA Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, March 2010. 22. ANP-3140(P) Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, August 2012. Reporting Requirements 5.6 BFN-UNIT 1 5.0-25 Amendment No. 234, 239, 252, 281 April 27, 2012 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met. d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC. 5.6.6 PAM Report When a report is required by Condition B or G of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. TS 5.6 Insert A 5.6.7 OPRM Report When an OPRM report is required by Condition I of LCO 3.3.1.1, "RPS Instrumentation," the report shall be submitted within the following 90 days. The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to OPERABLE status. -6e-(h) The results of the visual inspections of the steam dryer shall be submitted to the NRC staff in a report in accordance with 10 CFR 50.4. The report shall be submitted to the NRC within 90 days following startup from each of the first two respective refueling outages. (i) Within 6 months following completion of the second refueling outage, after the implementation of the EPU, the licensee shall submit a long-term steam dryer inspection plan based on industry operating experience along with the baseline inspection results. The license condition described above shall expire: (1) upon satisfaction of the requirements in items (g) and (h), provided that a visual inspection of the steam dryer does not reveal any new unacceptable flaw{s) or unacceptable flaw growth that is due to fatigue, and; {2) upon satisfaction of the requirements specified in Item (i). (19) Neutron Absorber Monitoring Program The licensee shall, at least once every ten years, withdraw a neutron absorber coupon from the spent fuel pool and perform Boron-10 (B-10) areal density measurement on the coupon. Based on the results of the B-10 areal density measurement, the licensee shall perform any technical evaluations that may be necessary and take appropriate actions using relevant regulatory and licensing processes. {20) Radiological Consequences Analyses Using Alternative Source Terms TVA shall perform facility and licensing basis modifications to resolve the non-conforming/degraded condition associated with the Alternate Leakage Treatment pathway such that the current licensing basis dose calculations (approved in License Amendment Nos. 251/282 (Unit 1 ), 290/308 {Unit 2) and 249/267 (Unit 3)) would remain valid. These facility and licensing basis modifications shall be complete prior to initial power ascension above 3458 MWt. D. The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21(d), shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71(e)(4) following the issuance of this renewed operating license. Until that update is complete, TVA may make changes to the programs and activities described in the supplement without prior Commission approval, provided that TV A evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section. E. The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. TVA shall complete these activities no later than June 28, 2014, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. BFN-UNIT 2 Renewed License No. DPR-52 Amendment No. 323 Insert LC for BFN Unit 2 (21) Maximum Extended Load Line Limit Analysis Plus (MELLLA+) Special Consideration The licensee shall not operate the facility within the MELLLA+ operating domain with a feedwater heater out of service resulting in more than a 10°F reduction in feedwater temperature below the design feedwater temperature. RPS Instrumentation 3.3.1.1 BFN-UNIT 2 3.3-3 Amendment No. 258, 273 July 26, 2001 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME G. As required by Required Action D.1 and referenced in Table 3.3.1.1-1. G.1 Be in MODE 3. 12 hours H. As required by Required Action D.1 and referenced in Table 3.3.1.1-1. H.1 Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Immediately I. As required by Required Action D.1 and referenced in Table 3.3.1.1-1. I.1 Initiate alternate method to detect and suppress thermal hydraulic instability oscillations. 12 hours J. Required Action and associated Completion Time of Condition I not met. J.1 Be in Mode 2. 4 hours TS 3.3.1.1 Insert A I. As required by Required Action D.1 and referenced in Table 3.3.1.1-1 I.1 Initiate action to implement the Manual Backup Stability Protection (BSP Regions defined in the COLR. AND I.2 Implement the Automated BSP Scram Region using the modified APRM Flow Biased Simulated Thermal Power-High scram setpoints defined in the COLR. AND I.3 Initiate action to submit an OPRM report in accordance with Specification 5.6.7. Immediately 12 hours Immediately J. Required Action and associated Completion Time of Condition I not met. J.1 Initiate action to implement the Manual BSP Regions defined in the COLR. AND J.2 Reduce operation to below the BSP Boundary defined in the COLR. AND J.3 ---------------NOTE---------------- LCO 3.0.4 is not applicable. ---------------------------------------- Restore required channel to OPERABLE status. Immediately 12 hours 120 days K. Required Actions and associated Completion Time of Condition J not met. K.1 Reduce THERMAL POWER to < 18% RTP. 4 hours SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE SR 3.3.1.1.10 Perform CHANNEL CALIBRATION. SR 3.3.1.1.11 (Deleted) SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. SR 3.3.1.1.13 --------------------------NOTE-------------------------Neutron detectors are excluded. ------------------------------------------------------------Perform CHANNEL CALIBRATION. SR 3.3.1.1.14 Perform LOGIC SYSTEM FUNCTIONAL TEST. SR 3.3.1.1.15 Verify Turbine Stop Valve -Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure -Low Functions are not bypassed when THERMAL POWER is z 26% RTP. SR 3.3.1 .1.16 --------------------------NOTE-------------------------For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. ------------------------------------------------------------Perform CHANNEL FUNCTIONAL TEST. SR 3.3.1.1.17 Verify OPRM is not bypassed when APRM Simulated Thermal Power is z 23% and recirculation drive flow is< 60% of rated recirculation drive flow. RPS Instrumentation 3.3.1.1 FREQUENCY 184 days 24 months 24 months 24 months 24 months 184 days 24 months BFN-UNIT 2 3.3-6 Amendment No. 323 FUNCTION 1. Intermediate Range Monitors a. Neutron Flux -High b. lnop 2. Average Power Range Monitors a. Neutron Flux -High, (Setdown) b. Flow Biased Simulated Thermal Power -High c. Neutron Flux -High Table3.3.1.1-1(page1 of3) Reactor Protection System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 2 5(a) 2 5(a) 2 REQUIRED CHANNELS PER TRIP SYSTEM 3 3 3 3 CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 G H G H G F F (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. RPS Instrumentation 3.3.1.1 SURVEILLANCE ALLOWABLE REQUIREMENTS VALUE SR 3.3.1.1.1 SR 3.3.1.1.3 SR 3.3.1.1.5 SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3.3.1.1.14 SR 3.3.1.1.1 SR 3.3.1.1.4 SR 3.3.1.1.9 SR 3.3.1.1.14 SR 3.3.1.1.3 SR 3.3.1.1.14 SR 3.3.1.1.4 SR 3.3.1.1.14 SR 3.3.1.1.1 SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.1 SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.1 SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 <; 120/125 divisions of full scale <; 120/125 divisions of full scale NA NA <; 13% RTP <; 0.55 w + 65.5% RTP and<; 120% RTP(c) <; 120% RTP continued) (c) [0.55 W + 65.5% -0.55 ti W] RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating." BFN-UNIT 2 3.3-7 Amendment No. 323 RPS Instrumentation 3.3.1.1 BFN-UNIT 2 3.3-8 Amendment No. 253, 254, 258, 260, 296, 309 February 15, 2013 Table 3.3.1.1-1 (page 2 of 3) Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 2.Average Power RangeMonitors (continued)d.Inop1,2 3(b)G SR 3.3.1.1.16 NA e.2-Out-Of-4 Voter1,2 2 G SR 3.3.1.1.1 SR 3.3.1.1.14 SR 3.3.1.1.16 NA f.OPRM Upscale1 3(b)I SR 3.3.1.1.1 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.17 NA(e)3.Reactor Vessel Steam DomePressure - High(d)1,2 2 G SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 1090 psig 4.Reactor Vessel Water Level -Low, Level 3(d)1,2 2 G SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 528 inches above vessel zero 5.Main Steam Isolation Valve -Closure1 8 F SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 10% closed 6.Drywell Pressure - High1,2 2 G SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 2.5 psig 7.Scram Discharge VolumeWater Level - Higha.Resistance TemperatureDetector1,2 2 G SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 50 gallons 5(a)2 H SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 50 gallons (continued) (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. (d) During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable. Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable. The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report. (e) Refer to COLR for OPRM period based detection algorithm (PBDA) setpoint limits. TS 3.3.1.1 Insert B (f) Following Detect and Suppress Solution - Confirmation Density (DSS-CD) implementation, DSS-CD is not required to be armed while in the DSS-CD Armed Region during the first reactor startup and during the first controlled shutdown that passes completely through the DSS-CD Armed Region. However, DSS-CD is considered OPERABLE and shall be maintained OPERABLE and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region. Recirculation Loops Operating 3.4.1 BFN-UNIT 2 3.4-1 Amendment No. 258 March05,19993.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation. OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable: a.LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATIONRATE (APLHGR)," single loop operation limits specified in theCOLR;b.LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"single loop operation limits specified in the COLR;c.LCO 3.3.1.1, "Reactor Protection System (RPS)Instrumentation," Function 2.b (Average Power Range MonitorsFlow Biased Simulated Thermal Power - High), Allowable Valueof Table 3.3.1.1-1 is reset for single loop operation;APPLICABILITY: MODES 1 and 2. Recirculation Loops Operating 3.4.1 BFN-UNIT 2 3.4-2 Amendment No. 258 March 05, 1999 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO not met. A.1 Satisfy the requirements of the LCO. 24 hours B. Required Action and associated Completion Time of Condition A not met.

OR No recirculation loops in operation. B.1 Be in MODE 3. 12 hours TS 3.4.1 Insert A B. Operation in the MELLLA+ operating domain with a single recirculation loop in operation. B.1 Initiate action to exit the MELLLA+ operating domain. Immediately Reporting Requirements 5.6 (continued) BFN-UNIT 2 5.0-24 Amendment No. 287, 309 February 15, 2013 5.6 Reporting Requirements (continued) 5.6.4 (Deleted). 5.6.5 CORE OPERATING LIMITS REPORT (COLR) a.Core operating limits shall be established prior to each reload cycle,or prior to any remaining portion of a reload cycle, and shall bedocumented in the COLR for the following:(1) The APLHGRs for Specification 3.2.1;(2) The LHGR for Specification 3.2.3;(3) The MCPR Operating Limits for Specification 3.2.2;(4) The period based detection algorithm (PBDA) setpoint forFunction 2f Oscillation Power Range Monitor (OPRM) Upscale, for Specification 3.3.1.1; and (5) The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1. b.The analytical methods used to determine the core operating limitsshall be those previously reviewed and approved by the NRC,specifically those described in the following documents:1.XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2,RODEX2 Fuel Rod Thermal-Mechanical ResponseEvaluation Model, Exxon Nuclear Company, March 1984.2.XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Designfor Exxon Nuclear Jet Pump BWR Reload Fuel, ExxonNuclear Company, September 1986.3.EMF-85-74(P) Revision 0 Supplement 1(P)(A) andSupplement 2 (P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation,February 1998. Reporting Requirements 5.6 (continued) BFN-UNIT 2 5.0-24b Amendment No. 287, 309, 311 and 313 July 31, 2014 and February 26, 2015 5.6 Reporting Requirements (continued) 14.EMF-2245(P)(A) Revision 0, Application of Siemens PowerCorporation's Critical Power Correlations to Co-Resident Fuel,Siemens Power Corporation, August 2000.15.EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS EvaluationModel, Framatome ANP Inc., May 2001 as supplemented by the site-specific approval in NRC safety evaluation dated February 15, 2013and July 31, 2014.16.EMF-2292(P)(A) Revision 0, ATRIUMŽ-10: Appendix K Spray HeatTransfer Coefficients, Siemens Power Corporation, September 2000.17.EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis:Assessment of STAIF with Input from MICROBURN-B2, SiemensPower Corporation, August 2000.18.BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM MethodologyUsing the RAMONA5-FA Code, AREVA NP, May 2008.19.BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel RodMethodology for Boiling Water Reactors, AREVA NP, February 2008.20.ANP-10298PA Revision 0, ACE/ATRIUM 10XM Critical PowerCorrelation, AREVA NP, March 2010.21.ANP-3140(P) Revision 0, Browns Ferry Units 1, 2, and 3 ImprovedK-factor Model for ACE/ATRIUM 10XM Critical Power Correlation,AREVA NP, August 2012. Reporting Requirements 5.6 BFN-UNIT 2 5.0-25 Amendment No. 253 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) c.The core operating limits shall be determined such that allapplicable limits (e.g., fuel thermal mechanical limits, core thermalhydraulic limits, Emergency Core Cooling Systems (ECCS) limits,nuclear limits such as SDM, transient analysis limits, and accidentanalysis limits) of the safety analysis are met.d.The COLR, including any midcycle revisions or supplements, shallbe provided upon issuance for each reload cycle to the NRC.5.6.6 PAM Report When a report is required by Condition B or G of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. TS 5.6 Insert A 5.6.7 OPRM Report When an OPRM report is required by Condition I of LCO 3.3.1.1, "RPS Instrumentation," the report shall be submitted within the following 90 days. The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to OPERABLE status. -6f-(16) Radiological Consequences Analyses Using Alternative Source Terms TVA shall perform facility and licensing basis modifications to resolve the non-conforming/degraded condition associated with the Alternate Leakage Treatment pathway such that the current licensing basis dose calculations (approved in license Amendment Nos. 251/282 (Unit 1), 290/308 (Unit 2) and 249/267 (Unit 3)) would remain valid. These facility and licensing basis modifications shall be complete prior to initial power ascension above 3458 MWt. D. The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21(d), shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71 (e}(4} following the issuance of this renewed operating license. Until that update is complete, TVA may make changes to the programs and activities described in the supplement without prior Commission approval, provided that TVA evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section. The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. TVA shall complete these activities no later than July 2, 2016, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. BFN-UNIT 3 Renewed License No. DPR-68 Amendment No. 283 Insert LC for BFN Unit 3 (17) Maximum Extended Load Line Limit Analysis Plus (MELLLA+) Special Consideration The licensee shall not operate the facility within the MELLLA+ operating domain with a feedwater heater out of service resulting in more than a 10°F reduction in feedwater temperature below the design feedwater temperature. RPS Instrumentation 3.3.1.1 BFN-UNIT 3 3.3-3 Amendment No. 212, 213, 221, 231 September 13, 2001 ACTIONS (continued) CONDITIONREQUIRED ACTIONCOMPLETIONTIME G. As required by Required Action D.1 and referenced in Table 3.3.1.1-1. G.1 Be in MODE 3. 12 hours H. As required by Required Action D.1 and referenced in Table 3.3.1.1-1. H.1 Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Immediately I. As required by Required Action D.1 and referenced in Table 3.3.1.1-1. I.1 Initiate alternate method to detect and suppress thermal hydraulic instability oscillations. 12 hours J. Required Action and associated Completion Time of Condition I not met. J.1 Be in MODE 2 4 hours TS 3.3.1.1 Insert A I. As required by Required Action D.1 and referenced in Table 3.3.1.1-1 I.1 Initiate action to implement the Manual Backup Stability Protection (BSP Regions defined in the COLR. AND I.2 Implement the Automated BSP Scram Region using the modified APRM Flow Biased Simulated Thermal Power-High scram setpoints defined in the COLR. AND I.3 Initiate action to submit an OPRM report in accordance with Specification 5.6.7. Immediately 12 hours Immediately J. Required Action and associated Completion Time of Condition I not met. J.1 Initiate action to implement the Manual BSP Regions defined in the COLR. AND J.2 Reduce operation to below the BSP Boundary defined in the COLR. AND J.3 ---------------NOTE---------------- LCO 3.0.4 is not applicable. ---------------------------------------- Restore required channel to OPERABLE status. Immediately 12 hours 120 days K. Required Actions and associated Completion Time of Condition J not met. K.1 Reduce THERMAL POWER to < 18% RTP. 4 hours RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SR 3.3.1.1.10 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.16 SR 3.3.1.1.17 BFN-UNIT 3 SURVEILLANCE FREQUENCY Perform CHANNEL CALIBRATION. 184 days (Deleted) Perform CHANNEL FUNCTIONAL TEST. 24 months --------------------------NOTE-------------------------Neutron detectors are excluded. ------------------------------------------------------------Perform CHANNEL CALIBRATION. 24 months Perform LOGIC SYSTEM FUNCTIONAL 24 months TEST. Verify Turbine Stop Valve -Closure and 24 months Turbine Control Valve Fast Closure, Trip Oil Pressure -Low Functions are not bypassed when THERMAL POWER is ?: 26% RTP. --------------------------NOTE-------------------------For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. ------------------------------------------------------------Perform CHANNEL FUNCTIONAL TEST. 184 days Verify OPRM is not bypassed when APRM 24 months Simulated Thermal Power is ?: 23% and recirculation drive flow is< 60% of rated recirculation drive flow. 3.3-6 Amendment No. 212, 213, 215, FUNCTION 1. Intermediate Range Monitors a. Neutron Flux -High b. lnop 2. Average Power Range Monitors a. Neutron Flux -High, (Setdown) b. Flow Biased Simulated Thermal Power -High c. Neutron Flux -High Table 3.3.1.1-1(page1 of3) Reactor Protection System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 2 5(a) 2 5(a) 2 REQUIRED CHANNELS PER TRIP SYSTEM 3 3 3 3 CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 G H G H G F F (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. RPS Instrumentation 3.3.1.1 SURVEILLANCE ALLOWABLE REQUIREMENTS VALUE SR 3.3.1.1.1 SR 3.3.1.1.3 SR 3.3.1.1.5 SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3.3.1.1.14 SR 3.3.1.1.1 SR 3.3.1.1.4 SR 3.3.1.1.9 SR 3.3.1.1.14 SR 3.3.1.1.3 SR 3.3.1.1.14 SR 3.3.1.1.4 SR 3.3.1.1.14 SR 3.3.1.1.1 SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.1 SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.1 SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 <; 120/125 divisions of full scale <; 120/125 divisions of full scale NA NA <; 13% RTP 0.55W + 65.5% RTP and<; 120% RTP(c) <; 120% RTP (continued) (c) [0.55 W + 65.5% -0.55 i\ W] RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating." BFN-UNIT 3 3.3-7 Amendment No. 24-9, 283 RPS Instrumentation 3.3.1.1 BFN-UNIT 3 3.3-8 Amendment No. 212, 213, 214, 219, 221, 254, 268 February 15, 2013 Table 3.3.1.1-1 (page 2 of 3) Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 2. Average Power Range Monitors (continued) d. Inop 1,2

3(b)

G

SR 3.3.1.1.16

NA e. 2-Out-Of-4 Voter 1,2 2 G SR 3.3.1.1.1 SR 3.3.1.1.14 SR 3.3.1.1.16 NA f. OPRM Upscale 1 3(b) I SR 3.3.1.1.1 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.17 NA(e) 3. Reactor Vessel Steam Dome Pressure - High(d) 1,2 2 G SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 1090 psig 4. Reactor Vessel Water Level - Low, Level 3(d) 1,2 2 G SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 528 inches above vessel zero 5. Main Steam Isolation Valve - Closure 1 8 F SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 10% closed 6. Drywell Pressure - High 1,2 2 G SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 2.5 psig 7. Scram Discharge Volume Water Level - High a. Resistance Temperature Detector

1,2

2

G

SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 50 gallons 5(a) 2 H SR 3.3.1.1.8 SR 3.3.1.1.13 SR 3.3.1.1.14 50 gallons (continued) (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. (d) During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable. Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable. The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report. (e) Refer to COLR for OPRM period based detection algorithm (PBDA) setpoint limits. TS 3.3.1.1 Insert B (f) Following Detect and Suppress Solution - Confirmation Density (DSS-CD) implementation, DSS-CD is not required to be armed while in the DSS-CD Armed Region during the first reactor startup and during the first controlled shutdown that passes completely through the DSS-CD Armed Region. However, DSS-CD is considered OPERABLE and shall be maintained OPERABLE and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region. Recirculation Loops Operating 3.4.1 BFN-UNIT 3 3.4-1 Amendment No. 212, 216, 221 September 27, 1999 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating

LCO 3.4.1 Two recirculation loops with matched flows shall be in operation. OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable: a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR;

c. LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power - High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation; APPLICABILITY: MODES 1 and 2.

Recirculation Loops Operating 3.4.1 BFN-UNIT 3 3.4-2 Amendment No. 212, 216, 221 September 27, 1999 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO not met. A.1 Satisfy the requirements of the LCO. 24 hours B. Required Action and associated Completion Time of Condition A not met.

OR No recirculation loops in operation. B.1 Be in MODE 3. 12 hours TS 3.4.1 Insert A B. Operation in the MELLLA+ operating domain with a single recirculation loop in operation. B.1 Initiate action to exit the MELLLA+ operating domain.

Immediately

Reporting Requirements 5.6 (continued) BFN-UNIT 3 5.0-24 Amendment No. 212, 226, 245, 250, 268 February 15, 2013 5.6 Reporting Requirements (continued) 5.6.4 (Deleted). 5.6.5 CORE OPERATING LIMITS REPORT (COLR) a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: (1) The APLHGRs for Specification 3.2.1; (2) The LHGR for Specification 3.2.3; (3) The MCPR Operating Limits for Specification 3.2.2; (4) The period based detection algorithm (PBDA) setpoint for Function 2.f, Oscillation Power Range Monitor (OPRM) Upscale, for Specification 3.3.1.1; and (5) The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1. b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: 1. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984. 2. XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986. 3. EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998. Reporting Requirements 5.6 (continued) BFN-UNIT 3 5.0-24b Amendment No. 212, 226, 245, 250, 268, 270 and 272 July 31, 2014 and February 26, 2015 5.6 Reporting Requirements (continued) 12. ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, September 2005. 13. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.

14. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000. 15. EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001 as supplemented by the site-specific approval in NRC safety evaluations dated February 15, 2013, and July 31, 2014. 16. EMF-2292(P)(A) Revision 0, ATRIUMŽ-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000. 17. EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000. 18. BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008. 19 BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008. 20. ANP-10298PA Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, March 2010. 21. ANP-3140(P) Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, August 2012.

Reporting Requirements 5.6 BFN-UNIT 3 5.0-25 Amendment No. 212 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met. d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC. 5.6.6 PAM Report When a report is required by Condition B or G of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. TS 5.6 Insert A 5.6.7 OPRM Report When an OPRM report is required by Condition I of LCO 3.3.1.1, "RPS Instrumentation," the report shall be submitted within the following 90 days. The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to OPERABLE status.

RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 1 B 3.3-16a Revision 45 February 27, 2007 BASES APPLICABLE 2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, LCO, and The OPRM Upscale Function provides compliance with GDC 10 APPLICABILITY and GDC 12, thereby providing protection from exceeding the (continued) fuel MCPR safety limit (SL) due to anticipated thermal hydraulic power oscillations. References 13, 14, and 15 describe three algorithms for detecting thermal hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm. The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation of the OPRM algorithms. The OPRM Upscale Function is required to be OPERABLE when the plant is in a region of power flow operation where anticipated events could lead to thermal hydraulic instability and related neutron flux oscillations. Within this region, the automatic trip is enabled when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 25% RTP and reactor core flow, as indicated by recirculation drive flow is < 60% of rated flow, the operating region where actual thermal hydraulic oscillations may occur. Requiring the OPRM Upscale Function to be OPERABLE in MODE 1 provides consistency with operability requirements for other APRM functions and assures that the OPRM Upscale Function is OPERABLE whenever reactor power could increase into the region of concern without operator action. Insert 2.f 2.f. Oscillation Power Range Monitor (OPRM) Upscale The OPRM Upscale Function provides compliance with 10 CFR 50, Appendix A, General Design Criteria (GDC) 10 and 12, thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations. Reference 13 describes the Detect and Suppress - Confirmation Density (DSS-CD) long-term stability solution and the licensing basis Confirmation Density Algorithm (CDA). Reference 13 also describes the DSS-CD Armed Region and the three additional algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm (PBDA), the amplitude based algorithm (ABA), and the growth rate algorithm (GRA). All four algorithms are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the CDA. The remaining three algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY is based only on the CDA.

The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into cells for evaluation by the OPRM algorithms. DSS-CD operability requires at least 8 responsive OPRM cells per channel. The DSS-CD software includes a self-check for the responsive OPRM cells; therefore, no SR is necessary. The OPRM Upscale Function is required to be OPERABLE above a power level set at 5% of rated power below the lower boundary of the Armed Region, which is >18% RTP. This requirement is designed to encompass the region of power-flow operation where anticipated events could lead to thermal-hydraulic instability and related neutron flux oscillations. The OPRM Upscale Function is automatically trip-enabled when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 23% RTP corresponding to the MCPR monitoring threshold and reactor recirculation drive flow, is less than 75% of rated flow. This region is the OPRM Armed Region. Note (f) allows for entry into the DSS-CD Armed Region without automatic arming of DSS-CD prior to completely passing through the DSS-CD Armed Region during the first startup and the first shutdown following DSS-CD implementation. Insert 2.f (continued) 2.f. Oscillation Power Range Monitor (OPRM) Upscale If any OPRM auto-enable setpoint is in a non-conservative condition, i.e., the OPRM Upscale is not auto-enabled with RTP 23% and reactor recirculation drive flow 75% of rated, the associated channel is considered inopera for the OPRM Upscale function. Alternatively, the auto-enable setpoint may be adjusted to place the channel in a conservative condition (armed). If placed in the armed condition, the channel is considered OPERABLE. Note (f) reflects the need for plant data collection in order to test the DSS-CD equipment. Testing the DSS-CD equipment ensures its proper operation and prevents spurious reactor trips. Entry into the DSS-CD Armed Region without automatic arming of DSS-CD during this initial testing phase also allows for changes in plant operations to address maintenance or other operational needs. However, during this initial testing period, the OPRM upscale function is OPERABLE and DSS-CD operability and capability to automatically arm shall be maintained at recirculation drive flow rates above the DSS-CD Armed Region flow boundary. An OPRM Upscale trip is issued from an OPRM channel when the confirmation density algorithm in that channel detects oscillatory changes in the neutron flux, indicated by periodic confirmations and amplitude exceeding specified setpoints for a specified number of OPRM cells in the channel. An OPRM Upscale trip is also issued from the channel if any of the defense-in-depth algorithms (PBDA, ABA, GRA) exceed their trip condition for one or more cells in that channel. Three of the four channels are required to be operable. Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the SLMCPR is exceeded. There is no Allowable Value for this function. The OPRM Upscale Function is not LSSS SL-related (Reference 13) and Reference 14 confirms that the OPRM Upscale Function settings based on DSS-CD also do not have traditional instrumentation setpoints determined under an instrument setpoint methodology. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 1 B 3.3-16b Revision 45 February 27, 2007 BASES APPLICABLE 2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, (continued) LCO, and APPLICABILITY An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect growing oscillatory changes in the neutron flux for one or more cells in that channel. Three of the four channels are required to be OPERABLE. Each channel is capable of detecting thermal hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the MCPR SL is exceeded. There is no allowable value for this function. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 1 B 3.3-33 Revision 0, 45 February 27, 2007 BASES ACTIONSC.1 (continued) For Function 8 (Turbine Stop Valve - Closure), this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated trip system in trip). The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. D.1 Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition. E.1, F.1, G.1, and J.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Action E.1 is consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)." RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 1 B 3.3-34 Revision 0, 45 February 27, 2007 BASES ACTIONSH.1 (continued) If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. I.1 If OPRM Upscale trip capability is not maintained, Condition I exists. Reference 12 justified use of alternate methods to detect and suppress oscillations. The alternate methods are procedurally established consistent with the guidelines identified in Reference 17 requiring manual operator action to scram the plant if certain predefined events occur. The 12 hour allowed action time is based on engineering judgment to allow orderly transition to the alternate methods while limiting the period of time during which no automatic or alternate detect and suppress trip capability is formally in place. Based on the small probability of an instability event occurring at all, the 12 hours is judged to be reasonable. Insert Required Actions I, J, K ACTIONS I.1 If OPRM Upscale trip capability is not maintained, Condition I exists and Backup Stability Protection (BSP) is required. The Manual BSP Regions are described in Reference 13. The Manual BSP Regions are procedurally established consistent with the guidelines identified in Reference 13 and require specified manual operator actions if certain predefined operational conditions occur. The Completion Time of immediate is based on the importance of limiting the period of time during which no automatic or alternate detect and suppress trip capability is in place. I.2 and I.3 Required Actions I.2 and I.3 are both required to be taken in conjunction with Required Action I.1 if OPRM Upscale trip capability is not maintained. As described in Section 7.4 of Reference 13, the Automated BSP Scram Region is designed to avoid reactor instability by automatically preventing entry into the region of the power and flow operating map that is susceptible to reactor instability. The reactor trip would be initiated by the modified APRM Simulated Thermal Power-High scram setpoints for flow reduction events that would have terminated in the Manual BSP Region I. The Automated BSP Scram Region ensures an early scram and SLMCPR protection. The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, and based on the importance of restoring an automatic reactor trip for thermal hydraulic instability events. Backup Stability Protection is intended as a temporary means to protect against thermal-hydraulic instability events. The action should be initiated immediately to document the situation and prepare the report. The reporting requirements of Specification 5.6.7 document the corrective actions and schedule to restore the required channels to an OPERABLE status. The Completion Time of 90 days shown in Specification 5.6.7 is adequate to allow time to evaluate the cause of the inoperability and to determine the appropriate corrective actions and schedule to restore the required channels to OPERABLE status. J.1 If the Required Action I is not completed within the associated Completion Time, then Action J is required. The Bases for the Manual BSP Regions and associated Completion Time is addressed in the Bases for I.1. The Manual BSP Regions are required in conjunction with the BSP Boundary. J.2 The BSP Boundary, as described in Section 7.3 of Reference 13, defines an operating domain where potential instability events can be effectively addressed by the specified BSP manual operator actions. The BSP Boundary is constructed such that a flow reduction event initiated from this boundary and terminated at the core natural circulation line (NCL) would not exceed the Manual BSP Region I stability criterion. Potential instabilities would develop slowly as a result of the feedwater temperature transient (Reference 13). The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, to reach the specific condition from full power conditions in an orderly manner and without challenging plant system. Insert Required Actions I, J, K (continued) J.3 Backup Stability Protection (BSP) is a temporary means for protection against thermal-hydraulic instability events. An extended period of inoperability without automatic trip capability is not justified. Consequently, the required channels are required to be restored to OPERABLE status within 120 days. Based on engineering judgment, the likelihood of an instability event that could not be adequately handled by the use of the BSP Regions (See Required Action J.1) and the BSP Boundary (See Required Action J.2) during a 120-day period is negligibly small. The 120-day period is intended to allow for the case where limited design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment. This action is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to normally be accomplished within the completion times allowed for Actions for Conditions A and B. A Note is provided to indicate that LCO 3.0.4 is not applicable. The intent of the note is to allow plant startup while operating within the 120 day Completion Time for Required Action J.3. The primary purpose of this exclusion is to allow an orderly completion of design and verification activities, in the event of a required design change, without undue impact on plant operation. K.1 If the required channels are not restored to OPERABLE status and the Required Actions of Condition J are not met within the associated Completion Times, then the plant must be placed in an operating condition in which the LCO does not apply. To achieve this status, the plant must be brought to less than 18% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the specified operating power level from full power conditions in an orderly manner and without challenging plant systems. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 1 B 3.3-43a Revision 45 February 27, 2007 BASES SURVEILLANCE SR 3.3.1.1.17 REQUIREMENTS (continued) This SR ensures that scrams initiated from OPRM Upscale Function (Function 2.f) will not be inadvertently bypassed when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 25% RTP and core flow, as indicated by recirculation drive flow, is < 60% rated core flow. This normally involves confirming the bypass setpoints. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. The actual surveillance ensures that the OPRM Upscale Function is enabled (not bypassed) for the correct values of APRM Simulated Thermal Power and recirculation drive flow. Other surveillances ensure that the APRM Simulated Thermal Power and recirculation flow properly correlate with THERMAL POWER and core flow, respectively. If any bypass setpoint is nonconservative (i.e., the OPRM Upscale Function is bypassed when APRM Simulated Thermal Power 25% RTP and recirculation drive flow < 60% rated), then the affected channel is considered inoperable for the OPRM Upscale Function. Alternatively, the bypass setpoint may be adjusted to place the channel in a conservative condition (unbypass). If placed in the unbypassed condition, this SR is met and the channel is considered OPERABLE. The frequency of 24 months is based on engineering judgment and reliability of the components. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 1 B 3.3-44 Revision 0, 40, 45 February 27, 2007 BASES (continued) REFERENCES 1.FSAR, Section 7.2.2.FSAR, Chapter 14.3.NEDO-23842, "Continuous Control Rod Withdrawal in theStartup Range," April 18, 1978.4.FSAR, Appendix N. 5.FSAR, Section 14.6.2. 6.FSAR, Section 6.5.7.FSAR, Section 14.5. 8.P. Check (NRC) letter to G. Lainas (NRC), "BWR ScramDischarge System Safety Evaluation," December 1, 1980.9.NEDC-30851-P-A , "Technical Specification ImprovementAnalyses for BWR Reactor Protection System," March 1988.10.NRC No. 93-102, "Final Policy Statement on TechnicalSpecification Improvements," July 23, 1993.11.MED-32-0286, "Technical Specification ImprovementAnalysis for Browns Ferry Nuclear Plant, Unit 2," October1995. 12.NEDC-32410P-A, "Nuclear Measurement Analysis andControl Power Range Neutron Monitor (NUMAC PRNM)Retrofit Plus Option III Stability Trip Function," October 1995.13.NEDO-31960-A, "BWR Owners' Group Long-Term StabilitySolutions Licensing Methodology," November 1995. Insert Refs 13.GE Hitachi Nuclear Energy, "GE Hitachi Boiling Water Reactor Detect and Suppress Solution -Confirmation Density," NEDC-33075P-A, Revision 8, November 2013.14.GEH letter to NRC, "NEDC-33075P-A, Detect and Suppress Solution - Confirmation Density(DSS-CD) Analytical Limit (TAC No. MD0277)," October 29, 2008. (ADAMS Accession No.ML083040052) RPS Instrumentation B 3.3.1.1 BFN-UNIT 1 B 3.3-44a Revision 45 February 27, 2007 BASES REFERENCES 14.NEDO-31960-A, Supplement 1, "BWR Owners' Group(continued) Long-Term Stability Solutions Licensing Methodology,"November 1995.15.NEDO-32465-A, "BWR Owners' Group Long-Term StabilityDetect and Suppress Solutions Licensing Basis Methodologyand Reload Applications," August 1996.16.NEDC-32410P-A, Supplement 1, "Nuclear MeasurementAnalysis and Control Power Range Neutron Monitor(NUMAC PRNM) Retrofit Plus Option III Stability TripFunction," August 1996.17.Letter, L.A. England (BWROG) to M.J. Virgilio, "BWROwners' Group Guidelines for Stability Interim CorrectiveAction," June 6, 1994. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 1 B 3.4-4 Revision 45, 50 Amendment No. 236 May 03, 2007 BASES APPLICABLE Plant specific LOCA analyses have been performed assuming SAFETY ANALYSES only one operating recirculation loop. These analyses have (continued) demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Refs. 7 and 8). The transient analyses of Chapter 14 of the FSAR have also been performed for single recirculation loop operation (Ref. 7) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoint is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Flow Biased Simulated Thermal Power-High setpoint is in LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation." Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 6). Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 1 B 3.4-5 Revision 45 Amendment No. 236 February 27, 2007 BASES (continued) LCO Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation. With only one recirculation loop in operation, modifications to the required APLHGR Limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LCO 3.2.2, 'MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM Flow Biased Simulated Thermal Power-High Setpoint (LCO 3.3.1.1) may be applied to allow continued operation consistent with the assumptions of References 7 and 8. APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 1 B 3.4-6 Revision 0, 45 February 27, 2007 BASES (continued) ACTIONSA.1 With the requirements of the LCO not met, the recirculation loops must be restored to operation with matched flows within 24 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the inoperable loop to operating status. Alternatively, if the single loop requirements of the LCO are applied to the operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 1 B 3.4-7 Revision 45 Amendment No. 236 February 27, 2007 BASES ACTIONSA.1 (continued) The 24 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected. This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 1 B 3.4-8 Revision 45 Amendment No. 236 February 27, 2007 BASES ACTIONSB.1 (continued) With no recirculation loops in operation while in MODES 1 or 2 or the Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 1 B 3.4-9 Revision 0, 45 February 27, 2007 BASES (continued) SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 70% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 70% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop. The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered inoperable. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The 24 hour Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner. Recirculation Loops Operating B 3.4.1 BFN-UNIT 1 B 3.4-10 Revision 45, 87 Amendment No. 236 October 29, 2014 BASES (continued) REFERENCES 1.FSAR, Section 14.6.3.2.FSAR, Section 4.3.5.3.Deleted. 4.Deleted. 5.Deleted. 6.NRC No. 93-102, "Final Policy Statement on TechnicalSpecification Improvements," July 23, 1993.7.NEDO-24236, "Browns Ferry Nuclear Plant Units 1, 2, and 3,Single-Loop Operation," May 1981.8.NEDC-32484P, "Browns Ferry Nuclear Plant Units 1, 2, and3, SAFER/GESTR-LOCA Loss-of-Coolant AccidentAnalysis," Revision 6, February 2005.9.ANP-3015(P), "Browns Ferry Units 1, 2, and 3 LOCA BreakSpectrum Analysis," Revision 0, September 2011. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 2 B 3.3-15a Amendment No. 258 March 05, 1999 BASES APPLICABLE 2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, LCO, and The OPRM Upscale Function provides compliance with GDC 10 APPLICABILITY and GDC 12, thereby providing protection from exceeding the (continued) fuel MCPR safety limit (SL) due to anticipated thermal hydraulic power oscillations. References 13, 14, and 15 describe three algorithms for detecting thermal hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm. The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation of the OPRM algorithms. The OPRM Upscale Function is required to be OPERABLE when the plant is in a region of power flow operation where anticipated events could lead to thermal hydraulic instability and related neutron flux oscillations. Within this region, the automatic trip is enabled when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 25% RTP and reactor core flow, as indicated by recirculation drive flow is < 60% of rated flow, the operating region where actual thermal hydraulic oscillations may occur. Requiring the OPRM Upscale Function to be OPERABLE in MODE 1 provides consistency with operability requirements for other APRM functions and assures that the OPRM Upscale Function is OPERABLE whenever reactor power could increase into the region of concern without operator action. Insert 2.f 2.f. Oscillation Power Range Monitor (OPRM) Upscale The OPRM Upscale Function provides compliance with 10 CFR 50, Appendix A, General Design Criteria (GDC) 10 and 12, thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations. Reference 13 describes the Detect and Suppress - Confirmation Density (DSS-CD) long-term stability solution and the licensing basis Confirmation Density Algorithm (CDA). Reference 13 also describes the DSS-CD Armed Region and the three additional algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm (PBDA), the amplitude based algorithm (ABA), and the growth rate algorithm (GRA). All four algorithms are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the CDA. The remaining three algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY is based only on the CDA.

The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into cells for evaluation by the OPRM algorithms. DSS-CD operability requires at least 8 responsive OPRM cells per channel. The DSS-CD software includes a self-check for the responsive OPRM cells; therefore, no SR is necessary. The OPRM Upscale Function is required to be OPERABLE above a power level set at 5% of rated power below the lower boundary of the Armed Region, which is >18% RTP. This requirement is designed to encompass the region of power-flow operation where anticipated events could lead to thermal-hydraulic instability and related neutron flux oscillations. The OPRM Upscale Function is automatically trip-enabled when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 23% RTP corresponding to the MCPR monitoring threshold and reactor recirculation drive flow, is less than 75% of rated flow. This region is the OPRM Armed Region. Note (f) allows for entry into the DSS-CD Armed Region without automatic arming of DSS-CD prior to completely passing through the DSS-CD Armed Region during the first startup and the first shutdown following DSS-CD implementation. Insert 2.f (continued) 2.f. Oscillation Power Range Monitor (OPRM) Upscale If any OPRM auto-enable setpoint is in a non-conservative condition, i.e., the OPRM Upscale is not auto-enabled with RTP 23% and reactor recirculation drive flow 75% of rated, the associated channel is considered inopera for the OPRM Upscale function. Alternatively, the auto-enable setpoint may be adjusted to place the channel in a conservative condition (armed). If placed in the armed condition, the channel is considered OPERABLE. Note (f) reflects the need for plant data collection in order to test the DSS-CD equipment. Testing the DSS-CD equipment ensures its proper operation and prevents spurious reactor trips. Entry into the DSS-CD Armed Region without automatic arming of DSS-CD during this initial testing phase also allows for changes in plant operations to address maintenance or other operational needs. However, during this initial testing period, the OPRM upscale function is OPERABLE and DSS-CD operability and capability to automatically arm shall be maintained at recirculation drive flow rates above the DSS-CD Armed Region flow boundary. An OPRM Upscale trip is issued from an OPRM channel when the confirmation density algorithm in that channel detects oscillatory changes in the neutron flux, indicated by periodic confirmations and amplitude exceeding specified setpoints for a specified number of OPRM cells in the channel. An OPRM Upscale trip is also issued from the channel if any of the defense-in-depth algorithms (PBDA, ABA, GRA) exceed their trip condition for one or more cells in that channel. Three of the four channels are required to be operable. Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the SLMCPR is exceeded. There is no Allowable Value for this function. The OPRM Upscale Function is not LSSS SL-related (Reference 13) and Reference 14 confirms that the OPRM Upscale Function settings based on DSS-CD also do not have traditional instrumentation setpoints determined under an instrument setpoint methodology. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 2 B 3.3-15b Amendment No. 258 March05,1999BASES APPLICABLE 2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, (continued) LCO, and APPLICABILITY An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect growing oscillatory changes in the neutron flux for one or more cells in that channel. Three of the four channels are required to be OPERABLE. Each channel is capable of detecting thermal hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the MCPR SL is exceeded. There is no allowable value for this function. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 2 B 3.3-34 Amendment No. 258 March05,1999BASES ACTIONSD.1 (continued) Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition. E.1, F.1, G.1, and J.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Action E.1 is consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)." RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 2 B 3.3-35 Revision 14 Amendment No. 258 July 26, 2001 BASES ACTIONSH.1 (continued) If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. I.1 If OPRM Upscale trip capability is not maintained, Condition I exists. Reference 12 justified use of alternate methods to detect and suppress oscillations. The alternate methods are procedurally established consistent with the guidelines identified in Reference 17 requiring manual operator action to scram the plant if certain predefined events occur. The 12 hour allowed action time is based on engineering judgment to allow orderly transition to the alternate methods while limiting the period of time during which no automatic or alternate detect and suppress trip capability is formally in place. Based on the small probability of an instability event occurring at all, the 12 hours is judged to be reasonable. Insert Required Actions I, J, K ACTIONS I.1 If OPRM Upscale trip capability is not maintained, Condition I exists and Backup Stability Protection (BSP) is required. The Manual BSP Regions are described in Reference 13. The Manual BSP Regions are procedurally established consistent with the guidelines identified in Reference 13 and require specified manual operator actions if certain predefined operational conditions occur. The Completion Time of immediate is based on the importance of limiting the period of time during which no automatic or alternate detect and suppress trip capability is in place. I.2 and I.3 Required Actions I.2 and I.3 are both required to be taken in conjunction with Required Action I.1 if OPRM Upscale trip capability is not maintained. As described in Section 7.4 of Reference 13, the Automated BSP Scram Region is designed to avoid reactor instability by automatically preventing entry into the region of the power and flow operating map that is susceptible to reactor instability. The reactor trip would be initiated by the modified APRM Simulated Thermal Power-High scram setpoints for flow reduction events that would have terminated in the Manual BSP Region I. The Automated BSP Scram Region ensures an early scram and SLMCPR protection. The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, and based on the importance of restoring an automatic reactor trip for thermal hydraulic instability events. Backup Stability Protection is intended as a temporary means to protect against thermal-hydraulic instability events. The action should be initiated immediately to document the situation and prepare the report. The reporting requirements of Specification 5.6.7 document the corrective actions and schedule to restore the required channels to an OPERABLE status. The Completion Time of 90 days shown in Specification 5.6.7 is adequate to allow time to evaluate the cause of the inoperability and to determine the appropriate corrective actions and schedule to restore the required channels to OPERABLE status. J.1 If the Required Action I is not completed within the associated Completion Time, then Action J is required. The Bases for the Manual BSP Regions and associated Completion Time is addressed in the Bases for I.1. The Manual BSP Regions are required in conjunction with the BSP Boundary. J.2 The BSP Boundary, as described in Section 7.3 of Reference 13, defines an operating domain where potential instability events can be effectively addressed by the specified BSP manual operator actions. The BSP Boundary is constructed such that a flow reduction event initiated from this boundary and terminated at the core natural circulation line (NCL) would not exceed the Manual BSP Region I stability criterion. Potential instabilities would develop slowly as a result of the feedwater temperature transient (Reference 13). The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, to reach the specific condition from full power conditions in an orderly manner and without challenging plant system. Insert Required Actions I, J, K (continued) J.3 Backup Stability Protection (BSP) is a temporary means for protection against thermal-hydraulic instability events. An extended period of inoperability without automatic trip capability is not justified. Consequently, the required channels are required to be restored to OPERABLE status within 120 days. Based on engineering judgment, the likelihood of an instability event that could not be adequately handled by the use of the BSP Regions (See Required Action J.1) and the BSP Boundary (See Required Action J.2) during a 120-day period is negligibly small. The 120-day period is intended to allow for the case where limited design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment. This action is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to normally be accomplished within the completion times allowed for Actions for Conditions A and B. A Note is provided to indicate that LCO 3.0.4 is not applicable. The intent of the note is to allow plant startup while operating within the 120 day Completion Time for Required Action J.3. The primary purpose of this exclusion is to allow an orderly completion of design and verification activities, in the event of a required design change, without undue impact on plant operation. K.1 If the required channels are not restored to OPERABLE status and the Required Actions of Condition J are not met within the associated Completion Times, then the plant must be placed in an operating condition in which the LCO does not apply. To achieve this status, the plant must be brought to less than 18% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the specified operating power level from full power conditions in an orderly manner and without challenging plant systems. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 2 B 3.3-45a Amendment No. 258 March05,1999BASES SURVEILLANCE SR 3.3.1.1.17 REQUIREMENTS (continued) This SR ensures that scrams initiated from OPRM Upscale Function (Function 2.f) will not be inadvertently bypassed when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 25% RTP and core flow, as indicated by recirculation drive flow, is < 60% rated core flow. This normally involves confirming the bypass setpoints. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. The actual surveillance ensures that the OPRM Upscale Function is enabled (not bypassed) for the correct values of APRM Simulated Thermal Power and recirculation drive flow. Other surveillances ensure that the APRM Simulated Thermal Power and recirculation flow properly correlate with THERMAL POWER and core flow, respectively. If any bypass setpoint is nonconservative (i.e., the OPRM Upscale Function is bypassed when APRM Simulated Thermal Power 25% RTP and recirculation drive flow < 60% rated), then the affected channel is considered inoperable for the OPRM Upscale Function. Alternatively, the bypass setpoint may be adjusted to place the channel in a conservative condition (unbypass). If placed in the unbypassed condition, this SR is met and the channel is considered OPERABLE. The frequency of 24 months is based on engineering judgment and reliability of the components. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 2 B 3.3-46 Amendment No. 258 BASES (continued) REFERENCES 1.FSAR, Section 7.2.2.FSAR, Chapter 14.3.NEDO-23842, "Continuous Control Rod Withdrawal in theStartup Range," April 18, 1978.4.FSAR, Appendix N. 5.FSAR, Section 14.6.2. 6.FSAR, Section 6.5.7.FSAR, Section 14.5. 8.P. Check (NRC) letter to G. Lainas (NRC), "BWR ScramDischarge System Safety Evaluation," December 1, 1980.9.NEDC-30851-P-A , "Technical Specification ImprovementAnalyses for BWR Reactor Protection System," March 1988.10.NRC No. 93-102, "Final Policy Statement on TechnicalSpecification Improvements," July 23, 1993.11.MED-32-0286, "Technical Specification ImprovementAnalysis for Browns Ferry Nuclear Plant, Unit 2," October1995. 12.NEDC-32410P-A, "Nuclear Measurement Analysis andControl Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," October 1995.13.NEDO-31960-A, "BWR Owners' Group Long-Term StabilitySolutions Licensing Methodology," November 1995. Insert Refs 13.GE Hitachi Nuclear Energy, "GE Hitachi Boiling Water Reactor Detect and Suppress Solution -Confirmation Density," NEDC-33075P-A, Revision 8, November 2013.14.GEH letter to NRC, "NEDC-33075P-A, Detect and Suppress Solution - Confirmation Density(DSS-CD) Analytical Limit (TAC No. MD0277)," October 29, 2008. (ADAMS Accession No.ML083040052) RPS Instrumentation B 3.3.1.1 BFN-UNIT 2 B 3.3-46a Amendment No. 258 March05,1999BASES REFERENCES 14. NEDO-31960-A, Supplement 1, "BWR Owners' Group(continued) Long-Term Stability Solutions Licensing Methodology," November1995.15.NEDO-32465-A, "BWR Owners' Group Long-Term StabilityDetect and Suppress Solutions Licensing Basis Methodologyand Reload Applications," August 1996.16.NEDC-32410P-A, Supplement 1, "Nuclear MeasurementAnalysis and Control Power Range Neutron Monitor(NUMAC PRNM) Retrofit Plus Option III Stability TripFunction," August 1996.17.Letter, L.A. England (BWROG) to M.J. Virgilio, "BWROwners' Group Guidelines for Stability Interim CorrectiveAction," June 6, 1994. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 2 B 3.4-4 Amendment No. 258 and Revision 3 March 05, 1999/March 19, 1999 BASES APPLICABLE Plant specific LOCA analyses have been performed assuming SAFETY ANALYSES only one operating recirculation loop. These analyses have (continued) demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Refs. 7 and 8). The transient analyses of Chapter 14 of the FSAR have also been performed for single recirculation loop operation (Ref. 7) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoint is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Flow Biased Simulated Thermal Power-High setpoint is in LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation." Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 6). Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 2 B 3.4-5 Amendment No. 258 March05,1999BASES (continued) LCO Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation. With only one recirculation loop in operation, modifications to the required APLHGR Limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LCO 3.2.2, 'MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM Flow Biased Simulated Thermal Power-High Setpoint (LCO 3.3.1.1) may be applied to allow continued operation consistent with the assumptions of References 7 and 8. APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 2 B 3.4-6 Amendment No. 258 March05,1999BASES (continued) ACTIONSA.1With the requirements of the LCO not met, the recirculation loops must be restored to operation with matched flows within 24 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the inoperable loop to operating status. Alternatively, if the single loop requirements of the LCO are applied to the operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 2 B 3.4-7 Amendment No. 258 March05,1999BASES ACTIONS A.1 (continued) The 24 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected. This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 2 B 3.4-8 Amendment No. 258 March 05, 1999 BASES ACTIONS B.1 (continued) With no recirculation loops in operation while in MODES 1 or 2 or the Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 2 B 3.4-9 Amendment No. 258 March 05, 1999 BASES (continued) SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 70% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 70% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop. The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered inoperable. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The 24 hour Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner. Recirculation Loops Operating B 3.4.1 BFN-UNIT 2 B 3.4-10 Amendment No. 258 Revision 87 October 29, 2014 BASES (continued) REFERENCES 1. FSAR, Section 14.6.3.

2. FSAR, Section 4.3.5.

3. Deleted.

4. Deleted.

5. Deleted.

6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

7. NEDO-24236, "Browns Ferry Nuclear Plant Units 1, 2, and 3, Single-Loop Operation," May 1981.

8. NEDC-32484P, "Browns Ferry Nuclear Plant Units 1, 2, and 3, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," Revision 2, December 1997. 9. ANP-3015(P), "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis," Revision 0, September 2011.

RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 3 B 3.3-15a Amendment No. 221 September 27, 1999 BASES APPLICABLE 2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, LCO, and The OPRM Upscale Function provides compliance with GDC 10 APPLICABILITY and GDC 12, thereby providing protection from exceeding the (continued) fuel MCPR safety limit (SL) due to anticipated thermal hydraulic power oscillations. References 13, 14, and 15 describe three algorithms for detecting thermal hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm. The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation of the OPRM algorithms. The OPRM Upscale Function is required to be OPERABLE when the plant is in a region of power flow operation where anticipated events could lead to thermal hydraulic instability and related neutron fux oscillations. Within this region, the automatic trip is enabled when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 25% RTP and reactor core flow, as indicated by recirculation drive flow is < 60% of rated flow, the operating region where actual thermal hydraulic oscillations may occur. Requiring the OPRM Upscale Function to be OPERABLE in MODE 1 provides consistency with operability requirements for other APRM functions and assures that the OPRM Upscale Function is OPERABLE whenever reactor power could increase into the region of concern without operator action. Insert 2.f 2.f. Oscillation Power Range Monitor (OPRM) Upscale The OPRM Upscale Function provides compliance with 10 CFR 50, Appendix A, General Design Criteria (GDC) 10 and 12, thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations. Reference 13 describes the Detect and Suppress - Confirmation Density (DSS-CD) long-term stability solution and the licensing basis Confirmation Density Algorithm (CDA). Reference 13 also describes the DSS-CD Armed Region and the three additional algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm (PBDA), the amplitude based algorithm (ABA), and the growth rate algorithm (GRA). All four algorithms are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the CDA. The remaining three algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY is based only on the CDA.

The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into cells for evaluation by the OPRM algorithms. DSS-CD operability requires at least 8 responsive OPRM cells per channel. The DSS-CD software includes a self-check for the responsive OPRM cells; therefore, no SR is necessary. The OPRM Upscale Function is required to be OPERABLE above a power level set at 5% of rated power below the lower boundary of the Armed Region, which is >18% RTP. This requirement is designed to encompass the region of power-flow operation where anticipated events could lead to thermal-hydraulic instability and related neutron flux oscillations. The OPRM Upscale Function is automatically trip-enabled when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 23% RTP corresponding to the MCPR monitoring threshold and reactor recirculation drive flow, is less than 75% of rated flow. This region is the OPRM Armed Region. Note (f) allows for entry into the DSS-CD Armed Region without automatic arming of DSS-CD prior to completely passing through the DSS-CD Armed Region during the first startup and the first shutdown following DSS-CD implementation. Insert 2.f (continued) 2.f. Oscillation Power Range Monitor (OPRM) Upscale If any OPRM auto-enable setpoint is in a non-conservative condition, i.e., the OPRM Upscale is not auto-enabled with RTP 23% and reactor recirculation drive flow 75% of rated, the associated channel is considered inopera for the OPRM Upscale function. Alternatively, the auto-enable setpoint may be adjusted to place the channel in a conservative condition (armed). If placed in the armed condition, the channel is considered OPERABLE. Note (f) reflects the need for plant data collection in order to test the DSS-CD equipment. Testing the DSS-CD equipment ensures its proper operation and prevents spurious reactor trips. Entry into the DSS-CD Armed Region without automatic arming of DSS-CD during this initial testing phase also allows for changes in plant operations to address maintenance or other operational needs. However, during this initial testing period, the OPRM upscale function is OPERABLE and DSS-CD operability and capability to automatically arm shall be maintained at recirculation drive flow rates above the DSS-CD Armed Region flow boundary. An OPRM Upscale trip is issued from an OPRM channel when the confirmation density algorithm in that channel detects oscillatory changes in the neutron flux, indicated by periodic confirmations and amplitude exceeding specified setpoints for a specified number of OPRM cells in the channel. An OPRM Upscale trip is also issued from the channel if any of the defense-in-depth algorithms (PBDA, ABA, GRA) exceed their trip condition for one or more cells in that channel. Three of the four channels are required to be operable. Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the SLMCPR is exceeded. There is no Allowable Value for this function. The OPRM Upscale Function is not LSSS SL-related (Reference 13) and Reference 14 confirms that the OPRM Upscale Function settings based on DSS-CD also do not have traditional instrumentation setpoints determined under an instrument setpoint methodology. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 3 B 3.3-15b Amendment No. 221 September 27, 1999 BASES APPLICABLE 2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, (continued) LCO, and APPLICABILITY An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect growing oscillatory changes in the neutron flux for one or more cells in that channel. Three of the four channels are required to be OPERABLE. Each channel is capable of detecting thermal hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the MCPR SL is exceeded. There is no allowable value for this function. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 3 B 3.3-34 Amendment No. 213, 221 Revision0 September 27, 1999 BASES ACTIONSD.1 (continued) Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition. E.1, F.1, G.1, and J.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Action E.1 is consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)." RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 3 B 3.3-35 Amendment No. 221, 231 Revision 0 September 13, 2001 BASES ACTIONSH.1 (continued) If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. I.1 If OPRM Upscale trip capability is not maintained, Condition I exists. Reference 12 justified use of alternate methods to detect and suppress oscillations. The alternate methods are procedurally established consistent with the guidelines identified in Reference 17 requiring manual operator action to scram the plant if certain predefined events occur. The 12 hour allowed action time is based on engineering judgment to allow orderly transition to the alternate methods while limiting the period of time during which no automatic or alternate detect and suppress trip capability is formally in place. Based on the small probability of an instability event occurring at all, the 12 hours is judged to be reasonable. Insert Required Actions I, J, K ACTIONS I.1 If OPRM Upscale trip capability is not maintained, Condition I exists and Backup Stability Protection (BSP) is required. The Manual BSP Regions are described in Reference 13. The Manual BSP Regions are procedurally established consistent with the guidelines identified in Reference 13 and require specified manual operator actions if certain predefined operational conditions occur. The Completion Time of immediate is based on the importance of limiting the period of time during which no automatic or alternate detect and suppress trip capability is in place. I.2 and I.3 Required Actions I.2 and I.3 are both required to be taken in conjunction with Required Action I.1 if OPRM Upscale trip capability is not maintained. As described in Section 7.4 of Reference 13, the Automated BSP Scram Region is designed to avoid reactor instability by automatically preventing entry into the region of the power and flow operating map that is susceptible to reactor instability. The reactor trip would be initiated by the modified APRM Simulated Thermal Power-High scram setpoints for flow reduction events that would have terminated in the Manual BSP Region I. The Automated BSP Scram Region ensures an early scram and SLMCPR protection. The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, and based on the importance of restoring an automatic reactor trip for thermal hydraulic instability events. Backup Stability Protection is intended as a temporary means to protect against thermal-hydraulic instability events. The action should be initiated immediately to document the situation and prepare the report. The reporting requirements of Specification 5.6.7 document the corrective actions and schedule to restore the required channels to an OPERABLE status. The Completion Time of 90 days shown in Specification 5.6.7 is adequate to allow time to evaluate the cause of the inoperability and to determine the appropriate corrective actions and schedule to restore the required channels to OPERABLE status. J.1 If the Required Action I is not completed within the associated Completion Time, then Action J is required. The Bases for the Manual BSP Regions and associated Completion Time is addressed in the Bases for I.1. The Manual BSP Regions are required in conjunction with the BSP Boundary. J.2 The BSP Boundary, as described in Section 7.3 of Reference 13, defines an operating domain where potential instability events can be effectively addressed by the specified BSP manual operator actions. The BSP Boundary is constructed such that a flow reduction event initiated from this boundary and terminated at the core natural circulation line (NCL) would not exceed the Manual BSP Region I stability criterion. Potential instabilities would develop slowly as a result of the feedwater temperature transient (Reference 13). The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, to reach the specific condition from full power conditions in an orderly manner and without challenging plant system. Insert Required Actions I, J, K (continued) J.3 Backup Stability Protection (BSP) is a temporary means for protection against thermal-hydraulic instability events. An extended period of inoperability without automatic trip capability is not justified. Consequently, the required channels are required to be restored to OPERABLE status within 120 days. Based on engineering judgment, the likelihood of an instability event that could not be adequately handled by the use of the BSP Regions (See Required Action J.1) and the BSP Boundary (See Required Action J.2) during a 120-day period is negligibly small. The 120-day period is intended to allow for the case where limited design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment. This action is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to normally be accomplished within the completion times allowed for Actions for Conditions A and B. A Note is provided to indicate that LCO 3.0.4 is not applicable. The intent of the note is to allow plant startup while operating within the 120 day Completion Time for Required Action J.3. The primary purpose of this exclusion is to allow an orderly completion of design and verification activities, in the event of a required design change, without undue impact on plant operation. K.1 If the required channels are not restored to OPERABLE status and the Required Actions of Condition J are not met within the associated Completion Times, then the plant must be placed in an operating condition in which the LCO does not apply. To achieve this status, the plant must be brought to less than 18% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the specified operating power level from full power conditions in an orderly manner and without challenging plant systems. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 3 B 3.3-45a Amendment No. 221 September 27, 1999 BASES SURVEILLANCE SR 3.3.1.1.17 REQUIREMENTS (continued) This SR ensures that scrams initiated from OPRM Upscale Function (Function 2.f) will not be inadvertently bypassed when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 25% RTP and core flow, as indicated by recirculation drive flow, is < 60% rated core flow. This normally involves confirming the bypass setpoints. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. The actual surveillance ensures that the OPRM Upscale Function is enabled (not bypassed) for the correct values of APRM Simulated Thermal Power and recirculation drive flow. Other surveillances ensure that the APRM Simulated Thermal Power and recirculation flow properly correlate with THERMAL POWER and core flow, respectively. If any bypass setpoint is nonconservative (i.e., the OPRM Upscale Function is bypassed when APRM Simulated Thermal Power 25% RTP and recirculation drive flow < 60% rated), then the affected channel is considered inoperable for the OPRM Upscale Function. Alternatively, the bypass setpont may be adjusted to place the channel in a conservative condition (unbypass). If placed in the unbypassed condition, this SR is met and the channel is considered OPERABLE. The frequency of 24 months is based on engineering judgment and reliability of the components. RPS Instrumentation B 3.3.1.1 (continued) BFN-UNIT 3 B 3.3-46 Amendment No. 213, 215, 221 Revision0 September 27, 1999 BASES (continued) REFERENCES 1.FSAR, Section 7.2.2.FSAR, Chapter 14.3.NEDO-23842, "Continuous Control Rod Withdrawal in theStartup Range," April 18, 1978.4.FSAR, Appendix N. 5.FSAR, Section 14.6.2. 6.FSAR, Section 6.5.7.FSAR, Section 14.5. 8.P. Check (NRC) letter to G. Lainas (NRC), "BWR ScramDischarge System Safety Evaluation," December 1, 1980.9.NEDC-30851-P-A , "Technical Specification ImprovementAnalyses for BWR Reactor Protection System," March 1988.10.NRC No. 93-102, "Final Policy Statement on TechnicalSpecification Improvements," July 23, 1993.11.MED-32-0286, "Technical Specification ImprovementAnalysis for Browns Ferry Nuclear Plant, Unit 2," October1995. 12.NEDC-32410P-A, "Nuclear Measurement Analysis andControl Power Range Neutron Monitor (NUMAC PRNM)Retrofit Plus Option III Stability Trip Function," October 1995.13.NEDO-31960-A, "BWR Owners' Group Long-Term StabilitySolutions Licensing Methodology," November 1995. Insert Refs 13.GE Hitachi Nuclear Energy, "GE Hitachi Boiling Water Reactor Detect and Suppress Solution -Confirmation Density," NEDC-33075P-A, Revision 8, November 2013.14.GEH letter to NRC, "NEDC-33075P-A, Detect and Suppress Solution - Confirmation Density(DSS-CD) Analytical Limit (TAC No. MD0277)," October 29, 2008. (ADAMS Accession No.ML083040052) RPS Instrumentation B 3.3.1.1 BFN-UNIT 3 B 3.3-46a Amendment No. 221 September 27, 1999 BASES REFERENCES 14. NEDO-31960-A, Supplement 1, "BWR Owners' Group (continued) Long-Term Stability Solutions Licensing Methodology,"November 199515.NEDO-32465-A, "BWR Owners' Group Long-Term StabilityDetect and Suppress Solutions Licensing Basis Methodologyand Reload Applications," August 1996.16.NEDC-32410P-A, Supplement 1, "Nuclear MeasurementAnalysis and Control Power Range Neutron Monitor(NUMAC PRNM) Retrofit Plus Option III Stability TripFunction," August 1996.17.Letter, L. A. England (BWROG) to M. J. Virgilio, "BWROwners' Group Guidelines for Stability Interim CorrectiveAction," June 6, 1994. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 3 B 3.4-4 Amendment No. 216, 221 Revision 0, 3 September 27, 1999 BASES APPLICABLE Plant specific LOCA analyses have been performed assuming SAFETY ANALYSES only one operating recirculation loop. These analyses have (continued) demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Refs. 7 and 8). The transient analyses of Chapter 14 of the FSAR have also been performed for single recirculation loop operation (Ref. 7) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoint is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Flow Biased Simulated Thermal Power-High setpoint is in LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation." Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 6). Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 3 B 3.4-5 Amendment No. 213, 214, 216, 221 Revision0 September 27, 1999 BASES (continued) LCO Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation. With only one recirculation loop in operation, modifications to the required APLHGR Limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LCO 3.2.2, 'MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM Flow Biased Simulated Thermal Power-High Setpoint (LCO 3.3.1.1) may be applied to allow continued operation consistent with the assumptions of References 7 and 8. APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 3 B 3.4-6 Amendment No. 216, 221 Revision0 September 27, 1999 BASES (continued) ACTIONSA.1 With the requirements of the LCO not met, the recirculation loops must be restored to operation with matched flows within 24 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the inoperable loop to operating status. Alternatively, if the single loop requirements of the LCO are applied to the operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 3 B 3.4-7 Amendment No. 216, 221 Revision0 September 27, 1999 BASES ACTIONSA.1 (continued) The 24 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected. This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 3 B 3.4-8 Amendment No. 216, 221 Revision0 September 27, 1999 BASES ACTIONSB.1 (continued) With no recirculation loops in operation while in MODES 1 or 2 or the Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. Recirculation Loops Operating B 3.4.1 (continued) BFN-UNIT 3 B 3.4-9 Amendment No. 221 Revision0 September 27, 1999 BASES (continued) SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 70% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 70% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop. The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered inoperable. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The 24 hour Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner. Recirculation Loops Operating B 3.4.1 BFN-UNIT 3 B 3.4-10 Amendment No. 213, 216, 221 Revision0, 87 October 29, 2014 BASES (continued) REFERENCES 1.FSAR, Section 14.6.3.2.FSAR, Section 4.3.5.3.Deleted. 4.Deleted. 5.Deleted. 6.NRC No. 93-102, "Final Policy Statement on TechnicalSpecification Improvements," July 23, 1993.7.NEDO-24236, "Browns Ferry Nuclear Plant Units 1, 2, and 3,Single-Loop Operation," May 1981.8.NEDC-32484P, "Browns Ferry Nuclear Plant Units 1, 2, and3, SAFER/GESTR-LOCA Loss-of-Coolant AccidentAnalysis," Revision 2, December 1997.9.ANP-3015(P), "Browns Ferry Units 1, 2, and 3 LOCA BreakSpectrum Analysis," Revision 0, September 2011. BASES (continued) REFERENCES BFN-UNIT 3 1. FSAR, Section 14.6.3. 2. FSAR, Section 4.3.5. 3. Deleted. 4. Deleted. 5. Deleted. Recirculation Loops Operating B 3.4.1 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993. 7. NED0-24236, "Browns Ferry Nuclear Plant Units 1, 2, and 3, Single-Loop Operation," May 1981. 8. NEDC-32484P, "Browns Ferry Nuclear Plant Units 1, 2, and 3, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," Revision 2, December 1997. 9. ANP-3015(P), "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis," Revision 0, September 2011. 10. NEDC-33006P-A, "General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report," Revision 3, June 2009. B 3.4-10 Amendment Revision 0, 87,

GE Hitachi Nuclear EnergyNon-Proprietary Information - Class I (Public) SAFETY ANALYSIS REPORT FOR BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 MAXIMUM EXTENDED LOAD LINE LIMIT ANALYSIS PLUS

Copyright 2018 GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved INFORMATION NOTICE IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully

TABLE OF CONTENTS Page Executive Summary ..................................................................................................................... ixAcronyms ...................................................................................................................................... xi1.0Introduction ..................................................................................................................... 1-12.0Reactor Core and Fuel Performance ............................................................................ 2-13.0Reactor Coolant and Connected Systems ..................................................................... 3-14.0Engineered Safety Features ............................................................................................ 4-15.0Instrumentation and Control ......................................................................................... 5-1 6.0Electrical Power and Auxiliary Systems ....................................................................... 6-17.0Power Conversion Systems ............................................................................................ 7-18.0Radwaste Systems and Radiation Sources .................................................................... 8-19.0Reactor Safety Performance Evaluations ..................................................................... 9-110.0Other Evaluations ......................................................................................................... 10-111.0Licensing Evaluations ................................................................................................... 11-112.0References ...................................................................................................................... 12-1 Appendices

List of Tables Table Title Page

List of Figures Figure Title Page EXECUTIVE SUMMARY

ACRONYMS Term Definition Term Definition Term Definition Term Definition Term Definition Term Definition Term Definition Term Definition

1.0INTRODUCTION

1.1REPORT APPROACH 1.1.1Generic Assessments

1.1.2Plant-Specific Evaluation 1.1.3Computer Codes and Methods

1.1.4Scope of Evaluations Section 2.0, Reactor Core and Fuel Performance: Section 3.0, Reactor Coolant and Connected Systems: Section 4.0, Engineered Safety Features: Section 5.0, Instrumentation and Control: Section 6.0, Electrical Power and Auxiliary Systems: Section 7.0, Power Conversion Systems: Section 8.0, Radwaste Systems and Radiation Sources: Section 9.0, Reactor Safety Performance Evaluations: Section 10.0, Other Evaluations: Section 11.0, Licensing Evaluations:1.1.5Product Line Applicability 1.1.6Report Generation and Review Process - GEH Scope

1.1.7Report Generation and Review Process - TVA Scope 1.2OPERATING CONDITIONS AND CONSTRAINTS 1.2.1Power/Flow Map

1.2.2Reactor Heat Balance 1.2.3Core and Reactor Conditions

1.2.4Operational Enhancements Operational Enhancements MELLLA+ BFN M+SAR

1.3SUMMARY AND CONCLUSIONS

Table 1-1 Computer Codes Used in the M+SAR Evaluations Task Computer Code* Version orRevision NRC Approved Comments

Notes for Table 1-1:

Table 1-2 Core Thermal Power to Core Flow RatiosSteady-State Operation Point on the P/F Map Core Thermal Power (MWt/%CLTP)Core Flow (Mlbm/hr/%rated) Power-to-Flow Ratio (MWt/Mlbm/hr)

Table 1-3 Comparison of Thermal-Hydraulic Parameters Parameter MELLLA 100% CLTP, 99% Core Flow MELLLA 100% CLTP, 99% Core Flow,FFWTR/ FWHOOS MELLLA+ 100% CLTP, 85% Core Flow MELLLA+ 100% CLTP, 85% Core FlowNFWT1 -10F MELLLA+ 77.6% CLTP, 55% Core Flow MELLLA+ 77.6% CLTP, 55% Core Flow NFWT -10F Note: Figure 1-1 Power/Flow Operating Map for MELLLA+ 2.0REACTOR CORE AND FUEL PERFORMANCE 2.1FUEL DESIGN AND OPERATION 2.1.1Fuel Product Line 2.1.2Core Design and Fuel Thermal Monitoring Threshold 2.2THERMAL LIMITS ASSESSMENT 2.2.1Safety Limit Minimum Critical Power Ratio 2.2.2Operating Limit Minimum Critical Power Ratio 2.2.3Maximum Average Planar Linear Heat Generation Rate Limits 2.2.4 Linear Heat Generation Rate Limits

2.2.5Power-to-Flow Ratio 2.3REACTIVITY CHARACTERISTICS 2.3.1Hot Excess Reactivity 2.3.2Strong Rod Out Shutdown Margin 2.3.3SLCS Shutdown Margin 2.4STABILITY Topic M+LTR Disposition BFN Result 2.4.1DSS-CD Setpoints

.

2.4.1.1DSS-CD Diversity

2.4.2Armed Region 2.4.3Backup Stability Protection

2.4.4M+LTR SER Limitation and Condition 12.5.b 2.5REACTIVITY CONTROL Topic M+LTR Disposition BFN Result 2.5.1Control Rod Scram 2.5.2Control Rod Drive Positioning and Cooling 2.5.3Control Rod Drive Integrity

2.6ADDITIONAL LIMITATIONS AND CONDITIONS RELATED TO REACTOR CORE AND FUEL PERFORMANCE 2.6.1TGBLA/PANAC Version 2.6.2M+LTR SER Limitation and Condition 12.24.1 Table 2-1 [[ ]] [[ ]] Table 2-2 [[ ]] Note: Table 2-3 [[ ]] Note: Figure 2-1 Required OPRM Armed Region 3.0REACTOR COOLANT AND CONNECTED SYSTEMS 3.1NUCLEAR SYSTEM PRESSURE RELIEF AND OVERPRESSURE PROTECTION Topic M+LTR Disposition BFN Result 3.1.1Flow-Induced Vibration 3.1.2Overpressure Relief Capacity

3.2REACTOR VESSEL Topic M+LTR Disposition BFN Result 3.2.1Fracture Toughness

3.2.2Reactor Vessel Structural Evaluation

3.3REACTOR INTERNALS 3.3.1Reactor Internal Pressure Differences Topic M+LTR Disposition BFN Result

3.3.1.1Fuel Assembly Lift Forces

3.3.1.2Reactor Internal Pressure Differences for Normal, Upset, Emergency and Faulted Conditions

3.3.1.3Reactor Internal Pressure Differences (Acoustic and Flow-Induced Loads) for Faulted Conditions 3.3.2Reactor Internals Structural Evaluation Load Category MELLLA+ Results for Normal, Upset and Emergency Conditions Load Category MELLLA+ Results for Normal, Upset and Emergency Conditions 3.3.2.1Reactor Internals Structural Evaluation for Faulted Conditions Load Category MELLLA+ Results for Faulted Conditions

3.3.3Steam Separator and Dryer Performance 3.4FLOW-INDUCED VIBRATION Topic M+LTR Disposition BFN Result

3.4.1FIV Influence on Piping

3.4.2FIV Influence on Reactor Internals Component(s) MELLLA+ Results

Component(s) MELLLA+ Results 3.5PIPING EVALUATION 3.5.1Reactor Coolant Pressure Boundary Piping Topic M+LTR Disposition BFN Result Topic M+LTR Disposition BFN Result

3.5.1.1Main Steam and Feedwater Piping Inside Containment 3.5.1.2Reactor Recirculation and Control Rod Drive Systems 3.5.1.3Other RCPB Piping Systems 3.5.1.3.1 Other RCPB Piping Systems - CS, RHR/LPCI, and SLCS 3.5.1.3.2 Other RCPB Piping Systems - RPV Head Vent Line and MSRV Discharge Lines 3.5.1.3.3 Other RCPB Piping Systems - RWCU 3.5.1.3.4 Other RCPB Piping Systems - Safety-Related Thermowells

3.5.1.3.5 Other RCPB Piping Systems - Conclusion 3.5.1.4Other-Than-Category "A" RCPB Material

3.5.2Balance-of-Plant Piping Topic M+LTR Disposition BFN Result

3.5.2.1Main Steam and Feedwater (Outside Containment)

3.5.2.2Other BOP Piping Systems 3.5.2.2.1 Other BOP Piping Systems - RCIC, HPCI, CS, and RHR 3.5.2.2.2 Other BOP Piping Systems - Offgas, Containment Air Monitoring, and Neutron Monitoring Systems 3.6REACTOR RECIRCULATION SYSTEM Topic M+LTR Disposition BFN Result 3.6.1System Evaluation 3.6.2Net Positive Suction Head

3.6.3Single Loop Operation 3.6.4Flow Mismatch 3.7MAIN STEAM LINE FLOW RESTRICTORS Topic M+LTR Disposition BFN Result 3.8MAIN STEAM ISOLATION VALVES Topic M+LTR Disposition BFN Result

3.9REACTOR CORE ISOLATION COOLING Topic M+LTR Disposition BFN Result 3.9.1System Hardware 3.9.2System Initiation

3.9.3Net Positive Suction Head

3.9.4Inventory Makeup Level Margin to TAF 3.10RESIDUAL HEAT REMOVAL SYSTEM Topic M+LTR Disposition BFN Result 3.10.1 LPCI Mode 3.10.2Suppression Pool and Containment Spray Cooling Modes 3.10.3Shutdown Cooling Mode 3.10.4Steam Condensing Mode 3.10.5Fuel Pool Cooling Assist Mode 3.11REACTOR WATER CLEANUP SYSTEM Topic M+LTR Disposition BFN Result 3.11.1System Performance 3.11.2Containment Isolation

Table 3-1 Key Results for RPV Flux/Fluence Item Parameter Unit CLTP Value MELLLA+ Value MELLLA+ to CLTP Comparison

Table 3-2a BFN Unit 1 Equivalent Margin Analysis - Plate Material EMA Plant Applicability Verification Form for BFN Unit 1 60 Years (38 EFPY) BWR/3-6 PLATE Surveillance Plate USE: Limiting Beltline Plate USE ([[ ]]): < [[ ]] Therefore, vessel plates are bounded by EMA Table 3-2b BFN Unit 1 Equivalent Margin Analysis - Nozzle Material EMA Plant Applicability Verification Form for BFN Unit 1 60 Years (38 EFPY) BWR/3-6 NOZZLE [1] Surveillance Plate USE: Limiting Beltline Plate USE ([[ ]]): [[ ]]< [[ ]] Therefore, vessel nozzles are bounded by EMA Note 1 Table 3-2c BFN Unit 1 Equivalent Margin Analysis - Weld Material EMA Plant Applicability Verification Form for BFN Unit 1 60 Years (38 EFPY) BWR/2-6 WELD Surveillance Weld USE (406L44): Limiting Beltline Weld USE ([[ ]]): [[ ]] < [[ ]] Therefore, vessel welds are bounded by EMA Table 3-3a BFN Unit 2 Equivalent Margin Analysis - Plate Material EMA Plant Applicability Verification Form for BFN Unit 2 60 Years (48 EFPY) BWR/3-6 PLATE Surveillance Plate USE (A0981-1): Limiting Beltline Plate USE ([[ ]]): [[ ]] < [[ ]] Therefore, vessel plates are bounded by EMA Note 1 Table 3-3b BFN Unit 2 Equivalent Margin Analysis - Nozzle Material EMA Plant Applicability Verification Form for BFN Unit 2 60 Years (48 EFPY) BWR/3-6 NOZZLE [1] Surveillance Plate USE: Limiting Beltline Plate USE ([[ ]]): [[ ]] < [[ ]] Therefore, vessel nozzles are bounded by EMA Note 1 Table 3-3c BFN Unit 2 Equivalent Margin Analysis - Weld Material EMA Plant Applicability Verification Form for BFN Unit 2 60 Years (48 EFPY) BWR/2-6 WELD Surveillance Weld USE (BF2 ESW): Limiting Beltline Weld USE ([[ ]]): [[ ]] < [[ ]] Therefore, vessel welds are bounded by EMA Table 3-4a BFN Unit 3 Equivalent Margin Analysis - Plate Material EMA Plant Applicability Verification Form for BFN Unit 3 60 Years (54 EFPY) BWR/3-6 PLATE Surveillance Plate USE: Limiting Beltline Plate USE ([[ ]]): [[ ]] < [[ ]] Therefore, vessel plates are bounded by EMA Table 3-4b BFN Unit 3 Equivalent Margin Analysis - Nozzle Material EMA Plant Applicability Verification Form for BFN Unit 3 60 Years (54 EFPY) BWR/3-6 NOZZLE [1] Surveillance Plate USE: Limiting Beltline Plate USE ([[ ]]): [[ ]] < [[ ]] Therefore, vessel nozzles are bounded by EMA Note 1 Table 3-4c BFN Unit 3 Equivalent Margin Analysis - Weld Material EMA Plant Applicability Verification Form for BFN Unit 3 60 Years (54 EFPY) BWR/2-6 WELD Surveillance Weld USE: Limiting Beltline Weld USE ([[ ]]): [[ ]] < [[ ]] Therefore, vessel welds are bounded by EMA Table 3-5a Adjusted Reference Temperatures for Unit 1 (38 EFPY) Lower-Intermediate Plates Lower Plates & Lower to Lower-Intermediate Girth Weld Axial Welds Water Level Instrumentation Nozzle Table 3-5a Adjusted Reference Temperatures for Unit 1 (38 EFPY) (Continued) Component Heat % Cu % Ni CF Adjusted CF Initial RTNDT°F 1/4T Fluence n/cm2 38 EFPY RTNDT°F I Margin°F 38 EFPY Shift °F 38 EFPY ART °F PLANT-SPECIFIC CHEMISTRIES Plates:Welds:Nozzles:BEST ESTIMATE CHEMISTRIES:Integrated Surveillance Program (ISP) (BWRVIP-135 R3 (Reference 30)): Notes for Table 3-5a Table 3-5b Adjusted Reference Temperatures for Unit 2 (48 EFPY) Lower-Intermediate Plates Lower Plates & Lower to Lower-Intermediate Girth Weld Intermediate Plates & Intermediate to Lower-Intermediate Girth Weld Axial Welds Water Level Instrumentation Nozzle Table 3-5b Adjusted Reference Temperatures for Unit 2 (48 EFPY) (Continued) Component Heat % Cu % Ni CF Adjusted CF Initial RTNDT°F 1/4T Fluence n/cm2 48 EFPY RTNDT°F I Margin°F 48 EFPY Shift °F 48 EFPY ART °F PLANT-SPECIFIC CHEMISTRIES Plates:Welds: Nozzles:BEST ESTIMATE CHEMISTRIES:ISP (BWRVIP-135 R3 (Reference 30)): Notes for Table 3-5b Table 3-5c Adjusted Reference Temperatures for Unit 3 (54 EFPY) Lower-Intermediate Plates Lower Plates & Lower to Lower-Intermediate Girth Weld Intermediate Plates & Intermediate to Lower-Intermediate Girth Weld Axial Welds Water Level Instrumentation Nozzle Table 3-5c Adjusted Reference Temperatures for Unit 3 (54 EFPY) (Continued) Component Heat % Cu % Ni CF Adjusted CF Initial RTNDT °F 1/4T Fluence n/cm2 54 EFPY RTNDT°F I Margin°F 54 EFPY Shift °F 54 EFPY ART °F PLANT-SPECIFIC CHEMISTRIES Plates: Welds:Nozzles:BEST ESTIMATE CHEMISTRIES:ISP (BWRVIP-135 R3 (Reference 30)): Notes for Table 3-5c Table 3-6 Flow Rate Comparisons Item Key Parameter Unit CLTP Value MELLLA+Value Table 3-7 Summary of Reactor Coolant Pressure Boundary Welds Summary of Weld Types per GL 88-01/BWRVIP-75-A Classifications IGSCC Weld CategoryUnit 1Unit 2Unit 3 Table 3-8 RCPB Piping and Safe End Materials of Construction Nozzle Designation / SystemUnit 1 Unit 2 Unit 3

Nozzle Designation / SystemUnit 1 Unit 2 Unit 3

Nozzle Designation / SystemUnit 1 Unit 2 Unit 3

Note:

4.0ENGINEERED SAFETY FEATURES 4.1CONTAINMENT SYSTEM PERFORMANCE Topic M+LTR Disposition BFN Result

4.1.1Short-Term Pressure and Temperature Response

4.1.1.1Long-Term Suppression Pool Cooling Temperature Response

4.1.2Containment Dynamic Loads 4.1.2.1LOCA Loads

Vent Thrust Loads Pool Swell Loads Condensation Oscillation Chugging Load 4.1.2.2Subcompartment Pressurization Subcompartment Pressurization Evaluation Results for SSW Subcompartment Pressurization Evaluation Results for SSW Doors Conclusions for Subcompartment Pressurization Evaluations

4.1.2.3MSRV Piping - Containment Dynamic Loads

4.1.2.4MSRV Containment Dynamic Loads

4.1.3Containment Isolation

4.1.4Generic Letter 89-10 4.1.5Generic Letter 89-16 4.1.6Generic Letter 95-07 4.1.7Generic Letter 96-06 4.2EMERGENCY CORE COOLING SYSTEMS Topic M+LTR Disposition BFN Result 4.2.1High Pressure Coolant Injection

4.2.2High Pressure Core Spray 4.2.3Core Spray

4.2.4Low Pressure Coolant Injection

4.2.5Automatic Depressurization System 4.2.6ECCS Net Positive Suction Head

4.2.6.1ECCS NPSH During ATWS

NPSHReff = (1 + uncertainty)NPSHR3%

4.3EMERGENCY CORE COOLING SYSTEM PERFORMANCE 4.3.1Break Spectrum Response and Limiting Single Failure 4.3.2Large Break Peak Clad Temperature 4.3.3Small Break Peak Clad Temperature 4.3.4Local Cladding Oxidation 4.3.5Core Wide Metal Water Reaction 4.3.6Coolable Geometry 4.3.7Long Term Cooling 4.3.8Flow Mismatch Limits 4.4MAIN CONTROL ROOM ATMOSPHERE CONTROL SYSTEM Topic M+LTR Disposition BFN Result 4.5STANDBY GAS TREATMENT SYSTEM Topic M+LTR Disposition BFN Result 4.5.1Flow Capacity 4.5.2Iodine Removal Capability 4.6MSIV LEAKAGE CONTROL SYSTEM 4.7POST-LOCA COMBUSTIBLE GAS CONTROL SYSTEM Table 4-1 Comparison of MELLLA+ Short Term Containment Response to CLTP Power (%) Flow (%) FWT (1) Peak Drywell Pressure (psig) Maximum Drywell Temperature (°F) Initial Drywell Temperature (°F) (2) Notes:

Table 4-2 Comparison of MELLLA+ Inputs to CLTP Parameter CLTP Analysis Input Value MELLLA+ Analysis Input Value Justification if CLTP Value is Less Conservative

Parameter CLTP Analysis Input Value MELLLA+ Analysis Input Value Justification if CLTP Value is Less Conservative

Notes:

Table 4-3 Results Summary for BFN CO Loads Point in Figure 1-1 Power (%) Flow (%) Calculated Maximum CO PRMS, psi Table 4-4 SSW DP Analysis Results for MELLLA+ in Comparison with MPS Parameter MPS 55.4% CLTP, 37.3% Flow, FFWTR MELLLA+ 102% CLTP, 85% Flow, NFWT -10ºF MELLLA+ 79.17% CLTP, 55% Flow, NFWT -10ºF Design Limit (psid) Table 4-5a SSW Door Jet Force Analysis Results for MELLLA+ in Comparison with CLTP Parameter Unit 102% CLTP 85% RCF NFWT -10°F 102% of 77.6% CLTP 55% RCF NFWT -10°F 102% CLTP RCF FFWTR Table 4-5b SSW Door Drag Load Analysis Results for MELLLA+ in Comparison with CLTP Parameter Unit 102% CLTP 85% RCF NFWT -10°F 102% of 77.6% CLTP 55% RCF NFWT -10°F 102% CLTP RCF FFWTR

Table 4-6 RHR Pump NPSH - MELLLA+ ATWS Event ATWS (Non-LOOP) PRFO EOC ATWS LOOP at EOC Event Type Special Event Special Event Parameter Units RHR Pump NPSH Summary Note:

Figure 4-1 MELLLA+ Short-Term DBA-LOCA Containment Pressure Response - BOUNDING (130F Initial Drywell Temperature) Figure 4-2 MELLLA+ Short-Term DBA-LOCA Containment Temperature Response - BOUNDING (130F Initial Drywell Temperature) Figure 4-3 MELLLA+ Short-Term DBA-LOCA Containment Pressure Response - DESIGN (70F Initial Drywell Temperature) Figure 4-4 MELLLA+ Short-Term DBA-LOCA Containment Temperature Response - DESIGN (70F Initial Drywell Temperature)

Figure 4-5 ATWS Event RHR Pump NPSH vs. Time 5.0INSTRUMENTATION AND CONTROL 5.1NSSS MONITORING AND CONTROL Topic M+LTR Disposition BFN Result 5.1.1Average Power Range, Intermediate Range, and Source Range Monitors 5.1.2Local Power Range Monitors 5.1.3Rod Block Monitors 5.1.4Rod Worth Minimizer 5.1.5Traversing Incore Probes 5.2BOP MONITORING AND CONTROL Topic M+LTR Disposition BFN Result 5.2.1Pressure Control System

5.2.2Turbine Steam Bypass System (Normal Operation) 5.2.3Turbine Steam Bypass System (Safety Analysis) 5.2.4Feedwater Control System (Normal Operation) 5.2.5Feedwater Control System (Safety Analysis) 5.2.6Leak Detection System

5.3TECHNICAL SPECIFICATION INSTRUMENT SETPOINTS Topic M+LTR Disposition BFN Result 5.3.1APRM Flow-Biased Scram

5.3.2Rod Block Monitor 6.0ELECTRICAL POWER AND AUXILIARY SYSTEMS 6.1AC POWER Topic M+LTR Disposition BFN Result

6.2DC POWER Topic M+LTR Disposition BFN Result 6.3FUEL POOL Topic M+LTR Disposition BFN Result 6.3.1Fuel Pool Cooling

6.3.2Crud Activity and Corrosion Products

6.3.3Radiation Levels 6.3.4Fuel Racks

6.3.4.1New and Spent Fuel Storage Criticality Review 6.4WATER SYSTEMS Topic M+LTR Disposition BFN Result

6.5STANDBY LIQUID CONTROL SYSTEM Topic M+LTR Disposition BFN Result 6.5.1Shutdown Margin 6.5.2System Hardware

6.5.3ATWS Requirements

6.6HEATING, VENTILATION AND AIR CONDITIONING Topic M+LTR Disposition BFN Result 6.7FIRE PROTECTION Topic M+LTR Disposition BFN Result 6.8OTHER SYSTEMS AFFECTED Topic M+LTR Disposition BFN Result

7.0POWER CONVERSION SYSTEMS 7.1TURBINE-GENERATOR Topic M+LTR Disposition BFN Result 7.2CONDENSER AND STEAM JET AIR EJECTORS Topic M+LTR Disposition BFN Result 7.3TURBINE STEAM BYPASS Topic M+LTR Disposition BFN Result

7.4FEEDWATER AND CONDENSATE SYSTEMS Topic M+LTR Disposition BFN Result

8.0RADWASTE SYSTEMS AND RADIATION SOURCES 8.1LIQUID AND SOLID WASTE MANAGEMENT Topic M+LTR Disposition BFN Result 8.1.1Coolant Fission and Corrosion Product Levels 8.1.2Waste Volumes

8.2GASEOUS WASTE MANAGEMENT Topic M+LTR Disposition BFN Result 8.2.1Off-Site Release Rate 8.2.2Recombiner Performance 8.3RADIATION SOURCES IN THE REACTOR CORE Topic M+LTR Disposition BFN Result

8.4RADIATION SOURCES IN REACTOR COOLANT Topic M+LTR Disposition BFN Result 8.4.1Coolant Activation Products 8.4.2Fission and Activated Corrosion Products

8.5RADIATION LEVELS Topic M+LTR Disposition BFN Result 8.5.1Normal Operational Radiation Levels

8.5.2Post-Shutdown Radiation Levels 8.5.3Post-Accident Radiation Levels 8.6NORMAL OPERATION OFF-SITE DOSES Topic M+LTR Disposition BFN Result 8.6.1Plant Gaseous Emissions 8.6.2Gamma Shine from the Turbine

9.0REACTOR SAFETY PERFORMANCE EVALUATIONS 9.1ANTICIPATED OPERATIONAL OCCURRENCES 9.1.1Fuel Thermal Margin Events 9.1.2Power and Flow Dependent Limits 9.1.3Non-Limiting Events 9.2DESIGN BASIS ACCIDENTS AND EVENTS OF RADIOLOGICAL CONSEQUENCE 9.2.1Design Basis Events Topic M+LTR Disposition BFN Result

9.2.1.1Control Rod Drop Accident

9.2.1.2Instrument Line Break Accident 9.2.1.3Main Steam Line Break Accident (Outside Containment)

9.2.1.4Loss of Coolant Accident (Inside Containment)

9.2.1.5Large Line Break (Feedwater or Reactor Water Cleanup) 9.2.1.6Liquid Radwaste Tank Failure 9.2.1.7Fuel Handling Accident 9.2.1.8Offgas System Failure 9.2.1.9Cask Drop 9.2.2Other Events with Radiological Consequences 9.3SPECIAL EVENTS Topic M+LTR Disposition BFN Result 9.3.1Anticipated Transients Without Scram

9.3.1.1Anticipated Transients without Scram (Licensing Basis)

9.3.1.2Anticipated Transients without Scram (Best-Estimate Calculation)

9.3.2Station Blackout

9.3.3ATWS with Core Instability

Table 9-1 Key Input Parameters for ATWS Analyses Parameter CLTP MELLLA+Basis Notes for Table 9-1: Table 9-2 Key Results for Licensing Basis ATWS Analysis ATWS Acceptance Criteria CLTP MELLLA+ Design Limit ° Note: Table 9-3 ATWS Analysis Limiting Event Results Parameter1 Limiting Event Peak Value Time Trace Note: Table 9-4 ODYN PRFO Sequence of Events Item Event Response MELLLA+ Event Time (sec) BOC EOC MOC

Note:

Table 9-5 ODYN MSIVC Sequence of Events Item Event Response EPU/MELLLA+ Event Time (sec) BOC EOC MOC

Note: Table 9-6 ODYN LOOP Sequence of Events Item Event Response EPU/MELLLA+ Event Time (sec) BOC EOC MOC Note: Table 9-7 Key Results for ATWSI Analysis from MELLLA+ Operating Domain ATWS Acceptance Criteria MELLLA+ Design Limit Notes:

Table 9-8 TRACG ATWSI TTWBP Sequence of Events - MOC Exposure, Regional Oscillations Item Event Response Time (sec) Table 9-9 TRACG ATWSI RPT Sequence of Events - MOC Exposure, Regional Oscillations Item Event Response Time (sec) Table 9-10 PCT Results for ATWSI Sensitivity Analysis Event PCT (K) / (°F) Table 9-11 PCT Results for ATWSI Fuel Parameter Sensitivity Analysis Sensitivity Parameter Sensitivity Multiplier PCT (K) / (°F) Notes: Figure 9-1 HCTL as a Function of Reactor Pressure and Suppression Pool Water Level Figure 9-2 ATWSI from MELLLA+ Operating Domain - 2RPT Figure 9-3 ATWSI from MELLLA+ Operating Domain - 2RPT Figure 9-4 ODYN ATWS Analysis - LOOP at EOC Figure 9-5 ODYN ATWS Analysis - PRFO at EOC - PCT 10.0OTHER EVALUATIONS 10.1HIGH ENERGY LINE BREAK Topic M+LTR Disposition BFN Result 10.1.1Steam Lines

10.1.2Balance of Plant Liquid Lines

10.1.3Other Liquid Lines 10.1.3.1RWCU Line Break °° 10.1.3.2Pipe Whip and Jet Impingement 10.1.3.3Flooding 10.2MODERATE ENERGY LINE BREAK Topic M+LTR Disposition BFN Result 10.2.1MELB Flooding 10.2.2MELB Environmental Qualification 10.3ENVIRONMENTAL QUALIFICATION Topic M+LTR Disposition BFN Result 10.3.1Electrical Equipment 10.3.2Mechanical Equipment with Non-Metallic Components

10.3.3Mechanical Component Design Qualification 10.4TESTING Topic M+LTR Disposition BFN Result 10.4.1Steam Separator-Dryer Performance 10.4.2Average Power Range Monitor Calibration 10.4.3Core Performance

10.4.4Pressure Regulator 10.4.5Water Level Setpoint Changes 10.4.6Neutron Flux Noise Surveillance 10.5INDIVIDUAL PLANT EXAMINATION Topic M+LTR Disposition BFN Result 10.5.1Initiating Event Categories and Frequency

10.5.2Component and System Reliability 10.5.3Operator Response

10.5.4Success Criteria 10.5.5External Events 10.5.6Shutdown Risks 10.5.7PRA Quality 10.5.8PRA Conclusion

10.6OPERATOR TRAINING AND HUMAN FACTORS Topic M+LTR Disposition BFN Result

10.7PLANT LIFE Topic M+LTR Disposition BFN Result 10.7.1Irradiation Assisted Stress Corrosion Cracking

10.7.2Flow Accelerated Corrosion

10.8NRC AND INDUSTRY COMMUNICATIONS Topic M+LTR Disposition BFN Result 10.9EMERGENCY AND ABNORMAL OPERATING PROCEDURES Topic M+LTR Disposition BFN Result 10.9.1Emergency Operating Procedures 10.9.2Abnormal Operating Procedures

11.0LICENSING EVALUATIONS

12.0REFERENCES

Appendix A - Limitations from Safety Evaluation for LTR NEDC-33173P Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

[sic] Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Notes for Appendix A: Appendix B - Limitations from Safety Evaluation for LTR NEDC-33006P Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Limitation and Condition Number from NRC SER Limitation and Condition Title Limitation and Condition Description Disposition Section of BFN M+SAR which addresses the Limitation and Condition

Notes for Appendix B: Appendix C - Limitations from Safety Evaluation for LTR NEDC-33075P Limitation and Condition Number from NRC SER Limitation and Condition Description DispositionSection of BFN M+SAR which addresses the Limitation and Condition

Notes for Appendix C:

ANP-3551NP Revision 0 December 2017 AREVA Inc. © 2017 AREVA Inc. ANP-3551NP Revision 0 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page i Item Section(s) or Page(s) Description and Justification 1 All This is the initial release. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page ii INTRODUCTION ............................................................................................................. 1 REPORT APPROACH .................................................................................................... 1 EVALUATION BASIS ...................................................................................................... 1 REPORT GENERATION AND REVIEW PROCESS....................................................... 2 1.1 APPROVED METHODOLOGIES .............................................................. 3 1.1.3 COMPUTER CODES AND METHODS ..................................................... 3 1.2 OPERATING CONDITIONS AND CONSTRAINTS ................................... 6 1.2.1 POWER/FLOW MAP ................................................................................. 6 2.0 REACTOR CORE AND FUEL PERFORMANCE ...................................... 9 2.1 FUEL DESIGN AND OPERATION ............................................................ 9 2.1.1 FUEL PRODUCT LINE ............................................................................ 10 2.1.2 CORE DESIGN AND FUEL THERMAL MONITORING THRESHOLD ........................................................................................... 11 2.2 THERMAL LIMITS ASSESSMENT .......................................................... 14 2.2.1 SAFETY LIMIT MINIMUM CRITICAL POWER RATIO ............................ 15 2.2.2 OPERATING LIMIT MINIMUM CRITICAL POWER RATIO ..................... 16 2.2.3 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITS .................................................................. 16 2.2.4 LINEAR HEAT GENERATION RATE LIMITS .......................................... 17 2.2.5 POWER-TO-FLOW RATIO ...................................................................... 17 2.3 REACTIVITY CHARACTERISTICS ......................................................... 18 2.3.1 HOT EXCESS REACTIVITY .................................................................... 18 2.3.2 STRONG ROD OUT SHUTDOWN MARGIN ........................................... 20 2.3.3 SLCS SHUTDOWN MARGIN .................................................................. 22 2.6 ADDITIONAL LIMITATIONS AND CONDITIONS RELATING TO REACTOR CORE AND FUEL PERFORMANCE............................... 23 3.1.2 OVERPRESSURE RELIEF CAPACITY ................................................... 46 4.3 EMERGENCY CORE COOLING SYSTEM PERFORMANCE ................ 52 4.3.1 BREAK SPECTRUM RESPONSE AND LIMITING SINGLE FAILURE .................................................................................................. 52 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page iii 4.3.2 LARGE BREAK PEAK CLAD TEMPERATURE ...................................... 55 4.3.3 SMALL BREAK PEAK CLAD TEMPERATURE ....................................... 55 4.3.4 LOCAL CLADDING OXIDATION ............................................................. 57 4.3.5 CORE WIDE METAL WATER REACTION .............................................. 57 4.3.6 COOLABLE GEOMETRY ........................................................................ 57 4.3.7 LONG TERM COOLING .......................................................................... 57 4.3.8 FLOW MISMATCH LIMITS ...................................................................... 58 4.7 POST-LOCA COMBUSTIBLE GAS CONTROL SYSTEM ....................... 60 5.1.3 ROD BLOCK MONITORS ....................................................................... 61 5.1.4 ROD WORTH MINIMIZER ....................................................................... 61 5.1.5 TRAVERSING INCORE PROBES ........................................................... 62 5.3.2 ROD BLOCK MONITOR .......................................................................... 63 6.3.4.1 NEW AND SPENT FUEL STORAGE CRITICALITY REVIEW ................ 65 8.3 RADIATION SOURCES IN THE REACTOR CORE ................................ 68 9.0 REACTOR SAFETY PERFORMANCE EVALUATIONS.......................... 70 9.1 ANTICIPATED OPERATIONAL OCCURRENCES .................................. 70 9.1.1 FUEL THERMAL MARGIN EVENTS ....................................................... 70 9.1.2 POWER AND FLOW DEPENDENT LIMITS ............................................ 72 9.1.3 NON-LIMITING EVENTS ......................................................................... 74 9.2 DESIGN BASIS ACCIDENTS AND EVENTS OF RADIOLOGICAL CONSEQUENCE ......................................................... 91 9.2.1 DESIGN BASIS EVENTS ........................................................................ 91 9.2.1.1 CONTROL ROD DROP ACCIDENT ........................................................ 91 9.3 SPECIAL EVENTS .................................................................................. 93 9.3.1 ANTICIPATED TRANSIENTS WITHOUT SCRAM .................................. 93 9.3.1.1 ATWS (OVERPRESSURE) ..................................................................... 94 9.3.1.2 ATWS (PEAK CLADDING TEMPERATURE) .......................................... 96 9.3.2 STATION BLACKOUT ............................................................................. 96 REFERENCES ............................................................................................................ 110 Appendix A Limitations from the Final Safety Evaluation for LTR NEDC-33173P .......................................................................... 112 Appendix B Limitations from the Final Safety Evaluation for LTR NEDC-33006P .......................................................................... 131 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page iv Table 1-1 Computer Codes Used in the AMSAR Evaluations by AREVA .................. 5 Table 1-3 Core Thermal Power to Core Flow Ratio at Steady-State and Off-Rated Conditions ....................................................................................... 7 Table 2-1 Peak Nodal Exposures ............................................................................ 26 Table 2-2 Steady-State Bypass Voiding .................................................................. 27 Table 2-3 Core Thermal Power to Core Flow Ratio at Steady-State and Off-Rated Conditions ..................................................................................... 28 Table 4-1 Summary of TLO Recirculation Line Break Results Highest PCT Cases ....................................................................................................... 59 Table 5-1 Hot Channel Bypass Voiding at Steady-State and Off-Rated Conditions ................................................................................................ 64 Table 9-1 Transient Event Results Summary .......................................................... 75 Table 9-2 Transient Results Comparison (100% Power) ......................................... 76 Table 9-3 Base Case Transient Results .................................................................. 77 Table 9-4 EOOS Transient Results .......................................................................... 78 Table 9-5 Comparison Slow Recirculation Flow Increase Results and MCPR Flow Limit ................................................................................................. 80 Table 9-6 Key Input Parameters for ATWS Analyses .............................................. 97 Table 9-7 Key Results for ATWS Analyses .............................................................. 98 Table 9-8 ATWS Analysis Limiting Event Results .................................................... 98 Table 9-9 MSIVC ATWS Sequence of Events at BOC ............................................ 99 Table 9-10 PRFO ATWS Sequence of Events at BOC .............................................. 99 Table 9-11 MSIVC ATWS Sequence of Events at EOC .......................................... 100 Table 9-12 PRFO ATWS Sequence of Events at EOC ............................................ 100 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page v Figure 1-1 Power/Flow Operating Map for MELLLA+ ................................................. 8 Figure 2-1 Peak Bundle Power vs Cycle Exposure ................................................... 29 Figure 2-2 Flow in Peak Bundle vs Cycle Exposure .................................................. 30 Figure 2-3 Exit Void Fraction for Peak Bundle vs Cycle Exposure ............................ 31 Figure 2-4 Maximum Channel Exit Void Fraction vs Cycle Exposure ....................... 32 Figure 2-5 Core Average Exit Void Fraction vs Cycle Exposure ............................... 33 Figure 2-6 Peak LHGR vs Cycle Exposure ............................................................... 34 Figure 2-7 Nominal Bundle Power Map at BOC (0 MWd/MTU) ................................ 35 Figure 2-8 Nominal Bundle Power Map at MOC (9,250 MWd/MTU) ......................... 36 Figure 2-9 Nominal Bundle Power Map at EOR (End of Rated, 18,525 MWd/MTU) .............................................................................................. 37 Figure 2-10 Maximum LHGR (kW/ft) Map at BOC (0 MWd/MTU) ............................... 38 Figure 2-11 Maximum LHGR (kW/ft) Map at MOC (9,250 MWd/MTU) ....................... 39 Figure 2-12 Maximum LHGR (kW/ft) Map at EOR (End of Rated, 18,525 MWd/MTU) .............................................................................................. 40 Figure 2-13 MCPR Map at BOC (0 MWd/MTU) .......................................................... 41 Figure 2-14 MCPR Map at MOC (9,250 MWd/MTU) ................................................... 42 Figure 2-15 MCPR at EOR (End of Rated, 18,525 MWd/MTU) .................................. 43 Figure 2-16 Maximum LHGR (kW/ft) Map at Exposure Where Peak Cycle FDLRX (or MFLPD) Occurs (18,100 MWd/MTU) ..................................... 44 Figure 2-17 MCPR Map at Exposure Where Peak Cycle MFLCPR Occurs (17,591 MWd/MTU) ................................................................................. 45 Figure 3-1 MSIV Closure Overpressurization Event at 102P/105F - Key Parameters .............................................................................................. 48 Figure 3-2 MSIV Closure Overpressurization Event at 102P/105F - Sensed Water Level .............................................................................................. 49 Figure 3-3 MSIV Closure Overpressurization Event at 102P/105F - Vessel Pressures ................................................................................................. 50 Figure 3-4 MSIV Closure Overpressurization Event at 102P/105F - Safety/Relief Valve Flow Rates ................................................................ 51 Figure 9-1 EOC LRNB at 100P/105F - TSSS Key Parameters ................................ 81 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page vi Figure 9-2 EOC LRNB at 100P/105F - TSSS Sensed Water Level.......................... 82 Figure 9-3 EOC LRNB at 100P/105F - TSSS Vessel Pressures .............................. 83 Figure 9-4 EOC TTNB at 100P/105F - TSSS Key Parameters ................................ 84 Figure 9-5 EOC TTNB at 100P/105F - TSSS Sensed Water Level .......................... 85 Figure 9-6 EOC TTNB at 100P/105F - TSSS Vessel Pressures .............................. 86 Figure 9-7 EOC FWCF at 100P/105F - TSSS Key Parameters ............................... 87 Figure 9-8 EOC FWCF at 100P/105F - TSSS Sensed Water Level ......................... 88 Figure 9-9 EOC FWCF at 100P/105F - TSSS Vessel Pressures ............................ 89 Figure 9-10 MCPRf Limits for Maximum Flow Run-up of 107% of Rated .................... 90 Figure 9-11 MSIVC ATWS Overpressurization Event at 100P/85F - Key Parameters ............................................................................................ 101 Figure 9-12 MSIVC ATWS Overpressurization Event at 100P/85F - Sensed Water Level ............................................................................................ 102 Figure 9-13 MSIVC ATWS Overpressurization Event at 100P/85F - Vessel Pressures ............................................................................................... 103 Figure 9-14 MSIVC ATWS Overpressurization Event at 100P/85F - Safety/Relief Valve Flow Rates .............................................................. 104 Figure 9-15 PRFO ATWS Overpressurization Event at 100P/85F - Key Parameters ............................................................................................ 105 Figure 9-16 PRFO ATWS Overpressurization Event at 100P/85F - Sensed Water Level ............................................................................................ 106 Figure 9-17 PRFO ATWS Overpressurization Event at 100P/85F - Vessel Pressures ............................................................................................... 107 Figure 9-18 PRFO ATWS Overpressurization Event at 100P/85F - Safety/Relief Valve Flow Rates ................................................................................... 108 Figure 9-19 PRFO ATWS Overpressurization Event at 100P/85F - Key Parameters at Coastdown ..................................................................... 109 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page vii AMSAR AREVA MELLLA+ Safety Analysis Report AOO Anticipated Operational Occurrence AOT Abnormal Operational Transient APRM Average Power Range Monitor ARI Alternate Rod Insertion ARTS APRM / RBM / Technical Specifications ASME American Society of Mechanical Engineers AST Alternate Source Term ATWS Anticipated Transient Without Scram ATWS-MSIV Anticipated Transient Without Scram - Main Steam Isolation Valve ATWS-PRFO Anticipated Transient Without Scram - Pressure Regulator Failure Open ATWS-RPT Anticipated Transient Without Scram - Recirculation Pump Trip BFN Browns Ferry Nuclear Plant BOC Beginning-of-Cycle BOL Beginning of Life BWR Boiling Water Reactor CF Core Flow CFR Code of Federal Regulations CHF Critical Heat Flux CLTP Current Licensed Thermal Power (3952 MWt, same as EPU) CMS Core Monitoring System COAST Coastdown exposure COLR Core Operating Limits Report CPR Critical Power Ratio CRDA Control Rod Drop Accident CRWE Control Rod Withdrawal Error CSA Criticality Safety Analysis DBA Design Basis Accident DC Direct Current DEG Double Ended Guillotine Break DSS-CD Detect and Suppress Solution - Confirmation Density ECCS Emergency Core Cooling System EFW Extended Flow Window EOC End-of-Cycle AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page viii EOOS Equipment Out-of-Service EOP Emergency Operating Procedure EOR End of Rated EPFOD Extended Power / Flow Operating Domain EPG Emergency Procedure Guideline EPU Extended Power Uprate (defined as 120% OLTP, 3952 MWt) FDL Fuel Design Limit FDLRX Maximum Fuel Design Limit Ratio FHOOS Feedwater Heaters Out-of-Service FSAR Final Safety Analysis Report FW Feedwater FWCF Feedwater Controller Failure FWHOOS Feedwater Heaters Out-of-Service GE General Electric GEH GE-Hitachi Nuclear Energy Americas LLC GESTAR General Electric Standard Application for Reactor Fuel GNF Global Nuclear Fuel GWd Gigawatt Days HCTL Heat Capacity Temperature Limit HPCI High Pressure Coolant Injection HSBW Hot Shutdown Boron Worth IASCC Irradiation Assisted Stress Corrosion Cracking ICF Increased Core Flow IORV Inadvertent Opening of a Main Steam Relief Valve kW Kilowatt L&C Limitation and Condition LAR License Amendment Request LCO Limiting Condition of Operation LFWH Loss of Feedwater Heating LHGR Linear Heat Generation Rate LHGRFACf Flow-Dependent Linear Heat Generation Rate Multipliers LHGRFACp Power-Dependent Linear Heat Generation Rate Multipliers LOCA Loss of Coolant Accident LOFW Loss of Feedwater Flow LOOP Loss of Offsite Power LPRM Local Power Range Monitor AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page ix LRNB Load Rejection Without Bypass LTR Licensing Topical Report MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MCPRf Flow-Dependent Minimum Critical Power Ratio MCPRp Power-Dependent Minimum Critical Power Ratio MELLLA Maximum Extended Load Line Limit Analysis MELLLA+ Maximum Extended Load Line Limit Analysis Plus MFLCPR Maximum Fraction of Limiting CPR MFLPD Maximum Fraction of Limiting Power Density (same as FDLRX) M+LTR MELLLA+ Licensing Topical Report M+SAR MELLLA+ Safety Analysis Report MOC Middle-of-Cycle MS Main Steam MSIV Main Steam Isolation Valve MSIVC Main Steam Isolation Valve Closure MSIVF Main Steam Isolation Valve Closure with Scram on High Flux MSRV Main Steam Relief Valve MSRVOOS Main Steam Relief Valve Out-of-Service MTU Metric Tons of Uranium MVP Mechanical Vacuum Pump MWt Megawatt Thermal NFSV New Fuel Storage Vault NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OLMCPR Operating Limit Minimum Critical Power Ratio OLTP Original Licensed Thermal Power (3293 MWt) OOS Out-of-Service PCT Peak Clad Temperature PLUOOS Power Load Unbalance Out-of-Service PRFO Pressure Regulator Failure Open RAI Response to Additional Information RBM Rod Block Monitor RCF Rated Core Flow RCIC Reactor Core Isolation Cooling RG Regulatory Guide AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page x RHR Residual Heat Removal RMCS Reactor Manual Control System RPT Recirculation Pump Trip RPTOOS Recirculation Pump Trip Out-of-Service RPV Reactor Pressure Vessel RSAR Reload Safety Analysis Report RWE Rod Withdrawal Error RWM Rod Worth Minimizer SAG Severe Accident Guideline SAR Safety Analysis Report SBO Station Blackout SDM Shutdown Margin SER Safety Evaluation Report SF-BATTlBA Single Failure of Battery (DC) Power, Board A SF-BATTlBB Single Failure of Battery (DC) Power, Board B SFSP Spent Fuel Storage Pool SLMCPR Safety Limit Minimum Critical Power Ratio SLC Standby Liquid Control SLCS Standby Liquid Control System SLO Single-Loop Operation SRLR GEH Supplemental Reload Licensing Report SRO Strong Rod Out SRV Safety Relief Valve TBVOOS Turbine Bypass Valve Out-of-Service TCV Turbine Control Valve TIP Traversing Incore Probe TLO Two-Loop Operation T-M Thermal-Mechanical TS Technical Specifications TSSS Technical Specifications Scram Speed TTNB Turbine Trip Without Bypass TVA Tennessee Valley Authority AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 1 This report summarizes the results of a subset of the safety evaluations performed to support the maximum extended load line limit analysis plus (MELLLA+) license amendment request (LAR) for the Browns Ferry Nuclear Plant (BFN). The BFN MELLLA+ LAR requests an increase from the current licensed power / flow map, the extended power flow operating domain (EPFOD). This report supplements the General Electric Hitachi (GEH) supplied MELLLA+ safety analysis report (M+SAR). The BFN units will utilize AREVA ATRIUMŽ 10XM* fuel, with the potential for some legacy ATRIUM-10 fuel, under MELLLA+ conditions. Therefore, the GEH M+SAR uses an equilibrium ATRIUM 10XM core modeled by GEH using AREVA supplied data for the key fuel related evaluations. The subset of the safety analyses that are addressed in this report were identified by the Tennessee Valley Authority (TVA) as being required to support the continued use of AREVA fuel designs in the BFN reactor cores in the MELLLA+ LAR. This report follows the general format of Reference 1 for the sections being addressed. For example, where possible, the numbering of the sections included in this report is based upon the corresponding sections of Reference 1. Because only selected portions of Reference 1 are being addressed, the numbering of the individual sections of this report is not continuous (i.e. section numbers skip over sections of Reference 1 that are not directly addressed). The safety evaluations documented in this report are based upon the continued use of the ATRIUM 10XM fuel design. Where appropriate, evaluations for the existing ATRIUM-10 fuel design have also been included because the potential exists that some of these previously loaded assemblies may still be resident in a BFN reactor core that is operated at MELLLA+ conditions. A number of the safety evaluations supporting MELLLA+ operation are cycle-specific because they are dependent upon the specific bundle and core designs. These are addressed in the reload licensing analyses performed for each cycle.

• ATRIUM is a trademark of AREVA Inc.

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 2 The safety evaluations in this document are based upon an ATRIUM 10XM representative equilibrium cycle fuel cycle design (Reference 2, Attachment 9 of the MELLLA+ LAR). Additional cycle-specific safety evaluations have been provided for a reference cycle (Cycle 19) as Attachment 17 of the MELLLA+ LAR. Consistent with the division of responsibility established between TVA and AREVA, the AREVA work scope focused on the reload fuel analyses. In broad terms, this included fuel and core design, the ASME overpressure evaluation, and establishing the thermal operating limits and backup stability regions. The tasks performed by AREVA were prepared, reviewed and approved consistent with AREVA's quality assurance procedures. In support of the AREVA MELLLA+ work scope, design and analysis inputs were provided in approved documents. These inputs were used to perform the AREVA work scope to support the AMSAR. Descriptions of the analyses performed, inputs, methodology, and results were transmitted to TVA for review and approval. TVA personnel reviewed the AMSAR input and the supporting documentation and provided comments. Responses to the comments were prepared by AREVA and subsequently accepted by TVA. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 3 The safety evaluations utilize a series of NRC-approved AREVA methods which are summarized in a Licensing Compendium, Reference 3. Application of this approved methodology to EPU MELLLA+ conditions remains within its approval basis and safety evaluation report (SER) restrictions, as addressed in ANP-2860P Revision 2 (Reference 4). This methods applicability was extended to the ATRIUM 10XM design for BFN through Supplement 1P (Reference 5) which addressed conditions at 3458 MWt. This is further extended to the ATRIUM 10XM at EPU conditions in Supplement 2P (Reference 6). In addition, it is extended to the ATRIUM 10XM at MELLLA+ conditions in Supplement 3P (Reference 7) which is included as Attachment 27 of the MELLLA+ LAR. Nuclear Regulatory Commission (NRC) approved or industry-accepted computer codes and calculational techniques are used in the evaluations for the MELLLA+ operating domain. The primary AREVA computer codes used for BFN evaluations are listed in Table 1-1. The application of these codes complies with the limitations, restrictions, and conditions specified in the approving NRC SER. Exceptions to the use of the code or special conditions of the applicable SER are included as notes to Table 1-1. The continued applicability of AREVA methodology to the BFN MELLLA+ EPFOD extension is addressed in ANP-2860P Revision 2 Supplement 3P (Reference 7, included as Attachment 27 of the BFN MELLLA+ LAR). In accordance with the MELLLA+ licensing topical report (M+LTR) SER Limitation and Condition 12.1 (Reference 1), the range of mass fluxes and power/flow ratios in the AREVA ACE/ ATRIUM 10XM critical power correlation database and the GEXL equivalent database for ATRIUM 10XM fuel covers the intended MELLLA+ operating domain. The database includes low flow, high qualities, and void fractions beyond the application of the GEXL-PLUS equivalent or the AREVA ACE correlations in the MELLLA+ operating domain. To meet the intent of M+LTR SER Limitation and Condition 12.23.2, the BFN-specific ATWS calculation input parameters are provided in Table 9-6 and calculation results are provided in Tables 9-7 and 9-8. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 4 Consistent with M+LTR SER Limitation and Condition 12.2, the specific limitations and conditions associated with the M+LTR and GE/GEH Methods licensing topical report (LTR), Reference 13, are discussed along with the relevant sections of this report. A complete listing of the required M+LTR SER, Methods and GE/GEH Methods LTR SER limitations and conditions and the sections of this report which address them, as well as additional comments, is presented in Appendices A and B. Limitations associated with the DSS-CD licensing topical report are addressed in the GEH M+ SAR. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 5 Task Computer Code NRC Approved Comments Reactor Heat Balance HTBAL MICROBURN-B2 (1) Y EMF-2158(P)(A) R0, 10/99 Reactor Core and Fuel Performance CASMO-4 MICROBURN-B2 XCOBRA SAFLIM3D RODEX4 Y Y Y (2) Y Y EMF-2158(P)(A) R0, 10/99 EMF-2158(P)(A) R0, 10/99 XN-NF-80-19(P)(A) V3, R2, 1/87 ANP-10307(P)(A) R0, 6/11 BAW-10247PA R0, 2/08 Thermal Hydraulic Stability STAIF Y EMF-CC-074(P)(A) V4, R0 ECCS-Loss of Coolant Accident (LOCA) RELAX HUXY RODEX2 Y Y Y (4) EMF-2361(P)(A) R0, 5/01 XN-CC-33(A) R1, 11/75 XN-NF-81-58(P)(A) R2, 3/84 EMF-85-74(P) S1(P)(A) and S2(P)(A), 8/86 AOO Transient Analysis MICROBURN-B2 COTRANSA2 XCOBRA XCOBRA-T RODEX2 Y Y (3) Y Y (3) Y (4) EMF-2158(P)(A) R0, 10/99 ANF-913(P)(A) V1, R1, 8/90 XN-NF-80-19(P)(A) V3, R2, 1/87 XN-NF-84-105(P)(A) V1, S1 and S2, 2/87 XN-NF-81-58(P)(A) R2, 3/84 EMF-85-74(P) S1 (P)(A) and S2(P)(A), 8/86 ASME and ATWS Overpressurization COTRANSA2 RODEX2 Y Y (4) ANF-913(P)(A) V1, R1, 8/90 XN-NF-81-58(P)(A) R2, 3/84 EMF-85-74(P) S1 (P)(A) and S2(P)(A), 8/86 The application of these codes to the MELLLA+ analyses complies with the limitations, restrictions, and conditions specified in the approving NRC SER, where applicable, for each code. The application of the codes also complies with SERs for the extended power uprate programs. 1. HTBAL is not explicitly approved by the NRC but it is a stand-alone version of the heat balance routine included in the NRC-approved MICROBURN-B2 code documented in EMF-2158(P)(A). 2. The approval of XCOBRA is included in the approval of the THERMEX methodology in XN-NF-80-19 (P)(A) Vol3 Rev. 2. 3. The list of events for which COTRANSA2 and XCOBRA-T can be used was expanded in the clarification acceptance in Letter, S. Richards (NRC) to J. F. Mallay (FANP), "Siemens Power Corporation RE: Request for Concurrence on Safety Evaluation Report Clarifications (TAC No. MA6160)," May 31, 2000. 4. The impact of thermal conductivity degradation on licensing analyses is discussed in Reference 7. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 6 The BFN power/flow map including the MELLLA+ operating domain expansion is shown in Figure 1-1. All lines on the power/flow map in Figure 1-1, other than those associated with the MELLLA+ operating domain expansion, are unchanged by MELLLA+. In accordance with M+LTR SER Limitation and Condition 12.5.c, BFN will include the power/flow maps in the core operating limits report (COLR) once the MELLLA+ operating domain expansion is approved. The MELLLA+ operating domain extends from 55% of rated core flow (RCF) at 77.6% of current licensed thermal power (CLTP) to 85% of RCF at 100% of CLTP. Normal core performance characteristics for plant power/flow maneuvers at near full power can be accomplished above 55% of RCF. Due to stability considerations at high power and low core flow (CF), the MELLLA+ operating domain was not extended below 55% of RCF. The reactor operating conditions following an unplanned event could stabilize at a power/flow point outside the allowed operating domain. If this occurs the operator must reduce power or increase flow in accordance with plant procedures to place the plant back into the allowed operating domain. The steady state core thermal power to CF ratio for operation in the MELLLA+ operating domain is listed in Table 1-3. Each point listed is in compliance with the Methods LTR SER Limitation and Condition 9.3 of 50 MWt / Mlbm/hr with the exception of the point of low flow high power, point 'O' (55% of RCF 77.6% of CLTP), on Figure 1-1. The point on the power/flow map is only marginally above the limit and is not used for extended periods of operation. Because the limitation is not intended to place operational restrictions on the plant (Reference 7), the BFN MELLLA+ Power/Flow map remains as shown in Figure 1-1, without any additional restrictions. When BFN exceeds the power-to-flow ratio of 50 MWt / Mlbm/hr at 55% of RCF, the limitation with respect to the conservatism of the power distribution uncertainties were evaluated and do not apply. The results of this assessment are provided in Section 2.2.5. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 7 Current Operating Domain 100% Rated Core Flow "E" 3952 / 100.0 102.5 / 100.0 38.56 Current Operating Domain 99% Rated Core Flow "D" 3952 / 100.0 101.475 / 99.0 38.95 MELLLA+ Operating Domain 85% Rated Core Flow "N" 3952 / 100.0 87.125 / 85.0 45.36 MELLLA+ Operating Domain 55% Rated Core Flow "O" 3067 / 77.6 56.375 / 55.0 54.40 Current Operating Domain / MELLLA+ 55% Rated Core Flow "P" 2703 / 68.4 56.375 / 55.0 47.95 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 8 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 9 This section addresses applicable evaluations involving reactor core and fuel performance for the MELLLA+ EPFOD. The major evaluations are as follows: 2.1 Fuel Design and Operation Plant Specific Acceptable 2.2 Thermal Limits Assessment Plant Specific Acceptable 2.3 Reactivity Characteristics Plant Specific Acceptable BFN currently uses the AREVA ATRIUMŽ* 10XM fuel design and at the time of MELLLA+ initial implementation the reactor cores will predominantly contain this design. Any co-resident fuel will be of the ATRIUM-10 design that will be located in non-limiting locations on or near the core periphery.The effect of MELLLA+ on the fuel design and operation is described below. The primary topics addressed in this evaluation are: Fuel Product Line Design Plant Specific Acceptable (see Section 2.1.1) Core Design Plant Specific Acceptable (see Section 2.1.2) Fuel Thermal Margin Monitoring Threshold Plant Specific Acceptable (see Section 2.1.2)

• ATRIUM is a trademark of AREVA NP AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 10 Implementation of the MELLLA+ EPFOD does not require any changes to the fuel mechanical design and therefore no changes in fuel product line are required. BFN will continue to use the AREVA ATRIUM 10XM fuel design for future MELLLA+ operating cycles. As stated previously, at the time of initial MELLLA+ implementation the BFN reactor cores will contain only AREVA fuel. The predominant fuel type will be the ATRIUM 10XM fuel design with any remaining ATRIUM-10 fuel limited to non-limiting locations on or near the core periphery. Therefore, the following evaluation addresses the potential impact of the MELLLA+ EPFOD on the ATRIUM 10XM fuel design. The M+ LTR generically addresses the impact of MELLLA+ on the fuel product line by stating that the fuel design limits (FDL) are established for all new fuel product lines as part of the fuel introduction. This is also true for the AREVA ATRIUM 10XM fuel design; however, it should be noted thatcontinued applicability of the thermal-mechanical FDL is confirmed for each operating cycle as part of the reload licensing process. The ATRIUM 10XM thermal-mechanical FDL for BFN has been established using the NRC approved RODEX4 methodology (Reference 8). This methodology has been included as part of BFN Technical Specification (TS) 5.6.5.b since the initial introduction of the ATRIUM 10XM fuel design. The application of this methodology includes a cycle-specific confirmation of the continued applicability of the FDL which includes validation that the underlying thermal-mechanical criteria continue to be met during expected operation. For a MELLLA+ operating cycle this reload evaluation will include the impact of the depletion with the EPFOD using the specific core configuration for that cycle. This is consistent with SER Limitation and Condition 12.3.e* of the M+ LTR. The BFN ATRIUM 10XM thermal-mechanical FDL has been confirmed to remain applicable to the MELLLA+ EPFOD for both an equilibrium cycle and the reference cycle core designs.
• SER Limitation and Conditions section 12.3 addresses concurrent changes. While the use of ATRIUM 10XM is not a change to the BFN licensing basis it does represent a change from the approval basis of NEDC-33006-A which assumed the use of the GNF GE-14 fuel design.

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 11 The basis for not monitoring thermal limits below the thermal monitoring threshold is the large margin to thermal limits as described in the TS Bases. Therefore, with these large margins, there are no transients that have limiting consequences when initiated from powers below the established threshold. The maximum licensed power level and the fuel design do not change as a result of MELLLA+ implementation. The CLTP will remain at the power level of 3952 MWt and BFN will continue to use the AREVA ATRIUM 10XM fuel design. Since there is no change in core power, the average bundle power and average core power density remain unchanged. Because there is no change in the average power density there is no change required in the fuel thermal monitoring threshold. MELLLA+ implementation does not require changes to the BFN fuel design and the thermal-mechanical FDL is confirmed to remain applicable for each operating cycle. This also supports the conclusion that no change is required to the fuel thermal monitoring threshold. The MELLLA+ EPFOD allows for higher bundle power to flow conditions. The M+ LTR recognizes that this may cause the range of void fraction, axial and radial power shape, and control rod positions in the core to change slightly. Even with these changes, the individual fuel bundles are required to remain within the allowable thermal limits. These thermal limits are calculated and/or confirmed on a cycle-specific basis and provided in the reload safety analysis report (RSAR). These limits are then included in the cycle-specific COLR which is used to support operation of the core during that cycle. Operation with the lower flows allowed by the MELLLA+ EPFOD will increase the core average void fraction when compared to operation at the same power level in the maximum extended load line limit analysis (MELLLA) operating domain. This increase in core average void fraction is driven by the corresponding increase in the in-channel power to flow ratio. The potential for bypass region voiding may increase with MELLLA+ EPFOD operation due to the lower total core flow. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 12 The most significant impact of voiding in the bypass would be the impact on the local power range monitor (LPRM) reading. Limitation and Condition 9.17, as described in Appendix A of this report, limits the allowable bypass voiding at the LPRM levels. A confirmatory evaluation of bypass voiding for MELLLA+ conditions has been performed for an equilibrium cycle which uses to estimate the potential for localized bypass boiling. This to specifically determine a bounding local void distribution in the core. The model is conservative in that it . The impact on an LPRM location is calculated the location. These results are provided in Table 2-2 and correspond to state points on the MELLLA+ upper boundary at 100% power. As indicated in the discussion on Limitation and Condition 9.17 in Appendix A of this report, cycle-specific validation that the allowable bypass voiding at the LPRM levels has been met will be included in the RSAR. The following information is provided to document various core design and fuel monitoring parameters for each cycle exposure state point consistent with Limitation and Condition 9.24 of the GEH Methods LTR SER. The parameters are compared against an experience base taken from a subset of the plants addressed in the BFN EPFOD methods applicability document (Reference 7). Table 2-1 Peak Nodal Exposures Figure 2-1 Peak Bundle Power vs Cycle Exposure Figure 2-2 Flow in Peak Bundle vs Cycle Exposure Figure 2-3 Exit Void Fraction for Peak Bundle vs Cycle Exposure AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 13 Figure 2-4 Maximum Channel Exit Void Fraction vs Cycle Exposure Figure 2-5 Core Average Exit Void Fraction vs Cycle Exposure Figure 2-6 Peak LHGR vs Cycle Exposure In accordance with M+LTR SER Limitation and Condition 12.24.2, the exit void fraction for the peak power bundle is included in Figure 2-3 for both MELLLA+ and pre-MELLLA+ conditions. Quarter core maps assuming mirror symmetry are provided in Figure 2-7 through Figure 2-15 showing the bundle power, bundle operating linear heat generation rate (LHGR), and minimum critical power ratio (MCPR) for beginning of cycle (BOC) (0 MWd/MTU), middle of cycle (MOC) (9,250 MWd/MTU), and end of rated (EOR) (18,525 MWd/MTU) conditions. These maps represent an ATRIUM 10XM MELLLA+ equilibrium cycle design for BFN. The maximum fuel design limit ratio (FDLRX) occurs at 18,100 MWd/MTU (Figure 2-16). The maximum fraction of limiting critical power ratio (MFLCPR) occurs at 17,591 MWd/MTU (Figure 2-17). In Figure 2-7 through Figure 2-9, the bundle power is dimensionless. To obtain the bundle power in MWt, multiply each number by a factor of 5.17. This factor equals 3,952/764, where 3,952 MWt is the rated thermal power and 764 is the total number of fuel bundles in the core. Table 2-1 includes a comparison of the peak nodal exposure for various plants including both the BFN reference Cycle 19 (for both MELLLA and MELLLA+ depletions) and a BFN equilibrium cycle at MELLLA+ conditions. The BFN MELLLA+ equilibrium cycle has a slightly higher peak nodal exposure compared to the reference cycle 19; however, it remains bounded by a comparison BWR/6 plant. Equilibrium cycles are often used as a demonstration to show the potential fuel utilization for a specific bundle design. This generally means targeting smaller batch sizes with a corresponding increase in the batch average discharge exposure. The AREVA ATRIUM 10XM fuel design is subject to a Also included in this table is a MELLLA depletion for the same reference Cycle 19 design which provides a more direct demonstration of the impact of the change in operating domain on the peak nodal exposure. Comparison of the MELLLA AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 14 versus MELLLA+ depletion for the BFN reference cycle shows only small and insignificant changes in the peak burnup due to MELLLA+ operation. Figure 2-1, Figure 2-2, and Figure 2-6 show that the BFN MELLLA+ EPFOD is in the expected range as compared to the reference plants. Figure 2-3 through Figure 2-5 show that exit voiding at BFN when operating with MELLLA+ EPFOD is also within the range and bounded by the higher range of the comparison plants. The expected impact of MELLLA+ EPFOD on BFN can also be seen with an increase in void fraction when compared with MELLLA operation. Figure 2-7 through Figure 2-9 show the relative bundle power for BOC, MOC, and EOR, respectively. Figure 2-10 through Figure 2-12 show the operating LHGR for BOC, MOC, and EOR, respectively. Figure 2-13 through Figure 2-15 show the MCPR for BOC, MOC, and EOR, respectively. Figure 2-7 through Figure 2-17 show that the general operational conditions for BFN in the MELLLA+ operating domain are within expected parameters. Therefore, BFN meets the intent of the M+LTR in regard to core design and the fuel thermal monitoring threshold. Assurance that regulatory limits are not exceeded during postulated anticipated operational occurrences and accidents is accomplished by applying operating limits on the fuel. This section discusses the impact that MELLLA+ EPFOD operation has on these thermal limits. Consistent with the current practice, cycle-specific thermal limits are established or confirmed each reload based on the cycle-specific core configuration. The effect of MELLLA+ EPFOD on the MCPR safety and operating limits, maximum average planar linear heat generation rate (MAPLHGR) limits, and LHGR limits is described below. The BFN MELLLA+ Equilibrium Cycle (Reference 2) and the reference Cycle 19 (Reference 9) critical power analyses were performed using the ACE/ATRIUM 10XM critical power correlation (References 10 and 11), consistent with the BFN Technical Specifications. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 15 The topics addressed in this evaluation are: Safety Limit MCPR Plant Specific Acceptable (see Section 2.2.1) Operating Limit MCPR Plant Specific Acceptable (see Section 2.2.2) MAPLHGR Limit Plant Specific Acceptable (see Section 2.2.3) LHGR Limit Plant Specific Acceptable (see Section 2.2.4) Power-to-Flow Ratio Plant Specific Acceptable (see Section 2.2.5) The safety limit minimum critical power ratio (SLMCPR) analysis reflects the actual plant core loading pattern and planned operation and is performed for each reload core. In the event that the cycle-specific SLMCPR is not bounded by the current BFN TS value, a LAR will be submitted to change the TS. The cycle-specific SLMCPR will be determined using the methodology defined in Reference 12 which is already included as part of BFN TS 5.6.5.b. The two-loop operation (TLO) SLMCPR analyses will be performed for the minimum and maximum core flow conditions associated with rated power (85% and 105% of rated core flow), as well as the maximum core power at 55% core flow for the BFN operating power/flow map. For the maximum core flow state point, the TLO core flow uncertainty will be used. The analyses for the minimum core flow at full power and the 55% core flow state points will use the single-loop operation core flow uncertainty as discussed in Section 2.2.1.1 of the M+ LTR SER and is consistent with the M+ LTR SER Limitation and Condition 12.6 and 12.24.3. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 16 Limitation and Condition 9.5 to the GE methodology applicability LTR addresses a concern related to a lack of plant operating incore instrumentation data (i.e. TIP/LPRM data) to validate the current power distribution uncertainties when operating in high power/ low flow regions of the power/flow operating map. This concern is addressed in the AREVA methods applicability report (Reference 7, Attachment 27 of the BFN MELLLA+ LAR). The Reference 7 evaluation concludes that no additional SLMCPR adder is needed for AREVA methodology to address this concern. Consistent with the generic disposition in the M+ LTR and current reload requirements; the SLMCPR for BFN will be evaluated for the reload core prior to MELLLA+ implementation. The results of this evaluation are provided in the RSAR for the representative cycle 19 (Attachment 17 of the BFN MELLLA+ LAR). The operating limit minimum critical power ratio (OLMCPR) is calculated by adding the change in MCPR due to the limiting anticipated operational occurrence (AOO) event to the SLMCPR. The OLMCPR is determined on a cycle-specific basis from the results of the reload transient analysis. The cycle-specific analysis results are documented in the RSAR and the operating limits are included in the cycle-specific COLR. The MELLLA+ EPFOD operating conditions do not change the methods used to determine this limit. The OLMCPR is evaluated for each BFN operating cycle. Consistent with the generic disposition in the M+ LTR, the OLMCPR for BFN will be evaluated for the reload core prior to MELLLA+ implementation. The results of this evaluation are provided in the RSAR for the representative Cycle 19 (Attachment 17 of the BFN MELLLA+ LAR). The MAPLHGR limits ensure that the plant does not exceed regulatory limits established in 10 CFR 50.46. Section 4.3, Emergency Core Cooling System Performance, presents the evaluation to demonstrate that the plant meets the regulatory limits in the MELLLA+ operating domain. The reload analysis confirms applicability of the MAPLHGR operating limit for each reload fuel bundle design and the limits are documented in the cycle-specific COLR. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 17 The MAPLHGR limits for BFN will be evaluated for the reload core prior to MELLLA+ implementation. The results of this evaluation are provided in the RSAR for the representative cycle 19 (Attachment 17 of the BFN MELLLA+ LAR). The M+LTR describes that LHGR limits ensure that the plant does not exceed the fuel thermal-mechanical design limits. The LHGR limit is determined by the AREVA fuel rod thermal-mechanical RODEX4 methodology (Reference 8) which is already included as part of BFN TS 5.6.5.b. A cycle-specific validation of the LHGR limits is performed that includes the impact of expected operation. This validation includes the use of operation in the MELLLA+ EPFOD. The reload analysis confirms that the LHGR limits for each reload fuel bundle design are acceptable and establishes any required power or flow dependent LHGR setdowns to ensure that the thermal-mechanical fuel design limits are protected. These results, including the LHGR limits and any required power or flow dependent LHGR setdowns, are documented in the cycle-specific RSAR and provided in the COLR. Limitation and Condition 9.12 to the GE methodology applicability LTR provides requirements specific to that methodology. These requirements do not apply to the AREVA RODEX4 methodology. Implementation of the BFN LHGR limits and any required setdowns ensure the plant does not exceed the thermal-mechanical FDL. BFN continues to use the ATRIUM 10XM fuel product line. The MELLLA+ operating conditions do not change the methods used to determine the LHGR limits. The LHGR limits for BFN are evaluated for each reload core which will include cores designed for MELLLA+ EPFOD operation. The results of this evaluation are provided in the RSAR for the representative cycle 19 (Attachment 17 of the BFN MELLLA+ LAR). MELLLA+ submittals for other plants have addressed the potential for exceeding a core power to flow ratio of 50 MWt / Mlbm/hr at any state point in the allowed operating domain. BFN may exceed this figure of merit only near point "O" of the P/F map, Figure 1-1, as presented in Table 2-3. This concern was later extended to power to flow ratios as low as 42 MWt / Mbm/hr which in turn expands the potential affected region of the power / flow map. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 18 It is noted that the 50 MWt / Mlbm/hr power to flow ratio figure of merit, that addresses Limitation and Condition 9.3 of the GE MELLLA+ methods applicability LTR (Reference 13), is not directly applicable to the AREVA methods used to design and license the BFN reactor cores. The intent of this figure of merit is to provide assurance that the methodology specific power distribution uncertainties remain applicable to the expanded operating domain. For the use of AREVA methodology at BFN this is addressed in a separate EPFOD methods applicability report (Reference 7, provided as Attachment 27 to the BFN MELLLA+ LAR). Specifically, Reference 7 addresses the applicability of uncertainties including gamma scan comparisons and concludes that the power distribution uncertainties used with AREVA methods remain applicable to BFN when operating in the MELLLA+ EPFOD. Reference 7 concludes that the SLMCPR adder specified in Limitation and Condition 9.5 is not needed for AREVA methods. The effect of MELLLA+ on strong rod out (SRO) shutdown margin, standby liquid control system (SLCS) shutdown margin, and hot excess reactivity is described below. The topics addressed in this evaluation are: Hot Excess Reactivity Plant Specific Acceptable (see Section 2.3.1) SRO Shutdown Margin Plant Specific Acceptable (see Section 2.3.2) SLCS Shutdown Margin Plant Specific Acceptable (see Section 2.3.3) Hot excess reactivity is a parameter of interest to operation for two reasons: 1) the magnitude determines the required rod density, and 2) the rate of change of reactivity may determine when adjustments to this rod density are required to compensate. Both the core hot excess AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 19 magnitude and reactivity swing (i.e., flatness of the hot excess reactivity curve) are controlled during the cycle bundle and core design process. The hot excess reactivity magnitude is controlled in the core design process to ensure that enough rod density is available to compensate for unexpected variations in the core reactivity while maintaining the ability to control the margin to the licensed thermal limits. For example, an excessively high magnitude for hot excess reactivity may result in too many rods inserted to effectively control power peaking thereby affecting thermal limits during operation. On the opposite extreme, a very low hot excess reactivity could potentially result in a condition at which full power could not be achieved if the core reactivity was lower than predicted. Currently operating boiling water reactors (BWRs) maintain rated core power between control rod sequence exchanges by adjusting reactivity through changes in either the core flow or with minor adjustments in the control rod inventory. Reactors operating at EPU conditions without an extended flow window such as the MELLLA+ EPFOD have a reduced capability to accomplish these reactivity adjustments with changes in core flow. One method of compensating for this reduced capability is to perform additional minor rod adjustments between control rod sequence exchanges. An alternate approach has been to design these EPU cores with flatter hot excess reactivity curves versus exposure to reduce the need for frequent reactivity changes. Both methods have been employed to support current EPU operations. The larger flow window available with a MELLLA+ EPFOD allows for increased use of core flow adjustments for reactivity changes. The lower flow and corresponding higher average void conditions available with MELLLA+ operation can increase the observed hot excess magnitude at near end of rated conditions due to potentially higher plutonium production. This impact on hot excess reactivity is relatively small and does not affect safety. Furthermore, any potential increase is mitigated by a combination of both the cycle design process and the ability of the increased flow window to accommodate changes in reactivity during plant operation. The MELLLA+ EPFOD operating conditions do not change the methods used to evaluate hot excess reactivity. The continued applicability of the AREVA methods to the MELLLA+ EPFOD is addressed in Reference 7. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 20 Evaluations have been provided for both the MELLLA+ Equilibrium Cycle core as documented in Tables 3.4 through 3.6 of ANP-3544P (Reference 2, Attachment 9 of the BFN MELLLA+ LAR) and the reference Cycle 19 core as documented in Tables 3.4 through 3.6 of ANP-3553P (Reference 9, Attachment 19 of the BFN MELLLA+ LAR). These evaluations demonstrate acceptable hot excess reactivity results for BFN MELLLA+ EPFOD design. The licensing requirement to meet a minimum SRO cold shutdown margin is found in BFN TS 3.1.1. The specific requirement is to demonstrate that a minimum of 0.38 % k/k shutdown margin with the SRO. Compliance with this Technical Specification requirement is verified with a shutdown margin demonstration performed during the initial startup of each cycle. The TS also requires completion of this surveillance after criticality following any fuel movements or control rod replacements within the core. Each operating cycle is designed to meet the shutdown margin requirements specified by TS 3.1.1 to ensure that the core can be made subcritical with the highest reactivity worth control rod fully withdrawn (and the remaining blades inserted) at the most reactive condition throughout the cycle. This requirement is included in the generic fuel design criteria in Section 5.4 of ANF-89-98(P)(A) (Reference 14). Section 5.4 of Reference 15 indicates that compliance with this requirement for the ATRIUM 10XM fuel design is demonstrated with the cycle-specific calculations. The calculations involved in the cycle-specific analysis of the shutdown margin use the NRC approved CASMO-4 / MICROBURN-B2 methodology, EMF-2158(P)(A) (Reference 16). The continued applicability of this methodology to EPFOD conditions shows that it remains within its approval basis and SER restrictions, as addressed in ANP-2860P Revision 2 Supplement 3P (Reference 7, included as Attachment 27 of the BFN MELLLA+ LAR). Furthermore, significant margin exists to the existing TS criterion due to the practice of using a design target significantly larger than the criterion combined with small observed variations in the cold critical target from cycle to cycle. Therefore, it is concluded based upon observed results and the margin available that the acceptance criterion referred to in the BFN Technical Specifications remains valid for operation at MELLLA+ EPFOD conditions. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 21 The impact of MELLLA+ EPFOD operation is primarily related to operation with higher average void fraction. This results in a higher plutonium production and corresponding increased hot reactivity later in the operating cycle which in turn can decrease hot-to-cold reactivity differences. Consequently smaller cold shutdown margins may result from cores designed for operation with the MELLLA+ EPFOD. However, this potential loss in margin can be accommodated through the core design process within current design and TS cold shutdown margin requirements. All minimum SRO shutdown margin requirements apply to cold most reactive conditions, and are maintained without change. The most reactive temperature at and near end of cycle (EOC) conditions will have a tendency to increase with MELLLA+ EPFOD operation since an increased fraction of fissions is occurring in the plutonium isotopes. AREVA performs the shutdown margin evaluation at a range of temperatures to ensure that the most reactive temperature condition is captured. An evaluation has been provided for the MELLLA+ Equilibrium Cycle core as documented in Tables 2.1 and 3.4 - 3.6 of ANP-3544P (Reference 2, Attachment 9 of the BFN MELLLA+ LAR). The equilibrium core was shown to exhibit a minimum cold SRO shutdown margin of 1.21 %k/k at BOC assuming a short exposure basis for the previous cycle. The values shown in bold in Table 3.4 through 3.6 represent minimum shutdown margin values that occur at temperatures above 68 ºF, which included all cycle exposures greater than or equal to 13,000 MWd/MTU for the short exposure basis for the previous cycle. A similar evaluation of shutdown margin for the reference Cycle 19 core has been provided in Tables 2.1 and 3.4-3.6 of ANP-3553P (Reference 9, Attachment 19 of the BFN MELLLA+ LAR). The reference Cycle 19 core was shown to exhibit a minimum cold SRO shutdown margin of 1.17 %k/k at BOC assuming a short exposure basis for the previous cycle. The values shown in bold in Table 3.4 through 3.6 of Reference 9 represent minimum shutdown margin values that occur at temperatures above 68 ºF, which included all cycle exposures greater than or equal to 14,528 MWd/MTU. This core minimum SRO cold shutdown margin is reported in the cycle-specific fuel cycle design report and is included in Section 7.5 of Reference 2 (Attachment 9 of the BFN MELLLA+ LAR) and Reference 9 (Attachment 19 of the BFN MELLLA+ LAR) for the equilibrium cycle core and the reference Cycle 19 core, respectively. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 22 Based on the previous discussions and the demonstration provided with the ATRIUM 10XM equilibrium core in ANP-3544P and the reference Cycle 19 core in ANP-3553P, it has been shown that the required cold SRO shutdown margin can be achieved for MELLLA+ EPFOD through appropriate fuel and core design. Because plant reactivity margins are established in accordance with approved methodology for each core reload and the Technical Specification acceptance criterion remains applicable, the assessment of these topics for MELLLA+ EPFOD at BFN is acceptable. The SLCS is designed to shut down the reactor from rated power conditions to cold shutdown in the postulated situation that some or all of the control rods cannot be inserted. This manually operated system pumps a sodium pentaborate solution into the vessel, to provide neutron absorption and achieve a subcritical reactor condition. SLCS is designed to inject over a wide range of reactor operating pressures. The lower flows available with a MELLLA+ EPFOD can result in higher core average void fractions at rated power with a corresponding increase in plutonium production. As previously stated, this higher plutonium content can increase the core reactivity at or near EOC conditions when compared to non-MELLLA+ operation. While this increase in EOC reactivity may decrease SLCS margins at higher cycle exposures, this has minimal impact on the minimum SLCS shutdown margin since SLCS is typically limiting at BOC conditions. This BOC limiting behavior is due to the higher poison content of the fuel (i.e. gadolinia for current BWR designs) at beginning of life (BOL) since the gadolinia competes for each neutron. The effectiveness of the injected boron increases with cycle exposure as the gadolinia within the fuel is depleted. This behavior is not credited and SLCS analyses are performed for each reload core over the cycle to ensure that the most limiting condition is evaluated. The depletion impacts of MELLLA+ EPFOD operation are included in this evaluation so any potential increase in EOC reactivity is included. The calculations involved in the cycle-specific analysis of the SLCS shutdown margin use the NRC approved CASMO-4 / MICROBURN-B2 methodology, EMF-2158(P)(A) (Reference 16). Application of this methodology to MELLLA+ EPFOD conditions remains within its approval AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 23 basis and SER restrictions, as addressed in ANP-2860P Revision 2 Supplement 3P (Reference 7, included as Attachment 27 of the BFN MELLLA+ LAR). A SLCS shutdown margin evaluation was performed for the reference cycle assuming a system boron concentration of 720 ppm natural Boron equivalent at 70 oF. SLCS performance modifications that increase the weight percent of the Boron-10 isotope have no effect on the SLCS shutdown margin calculation as long as the minimum 720 ppm natural Boron equivalent at 70 oF is maintained. The results for the Equilibrium Cycle core using this boron concentration demonstratexposure basis for the previous cycle, as shown in Tables 2.1 and 3.4 of ANP-3544P (Reference 2, Attachment 9 of the BFN MELLLA+ EPFOD ). An additional SLCS shutdown margin (SDM) evaluation has been provided for the reference Cycle 19 core as documented in ANP-3553P (Reference 9, Attachment 19 of the BFN MELLLA+ LAR). The reference Cycle 19 core was shown to exhibit a minimum SLCS shutdown margin of 2.63 %k/k at BOC assuming a short exposure basis for the previous cycle. The minimum SLCS shutdown margin result is also documented in the RSAR for each reload cycle, Section 7.3 of the RSAR for the representative cycle 19 (Attachment 17 of the BFN MELLLA+ LAR). Therefore, the SLCS system will continue to meet the required shutdown margin capability for MELLLA+ EPFOD conditions. The continued applicability of AREVA methods to MELLLA+ EPFOD operation is addressed for BFN in ANP-2860P Revision 2 Supplement 3P (Reference 7, included as Attachment 27 of the BFN MELLLA+ LAR). This evaluation concludes that the NRC approved AREVA methodology remains within its approval basis and continues to meet all applicable SER Limitations and Conditions for the expanded operating domain. There are no generic restrictions or SER Limitations and Conditions applicable to AREVA methodology used at BFN that are specific to MELLLA+ EPFOD operation. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 24 Channel Bow Model, Fluence Gradient Range of Applicability During the addition of the SAFLIM3D safety limit methodology (Reference 12) to Section 5.6.5.b of the BFN TS, the following additional condition was added to the BFN operating license (Amendment 285 for Unit 1, Amendment 311 for Unit 2, and Amendment 270 for Unit 3) as documented in SER Section 4.0 of Reference 17.

This license condition addresses an NRC concern that a few channels may exhibit fluence gradients exceeding the measurement database used to develop the fluence based channel bow model. Fuel Cladding Peak Oxide Thickness During the addition of RODEX4 thermal-mechanical methodology (Reference 8) to Section 5.6.5.b of the BFN TS, the following regulatory commitment was made (SER Section 5.0 of Reference 17).

• The reference number for BFN Units 2 and 3 Technical Specification is shown above. The corresponding reference number for Unit 1 is TS 5.6.5.b.11.

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 25

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 26 A 17 46.121 GWd/ST (50.840 GWd/MT) B (BWR-6) 13 53.628 GWd/ST (59.115 GWd/MT) B (BWR-6) 14 56.937 GWd/ST (62.761 GWd/MT) C (BWR-6) 12 52.696 GWd/ST (58.086 GWd/MT) D 12 53.786 GWd/ST (59.288 GWd/MT) E

• 19 - 120% OLTP 52.341 GWd/ST (57.696 GWd/MT) BFN M+SAR (BWR/4 MELLLA+) Equilibrium - 120% OLTP 54.337 GWd/ST (59.896 GWd/MT) BFN MELLLA 19 - 120% OLTP 53.849 GWd/ST (59.358 GWd/MT) BFN MELLLA+ (Reference Cycle) 19 - 120% OLTP 53.875 GWd/ST (59.386 GWd/MT)
• Plant 'E' data is from a reference cycle for a MELLLA+ LAR submittal for a BWR/4 reactor.

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 27 "E" 100.0 100.0 "D" 100.0 99.0 "N" 100.0 85.0 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 28 "E" 3952 (100.0) 102.5 (100.0) 38.56 "D" 3952 (100.0) 101.475 (99.0) 38.95 "N" 3952 (100.0) 87.125 (85.0) 45.36 "O" 3067 (77.6) 56.375 (55.0) 54.40 "P" 2703 (68.4) 56.375 (55.0) 47.95

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 29 0.01.02.03.04.05.06.07.08.09.005101520Plant A Cycle 17Plant B Cycle 13Plant B Cycle 14Plant C Cycle 12Plant D Cycle 12BSEP U1 C19 MELLLA+BFN Eq Cycle MELLLA+BFN U3 C19 MELLLA+BFN U3 C19 MELLLA AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 30 2.03.04.05.06.07.08.09.010.011.012.013.014.005101520Plant A Cycle 17Plant B Cycle 13Plant B Cycle 14Plant C Cycle 12Plant D Cycle 12BSEP U1 C19 MELLLA+BFN Eq Cycle MELLLA+BFN U3 C19 MELLLA+BFN U3 C19 MELLLA AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 31 0.40.50.60.70.80.91.005101520Plant A Cycle 17Plant B Cycle 13Plant B Cycle 14Plant C Cycle 12Plant D Cycle 12BSEP U1 C19 MELLLA+BFN Eq Cycle MELLLA+BFN U3 C19 MELLLA+BFN U3 C19 MELLLA AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 32 0.40.50.60.70.80.9105101520Plant A Cycle 17Plant B Cycle 13Plant B Cycle 14Plant C Cycle 12Plant D Cycle 12BSEP U1 C19 MELLLA+BFN Eq Cycle MELLLA+BFN U3 C19 MELLLA+BFN U3 C19 MELLLA AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 33 0.20.30.40.50.60.70.80.905101520Plant A Cycle 17Plant B Cycle 13Plant B Cycle 14Plant C Cycle 12Plant D Cycle 12BSEP U1 C19 MELLLA+BFN Eq Cycle MELLLA+BFN U3 C19 MELLLA+BFN U3 C19 MELLLA AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 34 0.01.02.03.04.05.06.07.08.09.010.011.012.013.014.005101520Plant A Cycle 17Plant B Cycle 13Plant B Cycle 14Plant C Cycle 12Plant D Cycle 12BSEP U1 C19 MELLLA+BFN Eq Cycle MELLLA+BFN U3 C19 MELLLA+BFN U3 C19 MELLLA AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 35 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 36 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 37 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 38 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 39 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 40

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 41

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AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 46 The pressure relief system prevents overpressurization of the nuclear system during abnormal operational transients (AOTs), the plant American Society of Mechanical Engineers (ASME) upset overpressure protection event, and postulated anticipated transient without scram (ATWS) events. The SRVs along with other functions provide this protection. For BFN, the limiting overpressure event is the Main Steam Isolation Valve Closure with Scram on High Flux (MSIVF). The safety relief valve (SRV) setpoint tolerance is independent of the MELLLA+ operating domain expansion. The AOT, ASME overpressure, and ATWS overpressure reload licensing evaluations for MELLLA+ are performed using the current BFN SRV setpoint tolerance (3%) with an additional 5 psi uncertainty allowance, and assuming one SRV with the lowest opening setpoint pressure is out of service. The SRV setpoint tolerances are monitored at BFN for compliance to the TS requirements. There are no changes to the BFN current licensing basis for the ASME overpressure event. Assumptions and code input parameters are consistent with those in the current licensing basis, other than the addition of the MELLLA+ flow domain change, which allows full power operation at core flows as low as 85% of rated. As required by M+LTR SER Limitation and Condition 12.23.3, the limiting overpressure event for BFN was evaluated using SRV setpoints developed based on a 95/95 statistical treatment of plant-specific SRV performance. The statistical analysis concluded that there is a 99.7% probability, at a 95% confidence, that the peak RPV pressure is bounded by the current licensing basis SRV setpoint assumptions. The ASME overpressure analysis for BFN was performed at the 102% power and 105% increased core flow (ICF) core flow statepoint, and at the 102% power and 85% minimum core flow statepoint for a representative ATRIUM 10XM MELLLA+ core. The analysis of the limiting overpressure event for BFN demonstrates that no change in overpressure relief capacity is required. The peak reactor pressure vessel (RPV) bottom head pressure is 1,350 psig and remains less than the ASME limit of 1,375 psig. The peak RPV dome pressure is 1,319 psig and remains less than the ASME limit of 1,325 psig. The peak pressure values include AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 47 adjustments to address the NRC concerns associated with the void-quality correlation, exposure-dependent thermal conductivity, and Doppler effects. The results of the limiting ATRIUM 10XM overpressure analyses are presented in Figure 3-1 through Figure 3-4. To comply with M+ LTR SER L&C 12.3.d, AREVA analyzes the ASME event on a cycle specific basis. The ATWS analysis discussed in Section 9.3.1 concludes that no increase in the number of SRVs credited in the analysis is required to demonstrate acceptable results. No changes in the pressure relief system or SRV setpoints are required for MELLLA+. The analysis of the Overpressure Relief Capacity topic finds BFN overpressure relief capacity is acceptable. The ASME overpressure event continues to be analyzed each reload analysis and is reported in the RSAR. This process is unchanged by MELLLA+. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 48 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 49 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 50

• The pressures presented in this figure do not include the adjustments associated with NRC concerns with the void-quality correlation, exposure-dependent thermal conductivity, and Doppler effects.

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 51 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 52 The BFN emergency core cooling system (ECCS) is designed to provide protection against postulated loss of coolant accidents (LOCAs) caused by ruptures in the primary system piping. The ECCS performance characteristics do not change for the MELLLA + operating domain expansion. The topics addressed in this evaluation are: Large Break Peak Clad Temperature Plant Specific Acceptable Small Break Peak Clad Temperature Plant Specific Acceptable Local Cladding Oxidation Plant Specific Acceptable Core Wide Metal Water Reaction Plant Specific Acceptable Coolable Geometry Plant Specific Acceptable Long Term Cooling Plant Specific Acceptable Flow Mismatch Limits Plant Specific Acceptable The same LOCA analysis methodology (Reference 18) is used for all break sizes in the break spectrum analysis. As a result, sub-categories of small and large breaks are not used. The break spectrum analysis is performed to identify the characteristics of the break that result in the highest peak cladding temperature (PCT). The analysis examines variation in the following parameters: break location, break type, break size, limiting ECCS single failure and axial power shape. Split breaks from 0.05 ft2 up to the cross sectional area of the recirculation pipes, as well as double-ended guillotine (DEG) breaks with discharge coefficients ranging from 0.6 to 1.0, are evaluated. This ensures the analysis meets the M+ LTR SER Limitation and Condition 12.13 and 12.14 which require that a sufficient number of small break sizes are analyzed at rated power to ensure that the peak PCT break size is identified. To comply with M+ LTR SER L&C 12.3.a, AREVA LOCA analyses utilize the key plant parameters presented in Table 4.1 through 4.8 of Reference 19 (Attachment 11 of the MELLLA+ LAR). Details of the Browns Ferry AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 53 ATRIUM 10XM break spectrum analysis for MELLLA+ operation are presented in Reference 19 (Attachment 11 of the MELLLA+ LAR). It should be noted that the PCT results from the break spectrum analysis are used to determine the characteristics of the limiting break and do not necessarily reflect the licensing PCT. Even though the characteristics of the limiting break will not change with exposure or nuclear fuel design, the value of PCT calculated for any given set of break characteristics is dependent on exposure and power peaking. As a result, heatup analyses are performed to determine the PCT versus exposure for each nuclear fuel design in the core using the boundary conditions determined for the limiting break. The maximum or licensing PCT is documented in the LOCA-ECCS MAPLHGR report (References 20 and 21, Attachments 13 and 15 of the MELLLA+ LAR). To comply with M+ LTR SER L&C 12.3.d, AREVA LOCA analyses provided in References 19, 20, and 21 (Attachments 11, 13, and 15 of the MELLLA+ LAR) demonstrate that AREVA fuel support the 10 CFR 50.46 criteria. The ATRIUM 10XM LOCA break spectrum analyses were performed for initial conditions at 102% rated core power with This is consistent with the discussion presented in the M+ LTR SER Section 4.3.1.3 and complies with the M+ LTR Limitation and Conditions 9.8, 12.10.a, 12.10.b, 12.10.c, and 12.10.d. AREVA performed a full break spectrum analysis (Attachment 11 of the MELLLA+ LAR) and did not rely on any generic analyses or dispositions, satisfying M+ LTR SER Limitation and Conditions 12.3.b and 12.3.c. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 54 In accordance with approved AREVA methods, AREVA only calculates and reports Appendix K PCTs; thereby meeting Limitation and Condition 12.12.a. To comply with M+ LTR SER L&C 12.12.b, AREVA LOCA analyses utilize the key plant parameters presented in Tables 4.1 through 4.8 of Reference 19 (Attachment 11 of the MELLLA+ LAR). A summary of the TLO ATRIUM 10XM break spectrum analysis results is presented in Table 4-1. Based on the break spectrum results, the limiting break characteristics are identified below. Limiting LOCA Break Characteristics Location Recirculation discharge pipe Type / size Split / 0.23 ft2 Single failure Battery (DC) power, board A Axial power shape Top-Peaked Limiting power/flow state point 102%P/ As described in Reference 19 (Attachment 11 of the MELLLA+ LAR), the MAPLHGR multiplier for single-loop operation (SLO) is 0.85 for ATRIUM 10XM fuel. Applying this multiplier to the Reference 20 (Attachment 13 of the MELLLA+ LAR) MAPLHGR limits ensures a LOCA from SLO is less limiting than the limiting LOCA event from two-loop operation. The extension of the power/flow map to MELLLA+ does not impact SLO, as SLO is not allowed in the MELLLA+ operating domain. The limiting break supports operating with all the operational enhancements described in Section 1.2.4 of the GEH M+SAR. Effect of MELLLA+ at Rated Power The rated power PCT results presented in Table 4-1 show that decreasing the flow to the minimum core flow (85% of rated) results in a While the lower core flow can result in earlier critical heat flux (CHF) and a higher PCT, AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 55 Effect of MELLLA+ at Less Than Rated Power The results in Table 4-1 show that they are non-limiting; bound by the limiting results at . While the lower core flow can result in earlier CHF and a higher PCT, Effect of Axial Power Shape As required by M+ LTR SER Limitation and Condition 12.11 and Methods LTR SER Limitation and Condition 9.7, for MELLLA+ applications, the LOCA break spectrum analyses are required to include top-peaked and mid-peaked power shapes to establish the MAPLHGR limits and determine the PCT. Both top-peaked and mid-peaked axial power shapes were considered in the ATRIUM 10XM MELLLA+ LOCA break spectrum analyses for MELLLA+ operation. The boundary conditions from the limiting break were used in the follow-on LOCA-ECCS MAPLHGR analysis to determine the licensing PCT. A comparison of the more limiting top-peaked and mid-peak axial power shape results is presented in Table 4-1. More details are presented in Reference 19 (Attachment 11 of the MELLLA+ LAR). Table 2.1 of Reference 20 (Attachment 13 of the MELLLA+ LAR) reports a licensing PCT of 2008°F for the ATRIUM 10XM fuel to support operation in the Browns Ferry power/flow operating domain - including MELLLA+ operation. The BFN break spectrum response is determined by the ECCS network design that is common to all BWRs. The BFN reactors are small break limited as determined by AREVA's EXEM BWR-2000 evaluation model (Reference 18). This trend in the break spectrum analysis was not impacted by MELLLA+. Effects of MELLLA+ at Rated Power M+LTR SER Limitation and Condition 12.13 requires that the M+SAR include calculations for the limiting small break at rated power/MELLLA+ minimum core flow boundary if the small break AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 56 PCT at rated conditions is limiting. For BFN, the reactors are small break limited as determined by AREVA's EXEM BWR-2000 evaluation model (Reference 18). Appendices B and C of Reference 19 (Attachment 11 of the MELLLA+ LAR) present detailed results for the range of small break PCT calculations performed in support of MELLLA+. The limiting small break PCT for BFN is presented in Table 4-1. Effects of MELLLA+ at Less Than Rated Power M+LTR SER Limitation and Condition 12.10 requires the M+SAR provide a justification why the transition statepoint ECCS-LOCA response bounds the 55% low core flow statepoint. The PCT results summarized in Appendices B and C of Reference 19 (Attachment 11 of the MELLLA+ LAR) show that there are no unusual trends in PCT in the MELLLA+ region and that there is margin to the 2,200°F PCT limit. Effects of Axial Power Shape As required by M+LTR SER Limitation and Condition 12.11 and Methods LTR SER Limitation and Condition 9.7, for MELLLA+ applications, the small and large break ECCS-LOCA analyses have included top-peaked and mid-peaked power shape in establishing the MAPLHGR and determining the PCT. The BFN applications have confirmed the limiting break characteristics utilizing the axial power distributions presented in Figures 4.7 through 4.10 of Reference 19 (Attachment 11 of the MELLLA+ LAR). Appendices B and C of Reference 19 (Attachment 11 of the MELLLA+ LAR) present detailed results for the range of small break PCT calculations performed in support of MELLLA+ using these axial power distributions. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 57 Table 2.1 of Reference 20 (Attachment 13 of the MELLLA+ LAR) presents the results of the ATRIUM 10XM local cladding oxidation analysis. The maximum local cladding oxidation is 1.90%, much less than the 10 CFR 50.46 requirement of less than 17%. Table 2.1 of Reference 20 (Attachment 13 of the MELLLA+ LAR) presents the results of the ATRIUM 10XM core wide metal water reaction analysis. The results show that less than 0.40% of the hypothetical amount that would be generated if all the metal in the cladding cylinders surrounding the fuel, except the cladding surrounding the plenum volume, were to react. The result is less than the 10 CFR 50.46 requirement of less than 1%. Conformance with coolable geometry requirements is demonstrated by conformance with the 2200°F Licensing Basis PCT limit, local cladding oxidation limit of 17%, and total hydrogen generation limit of 1% of the total; therefore, the 10 CFR 50.46 requirement is met. Long-term coolability addresses the issue of reflooding the core and maintaining a water level adequate to cool the core and remove decay heat for an extended time period following a LOCA. For non-recirculation line breaks, the core can be reflooded to the top of the active fuel and be adequately cooled indefinitely. For recirculation line breaks, the core will initially remain covered following reflood due to the static head provided by the water filling the jet pumps to a level of approximately two-thirds core height. Eventually, the heat flux in the core will not be adequate to maintain a two-phase water level over the entire length of the core. Beyond this time, the upper third of the core will remain wetted and adequately cooled by core spray. Maintaining water level at two-thirds core height with one core spray system operating is sufficient to maintain long-term coolability as demonstrated by the Nuclear Steam Supply System (NSSS) vendor (Reference 22). AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 58 The Browns Ferry ATRIUM 10XM MELLLA+ break spectrum analyses AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 59 Single Failure Axial Power Shape Mid-Peaked Top-Peaked Break Size and Location PCT (°F) Break Size and Location PCT (°F) 102% Power / SF-BATTlBA SF-BATTlBB 102% Power / SF-BATTlBA SF-BATTlBB 102% Power / SF-BATTlBA SF-BATTlBB SF-BATTlBA SF-BATTlBB AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 60 The combustible gas control system is designed to ensure an inert atmosphere in the drywell and wetwell is maintained after a postulated LOCA. This is accomplished by injecting nitrogen into the drywell and wetwell to keep the oxygen concentration below 5% by volume. Following a LOCA, combustible gases may be produced through radiolytic decomposition of water and the metal-water reaction of the fuel cladding. The two primary factors that would impact the combustible gas control system is the cladding mass and the reactor thermal power, which are not impacted by EPFOD. Therefore, MELLLA+ has no effect on the post-LOCA combustible gas control system. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 61 The generic disposition of the Rod Block Monitor (RBM) topic in the M+LTR describes that the RBM uses LPRM instrumentation inputs that are combined and referenced to an average power range monitor (APRM) channel. AREVA performs a cycle specific Control Rod Withdrawal Error (CRWE) evaluation to ensure that the RBM will protect the fuel in the event of an inadvertent control rod withdrawal at power. BFN utilizes an RBM with an ARTS LPRM configuration which is included in the cycle-specific evaluation. The results of this evaluation allow for the selection of a RBM setpoint that will ensure that the fuel remains protected if the event were to occur. Details of this cycle specific evaluation are provided in Section 9.1.1. The generic disposition of the rod worth minimizer (RWM) topic in the M+LTR describes that the function of the RWM is to support the operator by enforcing rod patterns until reactor power has reached appropriate levels. The RWM functions to limit the local power in the core to control the effects of the postulated Control Rod Drop Accident (CRDA) at low power. The RWM precludes continuous control rod withdrawal errors during reactor startup by providing appropriate rod block signals to the Reactor Manual Control System (RMCS) rod block circuitry when an out-of-sequence rod is selected for withdrawal. Consistent with the generic disposition discussed above, the BFN RWM supports the operator by enforcing rod patterns until reactor power has reached appropriate levels. AREVA performs a cycle specific CRDA analysis for BFN using rod sequences consistent with or conservative to those enforced by the RWM. This evaluation confirms that the BFN CRDA licensing basis is met for each cycle. This is discussed in more detail in Section 9.2.1.1. The evaluation of the RWM system at BFN is confirmed to be consistent with the generic disposition of the M+LTR. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 62 The generic disposition of the Traversing Incore Probes (TIPs) in the M+LTR describes that there is no change in the neutron flux experienced by the TIPs resulting from the MELLLA+ EPFOD expansion. TVA has confirmed that the TIPs are installed at BFN in accordance with the requirements established by the GEH design specification. No further evaluation of this topic is required for MELLLA+. In accordance with Methods LTR SER Limitations and Condition 9.17 and M+LTR SER Limitation and Condition 12.15, for BFN, the predicted bypass and void fraction at the D-Level LPRMs satisfy the 5% design requirement. The RSAR will validate that the power distribution in the core is achieved while maintaining individual fuel bundles within the allowable thermal limits as defined in the COLR. When moving down and left on the MELLLA+ upper boundary, the hot channel exit void in the bypass region increases. The predicted hot channel exit void in the bypass region does not exceed in the MELLLA+ EPFOD as shown in Table 5-1. M+LTR SER Limitation and Condition 12.15 addresses the use of thermal neutron TIPs operating with predicted bypass voiding above the D-Level LPRMs in excess of the design requirement. BFN utilizes gamma TIPs and operates with bypass voiding at the TIP exit below the limit of concern. Thus, operator actions and procedures that mitigate the effect of bypass voiding on the thermal TIPs and the core simulator are not required for BFN. Therefore, based upon the above discussion,TVA has confirmed that BFN meets all M+LTR dispositions for the TIPs and no special operator actions or procedures are required. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 63 The generic disposition of the RBM topic in the M+LTR describes that the RBM setpoints are established to mitigate the CRWE event during power operation. BFN has APRM RBM TSs (ARTS) based RBM systems. Furthermore, there is no change in reactor power level as a result of the MELLLA+ operating domain expansion. Therefore, BFN is consistent with the M+LTR generic disposition and no further evaluation of the RBM TS values is required as a result of the MELLLA+ EPFOD.In accordance with the M+LTR SER Limitation and Condition 12.16, CRWE analyses have been performed for the MELLLA+ EPFOD. AREVA methods do not utilize generic RBM setpoints, but the analysis supports the BFN specific RBM setpoints. The analysis of this event is described in more detail in Section 9.1.1 with rated power results for the BFN MELLLA+ representative equilibrium cycle design provided in Table 9-1. As detailed in Section 9.1.1, the CRWE results for each cycle are provided in the RSAR (and COLR) and therefore the intent of Limitation and Condition 12.16 is met. Based upon the above,the evaluation of the BFN RBM TS setpoints is confirmed to be consistent with the generic disposition in the M+LTR. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 64 "D" 100.0 99.0 0.017 0.0020 0.000 "N" 100.0 85.0 0.016 0.0019 0.000 "O" 77.6 55.0 0.026 0.0040 0.000 "P" 68.4 55.0 0.023 0.0036 0.000 Note 1: There is no bypass voiding at LPRM levels A, B, or C. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 65 BFN Technical Specification 4.3.1.2 does not allow storage of fuel assemblies in the New Fuel Storage Vault (NFSV). No additional evaluation of NFSV fuel storage criticality is required. The BFN spent fuel storage pool (SFSP) criticality safety analysis (CSA) was previously reviewed by the NRC as part of the EPU LAR submittal (Reference 23) and review process (References 24 and 25). This evaluation focuses on the continued applicability of this CSA to the addition of the MELLLA+ EPFOD to BFN operations. ATRIUM 10XM is the current fuel design and is the design that will continue to be loaded for future MELLLA+ EPFOD operation at BFN. A small number of exposed ATRIUM-10 assemblies may also be co-resident in the initial MELLLA+ operating cycles. The SFSP CSA for these fuel designs are provided in ANP-3160P (Reference 26) and ANP-2945P (Reference 27), respectively. This evaluation focuses on storage of the ATRIUM 10XM fuel design based upon the following: The reference bounding lattices used in the ATRIUM-10 SFSP CSA (Reference 27) are bounded by those used for the ATRIUM 10XM SFSP CSA (Reference 26). Table B.2 of the Reference 26 report documents approximately 0.005 k margin to the corresponding ATRIUM-10 reference bounding lattices. Any co-resident ATRIUM-10 fuel assemblies will be limited to operating in the MELLLA+ operating domain to a single cycle or less in low power locations on or near the periphery of the core. For this reason the differences in depletion and void histories due to MELLLA+ operation are insignificant. No changes are required to the fuel design for future operation in the MELLLA+ expanded operating domain. ATRIUM 10XM fuel stored in the SFSP must meet the criticality storage requirements provided in Table 2.1 of ANP-3160P. The ATRIUM 10XM SFSP CSA document follows the intent of the NRC interim staff guidance document for SFSP criticality safety analyses, DSS-ISG-2010-01 Revision 0 (Reference 28). A number of sensitivities were performed and are documented in Section 6 of the ANP-3160P CSA that were used to ensure that the original CSA was based upon limiting lattices reactivity for a range of operational conditions. These sensitivities include AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 66 impact of power density, in-core depletion fuel temperature, controlled depletion, and void history. The assumed void history represents the greatest potential change with implementation of the MELLLA+ EPFOD due to the higher core average void fraction when operating at the lower flow conditions. Figure 6.5 of ANP-3160P shows the results of a sensitivity evaluation with respect to in-core depletion void history and its effect on the in-rack lattice k-infinity. The maximum in-rack reactivity condition from this sensitivity was utilized in establishing the reference bounding lattices used in the CSA. It is noted that this maximum reactivity condition occurs at low void history and therefore, potentially higher void values from MELLLA+ operation will not challenge this assumption. The control density assumed in the in-core depletion can also be potentially impacted with the implementation of MELLLA+ EPFOD. The increase in core average voids may make it necessary to operate with a lower control rod density in order to compensate for the reduction in reactivity due to the decrease in moderator. Table 6.6 of ANP-3160P provides the results of a sensitivity in which the in-rack reactivity is compared for cases assuming both controlled and uncontrolled in-core depletion. For ATRIUM 10XM fuel, the uncontrolled depletion results were shown to result in the higher in-rack k-infinities. The CSA was consequently based upon uncontrolled in-core depletions which remain bounding for MELLLA+ EPFOD operation. The core power density and assumed depletion fuel temperatures are dependent upon power level. While there is no change in the power level with the implementation of the MELLLA+ EPFOD, the SFSP CSA sensitivities include the impact of variations in power density of +/- 50% as well as sensitivity to variations in fuel temperature of +/- 100F. As noted above, the BFN criticality storage constraints are identified in Table 2.1 of the SFSP CSA (Reference 26) report. A cycle-specific confirmation of the SFSP criticality storage constraints is performed during the bundle and core design phase. The bundles used in the ATRIUM 10XM MELLLA+ equilibrium core design have been confirmed to comply with these criticality storage requirements, that is all lattices have an in-rack lattice k-infinity < 0.8825. An example of the documentation of the cycle-specific confirmation is provided for the Representative Cycle 19 core in Section 7.4 of the Reload Safety Analysis Report (Attachment 17 of the BFN MELLLA+ LAR). AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 67 Based upon the preceding evaluation, the criticality analyses supporting fuel storage at BFN is not adversely impacted with the implementation of MELLLA+ EPFOD and will remain valid for future cycles using the ATRIUM 10XM fuel design. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 68 During power operation, the radiation sources in the core are directly related to the fission rate. These sources include radiation from the fission process, accumulated fission products, and neutron activation reactions. The topics addressed in this evaluation are: Post Operational Radiation Sources for Radiological and Shielding Analysis Generic Confirmed For BFN, the radiological source terms are based upon the methods and assumptions for AST as defined in Regulatory Guide (RG) 1.183. The continued applicability of the BFN radiological source terms to the CLTP of 3952 MWt was previously reviewed as part of the EPU LAR submittal (Reference 23). Operation in the MELLLA+ EPFOD does not involve an increase in power or require a change in the fuel design from that previously reviewed. The post-operation radiation sources in the core are primarily the result of accumulated fission products which in turn are dependent upon the fission rate. The fission rate is not affected since there is no change in power level. Based upon the above, the BFN AST radiological source terms remain applicable for operation in the MELLLA+ EPFOD. Furthermore, the ATRIUM 10XM MELLLA+ reference equilibrium cycle design documented in ANP-3544P (Reference 2, provided as Attachment 9 of the BFN MELLLA+ LAR) has been reviewed to ensure that MELLLA+ cores for BFN will remain within the source term calculation basis. This review is summarized in the table below. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 69 Core Power (MWt) < 3952 3952 Meets Criteria. Average Enrichment Meets Criteria. Peak Bundle Average Exposure (GWd/MTU) Meets Criteria. Maximum Rod Average Exposure (GWd/MTU)f Meets Criteria. AST Core Average Exposure Limit (GWd/MTU) Meets Criteria. Maximum rod average LHGR for rods with burnups exceeding 54 GWD/MTU. (kw/ft) < 6.3 < 6.3 Meets Criteria. Criteria from footnote 11 of RG 1.183.

• Unless otherwise noted, the criteria provided in the table above are based upon the inputs assumed for the Browns Ferry AST source term calculations. f Maximum Rod Average Exposure refers to the maximum value for the 'rod average exposure' of all fuel rods in the core.

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 70 This section addresses the evaluations that are applicable to MELLLA+. AREVA analysis supporting Sections 9.1, 9.2, and 9.3 utilize void quality correlations addressed in Reference 7 Appendix A (Attachment 27 of the MELLLA+ LAR) thereby meeting SER Limitation and Condition 9.19. The BFN FSAR defines the licensing basis AOTs. Sections 9.1.1 through 9.1.3 provide an assessment of the effect of the MELLLA+ operating domain expansion on each of the BFN final safety analysis report (FSAR) limiting AOT events and key non-limiting events. Sections 9.1.1 through 9.1.3 include fuel thermal margin, pressurization, and loss of water level events. The fuel related overpressure protection analysis events are addressed in Section 3.1.2. The topics addressed in this evaluation are: Fuel Thermal Margin Events Plant Specific Acceptable Power and Flow Dependent Limits Plant Specific Acceptable Non-Limiting Events Plant Specific Acceptable Operating limits are established to ensure the AOT and accident acceptance criteria are satisfied. A review of the licensing basis of the Browns Ferry plant was performed to identify the potentially limiting thermal margin events, i.e., those events which set the thermal limits. The potentially limiting thermal margin events include: Generator Load Rejection No Bypass (LRNB) Turbine Trip No Bypass (TTNB) Feedwater Controller Failure (Maximum Demand) (FWCF) Loss of Feedwater Heating (LFWH) Control Rod Withdrawal Error (CRWE) AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 71 The fuel loading error is categorized as an Infrequent Event under AREVA methodology and therefore the ultimate requirement is that the offsite dose consequences shall not exceed a small fraction of the 10 CFR 50.67 dose limits. A dose consequence evaluation is not required if it is shown that fuel failures do not occur as a result of the loading error. For example, if the fuel loading error is bounded by the AOTs listed above no fuel failure would be expected to occur. The transient analysis codes presented in Table 1-1 are used to perform the AOT analyses. Analyses for an equilibrium Browns Ferry MELLLA+ cycle were performed at ICF and minimum flow conditions and are presented in this report. Attachment 17 of the MELLLA+ LAR provides AREVA's Reload Analysis report created to present licensing results of the MELLLA+ demonstration cycle, thereby meeting M+SAR Limitation and Condition 12.4. Transient event calculations are performed on a cycle specific basis for the actual core configuration fuel types thereby meeting M+ LTR SER L&C 12.3.d. Table 9-1 provides a comparison of the rated power results with ICF and minimum core flow for the potentially limiting events. Table 9-2 shows a comparison of the potentially limiting rated power transients between MELLLA and MELLLA+ conditions. The potentially limiting pressurization events (LRNB, TTNB, and FWCF) are caused by a rapid closure of the turbine control valves or the turbine stop valves. In the case of the FWCF event, the rapid valve closure occurs after an overcooling phase as the water level increases due to a significant increase in feedwater flow. The rapid turbine valve closure causes a compression wave to travel through the steam lines into the vessel and create a rapid pressurization of the core. The increase in pressure causes a decrease in core voids, which in turn causes a rapid increase in power. The rapid closure of the turbine valves also causes a scram. The core power excursion is terminated primarily by the reactor scram and revoiding of the core. Results of pressurization events are presented in Table 9-3 for base case transients and Table 9-4 for equipment out-of-service (EOOS) scenarios. The responses of various reactor and plant parameters during the LRNB, TTNB, and FWCF events for the rated power ICF core flow statepoints at EOC are presented in Figures 9-1 through 9-9. For the pressurization events, the limiting transient event is initiated from ICF conditions. Table 9-2 shows that depleting with AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 72 The LFWH event was analyzed for the equilibrium Browns Ferry MELLLA+ cycle and the results are presented in Tables 9-1 and 9-2. The LFWH event analysis supports an assumed 100°F decrease in the feedwater temperature. This results in an increase in core inlet subcooling, thereby reducing voids, increasing power and causing a shift in the axial power distribution. As a result of the axial power shift and increased core power, voids begin to build up in the bottom region of the core, acting as negative feedback to the increased subcooling effect and moderating the core power increase. The CRWE transient is an inadvertent reactor operator initiated withdrawal of a control rod. This withdrawal increases local power and core thermal power, lowering the core MCPR. The CRWE transient is typically terminated by control rod blocks initiated by the RBM. The CRWE event is evaluated on a cycle-specific basis with results provided in the cycle-specific RSAR. The CRWE event was analyzed for the equilibrium Browns Ferry MELLLA+ cycle and results are presented in Tables 9-1 through 9-3. Analyses were performed for a range of potential BFN RBM setpoints from 107% to 117%, including the unblocked condition. The rated power results presented in Tables 9-1 through 9-3 are for the unblocked condition. There are two types of fuel loading errors possible in a BWR: the mislocation of a fuel assembly in a core position prescribed to be loaded with another fuel assembly and the misorientation of a fuel assembly with respect to the control blade. The fuel loading errors are evaluated on a cycle-specific basis with results provided in the cycle-specific RSAR. The mislocation and misorientation analysis results for the equilibrium Browns Ferry MELLLA+ cycle show that they remain bounded by the results of the limiting AOTs and therefore a dose consequence evaluation was not required. As discussed in Section 2.2, MCPR and LHGR limits are established to ensure that the steady state and AOT acceptance criteria are met. These limits are modified by power- and flow-dependent limits or multipliers when the plant is operating at less than 100% power and/or 100% core flow. The flow-dependent MCPR (MCPRf) limits and LHGR multipliers (LHGRFACf) are based on CPR and heat flux changes experienced by the fuel during postulated slow recirculation flow AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 73 increase analyses. The analysis assumes a failure of the recirculation flow control system such that the core flow increases slowly to the maximum flow physically permitted by the equipment. An uncontrolled increase in flow creates the potential for a significant increase in core power and heat flux. The MCPRf limits are set such that the increase in core power, resulting from the maximum increase in core flow, does not result in a violation of the SLMCPR. Table 9-5 summarizes the results of the slow recirculation flow increase analysis and compares them with the MCPRf limits. The LHGRFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a slow flow runup. The power-dependent MCPR (MCPRp) limits are established or confirmed on a cycle-specific basis to ensure that less than 0.1% of the fuel rods in the core are expected to experience boiling transition during an AOT initiated from rated or off-rated conditions. The MCPRp limits arSLMCPR. These limits can be established as a function of exposure and/or scram speed. An evaluation performed in Reference 7 concludes that the SLMCPR penalty specified in M+ LTR SER Limitation and Condition 9.5 is not required for Browns Ferry when using AREVA methods. The power-dependent LHGR multipliers (LHGRFACp) are established to protect against fuel melting and overstraining (i.e. thermal-mechanical response) of the cladding during an AOT. The LHGRFACp multipliers are determined using the RODEX4 (Table 1-1) methodology and are applied to both the UO2 and GdO2 fuel rods, thereby meeting the SER Limitation and Condition 9.9 discussed in Appendix A. Both the LHGRFACf and LHGRFACp multipliers are applied directly to the steady state LHGR limits. The lower value of LHGRFACf and LHGRFACp is used to determine the margin to limits for a given core statepoint. The power and flow dependent limits are established or confirmed on a cycle-specific basis and are reported in the RSAR, thereby meeting SER Limitations and Conditions 9.9 and 9.10 discussed in Appendix A, and M+LTR Limitation and Condition 12.4. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 74 A disposition of events was performed to support AREVA fuel at BFN for MELLLA+ operation. The results identify the events discussed in Section 9.1.1 as the potentially limiting AOTs. The other events listed below are non-limiting. Inadvertent HPCI Start Fast Recirculation Increase MSIV Closure, All Valves MSIV Closure, One Valve TTNB with Scram on High Flux (Failure of Direct Scram) Previous analyses have shown that MSIV closure with scram on high flux is limiting, mainly due to the smaller available volume to pressurize. MELLLA+ operation does not impact the non-limiting nature of the event. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 75 TTNB Peak Neutron Flux % Initial 315 254 Peak Heat Flux % Initial 122 117 Peak Vessel Pressure psia 1319 1314 CPR (TSSS) NA 0.31 0.27 LRNB Peak Neutron Flux % Initial 324 272 Peak Heat Flux % Initial 119 123 Peak Vessel Pressure psia 1314 1322 CPR (TSSS) NA 0.31 0.27 FWCF Peak Neutron Flux % Initial 322 257 Peak Heat Flux % Initial 125 122 Peak Vessel Pressure psia 1275 1272 CPR (TSSS) NA 0.35 0.30 LFWH CPR* NA 0.14 -- CRWE CPR* NA 0.29f 0.23f

• f Result presented is for a RBM setting of 117%, equivalent to an unblocked condition.

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 76 Load Rejection with Bypass Failure 0.31 0.31 Turbine Trip with Bypass Failure 0.31 0.31 Feedwater (FW) Controller Failure Max Demand 0.34 0.35 Loss of FW Heating 0.13 0.14 Rod Withdrawal Error 0.27 (Note 1) 0.29 (Note 1) Slow Recirculation Increase MCPRf (Note 2) MCPRf (Note 2) Notes: 1. Results presented for RBM setting of 117%, equivalent to an unblocked setting. 2. The MCPRf limits developed for both MELLLA and MELLLA+ operating domain protect plant operation within their respective power/flow maps. MCPRf limits required to protect MELLLA+ operation do not adversely affect the limits developed for MELLLA. Therefore, the MCPRf limits are the same for both MELLLA and MELLLA+, as shown in Figure 9-10. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 77 100% Power 0.35 1.00 90% Power 0.38 1.00 77.6% Power 0.43 1.00 70% Power 0.46 1.00 100% Power 0.31 1.00 90% Power 0.32 1.00 77.6% Power 0.33 1.00 70% Power 0.31 1.00 100% Power 0.31 1.00 90% Power 0.31 1.00 77.6% Power 0.32 1.00 70% Power NA NA 100% Power 0.29 1.00 85% Power 0.29 1.00 65% Power 0.36 1.00 40% Power 0.56 1.00

• MCPRp is caP value at 100% Power is 1.41 (1.06 + 0.35).

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 78 100% Power 0.33 1.00 90% Power 0.36 1.00 77.6% Power 0.40 1.00 70% Power 0.43 0.98 100% Power 0.38 1.00 90% Power 0.36 1.00 77.6% Power 0.32 1.00 70% Power 0.31 1.00 100% Power 0.37 1.00 90% Power 0.40 1.00 77.6% Power 0.46 1.00 70% Power 0.49 1.00 100% Power 0.30 1.00 90% Power 0.31 1.00 77.6% Power 0.31 1.00 70% Power NA NA 100% Power 0.35 1.00 90% Power 0.38 1.00 77.6% Power 0.42 1.00 70% Power 0.45 0.99 100% Power 0.29 1.00 90% Power 0.30 1.00 77.6% Power 0.29 1.00 70% Power NA NA 100% Power 0.32 1.00 90% Power 0.35 1.00 77.6% Power 0.41 1.00 70% Power 0.49 1.00

• MCPRp AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 79 100% Power 0.39 1.00 90% Power 0.38 1.00 77.6% Power 0.42 1.00 70% Power 0.48 1.00 100% Power 0.32 1.00 90% Power 0.33 1.00 77.6% Power 0.39 1.00 70% Power 0.46 1.00 100% Power 0.30 1.00 90% Power 0.31 1.00 77.6% Power 0.38 1.00 70% Power 0.45 1.00 100% Power 0.38 1.00 90% Power 0.42 1.00 77.6% Power 0.47 1.00 70% Power 0.50 1.00 100% Power 0.43 0.98 90% Power 0.42 0.98 77.6% Power 0.44 0.96 70% Power 0.47 0.95 100% Power 0.40 1.00 90% Power 0.44 1.00 77.6% Power 0.49 1.00 70% Power 0.53 1.00 100% Power 0.38 1.00 90% Power 0.41 0.97 77.6% Power 0.46 0.96 70% Power 0.49 0.94
• MCPRp AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 80 30 1.36 1.61 40 1.34 ----- 50 1.36 ----- 60 1.36 ----- 70 1.30 ----- 78 ----- 1.28 80 1.25 ----- 90 1.22 ----- 100 1.17 ----- 107 1.09 1.28 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 81 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 82 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 83 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 84 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 85 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 86 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 87 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 88 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 89 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 90 1.01.1 1.2 1.3 1.41.51.6 1.7 1.830405060708090100110 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 91 Control Rod Drop Accident Plant Specific Acceptable (Section 9.2.1.1) The radiological consequences of this design basis accident (DBA) are evaluated to determine off-site doses as well as control room operator doses. DBA calculations are generally based on core inventory sources or Technical Specification source terms. For BFN, two postulated CRDA events govern the analysis of radiological consequences. For Event 1, the release path is via the mechanical vacuum pump (MVP) at low power operation. For Event 2, the release path is via the condenser and the steam jet air ejectors. For Event 1, the plant is not operating in the MELLLA+ operating domain as shown by the Power/Flow map, and therefore there is no effect on the results. Similarly, the CRDA is only a concern at low power conditions so radiological release for Event 2 is also not affected by MELLLA+. Event 1 (MVP release path) represents the bounding radiological consequences for CRDA at BFN. The CRDA release is dependent on the source terms and maximum peaking factor. Operation in the MELLLA+ operating domain does not affect the Alternate Source Term (AST) CRDA source term and the peaking factor remains bounding. There are no changes to removal, transport, or dose conversion assumptions for this event. Therefore, the BFN CRDA evaluation for the MELLLA+ operating domain is bounded by the analysis for the current licensed operating domain, and no further evaluation is required. The CRDA analysis is performed for each reload core with the results provided in the cycle specific RSAR. These results are used to validate that the maximum deposited enthalpy meets the regulatory limit and the number of projected fuel rod failures remains below the value used in the licensing basis dose analysis, 850 rod failures for BFN. A fuel rod failure is determined to AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 92 occur if the deposited enthalpy exceeds the fuel damage threshold of 170 cal/g. The results of a CRDA analysis for an ATRIUM 10XM representative equilibrium cycle at MELLLA+ conditions (Reference 2, provided as Attachment 9 of the BFN MELLLA+ LAR) is provided in the following table. Maximum dropped control rod worth, mk 8.42 Core average Doppler coeffioF -10.5 x 10-6 Effective delayed neutron fraction 0.0052 Four-bundle local peaking factor 1.449 Maximum deposited fuel rod enthalpy, cal/g 145.4 Maximum number of rods exceeding 170 cal/g 0 Results for the representative Cycle 19 core are provided in Section 6.2 of the RSAR (Attachment 17 of the BFN MELLLA+ LAR).

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 93 This section considers two special events: ATWS overpressure and station blackout (SBO). The ATWS with core instability is addressed in the GE M+SAR. The operator actions required as a result of ATWS are reviewed and discussed as a part of Section 10.9 of the GEH M+SAR. The topics addressed in this evaluation are: ATWS (Overpressure) Plant Specific Acceptable ATWS (Suppression Pool Temperature and Containment Pressure) Plant Specific Evaluation provided in the GEH M+SAR ATWS (Peak Cladding Temperature and Oxidation) Plant Specific Acceptable Station Blackout Generic Confirmed ATWS with Core Instability Generic Evaluation provided in the GEH M+SAR There is no change in core power, decay heat, pressure, or steam flow at rated power and core flow conditions as a result of the MELLLA+ operating domain expansion. However, an ATWS recirculation pump trip (RPT) initiated from rated power on the MELLLA+ minimum core flow boundary results in a smaller power reduction than is achieved when the ATWS RPT is initiated from rated power on the MELLLA minimum core flow boundary. The ATWS evaluation acceptance criteria are to: Maintain reactor vessel integrity (i.e., peak vessel bottom pressure less than the ASME Service Level C limit of 1,500 psig) AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 94 Maintain containment integrity (i.e., maximum containment pressure lower than the design pressure of the containment structure and maximum suppression pool temperature lower than the pool temperature limit) Maintain coolable core geometry Plant-specific ATWS analyses are performed to demonstrate that the ATWS acceptance criteria are met for operation in the MELLLA+ operating domain. BFN meets the ATWS mitigation requirements in 10 CFR 50.62 for an alternate rod insertion (ARI) system, SLCS boron injection equivalent to 86 gpm, and automatic RPT logic (i.e., ATWS-RPT). The plant-specific ATWS analyses take credit for the ATWS-RPT and SLCS. However, ARI is not credited. AREVA only calculates the peak pressure from the initial overpressure phase of an ATWS. Transient event calculations are performed on a cycle specific basis for the actual core configuration fuel types thereby meeting M+ LTR SER L&C 12.3.d. In accordance with M+LTR SER Limitations and Conditions 12.18.e and 12.18.f, the key input parameters to the plant-specific ATWS analyses for the initial ATWS overpressure phase are provided in Table 9-6. For key input parameters that are important to simulating the ATWS analysis and are specified in the TSs (e.g., SLCS parameters, ATWS-RPT), the calculation assumptions are consistent with the proposed BFN TS values and plant configuration. In addition, the EOOS assumptions for ATWS are consistent with TS requirements. M+LTR SER Limitation and Condition 12.23.2 requires that the plant-specific automatic settings be modeled for ATWS. For BFN, the plant automatic settings, which include the ATWS-RPT, low steam line pressure isolation, and SRV actuation, are modeled based on the input parameters in Table 9-6. As required by M+LTR SER Limitation and Condition 12.23.8, the plant-specific ATWS analyses account for plant- and fuel-design-specific features including debris filters. This evaluation reviewed the results of the ATWS analyses considering the limiting cases for RPV overpressure and for suppression pool temperature / containment pressure. Previous evaluations considered four ATWS events. For the inadvertent opening of a relief valve (IORV) event, the reactor vessel is not pressurized from reactor isolation; therefore, this event is non-limiting. M+ LTR SER Limitation and Condition 12.17 is satisfied for the initial overpressure phase of the event. For the loss of offsite power (LOOP) event, the fast opening of the bypass AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 95 valves will reduce the pressure wave created by the reactor isolation; therefore, this event is non-limiting. The only two cases that need to be further analyzed include (1) Main Steam Isolation Valve Closure (MSIVC) and (2) pressure regulator failure open (PRFO). These events have been analyzed and the results are presented below. The plant-specific ATWS analysis is performed using the approved COTRANSA2 methodology documented in Reference 29. The ATWS analysis using the COTRANSA2 methodology is the plant's licensing basis for peak pressure criteria for this application. The key inputs to the Browns Ferry ATWS overpressurization analysis are provided in Table 9-6 in order to comply with M+ LTR SER L&C 12.3.a. The results of the analysis are presented in Tables 9-7 and 9-8, including comparisons between MELLLA and MELLLA+ conditions. Tables 9-9 and 9-10 present the ATWS-MSIV and ATWS-PRFO sequence of events, respectively, for the BOC exposure. Tables 9-11 and 9-12 present the ATWS-MSIV and ATWS-PRFO sequence of events, respectively, for the EOC exposure. The results of the MSIVC and PRFO ATWS events at BOC (limiting exposure for peak vessel pressure) are provided in Figures 9-11 through 9-18. The limiting ATWS event with respect to RPV overpressure for BFN is PRFO. The PRFO event, prior to SLCS initiation, produces the highest peak lower plenum pressure (1498 psig). The peak pressure value includes adjustments to address the NRC concerns associated with the void-quality correlation, exposure-dependent thermal conductivity, and Doppler effects. The results show that the ATWS overpressurization criteria are met for reactor conditions with one lowest setpoint main steam relief valve (MSRV) out of service. The analysis of the limiting overpressure event for BFN demonstrates that no change in overpressure relief capacity is required. No other changes in the pressure relief system or SRV setpoints are required for MELLLA+. The analysis finds BFN overpressure relief capacity is acceptable. The ATWS overpressure event continues to be analyzed each reload analysis and is reported in the RSAR. This process is unchanged by MELLLA+. Any cycle-specific changes will be captured for analyses supporting the cycle of interest, as required by M+ LTR SER Limitation and Conditions 12.3.b, 12.3.c, 12.18.d, 12.23.1, 12.23.9, and 12.24.4. As required by M+LTR SER Limitation and Condition 12.23.3, the limiting overpressure event for BFN was evaluated using SRV setpoints developed based on a 95/95 statistical treatment AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 96 plant-specific SRV performance which concluded that there is a 99.7% probability, with a 95% confidence, that the peak RPV pressure is bounded by the current licensing basis SRV configuration. The SRV setpoints shown in Table 9-6, which were assumed in this peak pressure calculation, bound the 95/95 statistical treatment of plant specific SRV performance. For ATWS events, the acceptance criteria for PCT and local cladding oxidation for ECCS, defined in 10 CFR 50.46, are adopted to ensure an ATWS event does not impede core cooling. Coolable core geometry is assured by meeting the 2200°F PCT and the 17% local cladding oxidation acceptance criteria stated in 10 CFR 50.46. There is no core uncovering associated with the ATWS event, hence the PCT and local cladding oxidation results will be bounded by LOCA (Section 4.3). Therefore, the PCT and local cladding oxidation for the BFN ATWS events is qualitatively evaluated to demonstrate compliance with the acceptance criteria of 10 CFR 50.46. This evaluation satisfies M+ LTR SER Limitation and Condition 12.18.c for the initial overpressure phase of the events. The generic disposition of the SBO topic in the M+LTR describes that there is no significant change in core power, decay heat, pressure, or steam flow as a result of the MELLLA+ operating domain expansion. Consistent with the generic disposition presented above, there is no change in the reactor power level as a result of the MELLLA+ operating domain expansion. As discussed in Section 1.2.3 of the GEH M+SAR, there is no significant change in decay heat as a result of the MELLLA+ operating domain expansion. For BFN, there are no increases in reactor operating pressure as a result of MELLLA+ operating domain expansion. For BFN, there are no significant changes in the main steam flow rate. The numerical values showing no significant changes to reactor operating power and main steam (MS) flow rate are presented in Table 1-2 of the GEH M+SAR. No further evaluation is required. The evaluation of the BFN SBO is confirmed to be consistent with the generic disposition in the M+LTR. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 97

Reactor Power (MWt) 3952 3952 Rated value Analyzed Power (MWt) 3952 3952 Rated value Analyzed Core Flow (Mlbm/hr / % Rated) 107.625 / 105% 101.475 / 99% 107.625 / 105% 87.125 / 85% Plant-specific value Reactor Dome Pressure (psia) 1050 1050 Rated value MSIV Closure Time (sec) 4 4 Nominal value High Pressure ATWS-RPT Setpoint (psig) 1177 1177 Analytical limit ATWS-RPT Delay Time (sec) 0.50 0.75 Plant configuration Number of MSRVs 13 13 Plant configuration Number of MSRVs OOS 1 1 Plant configuration Each MSRV Capacity at 103% of 1,090 psig (lbm/hr) 870,000 870,000 Nominal value MSRV Analytical Opening Setpoints (psig) 1174 (L) 1184 (M) 1194 (H) 1174 (L) 1184 (M) 1194 (H) Technical Specification Value plus Uncertainty Allowance AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 98 Peak Vessel Pressure (psig)

• 1469 1498 1500 Peak Cladding Temperature (deg F) f NA NA 2,200 Peak Local Cladding Oxidation (%) f NA NA 17 Reactor Power (Heat Flux, %) PRFO at COAST 150 Figure 9-19 Reactor Vessel Pressure (psig) PRFO at BOC 1498 Figure 9-17 Peak Cladding Temperature f NA NA NA
• The peak pressure results include adjustments to address the NRC concerns associated with the void-quality correlation, exposure-dependent thermal conductivity, and Doppler effects. In regards to the NRC concerns about exposure dependent thermal conductivity, a generic pressure penalty was conservatively calculated for the ATWS event, for both the MELLLA and MELLLA+ results shown, based on the change in dome pressure during the overpressure phase of the event. For the MELLLA+ analysis, this penalty was calculated to be 6 psi. However, a method exists to develop the thermal-conductivity penalty by decreasing the core average thermal-conductivity input into COTRANSA2. For the MELLLA+ analysis, the penalty calculated using this method was 2 psi. Therefore, there is 4 psi of margin to the reported number above for MELLLA+. f The PCT and local cladding oxidation for the BFN ATWS events are qualitatively evaluated to demonstrate compliance with the acceptance criteria of 10 CFR 50.46.

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 99 1 MSIV Isolation Initiated 0.00 2 MSIVs Fully Closed 4.00 3 Peak Neutron Flux 4.50 4 High Pressure ATWS Setpoint 4.951 5 Opening of the First Relief Valve 5.417 6 Recirculation Pumps Trip 5.701 7 Peak Heat Flux 5.972 8 Peak Vessel Pressure 12.239 1 TCVs and Bypass Valves Start Open 0.00 2 MSIV Closure Initiated by Low Steam Line Pressure 5.971 3 MSIVs Fully Closed 9.971 4 High Pressure ATWS Setpoint 13.370 5 Opening of the First Relief Valve 13.757 6 Peak Neutron Flux 10.598 7 Recirculation Pumps Trip 14.120 8 Peak Heat Flux 14.215 9 Peak Vessel Pressure 20.742 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 100 1 MSIV Isolation Initiated 0.00 2 MSIVs Fully Closed 4.00 3 Peak Neutron Flux 4.472 4 High Pressure ATWS Setpoint 4.879 5 Recirculation Pumps Trip 5.629 6 Opening of the First Relief Valve 5.396 7 Peak Heat Flux 5.685 8 Peak Vessel Pressure 11.311 1 TCVs and Bypass Valves Start Open 0.00 2 MSIV Closure Initiated by Low Steam Line Pressure 5.750 3 MSIVs Fully Closed 9.750 4 Peak Neutron Flux 10.363 5 High Pressure ATWS Setpoint 13.126 6 Opening of the First Relief Valve 13.538 7 Recirculation Pumps Trip 13.876 8 Peak Heat Flux 13.857 9 Peak Vessel Pressure 19.518 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 101 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 102 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 103

• The peak pressure results do not include adjustments to address the NRC concerns associated with the void-quality correlation, exposure-dependent thermal conductivity, and Doppler effects.

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 104 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 105 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 106 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 107

• The peak pressure results do not include adjustments to address the NRC concerns associated with the void-quality correlation, exposure-dependent thermal conductivity, and Doppler effects.

AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 108 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 109 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 110 1. NEDC-33006-A, Revision 3, "General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus", GE-Hitachi, Non-Proprietary Version, June 2009. (ADAMS Accession nos. ML091800513 and ML091800516)2. ANP-3544P Revision 0, , AREVA Inc. December 2016. 3. ANP-2637 Revision 6, , October 2014. 4. ANP-2860P Revision 2, , AREVA NP, October 2009. 5. Supplement 1P Revision 0 to ANP-2860P Revision 2, , AREVA NP, November 2012. 6. Supplement 2P Revision 1 to ANP-2860P Revision 2, , AREVA Inc., August 2015. 7. Supplement 3P Revision 2 to ANP-2860P Revision 2, , AREVA Inc., December 2017. 8. BAW-10247PA Revision 0, Reactors, AREVA NP, February 2008. 9. ANP-3553P Revision 0, , AREVA Inc., January 2017. 10. ANP-10298PA Revision 0, , AREVA NP, March 2010. 11. ANP-3140(P) Revision 0, "Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation," AREVA NP, August 2012. 12. ANP-10307PA Revision 0, , AREVA NP, June 2011. 13. NEDC-33173-A Revision 4, , November 2012. 14. ANF-89-98(P)(A) Revision 1 and Supplement 1, , Advanced Nuclear Fuels, May 1995. 15. ANP-2899P Revision 0, , AREVA NP, April 2010. 16. EMF-2158(P)(A) Revision 0, , Siemens Power Corporation, October 1999. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 111 17. Letter USNRC to TVA, July 31, 2014. (ML14113A286) 18. EMF-2361(P)(A) Revision 0, Framatome ANP, May 2001, as supplemented by the site specific approval in NRC safety evaluations dated April 27, 2012 (for Unit 1), February 15, 2013 (for Units 2 and 3), and July 31, 2014. 19. ANP-3546P Revision 0, March 2017. 20. ANP-3547P Revision 0, March 2017. 21. ANP-3548P Revision 0, , March 2017. 22. NEDO-20566AGeneral Electric Company, September 1986. 23. Letter, JW Shea (TVA) to USNRC, "Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 1, Spent Fuel Pool Criticality Safety Analysis Information", CNL-15-249, December 15, 2015. (Accession Number ML15351A097) 24. ANP-3465P Revision 0, Response to RAI for Browns Ferry Nuclear Plant EPU Submittal - SFSP Criticality Safety Analysis, AREVA Inc., February 2016. 25. ANP-3495 Revision 1, Response to RAI for Browns Ferry Nuclear Plant EPU Submittal - SFSP Criticality Safety Analysis, Round 2, AREVA Inc., July 2016. 26. ANP-3160(P) Revision 1, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUMŽ 10XM Fuel, December 2015. 27. ANP-2945(P) Revision 1, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis" July 2011. 28. Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01 Revision 0, Staff Guidance Regarding The Nuclear Safety Analysis for Spent Fuel Pools, Nuclear Regulatory Commission (ADAMS Accession Number ML110620086). 29. ANF-913(P)(A) Volume 1Revision 1 and Volume 1 Supplements 2, 3 and 4, Advanced Nuclear Fuels Corporation, August 1990. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 112 Disposition of additional limitations and conditions related to the final SE for NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains" There are 24 limitations and conditions listed in Section 9 of the Methods LTR SER. The table below lists each of the 24 limitations and conditions and where applicable, identifies which section of the AMSAR discusses compliance with each limitation and condition. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 113 Disposition of additional limitations and conditions related to the final SE for NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains" 9.1 TGBLA/PANAC Version The neutronic methods used to simulate the reactor core response and that feed into the downstream safety analyses supporting operation at EPU/MELLLA+ will apply TGBLA06/PANAC11 or later NRC-approved version of neutronic method. Comply While the identified methodology is specific to GEH/GNF, the intent of this L&C is met. AREVA has used the most current NRC approved neutronics methods: CASMO4/MICROBURN-B2 EMF-2158(P)(A) AREVA will continue to use these methodologies in supporting MELLLA+ EPFOD for BFN. Table 1-1 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 114 9.2 3D Monicore For EPU/MELLLA+ applications, relying on TGBLA04/PANAC10 methods, the bundle RMS difference uncertainty will be established from plant-specific core-tracking data, based on TGBLA04/PANAC10. The use of plant-specific trendline based on the neutronic method employed will capture the actual bundle power uncertainty of the core monitoring system. Comply While the specified Core Monitoring System (CMS) is specific to GEH/GNF, the intent of this L&C is met. TVA uses the POWERPLEX core monitoring system based on the NRC approved CASMO-4/ MICROBURN-B2 methodology. The uncertainties associated with the POWERPLEX CMS remain applicable to MELLLA+ EPFOD operation at BFN and performed by AREVA. Table 1-1 Reference 7 9.3 Power/Flow Ratio Plant-specific EPU and expanded operating domain applications will confirm that the core thermal power to core flow ratio will not exceed 50 MWt/Mlbm/hr at any statepoint in the allowed operating domain. For plants that exceed the power-to-flow value of 50 MWt/Mlbm/hr, the application will provide power distribution assessment to establish that neutronic methods axial and nodal power distribution uncertainties have not increased. Comply Sections 1.2.1 and 2.2.5 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 115 9.4 SLMCPR 1 For EPU operation, a 0.02 value shall be added to the cycle-specific SLMCPR value. This adder is applicable to SLO, which is derived from the dual loop SLMCPR value. N/A AREVA performs statistical analyses with SLO uncertainties to establish the SLO SLMCPR. NRC did not impose an adder during previous reviews and approvals of EPU at Browns Ferry or Monticello. N/A 9.5 SLMCPR 2 For operation at MELLLA+, including operation at the EPU power levels at the achievable core flow statepoint, a 0.03 value shall be added to the cycle-specific SLMCPR value. Comply While there are locations on the power flow map that exceed the 50 MWt/Mlbm/hr and 42 MWt/Mlbm/hr figures of merit applied to previous MELLLA+ LARs, an evaluation has determined that additional adders are not required for the SLMCPR calculated with AREVA methods for BFN operation at MELLLA+ conditions. (Reference 7, Attachment 27 of the MELLLA+ LAR). Sections 2.2.1, 2.2.5, and 9.1.2 Reference 7 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 116 9.6 R-Factor The plant specific R-factor calculation at a bundle level will be consistent with lattice axial void conditions expected for the hot channel operating state. The plant-specific EPU/MELLLA+ application will confirm that the R-factor calculation is consistent with the hot channel axial void conditions. Comply Although R-factors are not used for NRC approved AREVA MCPR correlations, the intent of this L&C is met. The corresponding factors in AREVA methods are K-factors. These factors account for the conditions calculated within the fuel assembly. N/A 9.7 ECCS-LOCA 1 For applications requesting implementation of EPU or expanded operating domains, including MELLLA+, the small and large break ECCS-LOCA analyses will include top-peaked and mid-peaked power shape in establishing the maximum average planar linear heat generation rate (MAPLHGR) and determining the PCT. This limitation is applicable to both the licensing bases PCT and the upper bound PCT. The plant-specific applications will report the limiting small and large break licensing basis and upper bound PCTs. Comply LOCA analyses include top-peaked and mid-peaked power shapes and both large and small breaks. The AREVA LOCA methodology does not include an upper bound PCT. Section 4.3.1 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 117 9.8 ECCS-LOCA 2 The ECCS-LOCA will be performed for all statepoints in the upper boundary of the expanded operating domain, including the minimum core flow statepoints, the transition statepoint, as defined in Reference 1 and the 55 percent core flow statepoint. The plant-specific application will report the limiting ECCS-LOCA results as well as the rated power and flow results. The SRLR will include both the limiting statepoint ECCS-LOCA results and the rated conditions ECCS-LOCA results. Comply AREVA LOCA calculations were performed for the maximum and minimum core flow at rated power and the minimum core flow MELLLA+ boundary. Results for all statepoints are summarized in Reference 19 (Attachment 11 of the MELLLA+ LAR). Section 4.3.1 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 118 9.9 Transient LHGR 1 Plant-specific EPU and MELLLA+ applications will demonstrate and document that during normal operation and core-wide AOOs, the thermal-mechanical (T-M) acceptance criteria as specified in Amendment 22 to GESTAR II will be met. Specifically, during an AOO, the licensing application will demonstrate that the: (1) loss of fuel rod mechanical integrity will not occur due to fuel melting and (2) loss of fuel rod mechanical integrity will not occur due to pellet-cladding mechanical interaction. The plant-specific application will demonstrate that the T-M acceptance criteria are met for the both the UO2 and the limiting GdO2 rods. Comply Power- and flow-dependent multipliers on the LHGR limits are established to ensure that the T-M acceptance criteria (fuel melting and strain) are met during normal operation and AOOs as part of the standard reload process using AREVA's NRC approved methods, Reference 8. Section 9.1.2 9.10 Transient LHGR 2 Each EPU and MELLLA+ fuel reload will document the calculation results of the analyses demonstrating compliance to transient T-M acceptance criteria. The plant T-M response will be provided with the SRLR or COLR, or it will be reported directly to the NRC as an attachment to the SRLR or COLR. Comply T-M analyses will be performed each cycle to establish or confirm the power- and flow-dependent LHGR multipliers. Results for the reference Cycle 19 are provided in Table 5.11 of the RSAR (Attachment 17 of the MELLLA+ LAR). Section 9.1.2 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 119 9.11 Transient LHGR 3 To account for the impact of the void history bias, plant-specific EPU and MELLLA+ applications using either TRACG or ODYN will demonstrate an equivalent to 10 percent margin to the fuel centerline melt and the 1 percent cladding circumferential plastic strain acceptance criteria due to pellet-cladding mechanical interaction for all of limiting AOO transient events, including equipment out-of-service. Limiting transients in this case, refers to transients where the void reactivity coefficient plays a significant role (such as pressurization events). If the void history bias is incorporated into the transient model within the code, then the additional 10 percent margin to the fuel centerline melt and the 1 percent cladding circumferential plastic strain is no longer required. N/A AREVA utilizes the most current NRC approved thermal mechanical methodology, Reference 8, which does not have a void history bias. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 120 9.12 LHGR and Exposure Qualification In MFN 06-481, GE committed to submit plenum fission gas and fuel exposure gamma scans as part of the revision to the T-M licensing process. The conclusions of the plenum fission gas and fuel exposure gamma scans of GE 10x10 fuel designs as operated will be submitted for NRC staff review and approval. This revision will be accomplished through Amendment to GESTAR II or in a T-M licensing LTR. PRIME (a newly developed T-M code) has been submitted to the NRC staff for review (Reference 46). Once the PRIME LTR and its application are approved, future license applications for EPU and MELLLA+ referencing LTR NEDC-33173P must utilize the PRIME T-M methods. N/A This L&C is specific to GEH methods. AREVA uses RODEX4 in the T-M evaluation for BFN. The RODEX4 methods are described in Reference 8. Section 2.2.4 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 121 9.13 Application of 10 Weight Percent Gd Before applying 10 weight percent Gd to licensing applications, including EPU and expanded operating domain, the NRC staff needs to review and approve the T-M LTR demonstrating that the T-M acceptance criteria specified in GESTAR II and Amendment 22 to GESTAR II can be met for steady-state and transient conditions. Specifically, the T-M application must demonstrate that the T-M acceptance criteria can be met for TOP and MOP conditions that bounds the response of plants operating at EPU and expanded operating domains at the most limiting statepoints, considering the operating flexibilities (e.g., equipment out-of-service). Before the use of 10 weight percent Gd for modern fuel designs, NRC must review and approve TGBLA06 qualification submittal. Where a fuel design refers to a design with Gd-bearing rods adjacent to vanished or water rods, the submittal should include specific information regarding acceptance criteria for the qualification and address any downstream impacts in terms of the safety analysis. The 10 weight percent Gd qualifications submittal can supplement this report. N/A AREVA uses the most current NRC approved T-M methods described in Reference 8. RODEX4 is approved to 10.0 wt% gadolinia for solid UO2 fuel pellet (Reference 8). BFN currently typically uses up to 8.0 wt% in ATRIUM 10XM fuel design. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 122 9.14 Part 21 Evaluation of GESTR-M Fuel Temperature Calculation Any conclusions drawn from the NRC staff evaluation of the GE's Part 21 report will be applicable to the GESTR-M T-M assessment of this SE for future license application. GE submitted the T-M Part 21 evaluation, which is currently under NRC staff review. Upon completion of its review, NRC staff will inform GE of its conclusions. N/A The evaluation of the impact of pellet thermal conductivity degradation on AREVA methods is described in Appendix A of Reference 7. (Attachment 27 of the MELLLA+ LAR) N/A 9.15 Void Reactivity 1 The void reactivity coefficient bias and uncertainties in TRACG for EPU and MELLLA+ must be representative of the lattice designs of the fuel loaded in the core. N/A Related information for AREVA methods is presented in Section 3.0 of Reference 7 (Attachment 27 of the MELLLA+ LAR). N/A 9.16 Void Reactivity 2 A supplement to TRACG /PANAC11 for AOO is under NRC staff review (Reference 39). TRACG internally models the response surface for the void coefficient biases and uncertainties for known dependencies due to the relative moderator density and exposure on nodal basis. Therefore, the void history bias determined through the methods review can be incorporated into the response surface "known" bias or through changes in lattice physics/core simulator methods for establishing the instantaneous cross-sections. Including the bias in the calculations negates the need for ensuring that plant-specific applications N/A AREVA has not identified any bias related to void history and has determined that the void coefficient determined by the methodology is accurate and provides the best possible information for the transient analysis. This assessment is provided in Section 3.0 of Reference 7 (Attachment 27 of N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 123 show sufficient margin. For application of TRACG to EPU and MELLLA+ applications, the TRACG methodology must incorporate the void history bias. The manner in which this void history bias is accounted for will be established by the NRC staff SE approving NEDE-32906P, Supplement 3, "Migration to TRACG04/PANAC11 from TRACG02/PANAC10," May 2006 (Reference 39). This limitation applies until the new TRACG/PANAC methodology is approved by the NRC staff. the MELLLA+ LAR). Specifically, AREVA methodology the reactivity coefficients that are used in the transient analysis. Conservatisms in the methodology are used to produce conservative results that bound the uncertainties in the reactivity coefficients. Data indicates that there are no significant differences between MELLLA+ and non-MELLLA+ conditions that have an impact on the reactivity coefficients. The transient results have been demonstrated to be conservative, so there is no penalty needed. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 124 9.17 Steady-State 5 Percent Bypass Voiding The instrumentation specification design bases limit the presence of bypass voiding to 5 percent (LRPM (sic) levels). Limiting the bypass voiding to less than 5 percent for long-term steady operation ensures that instrumentation is operated within the specification. For EPU and MELLLA+ operation, the bypass voiding will be evaluated on a cycle-specific basis to confirm that the void fraction remains below 5 percent at all LPRM levels when operating at steady-state conditions within the MELLLA+ upper boundary. The highest calculated bypass voiding at any LPRM level will be provided with the plant-specific SRLR. Comply Compliance will be included in the plant-specific RSAR as provided for the reference cycle. Sections 2.1.2 and 5.1.5 Reference 7 9.18 Stability Setpoints Adjustment The NRC staff concludes that the presence bypass voiding at the low-flow conditions where instabilities are likely can result in calibration errors of less than 5 percent for OPRM cells and less than 2 percent for APRM signals. These calibration errors must be accounted for while determining the setpoints for any detect and suppress long term methodology. The calibration values for the different long-term solutions are specified in the associated sections of this SE, discussing the stability methodology. N/A for AREVA Limitations associated with DSS-CD are addressed in the GEH M+SAR. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 125 9.19 Void-Quality Correlation 1 For applications involving PANCEA/ODYN/ISCOR/TASC for operation at EPU and MELLLA+, an additional 0.01 will be added to the OLMCPR, until such time that GE expands the experimental database supporting the Findlay-Dix void-quality correlation to demonstrate the accuracy and performance of the void-quality correlation based on experimental data representative of the current fuel designs and operating conditions during steady-state, transient, and accident conditions. Comply While the comment is specific to GE/GEH methodology, the intent of this L&C is met. The void quality correlations used in the AREVA methods are addressed in Section 9.0 and in Reference 7 Appendix A (Attachment 27 of the MELLLA+ LAR). Section 9.0 Appendix A of Reference 7 (Attachment 27 of the MELLLA+ LAR) 9.20 Void-Quality Correlation 2 The NRC staff is currently reviewing Supplement 3 to NEDE-32906P, "Migration to TRACG04/PANAC11 from TRACG02/PANAC10," dated May 2006 (Reference 39). The adequacy of the TRACG interfacial shear model qualification for application to EPU and MELLLA+ will be addressed under this review. Any conclusions specified in the NRC staff SE approving Supplement 3 to LTR NEDC-32906P (Reference 39) will be applicable as approved. N/A N/A for AREVA methods AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 126 9.21 Mixed Core Method 1 Plants implementing EPU or MELLLA+ with mixed fuel vendor cores will provide plant-specific justification for extension of GE's analytical methods or codes. The content of the plant-specific application will cover the topics addressed in this SE as well as subjects relevant to application of GE's methods to legacy fuel. Alternatively, GE may supplement or revise LTR NEDC-33173P (Reference 2) for mixed core application. N/A The initial MELLLA+ BFN operation will contain only fuel supplied by AREVA, primarily the ATRIUM 10XM fuel design. Limitations associated with extending GEH's analytical methods and codes to AREVA fuel are addressed in the GEH M+SAR. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 127 9.22 Mixed Core Method 2 For any plant-specific applications of TGBLA06 with fuel type characteristics not covered in this review, GE needs to provide assessment data similar to that provided for the GE fuels. The Interim Methods review is applicable to all GE lattices up to GE14. Fuel lattice designs, other than GE lattices up to GE14, with the following characteristics are not covered by this review: square internal water channels water crosses Gd rods simultaneously adjacent to water and vanished rods 11x11 lattices MOX fuel The acceptability of the modified epithermal slowing down models in TGBLA06 has not been demonstrated for application to these or other geometries for expanded operating domains. Significant changes in the Gd rod optical thickness will require an evaluation of the TGBLA06 radial flux and Gd depletion modeling before being applied. Increases in the lattice Gd loading that result in nodal reactivity biases beyond those previously established will require review before the GE methods may be applied. N/A Operation at BFN with MELLLA+ will be with cores containing only AREVA fuel designs with the predominant design being ATRIUM 10XM. The continued applicability of the AREVA lattice physics method to MELLLA+ operation is addressed in Reference 7. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 128 9.23 MELLLA+ Eigenvalue Tracking In the first plant-specific implementation of MELLLA+, the cycle-specific eigenvalue tracking data will be evaluated and submitted to NRC to establish the performance of nuclear methods under the operation in the new operating domain. The following data will be analyzed: Hot critical eigenvalue, Cold critical eigenvalue, Nodal power distribution (measured and calculated TIP comparison), Bundle power distribution (measured and calculated TIP comparison), Thermal margin, Core flow and pressure drop uncertainties, and The MIP Criterion (e.g., determine if core and fuel design selected is expected to produce a plant response outside the prior experience base). Provision of evaluation of the core-tracking data will provide the NRC staff with bases to establish if operation at the expanded operating domain indicates: (1) changes in the performance of nuclear methods outside the EPU experience base; (2) changes in the available thermal margins; (3) need for changes in the uncertainties and NRC-approved criterion used in the SLMCPR methodology; or (4) any anomaly that may require corrective actions. Comply After implementation of MELLLA+ at BFN, the requested data applicable to AREVA methods (e.g. not MIP criterion) will be evaluated and provided to the NRC as requested. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 129 9.24 Plant-Specific Application The plant-specific applications will provide prediction of key parameters for cycle exposures for operation at EPU (and MELLLA+ for MELLLA+ applications). The plant-specific prediction of these key parameters will be plotted against the EPU Reference Plant experience base and MELLLA+ operating experience, if available. For evaluation of the margins available in the fuel design limits, plant-specific applications will also provide quarter core map (assuming core symmetry) showing bundle power, bundle operating LHGR, and MCPR for BOC, MOC, and EOC. Since the minimum margins to specific limits may occur at exposures other than the traditional BOC, MOC, and EOC, the data will be provided at these exposures. Comply Section 2.1.2 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 130 Disposition of additional limitations and conditions related to the final SE for NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains" Notes: None. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 131 Disposition of additional limitations and conditions related to the final SE for NEDC-33006P, "Maximum Extended Load Line Limit Analysis Plus" There are 52 limitations and conditions listed in Section 12 of the M+LTR SER. The table below lists each of the 52 limitations and conditions and where applicable, identifies which section of the AMSAR discusses compliance with each limitation and condition. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 132 12.1 GEXL-PLUS The plant-specific application will confirm that for operation within the boundary defined by the MELLLA+ upper boundary and maximum CF range, the GEXL-PLUS experimental database covers the thermal-hydraulic conditions the fuel bundles will experience, including, bundle power, mass flux, void fraction, pressure, and subcooling. If the GEXL-PLUS experimental database does not cover the within bundle thermal-hydraulic conditions, during steady state, transient conditions, and DBA conditions, GHNE will inform the NRC at the time of submittal and obtain the necessary data for the submittal of the plant-specific MELLLA+ application. In addition, the plant-specific application will confirm that the experimental pressure drop database for the pressure drop correlation covers the pressure drops anticipated in the MELLLA+ range. With subsequent fuel designs, the plant-specific applications will confirm that the database supporting the CPR correlations covers the powers, flows and void fractions BWR bundles will experience for operation at and within the MELLLA+ domain, during steady state, transient, and DBA conditions. The plant-specific submittal will also confirm that the NRC staff reviewed and approved the associated CPR correlation if the changes in the correlation are outside the GESTAR II (Amendment 22) Comply While the Limitation and Condition is specific to GE/GEH methodologies, AREVA complies with the intent. The ACE/ATRIUM 10XM critical power correlation has a range of applicability for key inputs. The ranges and conservative actions to be applied should the range be exceeded are presented in ANP-10298PA Revision 0. The conservative actions yield conservative critical power results using the correlation within the appropriate ranges. A discussion on thermal hydraulic comparisons in the extended power flow operating domain is presented in Reference 7. Section 1.1.3 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 133 process. Similarly, the plant-specific application will confirm that the experimental pressure drop database does cover the range of pressures the fuel bundles will experience for operation within the MELLLA+ domain. 12.2 Related LTRs Plant-specific MELLLA+ applications must comply with the limitations and conditions specified in and be consistent with the purpose and content covered in the NRC staff SEs approving the latest version of the following LTRs: NEDC-33173P, NEDC-33075P-A, and NEDC-33147-A. Comply AREVA complied with the limitations and conditions associated with all of the NRC approved AREVA methods that are being applied for BFN as well as the applicable issues from NEDC-33173P and NEDC-33006P as discussed in Appendices A and B of this document. Section 1.1.3 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 134 12.3.a Concurrent Changes The plant-specific analyses supporting MELLLA+ operation will include all operating condition changes that are implemented at the plant at the time of MELLLA+ implementation. Operating condition changes include, but are not limited to, those changes that affect, an increase in the dome pressure, maximum CF, fuel cycle length, or any changes in the licensed operational enhancements. For example, with an increase in dome pressure, the following analyses must be analyzed: the ATWS analysis, the ASME overpressure analyses, the transient analyses, and the ECCS-LOCA analysis. Any changes to the safety system settings or any actuation setpoint changes necessary to operate with the increased dome pressure must be included in the evaluations (e.g., SRV setpoints). Comply Sections 4.3.1, and 9.3.1.1 Tables 4.1 through 4.8 of Reference 19 (Attachment 11 of the MELLLA+ LAR) presents the important plant parameters used in the LOCA analyses. Table 9-6 presents the key plant parameters used in the ATWS overpressure analyses. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 135 12.3.b For all topics in LTR NEDC-33006P that are reduced in scope or generically dispositioned, the plant-specific application will provide justification that the reduced scope or generic disposition is applicable to the plant. If changes that invalidate the LTR dispositions are to be implemented at the time of MELLLA+ implementation, the plant-specific application will provide analyses and evaluations that demonstrate the cumulative effect with MELLLA+ operation. For example, if the dome pressure is increased, the ECCS performance will be evaluated on a plant-specific basis. Comply Sections 4.3.1, and 9.3.1.1 Tables 4.1 through 4.8 of Reference 19 (Attachment 11 of the MELLLA+ LAR) presents the important plant parameters used in the LOCA analyses. Table 9-6 presents the key plant parameters used in the ATWS overpressure analyses. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 136 12.3.c Any generic bounding sensitivity analyses provided in LTR NEDC-33006P will be evaluated to ensure that the key plant-specific input parameters and assumptions are applicable and bounded. If these generic sensitivity analyses are not applicable or additional operating condition changes affect the generic sensitivity analyses, a plant-specific evaluation will be provided. For example, with an increase in the dome pressure, the ATWS sensitivity analyses that model operator actions (e.g., depressurization if the HCTL is reached) needs to be reanalyzed, using the bounding dome pressure condition. Comply Sections 4.3.1, and 9.3.1.1 Tables 4.1 through 4.8 of Reference 19 (Attachment 11 of the MELLLA+ LAR) presents the important plant parameters used in the LOCA analyses. Table 9-6 presents the key plant parameters used in the ATWS overpressure analyses. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 137 12.3.d If a new GE fuel product line or another vendor's fuel is loaded at the plant, the applicability of any generic sensitivity analyses supporting the MELLLA+ application shall be justified in the plant-specific application. If the generic sensitivity analyses cannot be demonstrated to be applicable, the analyses will be performed including the new fuel. For example, the ATWS instability analyses supporting the MELLLA+ condition are based on the GE14 fuel response. New analyses that demonstrate the ATWS instability performance of the new GE fuel or another vendor's fuel for MELLLA+ operation shall be provided to support the plant-specific application. Comply All fuel-related events have either been dispositioned as non-limiting, or analyzed to support MELLLA+ operation with ATRIUM 10XM fuel. Limitations associated with ATWSi are addressed in the GEH M+SAR. Sections 3.1.2, 4.3.1, 9.1.1, and 9.3.1 12.3.e If a new GE fuel product line or another vendor's fuel is loaded at the plant prior to a MELLLA+ application, the analyses supporting the plant-specific MELLLA+ application will be based on a specific core configuration or bounding core conditions. Any topics that are generically dispositioned or reduced in scope in LTR NEDC-33006P will be demonstrated to be applicable, or new analyses based on the specific core configuration or bounding core conditions will be provided. Comply Section 2.1.1 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 138 12.3.f If a new GE fuel product line or another vendor's fuel is loaded at the plant prior to a MELLLA+ application, the plant-specific application will reference an NRC-approved stability method supporting MELLLA+ operation, or provide sufficient plant-specific information to allow the NRC staff to review and approve the stability method supporting MELLLA+ operation. The plant-specific application will demonstrate that the analyses and evaluations supporting the stability method are applicable to the fuel loaded in the core. Limitations associated with DSS-CD are addressed in the GEH M+SAR. N/A 12.3.g For MELLLA+ operation, core instability is possible in the event a transient or plant maneuver places the reactor at a high power/low-flow condition. Therefore, plants operating at MELLLA+ conditions must have a NRC-approved instability protection method. In the event the instability protection method is inoperable, the applicant must employ an NRC-approved backup instability method. The licensee will provide technical specification (TS) changes that specify the instability method operability requirements for MELLLA+ operation, including any backup stability protection methods. Limitations associated with DSS-CD are addressed in the GEH M+SAR. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 139 12.4 Reload analysis submittal The plant-specific MELLLA+ application shall provide the plant-specific thermal limits assessment and transient analysis results. Considering the timing requirements to support the reload, the fuel and cycle-dependent analyses including the plant-specific thermal limits assessment may be submitted by supplementing the initial M+SAR. Additionally, the SRLR for the initial MELLLA+ implementation cycle shall be submitted for NRC staff confirmation. Comply AREVA RSAR for SRLR equivalent. Attachment 17 of the MELLLA+ LAR provides AREVA's Reload Analysis report created to present licensing results of the MELLLA+ demonstration cycle. Sections 9.1.1 and 9.1.2 12.5.a Operating Flexibility The licensee will amend the TS LCO for any equipment out-of-service (i.e., SLO) or operating flexibilities prohibited in the plant-specific MELLLA+ application. N/A for AREVA TVA has proposed License Conditions and Technical Specification changes for equipment-out-of-service and operating flexibilities that are prohibited in the BFN MELLLA+ domain (Attachments 1 and 2 of the BFN MELLLA+ LAR). N/A 12.5.b For an operating flexibility, such as FWHOOS, that is prohibited in the MELLLA+ plant-specific application but is not included in the TS LCO, the licensee will propose and implement a license condition. N/A for AREVA TVA has proposed License Conditions and Technical Specification changes for equipment-out-of-service and operating flexibilities that are prohibited in the BFN MELLLA+ domain (Attachments 1 and 2 of the BFN MELLLA+ LAR). N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 140 12.5.c The power flow map is not specified in the TS; however, it is an important licensed operating domain. Licensees may elect to be licensed and operate the plant under plant-specific-expanded domain that is bounded by the MELLLA+ upper boundary. Plant-specific applications approved for operation within the MELLLA+ domain will include the plant-specific power/flow map specifying the licensed domain in the COLR. Comply TVA will include the power/flow map in the COLR. Section 1.2.1 12.6 SLMCPR Statepoints and CF Uncertainty Until such time when the SLMCPR methodology (References 47 and 48) for off-rated SLMCPR calculation is approved by the staff for MELLLA+ operation, the SLMCPR will be calculated at the rated statepoint (120 percent P/100 percent CF), the plant-specific minimum CF statepoint (e.g., 120 percent P/80 percent CF), and at the 100 percent OLTP at 55 percent CF statepoint. The currently approved off-rated CF uncertainty will be used for the minimum CF and 55 percent CF statepoints. The uncertainty must be consistent with the CF uncertainty currently applied to the SLO operation or as NRC-approved for MELLLA+ operation. The calculated values will be documented in the SRLR. Comply Section 2.2.1 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 141 12.7 Stability Manual operator actions are not adequate to control the consequences of instabilities when operating in the MELLLA+ domain. If the primary stability protection system is declared inoperable, a non-manual NRC-approved backup protection system must be provided, or the reactor core must be operated below a NRC-approved backup stability boundary specifically approved for MELLLA+ operation for the stability option employed. Limitations associated with DSS-CD are addressed in the GEH M+SAR. N/A 12.8 Fluence Methodology and Fracture Toughness The applicant is to provide a plant-specific evaluation of the MELLLA+ RPV fluence using the most up-to-date NRC-approved fluence methodology. This fluence will then be used to provide a plant-specific evaluation of the RPV fracture toughness in accordance with RG 1.99, Revision 2. Limitations associated with fluence methodology and fracture toughness are addressed in the GEH M+SAR. N/A 12.9 Reactor Coolant Pressure Boundary MELLLA+ applicants must identify all other than Category "A" materials, as defined in NUREG-0313, Revision 2, that exist in its RCPB piping, and discuss the adequacy of the augmented inspection programs in light of the MELLLA+ operation on a plant-specific basis. Limitations associated with reactor coolant pressure boundary piping are addressed in the GEH M+SAR. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 142 12.10.a ECCS-LOCA Off-rated Multiplier The plant-specific application will provide the 10 CFR Part 50, Appendix K, and the nominal PCTs calculated at the rated EPU power/rated CF, rated EPU power/minimum CF, at the low-flow MELLLA+ boundary (Transition Statepoint). For the limiting statepoint, both the upper bound and the licensing PCT will be reported. The M+SAR will justify why the transition statepoint ECCS-LOCA response bounds the 55 percent CF statepoint. The M+SAR will provide discussion on what power/flow combination scoping calculations were performed to identify the limiting statepoints in terms of DBA-LOCA PCT response for the operation within the MELLLA+ boundary. The M+SAR will justify that the upper bound and licensing basis PCT provided is in fact the limiting PCT considering uncertainty applications to the non-limiting statepoints. Comply AREVA methods do not include a calculation of upper bound PCT. Section 4.3.1 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 143 12.10.b LOCA analysis is not performed on cycle-specific basis; therefore, the thermal limits applied in the M+SAR LOCA analysis for the 55 percent CF MELLLA+ statepoint and/or the transition statepoint must be either bounding or consistent with cycle-specific off-rated limits. The COLR and the SRLR will contain confirmation that the off-rated limits assumed in the ECCS-LOCA analyses bound the cycle-specific off-rated limits calculated for the MELLLA+ operation. Every future cycle reload shall confirm that the cycle-specific off-rated thermal limits applied at the 55 percent CF and/or the transition statepoints are consistent with those assumed in the plant-specific ECCS-LOCA analyses. Comply Section 4.3.1 12.10.c Off-rated limits will not be applied to the minimum CF statepoint. Comply Section 4.3.1 (Note 2) 12.10.d If credit is taken for these off-rated limits, the plant will be required to apply these limits during core monitoring. N/A Section 4.3.1 (Note 2) AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 144 12.11 ECCS-LOCA Axial Power Distribution Evaluation For MELLLA+ applications, the small and large break ECCS-LOCA analyses will include top-peaked and mid-peaked power shape in establishing the MAPLHGR and determining the PCT. This limitation is applicable to both the licensing bases PCT and the upper bound PCT. The plant-specific applications will report the limiting small and large break licensing basis and upper bound PCTs. Comply AREVA methods do not include a calculation of upper bound PCT. Section 4.3.1 12.12.a ECCS-LOCA Reporting Both the nominal and Appendix K PCTs should be reported for all of the calculated statepoints, and Comply In accordance with approved AREVA methods, AREVA only calculates and reports Appendix K PCTs. Section 4.3.1 (Note 1) 12.12.b The plant-variable and uncertainties currently applied will be used, unless the NRC staff specifically approves a different plant variable uncertainty method for application to the non-rated statepoints. Comply Section 4.3.1 Tables 4.1 through 4.8 of Reference 19 (Attachment 11 of the MELLLA+ LAR) presents the important plant parameters used in the LOCA analyses AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 145 12.13 Small Break LOCA Small break LOCA analysis will be performed at the MELLLA+ minimum CF and the transition statepoints for those plants that: (1) are small break LOCA limited based on small break LOCA analysis performed at the rated EPU conditions; or (2) have margins of less than or equal to l l l l l l l relative to the Appendix K or the licensing basis PCT. Comply Section 4.3.1 12.14 Break Spectrum The scope of small break LOCA analysis for MELLLA+ operation relies upon the EPU small break LOCA analysis results. Therefore, the NRC staff concludes that for plants that will implement MELLLA+, sufficient small break sizes should be analyzed at the rated EPU power level to ensure that the peak PCT break size is identified. Comply Section 4.3.1 12.15 Bypass Voiding Above the D-level Plant-specific MELLLA+ applications shall identify where in the MELLLA+ upper boundary the bypass voiding greater than 5 percent will occur above the D-level. The licensee shall provide in the plant-specific submittal the operator actions and procedures that will mitigate the impact of the bypass voiding on the TIPs and the core simulator used to monitor the fuel performance. The plant-specific submittal shall also provide discussion on what impact the bypass voiding greater than 5 percent will have on the NMS as defined in Section 5.1.1.5. The NRC staff will evaluate on plant-specific bases acceptability of bypass voiding above D level. Comply Section 5.1.5 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 146 12.16 RWE Plants operating at the MELLLA+ operating domain shall perform RWE analyses to confirm the adequacy of the generic RBM setpoints. The M+SAR shall provide a discussion of the analyses performed and the results. Comply AREVA analyses do not utilize generic RBM setpoints. Instead a CRWE analysis is performed each cycle using the BFN specific RBM setpoints. The cycle specific analysis includes the impact of operation in the MELLLA+ operating domain. Results for the reference Cycle 19 are provided in Section 5.1.5 of the RSAR (Attachment 17 of the MELLLA+ LAR). Section 5.3.2 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 147 12.17 ATWS LOOP As specified in LTR NEDC-33006P, at least two plant-specific ATWS calculations must be performed: MSIVC and PRFO. In addition, if RHR capability is affected by LOOP, then a third plant-specific ATWS calculation must be performed that includes the reduced RHR capability. To evaluate the effect of reduced RHR capacity during LOOP, the plant-specific ATWS calculation must be performed for a sufficiently large period of time after HSBW injection is complete to guarantee that the suppression pool temperature is cooling, indicating that the RHR capacity is greater than the decay heat generation. The plant-specific application should include evaluation of the safety system performance during the long-term cooling phase, in terms of available NPSH. Comply (initial overpressure) N/A (long term response) Limitations associated with ATWS long term response are addressed in the GEH M+SAR. Section 9.3.1.1 12.18.a ATWS TRACG Analysis For plants that do not achieve hot shutdown prior to reaching the heat capacity temperature limit (HCTL) based on the licensing ODYN code calculation, plant-specific MELLLA+ implementations must perform best-estimate TRACG calculations on a plant-specific basis. The TRACG analysis will account for all plant parameters, including water-level control strategy and all plant-specific emergency operating procedure (EOP) actions. N/A for AREVA Limitations associated with ATWS long term response are addressed in the GEH M+SAR. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 148 12.18.b The TRACG calculation is not required if the plant increases the boron-10 concentration/enrichment so that the integrated heat load to containment calculated by the licensing ODYN calculation does not change with respect to a reference OLTP/75 percent flow ODYN calculation. N/A for AREVA Limitations associated with ATWS long term response are addressed in the GEH M+SAR. N/A 12.18.c Peak cladding temperature (PCT) for both phases of the transient (initial overpressure and emergency depressurization) must be evaluated on a plant-specific basis with the TRACG ATWS calculation. Comply (Initial Overpressure) The ATWS event PCT is qualitatively evaluated to demonstrate compliance with the acceptance criteria of 10 CFR 50.46. N/A (Emergency Depressurization) Section 9.3.1.2 AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 149 12.18.d In general, the plant-specific application will ensure that operation in the MELLLA+ domain is consistent with the assumptions used in the ATWS analysis, including equipment out of service (e.g., FWHOOS, SLO, SRVs, SLC pumps, and RHR pumps, etc.). If assumptions are not satisfied, operation in MELLLA+ is not allowed. The SRLR will specify the prohibited flexibility options for plant-specific MELLLA+ operation, where applicable. For key input parameters, systems and engineering safety features that are important to simulating the ATWS analysis and are specified in the Technical Specification (TS) (e.g., SLCS parameters, ATWS RPT, etc.), the calculation assumptions must be consistent with the allowed TS values and the allowed plant configuration. If the analyses deviate from the allowed TS configuration for long term equipment out of service (i.e., beyond the TS LCO), the plant-specific application will specify and justify the deviation. In addition, the licensee must ensure that all operability requirements are met (e.g., NPSH) by equipment assumed operable in the calculations. Comply (initial overpressure) N/A (long term response) Limitations associated with ATWS long term response are addressed in the GEH M+SAR. Section 9.3.1.1 Table 9-6 presents the key plant parameters used in the ATWS overpressure analyses. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 150 12.18.e Nominal input parameters can be used in the ATWS analyses provided the uncertainty treatment and selection of the values of these input parameters are consistent with the input methods used in the original GE ATWS analyses in NEDE-24222. Treatment of key input parameters in terms of uncertainties applied or plant-specific TS value used can differ from the original NEDE-24222 approach, provided the manner in which it is used yields more conservative ATWS results. Comply (initial overpressure) N/A (long term response) Limitations associated with ATWS long term response are addressed in the GEH M+SAR. Section 9.3.1 Table 9-6 presents the key plant parameters used in the overpressure analyses. 12.18.f The plant-specific application will include tabulation and discussion of the key input parameters and the associated uncertainty treatment. Comply (initial overpressure) N/A (long term response) Limitations associated with ATWS long term response are addressed in the GEH M+SAR. Section 9.3.1 Table 9-6 presents the key plant parameters used in the ATWS overpressure analyses. AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 151 12.19 Plant-Specific ATWS Instability Until such time that NRC approves a generic solution for ATWS instability calculations for MELLLA+ operation, each plant-specific MELLLA+ application must provide ATWS instability analysis that satisfies the ATWS acceptance criteria listed in SRP Section 15.8. The plant-specific ATWS instability calculation must: (1) be based on the peak-reactivity exposure conditions, (2) model the plant-specific configuration important to ATWS instability response including mixed core, if applicable, and (3) use the regional-mode nodalization scheme. In order to improve the fidelity of the analyses, the plant-specific calculations should be based on latest NRC-approved neutronic and thermal-hydraulic codes such as TGBLA06/PANAC11 and TRACG04. Limitations associated with ATWSi are addressed in the GEH M+SAR. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 152 12.20 Generic ATWS Instability Once the generic solution is approved, the plant-specific applications must provide confirmation that the generic instability analyses are relevant and applicable to their plant. Applicability confirmation includes review of any differences in plant design or operation that will result in significantly lower stability margins during ATWS such as: turbine bypass capacity, fraction of steam-driven feedwater pumps, any changes in plant design or operation that will significantly increase core inlet subcooling during ATWS events, significant differences in radial and axial power distributions, hot-channel power-to-flow ratio, fuel design changes beyond GE14. Limitations associated with ATWSi are addressed in the GEH M+SAR. N/A 12.21 Individual Plant Evaluation Licensees that submit a MELLLA+ application should address the plant-specific risk impacts associated with MELLLA+ implementation, consistent with approved guidance documents (e.g., NEDC-32424P-A, NEDC-32523P-A, and NEDC-33004P-A) and the Matrix 13 of RS-001 and re-address the plant-specific risk impacts consistent with the approved guidance documents that were used in their approved EPU application and Matrix 13 of RS-001. If an EPU and MELLLA+ application come to the NRC in parallel, the expectation is that the EPU submittal will have incorporated the MELLLA+ impacts. Limitations associated with the plant-specific risk impacts are addressed in the GEH M+SAR. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 153 12.22 IASCC The applicant is to provide a plant-specific IASCC evaluation when implementing MELLLA+, which includes the components that will exceed the IASCC threshold of 5x1020 n/cm2 (E>1MeV), the impact of failure of these components on the integrity of the reactor internals and core support structures under licensing design bases conditions, and the inspections that will be performed on components that exceed the IASCC threshold to ensure timely identification of IASCC, should it occur. Limitations associated with the plant-specific IASCC evaluation are addressed in the GEH M+SAR. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 154 12.23.1 Limitations from the ATWS RAI Evaluations See limitation 12.18.d. Comply (initial overpressure) N/A (long term response) Limitations associated with ATWS long term response are addressed in the GEH M+SAR. Section 9.3.1.1 Table 9-6 presents the key plant parameters used in the overpressure analyses 12.23.2 The plant-specific ODYN and TRACG key calculation parameters must be provided to the staff so they can verify that all plant-specific automatic settings are modeled properly. Comply (initial overpressure) N/A (long term response) Limitations associated with ATWS long term response are addressed in the GEH M+SAR. Sections 1.1.3 and 9.3.1 Table 9-6 presents the key plant parameters used in the ATWS overpressure analyses AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 155 12.23.3 The ATWS peak pressure response would be dependent upon SRVs upper tolerances assumed in the calculations. For each individual SRV, the tolerances used in the analysis must be consistent with or bound the plant-specific SRV performance. The SRV tolerance test data would be statistically treated using the NRC's historical 95/95 approach or any new NRC-approved statistical treatment method. In the event that current EPU experience base shows propensity for valve drift higher than pre-EPU experience base, the plant-specific transient and ATWS analyses would be based on the higher tolerances or justify the reason why the propensity for the higher drift is not applicable the plant's SRVs. Comply Sections 3.1.2 and 9.3.1.1 12.23.4 EPG/SAG parameters must be reviewed for applicability to MELLLA+ operation in a plant-specific basis. The plant-specific MELLLA+ application will include a section that discusses the plant-specific EOPs and confirms that the ATWS calculation is consistent with the operator actions. N/A for AREVA Limitations associated with ATWS long term response are addressed in the GEH M+SAR. N/A 12.23.5 The conclusions of this LTR and associated SE are limited to reactors operating with a power density lower than 52.5 MW/MLBM/hr for operation at the minimum allowable CF at 120 percent OLTP. Verification that reactor operation will be maintained below this analysis limit must be performed for all plant-specific applications. Limitations associated with ATWSi are addressed in the GEH M+SAR. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 156 12.23.6 For MELLLA+ applications involving GE fuel types beyond GE14 or other vendor fuels, bounding ATWS Instability analysis will be provided to the staff. Note: this limitation does not apply to special test assemblies. Limitations associated with ATWSi are addressed in the GEH M+SAR. N/A 12.23.7 See limitation 12.23.6. Limitations associated with ATWSi are addressed in the GEH M+SAR. N/A 12.23.8 The plant-specific ATWS calculations must account for all plant- and fuel-design-specific features, such as the debris filters. Comply Section 9.3.1 12.23.9 Plant-specific applications must review the safety system specifications to ensure that all of the assumptions used for the ATWS SE indeed apply to their plant-specific conditions. The NRC staff review will give special attention to crucial safety systems like HPCI, and physical limitations like NPSH and maximum vessel pressure that RCIC and HPCI can inject. The plant-specific application will include a discussion on the licensing bases of the plant in terms of NPSH and system performance. It will also include NPSH and system performance evaluation for the duration of the event. Comply (initial overpressure) N/A (long term response) Limitations associated with ATWS long term response are addressed in the GEH M+SAR. Section 9.3.1.1 12.23.10 Plant-specific applications must ensure that an increase in containment pressure resulting from ATWS events with EPU/MELLLA+ operation does not affect adversely the operation of safety-grade equipment. N/A for AREVA Limitations associated with ATWS long term response are addressed in the GEH M+SAR. N/A AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 157 12.23.11 The plant-specific applications must justify the use of plant-specific suppression pool temperature limits for the ODYN and TRACG calculations that are higher than the HCTL limit for emergency depressurization. N/A for AREVA Limitations associated with ATWS long term response are addressed in the GEH M+SAR. N/A 12.24.1 Limitations from Fuel Dependent Analyses RAI Evaluations For EPU/MELLLA+ plant-specific applications that use TRACG or any code that has the capability to model in-channel water rod flow, the supporting analysis will use the actual flow configuration. Limitations associated with the modeling of in-channel water rod flow are addressed in the GEH M+SAR for GE analyses. For AREVA analyses the water channel flow for AREVA ATRIUM 10XM fuel will be modeled with AREVA codes that have that capability. N/A 12.24.2 The EPU/MELLLA+ application would provide the exit void fraction of the high-powered bundles in the comparison between the EPU/MELLLA+ and the pre-MELLLA+ conditions. Comply Section 2.1.2 12.24.3 See limitation 12.6. Comply Section 2.2.1 12.24.4 See limitation 12.18.d. Comply (initial overpressure) N/A (long term response) Limitations associated with ATWS long term response are addressed in the GEH M+SAR. Section 9.3.1.1 Table 9-6 presents the key plant parameters used in the overpressure analyses AREVA Inc. AREVA MELLLA+ Safety Analysis Report for Browns Ferry Units 1, 2, and 3 ANP-3551NP Revision 0 Page 158 Disposition of additional limitations and conditions related to the final SE for NEDC-33006P, "Maximum Extended Load Line Limit Analysis Plus" Notes: 1. AREVA provides LOCA-ECCS analyses with acceptance criteria based on 10 CFR 50.46 in accordance with approved methodologies contained in 10 CFR 50 Appendix K to determine Peak Cladding Temperatures for the full spectrum of LOCA break sizes. 2. In the LOCA-ECCS analyses, Therefore, the ATRIUM 10XM LOCA-ECCS analysis complies with Limitation and Conditions 12.10.c and 12.10.d.

0414-12-F04 (Rev. 001, 03/10/2016) ANP-3544NP Revision 0 December 2016 AREVA Inc. PROPRIETARY - COMMERCIAL © 2016 AREVA Inc. ANP-3544NP Revision 0 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NPRevision 0Page i AREVA Inc. Item Section(s) or Page(s) Description and Justification 1 All This is a new document.

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NPRevision 0Page ii AREVA Inc. 1.0 Introduction ........................................................................................................ 1-1 2.0 Summary ........................................................................................................... 2-1 3.0 Equilibrium Fuel Cycle Design ........................................................................... 3-1 3.1 General Description ................................................................................ 3-1 3.2 Control Rod Patterns and Thermal Limits ............................................... 3-1 3.3 Hot Excess Reactivity and Cold Shutdown Margin ................................. 3-1 4.0 References ........................................................................................................ 4-1 Appendix A Browns Ferry Equilibrium Cycle Step-through Depletion Summary, Control Rod Patterns and Core Average Axial Power and Exposure Distributions ................................................................. A-1 Appendix B Elevation Views of the Browns Ferry Equilibrium Cycle Fresh Reload Batch Fuel Assemblies ........................................................... B-1 Appendix C Browns Ferry Equilibrium Cycle Fresh Fuel Locations ...................... C-1 Appendix D Browns Ferry Equilibrium Cycle Radial Exposure and Power Distributions ....................................................................................... D-1 Appendix E Browns Ferry Cycle N-1 EOC Projection Control Rod Patterns and Core Average Axial Power and Exposure Distributions ............... E-1 Table 2.1 Browns Ferry Equilibrium Cycle Energy and Key Results Summary ........... 2-2 Table 3.1 Equilibrium Cycle Core Composition and Design Parameters ..................... 3-3 Table 3.2 Browns Ferry Equilibrium Cycle Hot Operating Target k-eff Versus Cycle Exposure ..................................................................................................... 3-4 Table 3.3 Browns Ferry Equilibrium Cycle Cold Critical Target k-eff Versus Cycle Exposure ..................................................................................................... 3-4 Table 3.4 Browns Ferry Equilibrium Cycle Reactivity Margin Summary (Short EOC N-1) ............................................................................................................. 3-5 Table 3.5 Browns Ferry Cycle Equilibrium Cycle Reactivity Margin Summary (Nominal EOC N-1) ..................................................................................... 3-6 Table 3.6 Browns Ferry Equilibrium Cycle Reactivity Margin Summary (Long EOC N-1) ............................................................................................................. 3-7 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NPRevision 0Page iii AREVA Inc. Figure 2.1 Browns Ferry Equilibrium Cycle Design Step-through k-eff versus Cycle Exposure ......................................................................................... 2-4 Figure 2.2 Browns Ferry Equilibrium Cycle Design Margin to Thermal Limits versus Cycle Exposure .............................................................................. 2-4 Figure 3.1 Browns Ferry Equilibrium Cycle Upper Left Quarter Core Layout by Fuel Type .................................................................................................. 3-8 Figure 3.2 Browns Ferry Equilibrium Cycle Upper Right Quarter Core Layout by Fuel Type .................................................................................................. 3-9 Figure 3.3 Browns Ferry Equilibrium Cycle Lower Left Quarter Core Layout by Fuel Type ................................................................................................ 3-10 Figure 3.4 Browns Ferry Equilibrium Cycle Lower Right Quarter Core Layout by Fuel Type ................................................................................................ 3-11 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NPRevision 0Page iv AREVA Inc. ACE AREVA's advanced critical power correlation BLEU blended low enriched uranium BOC beginning of cycle BOL beginning of life BWR boiling water reactor CGU commercial grade uranium CSDM cold shutdown margin EOC end of cycle EOFP end of full power capability EPU extended power uprate FFTR final feedwater temperature reduction GWd gigawatt days GWd/MTU gigawatt days per metric ton of initial uranium HEXR hot excess reactivity LHGR linear heat generation rate MCPR minimum critical power ratio MELLLA maximum extended load line limit analysis MELLLA+ maximum extended load line limit analysis plus MICROBURN-B2 AREVA Inc. advanced BWR core simulator methodology with PPR capability MWd/MTU megawatt days per metric ton of initial uranium NRC Nuclear Regulatory Commission, U. S. PLFR part length fuel rod PPR Pin Power Reconstruction. The PPR methodology accounts for variation in local rod power distributions due to neighboring assemblies and control state. The local rod power distributions are reconstructed based on the actual flux solution for each statepoint. R Value the larger of zero or the shutdown margin at BOC minus the minimum calculated shutdown margin in the cycle SLC standby liquid control

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 1-1 AREVA Inc. AREVA Inc. (AREVA) has performed an equilibrium fuel cycle design and fuel management calculations for the Browns Ferry BWRs with ATRIUMŽ* 10XM fuel at EPU (120% OLTP) conditions with MELLLA+. These analyses have been performed with the approved AREVA neutronics methodology (Reference 1). The CASMO-4 lattice depletion code was used to generate nuclear data including cross sections and local power peaking factors. The MICROBURN-B2 three dimensional core simulator code, utilizes the pin power reconstruction (PPR) model to determine the thermal margins presented in this report. The ACE critical power correlation (References 2 and 5) was utilized for all assemblies in the core. The following MICROBURN-B2 modeling features are included in this analysis: Control blade B-10 depletion Explicit neutronic treatment of the spacer grids Explicit modeling of the PLFR plenums Explicit modeling of the water rod flow Design results for the equilibrium cycle reactor core loading including projected control rod patterns and evaluations of thermal and reactivity margins are presented. The equilibrium cycle results are summarized in Table 2.1.

• ATRIUM is a trademark of AREVA Inc.

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 2-1 AREVA Inc. The equilibrium cycle fresh batch size (332 assemblies) and batch average enrichment were determined to meet the energy requirements provided by Tennessee Valley Authority (TVA) (Reference 4). For a complete description of the fresh reload assemblies, see Reference 3. The loading of the equilibrium cycle fuel as described in this report results in a projected equilibrium cycle full power energy capability of 2,563+/-42 GWd (18,525+/-300 MWd/MTU). As expected, this EOFP energy is larger (due to taking advantage of the MELLLA+ flow window) than that achieved with the equivalent EPU equilibrium cycle with MELLLA in Reference 6. Beyond the full power capability, the cycle is depleted with 178 GWd additional energy via FFTR and coast operation in order to reach the same EOC energy that was achieved with the EPU equilibrium cycle with MELLLA in Reference 6. In order to obtain optimum operating flexibility, the projected control rod patterns for the equilibrium cycle were developed to be consistent with a conservative margin to thermal limits. The cycle design calculations also demonstrate adequate hot excess reactivity and cold shutdown margin throughout the cycle. The hot and cold target k-effs used in the design are provided in Table 3.2 and 3.3. Key results from the design analysis are summarized in Table 2.1. Figures 2.1 and 2.2 provide a summary of the cycle design step-through projection. The peak rod average power was verified to not exceed 6.3 kW/ft at exposures of > 54 and 62 GWd/MTU (Regulatory Guide 1.183 requirement). Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 2-2 AREVA Inc. Cycle Energy, GWd (Cycle Exposure, MWd/MTU) Cycle N-1

• Best estimate depletion to Nominal EOC N-1 2,741 (19,811)
• Short window EOC N-1 2,658 (19,211)* Long window EOC N-1 2,796 (20,211)Cycle N
• EOFP Energy 2,563+/-42 (18,525+/-300)* FFTR and coast Energy 178 (1,286)* EOC Energy 2,741+/-42 (19,811+/-300) Key Results BOC CSDM, %k/k (based on short EOC N-1) 1.21Minimum CSDM, %k/k (based on short EOC N-1) 1.21Cycle Exposure of Minimum CSDM, MWd/MTU (short basis) 0Moderator Temperature of Minimum CSDM, °F (short basis) 68Cycle R Value, %k/k (short basis) 0.00BOC CSDM, %k/k (based on nominal EOC N-1) 1.51Minimum CSDM, %k/k (based on nominal EOC N-1) 1.51Cycle Exposure of Minimum CSDM, MWd/MTU (nominal basis) 0Moderator Temperature of Minimum CSDM, °F (nominal basis) 68Cycle R Value, %k/k (nominal basis) 0.00BOC CSDM, %k/k (based on long EOC N-1) 1.75Minimum CSDM, %k/k (based on long EOC N-1) 1.75Cycle Exposure of Minimum CSDM, MWd/MTU (long basis) 0Moderator Temperature of Minimum CSDM, °F (long basis) 68Cycle R Value, %k/k (long basis) 0.00Minimum SLC SDM, %k/k (based on short EOC N-1) 2.92Cycle Exposure of Minimum SLC SDM, MWd/MTU (short basis) 0 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 2-3 AREVA Inc. Key Results Minimum SLC SDM, %k/k (based on nominal EOC N-1) 3.27Cycle Exposure of Minimum SLC SDM, MWd/MTU (nominal basis) 0Minimum SLC SDM, %k/k (based on long EOC N-1) 3.51Cycle Exposure of Minimum SLC SDM, MWd/MTU (long basis) 0BOC HEXR, %k/k (based on short EOC N-1) 1.80Maximum HEXR, %k/k (based on short EOC N-1) 1.81Cycle Exposure of Maximum HEXR, MWd/MTU (short basis) 14,000BOC HEXR, %k/k (based on nominal EOC N-1) 1.48Maximum HEXR, %k/k (based on nominal EOC N-1) 1.49Cycle Exposure of Maximum HEXR, MWd/MTU (nominal basis) 14,000BOC HEXR, %k/k (based on long EOC N-1) 1.26Maximum HEXR, %k/k (based on long EOC N-1) 1.28Cycle Exposure of Maximum HEXR, MWd/MTU (long basis) 14,000Minimum MAPLHGR Margin, % 16.0Exposure of Minimum MAPLHGR Margin, MWd/MTU 18,100Minimum LHGR Margin, % 15.1Exposure of Minimum LHGR Margin, MWd/MTU 17,875Minimum CPR Margin, % 5.6Exposure of Minimum CPR Margin, MWd/MTU 17,591 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 2-4 AREVA Inc. 0.9900.995 1.000 1.005 1.0100246810121416182022TargetCycle 21 0.60.7 0.8 0.9 1.00246810121416182022MCPRLHGRAPLHGR Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 3-1 AREVA Inc. 3.1 GENERAL DESCRIPTION The assembly design for the equilibrium cycle fresh reload fuel for Browns Ferry is described in detail in Reference 3. Elevation views of the fresh reload fuel design axial enrichment and gadolinia distributions are shown in Appendix B, Figures B.1 through B.3. The loading pattern maintains full core symmetry within a scatter load fuel management scheme. This loading in conjunction with the control rod patterns presented in Appendix A shows acceptable power peaking and associated margins to limits for projected equilibrium cycle operation. The analyses supporting this fuel cycle design were based on the core parameters shown in Table 3.1. Figures 3.1 through 3.4, along with Table 3.1 define the reference loading pattern used in the fuel cycle design. The specific core location of the fresh assemblies in the equilibrium cycle is provided in Appendix C. Key results for the cycle are summarized in Table 2.1. The equilibrium cycle assumes the use of BLEU material for one fuel type to account for about 30%

of the fresh reload assemblies. This was done to account for the possibility that limited supplies of BLEU material may become available after the current inventory is exhausted. 3.2 CONTROL ROD PATTERNS AND THERMAL LIMITS Projected control rod patterns for the equilibrium cycle and resultant key operating parameters including thermal margins are shown in Appendix A. The thermal margins presented in this report were determined using the MICROBURN-B2 3D core simulator PPR model and meet the required design margin to thermal limits. A detailed summary of the core parameters resulting from the step-through projection analysis is provided in Tables A.1 and A.2. Limiting results from the step-through are summarized in Table 2.1 and in Figure 2.2. The hot operating target k-eff versus cycle exposure which was determined to be appropriate for the equilibrium cycle is shown in Table 3.2. The k-eff and margin to limits results from the design cycle depletion are presented graphically in Figures 2.1 and 2.2. The k-eff values presented in Figure 2.1 and in Appendix A are not bias corrected. Selected exposure and radial power distributions from the design step-through are presented in Appendix D. 3.3 HOT EXCESS REACTIVITY AND COLD SHUTDOWN MARGIN The equilibrium cycle design calculations demonstrate adequate hot excess reactivity, SLC shutdown margin, and cold shutdown margin throughout the cycle. Key shutdown margin and R-Value results are presented in Table 2.1. The shutdown margin for the equilibrium cycle is in Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 3-2 AREVA Inc. conformance with the Technical Specification limit of R + 0.38 %k/k at BOC. The cold target k-eff versus exposure determined to be appropriate for calculation of cold shutdown margin for the equilibrium cycle is shown in Table 3.3. The core hot excess reactivity was calculated at full power with all rods out, 102.5 Mlb/hr core flow, with equilibrium xenon. Tables 3.4 through 3.6 summarize the equilibrium cycle reactivity margins versus cycle exposure, including the SLC shutdown margin for the cycle. Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 3-3 AREVA Inc. Fuel Description Cycle Loaded Nuclear Fuel Type Number of Assemblies ATRIUM 10XM N-2 13 16 ATRIUM 10XM N-2 14 76 ATRIUM 10XM N-2 15 8 ATRIUM 10XM N-1 16 208 ATRIUM 10XM N-1 17 92 ATRIUM 10XM N-1 18 32 ATRIUM 10XM N 19 208 ATRIUM 10XM N 20 92 ATRIUM 10XM N 21 32 Number of Fuel Assemblies in Core 764 Total Number of Fresh Assemblies 332 Total Core Mass, MTU 138.35 Rated Thermal Power Level, MWt 3,952 Rated Core Flow, Mlb/hr 102.5 Reference Pressure, psia 1,050* Reference Inlet Subcooling, Btu/lbm 26.94f

• Value is representative of MICROBURN-B2 input for dome pressure at rated conditions and varies depending on core state point. f Value is typically determined by MICROBURN-B2 using a heat balance method based on nominal feedwater temperature and other parameters identified in the cycle specific plant parameters document.

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 3-4 AREVA Inc. Cycle Exposure (MWd/MTU) Hot Operating k-eff* 0.0 1.0000 7,500.0 0.9985 10,500.0 0.9985 15,000.0 1.0000 25,000.0 1.0000 Cycle Exposure (MWd/MTU) Cold Critical k-eff* 0.0 0.9940 25,000.0 0.9940

• Values are linearly interpolated between cycle exposure points.

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 3-5 AREVA Inc. Cycle Exposure (MWd/MTU) Cold Shutdown Margin* (% k/k) SLC Cold Shutdown Marginf (% k/k) Hot Excess Reactivity (% k/k) 0 1.21 2.92 1.80 250 1.32 3.18 1.71 1,000 1.71 3.57 1.56 2,300 2.14 3.88 1.54 3,300 2.49 4.11 1.51 4,300 2.84 4.34 1.47 5,500 3.04 4.59 1.44 6,500 3.13 4.75 1.44 7,500 3.23 4.87 1.47 8,500 3.31 4.96 1.50 10,000 3.41 4.98 1.61 11,000 3.43 4.94 1.68 12,000 3.19 4.89 1.74 13,000 4.86 1.79 14,000 4.88 1.81 15,100 5.04 1.71 16,100 5.32 1.51 17,000 5.62 1.14 17,590 5.81 0.79 18,525 6.17 0.10 19,211 6.49 -- 19,811 6.85 --

• Values in are limiting values at elevated moderator temperatures. f Calculated at 366.0 °F ARO conditions.

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 3-6 AREVA Inc. Cycle Exposure (MWd/MTU) Cold Shutdown Margin* (% k/k) SLC Cold ShutdownMarginf (% k/k) Hot Excess Reactivity (% k/k) 0 1.51 3.27 1.48 250 1.65 3.55 1.38 1,000 2.09 3.93 1.23 2,300 2.56 4.23 1.20 3,300 2.92 4.45 1.18 4,300 3.15 4.69 1.14 5,500 3.30 4.90 1.12 6,500 3.42 5.05 1.12 7,500 3.50 5.15 1.15 8,500 3.61 5.23 1.19 10,000 3.70 5.23 1.30 11,000 3.63 5.19 1.37 12,000 5.14 1.42 13,000 3.23 5.11 1.47 14,000 5.13 1.49 15,100 5.29 1.43 16,100 5.55 1.26 17,000 5.84 0.92 17,590 6.04 0.58 18,525 6.38 -0.09 19,211 6.69 -- 19,811 7.04 --

• Values in are limiting values at elevated moderator temperatures. f Calculated at 366.0 °F ARO conditions.

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 3-7 AREVA Inc. Cycle Exposure (MWd/MTU) Cold Shutdown Margin* (% k/k) SLC Cold ShutdownMarginf (% k/k) Hot Excess Reactivity (% k/k) 0 1.75 3.51 1.26 250 1.90 3.80 1.14 1,000 2.39 4.18 0.99 2,300 2.88 4.46 0.97 3,300 3.19 4.70 0.95 4,300 3.34 4.88 0.91 5,500 3.48 5.10 0.89 6,500 3.63 5.24 0.90 7,500 3.72 5.33 0.94 8,500 3.82 5.40 0.97 10,000 3.91 5.38 1.08 11,000 3.82 5.34 1.15 12,000 3.62 5.30 1.20 13,000 5.28 1.25 14,000 5.31 1.28 15,100 5.44 1.23 16,100 5.71 1.08 17,000 6.01 0.77 17,590 6.19 0.45 18,525 6.52 -0.22 19,211 6.81 -- 19,811 7.16 --

• Values in are limiting values at elevated moderator temperatures. f Calculated at 366.0 °F ARO conditions.

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 3-8 AREVA Inc. 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 60 14 14 14 14 14 13 14 40.5 36.6 36.6 36.7 36.6 36.0 35.9

58 Nuclear Fuel Type 14 15 17 16 17 17 17 18 BOC Exposure (GWd/MTU) 37.9 34.3 23.6 23.0 20.8 21.3 21.1 19.0 56 14 13 16 17 20 17 20 16 20 16 38.5 37.4 23.0 21.9 0.0 22.0 0.0 24.2 0.0 23.2

54 14 17 17 20 21 16 20 20 21 19 37.7 23.9 19.0 0.0 0.0 23.3 0.0 0.0 0.0 0.0 52 14 17 21 20 19 20 20 21 17 19 16 41.2 24.5 0.0 0.0 0.0 0.0 0.0 0.0 25.0 0.0 25.2

50 14 14 17 17 20 16 20 16 19 16 19 16 19 38.2 37.7 24.1 23.7 0.0 23.8 0.0 24.6 0.0 23.9 0.0 24.0 0.0 48 13 17 21 20 20 19 16 19 17 19 16 19 16 37.1 23.8 0.0 0.0 0.0 0.0 24.5 0.0 24.8 0.0 24.4 0.0 23.9

46 14 16 17 20 16 19 16 19 16 19 16 19 16 19 37.8 22.9 18.9 0.0 24.1 0.0 23.8 0.0 23.9 0.0 25.4 0.0 23.0 0.0

44 14 15 17 20 19 20 16 19 18 19 16 19 16 19 16 40.3 34.3 21.8 0.0 0.0 0.0 24.5 0.0 21.9 0.0 24.8 0.0 24.6 0.0 25.0 42 14 17 20 21 20 16 19 16 19 18 19 16 19 16 19 36.3 23.3 0.0 0.0 0.0 24.7 0.0 24.3 0.0 21.9 0.0 24.7 0.0 24.8 0.0

40 14 16 17 16 20 19 17 19 16 19 16 19 18 19 16 36.7 23.4 22.0 23.5 0.0 0.0 24.7 0.0 24.2 0.0 23.7 0.0 23.5 0.0 24.7 38 14 17 20 20 21 16 19 16 19 16 19 16 19 18 19 36.5 20.6 0.0 0.0 0.0 23.9 0.0 25.4 0.0 24.6 0.0 23.4 0.0 24.8 0.0

36 14 17 16 20 17 19 16 19 16 19 18 19 16 19 16 36.7 21.3 24.3 0.0 24.1 0.0 24.6 0.0 24.9 0.0 23.6 0.0 25.0 0.0 24.5 34 13 17 20 21 19 16 19 16 19 16 19 18 19 16 19 36.0 21.0 0.0 0.0 0.0 24.0 0.0 23.4 0.0 25.0 0.0 24.0 0.0 25.0 0.0

32 14 18 16 19 16 19 16 19 16 19 16 19 16 19 16 35.6 19.0 23.4 0.0 25.5 0.0 25.3 0.0 24.9 0.0 24.3 0.0 25.1 0.0 25.0

No. Per Fuel Type Description Cycle Loaded Quarter core _________ ___________ ____________ ____________ 13 [] N-2 4 14 [] N-2 19 15 [] N-2 2 16 [] N-1 52 17 [] N-1 23 18 [] N-1 8 19 [] N 52 20 [] N 23 21 [] N 8

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 3-9 AREVA Inc. 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 60 14 13 14 14 14 14 14 35.9 36.0 36.8 36.7 36.7 36.7 40.5

58 18 17 17 17 16 17 15 14 Nuclear Fuel Type 19.0 21.1 21.3 20.8 23.0 23.6 34.3 38.0 BOC Exposure (GWd/MTU) 56 16 20 16 20 17 20 17 16 13 14 23.5 0.0 23.5 0.0 22.0 0.0 21.9 23.0 37.1 38.5

54 19 21 20 20 16 21 20 17 17 14 0.0 0.0 0.0 0.0 23.3 0.0 0.0 19.0 23.9 37.7 52 16 19 17 21 20 20 19 20 21 17 14 25.2 0.0 25.0 0.0 0.0 0.0 0.0 0.0 0.0 24.5 41.2

50 19 16 19 16 19 16 20 16 20 17 17 14 14 0.0 24.1 0.0 23.9 0.0 24.6 0.0 23.8 0.0 23.7 24.1 37.8 38.2 48 16 19 16 19 17 19 16 19 20 20 21 17 13 23.8 0.0 24.4 0.0 24.8 0.0 24.5 0.0 0.0 0.0 0.0 23.8 37.4

46 19 16 19 16 19 16 19 16 19 16 20 17 16 14 0.0 23.0 0.0 25.4 0.0 23.8 0.0 24.1 0.0 24.1 0.0 18.9 22.9 37.8

44 16 19 16 19 16 19 18 19 16 20 19 20 17 15 14 25.0 0.0 24.6 0.0 24.8 0.0 22.2 0.0 24.5 0.0 0.0 0.0 21.8 34.3 40.3 42 19 16 19 16 19 18 19 16 19 16 20 21 20 17 14 0.0 24.8 0.0 24.8 0.0 22.2 0.0 24.2 0.0 24.8 0.0 0.0 0.0 23.3 36.4

40 16 19 18 19 16 19 16 19 17 19 20 16 17 16 14 24.7 0.0 23.5 0.0 24.1 0.0 25.0 0.0 24.7 0.0 0.0 23.5 22.0 23.4 36.7 38 19 18 19 16 19 16 19 16 19 16 21 20 20 17 14 0.0 24.8 0.0 23.5 0.0 24.7 0.0 25.4 0.0 23.9 0.0 0.0 0.0 20.6 36.5

36 16 19 16 19 18 19 16 19 16 19 17 20 16 17 14 24.5 0.0 24.2 0.0 23.6 0.0 24.9 0.0 24.6 0.0 24.1 0.0 23.5 21.4 36.6 34 19 16 19 18 19 16 19 16 19 16 19 21 20 17 13 0.0 25.0 0.0 24.0 0.0 25.0 0.0 23.4 0.0 24.0 0.0 0.0 0.0 21.0 36.0

32 16 19 16 19 16 19 16 19 16 19 16 19 16 18 14 25.0 0.0 25.1 0.0 24.3 0.0 24.9 0.0 25.2 0.0 25.5 0.0 23.1 18.9 35.7

No. Per Fuel Type Description Cycle Loaded Quarter core _________ ___________ ____________ ____________ 13 [] N-2 4 14 [] N-2 19 15 [] N-2 2 16 [] N-1 52 17 [] N-1 23 18 [] N-1 8 19 [] N 52 20 [] N 23 21 [] N 8

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 3-10 AREVA Inc. 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 30 14 18 16 19 16 19 16 19 16 19 16 19 16 19 16 35.6 19.0 23.1 0.0 25.5 0.0 25.3 0.0 24.9 0.0 24.3 0.0 25.1 0.0 25.0

28 13 17 20 21 19 16 19 16 19 16 19 18 19 16 19 36.2 21.0 0.0 0.0 0.0 24.0 0.0 23.4 0.0 25.0 0.0 24.0 0.0 25.0 0.0 26 14 17 16 20 17 19 16 19 16 19 18 19 16 19 16 36.7 21.3 24.3 0.0 24.1 0.0 24.6 0.0 24.9 0.0 23.6 0.0 24.2 0.0 24.5

24 14 17 20 20 21 16 19 16 19 16 19 16 19 18 19 36.5 20.6 0.0 0.0 0.0 23.9 0.0 25.4 0.0 24.7 0.0 23.4 0.0 24.8 0.0 22 14 16 17 16 20 19 17 19 16 19 16 19 18 19 16 36.6 23.4 22.0 23.5 0.0 0.0 24.7 0.0 24.8 0.0 23.8 0.0 23.5 0.0 24.7

20 14 17 20 21 20 16 19 16 19 18 19 16 19 16 19 36.3 23.3 0.0 0.0 0.0 24.7 0.0 24.2 0.0 22.0 0.0 24.8 0.0 24.8 0.0 18 14 15 17 20 19 20 16 19 18 19 16 19 16 19 16 40.3 34.3 21.8 0.0 0.0 0.0 24.5 0.0 22.0 0.0 25.0 0.0 24.7 0.0 25.0

16 14 16 17 20 16 19 16 19 16 19 16 19 16 19 37.8 22.9 18.9 0.0 24.1 0.0 23.8 0.0 23.9 0.0 25.4 0.0 23.0 0.0

14 13 17 21 20 20 19 16 19 17 19 16 19 16 37.4 23.8 0.0 0.0 0.0 0.0 24.5 0.0 24.8 0.0 24.4 0.0 23.9 12 14 14 17 17 20 16 20 16 19 16 19 16 19 38.2 37.9 24.1 23.7 0.0 23.8 0.0 24.6 0.0 23.9 0.0 24.1 0.0

10 14 17 21 20 19 20 20 21 17 19 16 41.2 24.5 0.0 0.0 0.0 0.0 0.0 0.0 25.0 0.0 25.2 8 14 17 17 20 21 16 20 20 21 19 37.9 23.9 19.0 0.0 0.0 23.3 0.0 0.0 0.0 0.0

6 14 13 16 17 20 17 20 16 20 16 38.5 37.1 23.0 21.9 0.0 22.0 0.0 24.2 0.0 23.4 4 Nuclear Fuel Type 14 15 17 16 17 17 17 18 BOC Exposure (GWd/MTU) 37.9 34.3 23.6 23.0 20.8 21.3 21.1 19.0

2 14 14 14 14 14 13 14 40.5 36.7 36.7 36.7 36.8 36.2 35.8

No. Per Fuel Type Description Cycle Loaded Quarter core _________ ___________ ____________ ____________ 13 [] N-2 4 14 [] N-2 19 15 [] N-2 2 16 [] N-1 52 17 [] N-1 23 18 [] N-1 8 19 [] N 52 20 [] N 23 21 [] N 8

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 3-11 AREVA Inc. 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 30 16 19 16 19 16 19 16 19 16 19 16 19 16 18 14 25.0 0.0 25.1 0.0 24.3 0.0 24.9 0.0 25.3 0.0 25.5 0.0 23.5 18.9 35.7

28 19 16 19 18 19 16 19 16 19 16 19 21 20 17 13 0.0 25.0 0.0 24.0 0.0 25.0 0.0 23.4 0.0 24.0 0.0 0.0 0.0 21.0 36.2 26 16 19 16 19 18 19 16 19 16 19 17 20 16 17 14 24.5 0.0 25.0 0.0 23.6 0.0 24.9 0.0 24.6 0.0 24.1 0.0 23.5 21.4 36.8

24 19 18 19 16 19 16 19 16 19 16 21 20 20 17 14 0.0 24.8 0.0 23.5 0.0 24.6 0.0 25.4 0.0 23.9 0.0 0.0 0.0 20.6 36.5 22 16 19 18 19 16 19 16 19 17 19 20 16 17 16 14 24.7 0.0 23.6 0.0 24.8 0.0 24.2 0.0 24.7 0.0 0.0 23.5 22.0 23.4 36.7

20 19 16 19 16 19 18 19 16 19 16 20 21 20 17 14 0.0 24.7 0.0 24.7 0.0 22.2 0.0 24.2 0.0 24.8 0.0 0.0 0.0 23.3 36.3 18 16 19 16 19 16 19 18 19 16 20 19 20 17 15 14 25.0 0.0 24.7 0.0 24.1 0.0 22.2 0.0 24.5 0.0 0.0 0.0 21.8 34.3 40.3

16 19 16 19 16 19 16 19 16 19 16 20 17 16 14 0.0 23.0 0.0 25.4 0.0 23.9 0.0 24.1 0.0 24.1 0.0 18.9 22.9 37.8

14 16 19 16 19 17 19 16 19 20 20 21 17 13 23.9 0.0 24.4 0.0 24.8 0.0 24.5 0.0 0.0 0.0 0.0 23.8 37.1 12 19 16 19 16 19 16 20 16 20 17 17 14 14 0.0 24.0 0.0 23.9 0.0 24.6 0.0 23.8 0.0 23.7 24.1 37.9 38.2

10 16 19 17 21 20 20 19 20 21 17 14 25.2 0.0 25.0 0.0 0.0 0.0 0.0 0.0 0.0 24.5 41.2 8 19 21 20 20 16 21 20 17 17 14 0.0 0.0 0.0 0.0 23.3 0.0 0.0 19.0 23.9 37.9

6 16 20 16 20 17 20 17 16 13 14 23.1 0.0 23.5 0.0 22.0 0.0 21.9 23.0 37.4 38.5 4 18 17 17 17 16 17 15 14 Nuclear Fuel Type 19.0 21.1 21.4 20.8 23.0 23.6 34.3 38.0 BOC Exposure (GWd/MTU)

2 14 13 14 14 14 14 14 35.9 36.2 36.7 36.7 36.7 36.7 40.5

No. Per Fuel Type Description Cycle Loaded Quarter core _________ ___________ ____________ ____________ 13 [] N-2 4 14 [] N-2 19 15 [] N-2 2 16 [] N-1 52 17 [] N-1 23 18 [] N-1 8 19 [] N 52 20 [] N 23 21 [] N 8

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page 4-1 AREVA Inc. 1. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October, 1999. 2. ANP-10298PA Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, March 2010. 3. ANP-3343P Revision 0, Nuclear Fuel Design Report Browns Ferry EPU (120% OLTP) Equilibrium Cycle ATRIUM 10XM Fuel, AREVA Inc., October 2014. 4. BFE-3706 Revision 0, Browns Ferry Unit 3 EPU Transition Energy Plan, TVA, June 2014. (38-9226272-000) 5. ANP-3140(P) Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, August 2012. 6. ANP-3342P Revision 1, Browns Ferry EPU (120% OLTP) Equilibrium Fuel Cycle Design, AREVA Inc., August 2015. Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-1 AREVA Inc. Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-2 AREVA Inc. Total Total Inlet Core Core Cycle Control Core Core Ref. Sub- Core Maximum Maximum Exposure Calculated Rod Power Flow Pressure Cooling Void Minimum LHGR APLHGR (GWd/MT) K-eff Density MWt (Mlb/hr) (psia) (Btu/lb) Fraction CPR (kW/ft) (kW/ft) ________ __________ _______ ______ ________ ________ ________ ________ _______ _______ _______

0.000 0.99999 5.05 3952.0 89.58 1050.04 31.15 0.505 1.655 11.87 8.84 0.250 0.99994 5.05 3952.0 93.48 1050.04 29.75 0.491 1.680 11.32 8.47 0.500 0.99983 5.05 3952.0 97.48 1050.04 28.44 0.480 1.708 11.04 8.29 1.000 0.99976 5.05 3952.0 99.73 1050.04 27.75 0.474 1.722 10.88 8.22 1.500 0.99966 5.05 3952.0 99.63 1050.04 27.78 0.472 1.723 10.77 8.20 1.820 0.99958 5.05 3952.0 99.53 1050.04 27.81 0.470 1.723 10.70 8.19 1.820 0.99957 4.28 3952.0 90.30 1050.04 30.88 0.495 1.682 11.45 8.78 2.300 0.99943 4.28 3952.0 90.40 1050.04 30.84 0.493 1.682 11.30 8.73 2.800 0.99940 4.28 3952.0 91.02 1050.04 30.62 0.489 1.685 11.17 8.70 3.300 0.99926 4.28 3952.0 91.33 1050.04 30.51 0.486 1.684 11.09 8.70 3.800 0.99917 4.28 3952.0 91.94 1050.04 30.29 0.483 1.685 11.02 8.71 4.300 0.99904 4.28 3952.0 92.25 1050.04 30.18 0.481 1.685 10.98 8.75 4.563 0.99899 4.28 3952.0 92.45 1050.04 30.11 0.479 1.685 10.97 8.78 4.564 0.99914 3.78 3952.0 92.45 1050.04 30.11 0.489 1.656 10.77 8.62 5.000 0.99908 3.78 3952.0 93.07 1050.04 29.89 0.486 1.664 10.77 8.68 5.500 0.99895 3.78 3952.0 92.97 1050.04 29.93 0.485 1.664 10.78 8.76 6.000 0.99888 3.78 3952.0 92.97 1050.04 29.93 0.484 1.665 10.81 8.87 6.500 0.99882 3.78 3952.0 92.76 1050.04 30.00 0.484 1.664 10.89 9.01 6.750 0.99874 3.78 3952.0 92.35 1050.04 30.14 0.485 1.662 10.96 9.11 7.110 0.99869 3.78 3952.0 91.94 1050.04 30.29 0.485 1.660 11.05 9.25 7.111 0.99849 4.59 3952.0 97.68 1050.04 28.38 0.465 1.652 10.62 8.67 7.500 0.99841 4.59 3952.0 96.96 1050.04 28.60 0.466 1.649 10.64 8.73 8.000 0.99842 4.59 3952.0 96.45 1050.04 28.77 0.468 1.646 10.76 8.92 8.500 0.99847 4.59 3952.0 95.74 1050.04 29.00 0.469 1.641 10.90 9.15 9.000 0.99841 4.59 3952.0 94.20 1050.04 29.51 0.473 1.628 11.14 9.43 9.250 0.99840 4.59 3952.0 93.48 1050.04 29.75 0.475 1.621 11.27 9.56 9.658 0.99842 4.59 3952.0 92.35 1050.04 30.14 0.478 1.611 11.47 9.78 9.659 0.99847 5.23 3952.0 98.60 1050.04 28.09 0.462 1.727 11.18 9.65 10.000 0.99845 5.23 3952.0 96.96 1050.04 28.60 0.466 1.712 11.26 9.77 10.500 0.99848 5.23 3952.0 95.43 1050.04 29.10 0.470 1.699 11.28 9.84 11.000 0.99862 5.23 3952.0 94.20 1050.04 29.51 0.472 1.687 11.16 9.77 11.500 0.99878 5.23 3952.0 93.07 1050.04 29.89 0.472 1.673 10.89 9.57 12.000 0.99892 5.23 3952.0 91.94 1050.04 30.29 0.471 1.657 10.60 9.37 12.205 0.99901 5.23 3952.0 91.53 1050.04 30.43 0.470 1.646 10.43 9.24 12.206 0.99888 5.41 3952.0 94.20 1050.04 29.51 0.469 1.679 11.12 9.82 12.500 0.99901 5.41 3952.0 93.79 1050.04 29.65 0.467 1.672 10.83 9.56 13.000 0.99919 5.41 3952.0 92.97 1050.04 29.93 0.462 1.657 10.22 9.04 13.500 0.99940 5.41 3952.0 92.45 1050.04 30.11 0.455 1.643 9.63 8.47 14.000 0.99956 5.41 3952.0 92.05 1050.04 30.25 0.447 1.626 9.43 8.23 14.500 0.99970 5.41 3952.0 92.15 1050.04 30.22 0.438 1.603 9.73 8.56 14.753 0.99978 5.41 3952.0 92.45 1050.04 30.11 0.433 1.594 9.88 8.72 14.754 1.00005 4.32 3952.0 89.38 1050.04 31.22 0.456 1.618 9.11 8.07 15.100 1.00011 4.32 3952.0 90.92 1050.04 30.66 0.446 1.616 9.10 8.30 15.600 1.00014 4.32 3952.0 93.28 1050.04 29.82 0.431 1.612 9.36 8.58 16.100 1.00011 4.32 3952.0 97.17 1050.04 28.54 0.415 1.616 9.68 8.81 16.500 1.00014 4.32 3952.0 101.99 1050.04 27.08 0.400 1.618 9.90 9.01 16.501 1.00009 3.78 3952.0 93.17 1050.04 29.86 0.408 1.570 9.64 8.79 17.000 1.00012 3.78 3952.0 101.27 1050.04 27.29 0.386 1.587 9.89 9.03 17.001 1.00011 3.60 3952.0 96.45 1050.04 28.77 0.391 1.563 9.77 8.92 17.300 1.00011 3.60 3952.0 102.40 1050.04 26.97 0.376 1.576 9.90 9.05 17.301 1.00005 2.75 3952.0 95.94 1050.04 28.93 0.391 1.532 9.85 9.03 17.590 1.00004 2.75 3952.0 102.50 1050.04 26.94 0.376 1.558 10.03 9.14 17.591 1.00008 1.80 3952.0 94.50 1050.04 29.41 0.392 1.515 9.79 9.06 17.875 1.00011 1.80 3952.0 101.89 1050.04 27.11 0.375 1.551 9.93 9.17 17.876 1.00009 1.53 3952.0 95.22 1050.04 29.17 0.382 1.520 10.77 9.74 18.100 1.00005 1.53 3952.0 101.37 1050.04 27.26 0.369 1.553 10.86 9.81 18.101 1.00004 0.00 3952.0 94.10 1050.04 29.54 0.392 1.538 9.53 8.83 18.400 1.00005 0.00 3952.0 103.53 1050.04 26.65 0.372 1.598 9.73 8.96 18.525 1.00018 0.00 3952.0 107.62 1050.04 25.55 0.364 1.615 9.81 9.01 18.540 1.00004 0.00 3952.0 94.92 1044.65 36.39 0.363 1.598 9.39 8.65 18.940 0.99989 0.00 3952.0 107.62 1044.65 31.83 0.338 1.639 9.69 8.85 18.970 1.00000 0.00 3926.7 107.62 1044.16 31.60 0.336 1.647 9.64 8.81 19.211 1.00012 0.00 3703.0 107.62 1039.83 29.58 0.323 1.728 9.29 8.47 19.494 1.00014 0.00 3474.6 107.62 1035.41 27.52 0.309 1.820 8.91 8.11 19.811 1.00019 0.00 3211.8 107.62 1030.33 25.17 0.293 1.944 8.44 7.67 20.411 0.99958 0.00 2719.0 107.62 1020.81 20.80 0.261 2.232 7.69 6.90 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-3 AREVA Inc. Fraction Core Fraction Core Fraction Cycle Control Core of Limiting of Limiting of Exposure Calculated Rod Limiting Limiting LHGR Limiting APLHGR Limiting (GWd/MT) K-eff Density CPR CPR (kW/ft) LHGR (kW/ft) APLHGR __________ __________ __________ __________ __________ __________ __________ __________ __________

0.000 0.99999 5.045 1.655 0.858 11.87 0.842 8.77 0.741 0.250 0.99994 5.045 1.680 0.845 11.32 0.803 8.44 0.716 0.500 0.99983 5.045 1.708 0.831 11.04 0.783 8.25 0.702 1.000 0.99976 5.045 1.722 0.825 10.88 0.772 8.10 0.694 1.500 0.99966 5.045 1.723 0.824 10.77 0.764 7.90 0.691 1.820 0.99958 5.045 1.723 0.824 10.70 0.759 7.84 0.688 1.820 0.99957 4.279 1.682 0.844 11.45 0.812 8.50 0.733 2.300 0.99943 4.279 1.682 0.844 11.30 0.801 8.35 0.726 2.800 0.99940 4.279 1.685 0.843 11.17 0.792 7.83 0.722 3.300 0.99926 4.279 1.684 0.843 11.09 0.787 7.77 0.722 3.800 0.99917 4.279 1.685 0.843 11.02 0.782 7.72 0.723 4.300 0.99904 4.279 1.685 0.843 10.98 0.779 7.69 0.726 4.563 0.99899 4.279 1.685 0.842 10.97 0.778 7.67 0.728 4.564 0.99914 3.784 1.656 0.858 10.77 0.764 7.51 0.712 5.000 0.99908 3.784 1.664 0.853 10.77 0.764 7.49 0.715 5.500 0.99895 3.784 1.664 0.854 10.78 0.765 7.48 0.720 6.000 0.99888 3.784 1.665 0.853 10.81 0.767 7.48 0.726 6.500 0.99882 3.784 1.664 0.853 10.89 0.772 7.49 0.733 6.750 0.99874 3.784 1.662 0.854 10.96 0.777 7.51 0.737 7.110 0.99869 3.784 1.660 0.855 11.05 0.783 7.53 0.744 7.111 0.99849 4.595 1.652 0.860 10.62 0.753 7.36 0.713 7.500 0.99841 4.595 1.649 0.861 10.64 0.755 7.32 0.716 8.000 0.99842 4.595 1.646 0.863 10.76 0.763 7.31 0.727 8.500 0.99847 4.595 1.641 0.866 10.89 0.773 7.39 0.741 9.000 0.99841 4.595 1.628 0.872 11.14 0.798 7.47 0.755 9.250 0.99840 4.595 1.621 0.876 11.27 0.811 7.52 0.763 9.658 0.99842 4.595 1.611 0.881 11.47 0.833 7.59 0.776 9.659 0.99847 5.225 1.727 0.822 11.15 0.813 7.49 0.763 10.000 0.99845 5.225 1.712 0.830 11.22 0.824 9.77 0.770 10.500 0.99848 5.225 1.699 0.836 11.28 0.834 9.84 0.782 11.000 0.99862 5.225 1.687 0.842 11.16 0.833 9.77 0.781 11.500 0.99878 5.225 1.673 0.849 10.85 0.819 9.54 0.769 12.000 0.99892 5.225 1.657 0.857 10.49 0.794 9.37 0.756 12.205 0.99901 5.225 1.646 0.862 10.32 0.784 9.24 0.748 12.206 0.99888 5.405 1.679 0.846 11.12 0.844 9.82 0.795 12.500 0.99901 5.405 1.672 0.849 10.78 0.824 9.56 0.778 13.000 0.99919 5.405 1.657 0.857 10.12 0.781 9.04 0.742 13.500 0.99940 5.405 1.643 0.865 9.53 0.745 6.57 0.707 14.000 0.99956 5.405 1.626 0.873 7.58 0.722 6.31 0.685 14.500 0.99970 5.405 1.603 0.886 7.79 0.749 7.02 0.704 14.753 0.99978 5.405 1.594 0.891 7.90 0.763 7.12 0.717 14.754 1.00005 4.324 1.618 0.878 7.20 0.718 6.60 0.691 15.100 1.00011 4.324 1.616 0.885 7.38 0.736 6.79 0.699 15.600 1.00014 4.324 1.612 0.887 7.63 0.769 6.99 0.725 16.100 1.00011 4.324 1.616 0.885 7.84 0.798 7.25 0.749 16.500 1.00014 4.324 1.618 0.884 8.02 0.821 7.39 0.768 16.501 1.00009 3.784 1.570 0.911 7.82 0.800 7.21 0.750 17.000 1.00012 3.784 1.587 0.901 8.05 0.828 7.36 0.771 17.001 1.00011 3.604 1.563 0.915 7.95 0.818 7.28 0.763 17.300 1.00011 3.604 1.576 0.907 8.05 0.834 7.35 0.775 17.301 1.00005 2.748 1.532 0.933 8.06 0.834 7.42 0.782 17.590 1.00004 2.748 1.558 0.918 8.14 0.848 7.48 0.792 17.591 1.00008 1.802 1.515 0.944 8.02 0.835 7.40 0.784 17.875 1.00011 1.802 1.551 0.922 8.10 0.849 7.45 0.793 17.876 1.00009 1.532 1.520 0.941 8.34 0.837 8.10 0.832 18.100 1.00005 1.532 1.553 0.921 8.51 0.849 8.23 0.840 18.101 1.00004 0.000 1.538 0.930 7.78 0.819 7.18 0.767 18.400 1.00005 0.000 1.598 0.895 7.88 0.835 7.45 0.780 18.525 1.00018 0.000 1.615 0.885 8.18 0.842 7.50 0.787 18.540 1.00004 0.000 1.598 0.895 7.70 0.819 7.07 0.761 18.940 0.99989 0.000 1.639 0.873 7.85 0.842 7.38 0.780 18.970 1.00000 0.000 1.647 0.868 7.80 0.837 7.34 0.777 19.211 1.00012 0.000 1.728 0.828 7.71 0.805 7.09 0.746 19.494 1.00014 0.000 1.820 0.786 7.47 0.774 6.76 0.715 19.811 1.00019 0.000 1.944 0.735 7.05 0.736 6.36 0.677 20.411 0.99958 0.000 2.232 0.641 6.23 0.659 5.87 0.610 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-4 AREVA Inc. Cycle: 21 Core Average Exposure: MWd/MTU 15106.5 Exposure: MWd/MTU (GWd) 0.0 ( 0.00 ) Delta E: MWd/MTU, (GWd) 0.0 ( 0.00 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.171 3.331 13 0.496 0.518 59 34 Inlet Subcooling: Btu/lbm -31.15 24 0.490 9.341 14 0.436 0.534 59 30 Flow: Mlb/hr 89.58 ( 87.40 %) 23 0.634 12.049 15 0.544 0.547 3 18 22 0.731 13.435 16 1.112 1.282 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.801 14.489 17 0.878 1.162 51 36 59 -- -- -- -- -- -- -- 59 20 0.848 15.294 18 1.121 1.219 21 36 55 -- -- -- -- -- -- -- -- -- 55 19 0.869 15.720 19 1.151 1.260 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.884 16.230 20 1.060 1.145 53 36 47 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 47 17 0.898 16.565 21 1.053 1.145 33 8 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.918 16.841 39 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 39 15 0.956 17.321 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.978 17.225 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.109 16.242 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.164 16.723 23 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 23 11 1.194 16.969 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.219 17.306 15 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 15 9 1.260 17.696 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.303 17.910 7 -- -- -- -- -- -- -- -- -- 7 7 1.358 18.110 3 -- -- -- -- -- -- -- 3 6 1.439 18.547* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.496 18.526 4 1.497* 17.810 Control Rod Density: % 5.05 3 1.410 16.239 2 1.087 12.144 k-effective: 0.99999 Bottom 1 0.288 3.506 Void Fraction: 0.505 Core Delta-P: psia 20.604  % AXIAL TILT -23.178 -6.817 Core Plate Delta-P: psia 16.047 AVG BOT 8ft/12ft 1.1282 1.0412 Coolant Temp: Deg-F 548.3 In Channel Flow: Mlb/hr 78.48 Active Channel Flow: Mlb/hr 75.52 Total Bypass Flow (%): 12.4 (of total core flow) Total Water Rod Flow (%): 3.3 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.282 16 29 32 1.655 0.858 19 29 28 8.77 0.741 26.2 16 31 14 4 11.87 0.842 0.0 19 31 50 4 1.264 16 23 24 1.658 0.856 19 27 30 8.43 0.729 28.8 16 31 10 4 11.77 0.835 0.0 19 31 16 4 1.260 19 21 38 1.672 0.849 19 21 38 8.33 0.723 29.3 16 51 30 4 11.76 0.834 0.0 19 49 32 4 1.259 19 23 22 1.677 0.847 19 23 22 8.56 0.720 25.7 16 33 50 4 11.74 0.833 0.0 19 33 10 4 1.259 16 31 48 1.690 0.840 19 23 36 8.30 0.716 28.5 16 47 32 4 11.66 0.827 0.0 19 51 34 4 1.255 19 29 16 1.691 0.840 19 25 38 8.50 0.713 25.5 16 49 34 4 11.65 0.826 0.0 20 25 8 4 1.254 19 29 28 1.692 0.839 19 29 46 8.05 0.705 30.3 17 51 36 4 11.63 0.825 0.0 20 7 26 4 1.253 16 21 40 1.696 0.837 16 31 32 7.91 0.703 31.9 17 35 10 4 11.61 0.823 0.0 19 33 48 4 1.252 19 27 30 1.701 0.835 19 45 30 8.34 0.699 25.3 16 33 16 4 11.58 0.821 0.0 19 31 8 4 1.243 19 15 32 1.704 0.833 19 19 36 8.20 0.690 25.7 16 21 40 4 11.52 0.817 0.0 20 33 56 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-5 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 15355.4 Exposure: MWd/MTU (GWd) 250.0 ( 34.59 ) Delta E: MWd/MTU, (GWd) 250.0 ( 34.59 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.178 3.377 13 0.497 0.520 59 34 Inlet Subcooling: Btu/lbm -29.75 24 0.510 9.475 14 0.438 0.535 59 30 Flow: Mlb/hr 93.48 ( 91.20 %) 23 0.660 12.223 15 0.546 0.548 3 18 22 0.759 13.636 16 1.113 1.283 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.830 14.709 17 0.881 1.164 51 36 59 -- -- -- -- -- -- -- 59 20 0.877 15.527 18 1.123 1.220 21 36 55 -- -- -- -- -- -- -- -- -- 55 19 0.897 15.958 19 1.149 1.257 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.910 16.472 20 1.057 1.142 53 36 47 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 47 17 0.922 16.811 21 1.050 1.142 33 8 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.939 17.091 39 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 39 15 0.976 17.581 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.995 17.491 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.127 16.503 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.180 16.997 23 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 23 11 1.206 17.249 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.228 17.592 15 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 15 9 1.262 17.991 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.297 18.214 7 -- -- -- -- -- -- -- -- -- 7 7 1.341 18.426 3 -- -- -- -- -- -- -- 3 6 1.409 18.880* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.448* 18.871 4 1.429 18.152 Control Rod Density: % 5.05 3 1.331 16.560 2 1.021 12.390 k-effective: 0.99994 Bottom 1 0.270 3.575 Void Fraction: 0.491 Core Delta-P: psia 21.808  % AXIAL TILT -20.823 -6.957 Core Plate Delta-P: psia 17.252 AVG BOT 8ft/12ft 1.1155 1.0419 Coolant Temp: Deg-F 548.4 In Channel Flow: Mlb/hr 82.02 Active Channel Flow: Mlb/hr 78.98 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.283 16 29 32 1.680 0.845 19 29 28 8.44 0.716 26.6 16 31 14 4 11.32 0.803 0.6 19 31 50 4 1.265 16 23 24 1.683 0.844 19 27 32 8.11 0.704 29.3 16 31 10 4 11.23 0.796 0.6 19 31 16 4 1.259 16 31 48 1.698 0.836 19 21 38 8.01 0.698 29.6 16 51 32 4 11.21 0.795 0.6 19 49 32 4 1.257 19 21 38 1.704 0.833 19 23 22 8.22 0.695 26.2 16 33 50 4 11.20 0.794 0.6 19 33 10 4 1.256 19 23 22 1.714 0.829 16 31 32 7.98 0.691 29.0 16 47 32 4 11.11 0.788 0.6 19 51 34 4 1.254 16 21 40 1.716 0.828 19 23 36 8.16 0.688 25.9 16 49 34 4 11.09 0.787 0.6 20 25 8 4 1.252 19 29 16 1.717 0.827 19 25 38 7.57 0.680 33.0 17 35 10 5 11.09 0.786 0.6 19 33 48 4 1.251 19 29 28 1.720 0.826 19 29 46 7.67 0.680 31.6 17 51 36 5 11.08 0.785 0.6 20 7 26 4 1.250 19 27 30 1.728 0.822 19 45 30 8.01 0.674 25.7 16 33 16 4 11.06 0.784 0.6 19 31 8 4 1.243 16 27 16 1.729 0.822 19 19 36 7.87 0.664 26.1 16 21 40 4 10.98 0.779 0.6 20 33 56 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-6 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 15605.5 Exposure: MWd/MTU (GWd) 500.0 ( 69.17 ) Delta E: MWd/MTU, (GWd) 250.0 ( 34.59 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.183 3.425 13 0.495 0.517 59 34 Inlet Subcooling: Btu/lbm -28.44 24 0.523 9.613 14 0.436 0.532 59 30 Flow: Mlb/hr 97.48 ( 95.10 %) 23 0.677 12.402 15 0.544 0.547 3 18 22 0.778 13.844 16 1.114 1.282 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.848 14.935 17 0.881 1.166 51 36 59 -- -- -- -- -- -- -- 59 20 0.894 15.766 18 1.124 1.221 21 36 55 -- -- -- -- -- -- -- -- -- 55 19 0.912 16.202 19 1.149 1.256 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.923 16.719 20 1.057 1.143 53 36 47 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 47 17 0.934 17.061 21 1.051 1.143 33 8 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.950 17.346 39 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 39 15 0.986 17.846 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.004 17.761 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.136 16.767 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.189 17.274 23 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 23 11 1.214 17.532 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.233 17.880 15 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 15 9 1.265 18.286 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.295 18.517 7 -- -- -- -- -- -- -- -- -- 7 7 1.334 18.738 3 -- -- -- -- -- -- -- 3 6 1.393 19.208* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.419* 19.206 4 1.388 18.482 Control Rod Density: % 5.05 3 1.282 16.865 2 0.982 12.625 k-effective: 0.99983 Bottom 1 0.261 3.640 Void Fraction: 0.480 Core Delta-P: psia 23.121  % AXIAL TILT -19.460 -7.062 Core Plate Delta-P: psia 18.563 AVG BOT 8ft/12ft 1.1078 1.0423 Coolant Temp: Deg-F 548.4 In Channel Flow: Mlb/hr 85.64 Active Channel Flow: Mlb/hr 82.51 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.282 16 29 32 1.708 0.831 19 29 28 8.25 0.702 27.1 16 31 14 4 11.04 0.783 1.2 19 31 50 4 1.266 16 23 24 1.710 0.830 19 27 32 7.93 0.691 29.7 16 31 10 4 10.93 0.776 1.2 19 31 16 4 1.261 16 31 48 1.726 0.823 19 21 38 7.76 0.685 31.2 16 51 30 5 10.92 0.774 1.2 19 49 32 4 1.256 19 21 38 1.733 0.820 19 23 22 8.04 0.681 26.6 16 33 50 4 10.91 0.774 1.2 19 33 10 4 1.256 19 23 22 1.739 0.817 16 31 32 7.72 0.679 30.6 16 47 30 5 10.82 0.767 1.2 19 51 34 4 1.256 16 21 40 1.743 0.815 19 23 36 7.97 0.674 26.4 16 49 34 4 10.81 0.767 1.2 19 33 48 4 1.252 19 29 16 1.744 0.814 19 25 38 7.47 0.673 33.4 17 35 10 5 10.81 0.766 1.2 20 25 8 5 1.250 19 29 28 1.748 0.812 19 29 46 7.56 0.673 32.0 17 51 36 5 10.78 0.765 1.2 20 7 26 4 1.248 19 27 30 1.755 0.809 19 19 36 7.76 0.663 27.5 16 33 16 5 10.78 0.765 1.2 19 31 8 4 1.245 16 27 16 1.755 0.809 19 45 30 7.63 0.655 28.0 16 21 40 5 10.67 0.757 1.2 20 33 56 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-7 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 16105.5 Exposure: MWd/MTU (GWd) 1000.0 ( 138.35 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.189 3.523 13 0.490 0.512 59 34 Inlet Subcooling: Btu/lbm -27.75 24 0.535 9.896 14 0.431 0.526 59 30 Flow: Mlb/hr 99.73 ( 97.30 %) 23 0.692 12.768 15 0.540 0.543 3 18 22 0.793 14.268 16 1.113 1.276 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.862 15.397 17 0.879 1.166 51 36 59 -- -- -- -- -- -- -- 59 20 0.907 16.252 18 1.120 1.218 21 36 55 -- -- -- -- -- -- -- -- -- 55 19 0.923 16.697 19 1.153 1.260 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.933 17.220 20 1.059 1.146 53 36 47 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 47 17 0.941 17.567 21 1.054 1.148 51 38 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.956 17.860 39 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 39 15 0.990 18.379 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.007 18.303 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.138 17.298 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.190 17.829 23 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 23 11 1.214 18.099 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.234 18.456 15 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 15 9 1.264 18.878 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.292 19.122 7 -- -- -- -- -- -- -- -- -- 7 7 1.327 19.361 3 -- -- -- -- -- -- -- 3 6 1.381 19.857 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.400* 19.865* 4 1.361 19.125 Control Rod Density: % 5.05 3 1.252 17.458 2 0.960 13.079 k-effective: 0.99976 Bottom 1 0.258 3.768 Void Fraction: 0.474 Core Delta-P: psia 23.873  % AXIAL TILT -18.459 -7.225 Core Plate Delta-P: psia 19.315 AVG BOT 8ft/12ft 1.1017 1.0431 Coolant Temp: Deg-F 548.4 In Channel Flow: Mlb/hr 87.69 Active Channel Flow: Mlb/hr 84.52 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.276 16 29 32 1.722 0.825 19 29 28 8.10 0.694 27.9 16 31 14 4 10.88 0.772 2.4 19 31 50 4 1.263 16 23 24 1.725 0.823 19 27 32 7.75 0.687 31.7 16 31 10 5 10.75 0.762 2.4 19 49 32 5 1.260 19 21 38 1.738 0.817 19 21 38 7.66 0.682 32.0 16 51 30 5 10.75 0.762 2.3 19 31 16 5 1.260 19 23 22 1.745 0.814 19 23 22 7.81 0.677 29.0 16 33 50 5 10.74 0.762 2.4 19 33 10 5 1.259 16 31 48 1.756 0.809 19 23 36 7.63 0.675 31.4 16 47 30 5 10.70 0.759 2.4 20 25 8 5 1.257 19 29 16 1.756 0.809 16 31 32 7.38 0.671 34.2 17 35 10 5 10.67 0.757 2.4 20 7 26 5 1.253 16 21 40 1.756 0.809 19 25 38 7.75 0.670 28.8 16 49 34 5 10.66 0.756 2.3 19 51 34 5 1.251 19 29 28 1.760 0.807 19 29 46 7.47 0.670 32.8 17 51 36 5 10.65 0.756 2.4 19 33 48 4 1.250 19 27 30 1.766 0.804 19 19 36 7.64 0.658 28.3 16 33 16 5 10.63 0.754 2.3 19 31 8 4 1.245 19 15 32 1.766 0.804 19 45 30 7.51 0.649 28.8 16 21 40 5 10.50 0.745 2.3 19 7 32 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-8 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 16605.4 Exposure: MWd/MTU (GWd) 1500.0 ( 207.52 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.194 3.624 13 0.486 0.507 59 34 Inlet Subcooling: Btu/lbm -27.78 24 0.545 10.186 14 0.428 0.520 59 30 Flow: Mlb/hr 99.63 ( 97.20 %) 23 0.704 13.142 15 0.536 0.539 3 18 22 0.805 14.699 16 1.110 1.269 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.873 15.865 17 0.876 1.165 51 36 59 -- -- -- -- -- -- -- 59 20 0.916 16.744 18 1.115 1.214 21 36 55 -- -- -- -- -- -- -- -- -- 55 19 0.931 17.197 19 1.157 1.264 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.939 17.725 20 1.063 1.150 53 36 47 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 47 17 0.946 18.077 21 1.058 1.154 51 38 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.959 18.377 39 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 39 15 0.993 18.914 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.008 18.847 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.137 17.829 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.188 18.385 23 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 23 11 1.212 18.666 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.231 19.033 15 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 15 9 1.261 19.469 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.288 19.726 7 -- -- -- -- -- -- -- -- -- 7 7 1.321 19.980 3 -- -- -- -- -- -- -- 3 6 1.371 20.500 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.387* 20.517* 4 1.345 19.757 Control Rod Density: % 5.05 3 1.237 18.040 2 0.950 13.525 k-effective: 0.99966 Bottom 1 0.258 3.895 Void Fraction: 0.472 Core Delta-P: psia 23.813  % AXIAL TILT -17.750 -7.353 Core Plate Delta-P: psia 19.255 AVG BOT 8ft/12ft 1.0973 1.0436 Coolant Temp: Deg-F 548.4 In Channel Flow: Mlb/hr 87.61 Active Channel Flow: Mlb/hr 84.45 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.269 16 29 32 1.723 0.824 19 29 34 7.90 0.691 30.0 16 31 14 5 10.77 0.764 3.6 19 31 50 4 1.264 19 21 38 1.725 0.823 19 27 32 7.67 0.685 32.5 16 31 10 5 10.64 0.755 3.5 19 49 32 5 1.264 19 23 22 1.735 0.818 19 21 38 7.58 0.680 32.8 16 51 30 5 10.64 0.754 3.5 19 33 10 5 1.261 19 29 16 1.742 0.815 19 23 22 7.72 0.674 29.8 16 33 50 5 10.62 0.753 3.5 19 31 16 5 1.258 16 23 24 1.754 0.810 19 23 36 7.54 0.672 32.2 16 47 30 5 10.61 0.753 3.5 20 25 8 5 1.256 16 31 48 1.754 0.809 19 25 38 7.31 0.669 35.0 17 35 10 5 10.58 0.750 3.5 20 7 26 5 1.253 19 29 28 1.755 0.809 19 29 46 7.40 0.668 33.6 17 51 36 5 10.55 0.749 3.5 19 51 34 5 1.251 19 27 30 1.760 0.807 16 31 32 7.66 0.667 29.6 16 49 34 5 10.54 0.748 3.5 19 33 48 5 1.249 19 15 32 1.762 0.806 19 45 30 7.55 0.655 29.1 16 33 16 5 10.53 0.747 3.5 19 31 8 4 1.249 16 21 40 1.763 0.805 19 19 36 7.41 0.645 29.6 16 21 40 5 10.42 0.739 3.4 19 7 32 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-9 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 16925.0 Exposure: MWd/MTU (GWd) 1819.6 ( 251.74 ) Delta E: MWd/MTU, (GWd) 319.6 ( 44.22 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.197 3.690 13 0.483 0.503 59 34 Inlet Subcooling: Btu/lbm -27.81 24 0.552 10.373 14 0.425 0.517 59 30 Flow: Mlb/hr 99.53 ( 97.10 %) 23 0.712 13.384 15 0.534 0.537 3 18 22 0.813 14.978 16 1.108 1.264 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.880 16.167 17 0.874 1.165 51 36 59 -- -- -- -- -- -- -- 59 20 0.923 17.061 18 1.112 1.211 21 36 55 -- -- -- -- -- -- -- -- -- 55 19 0.936 17.519 19 1.160 1.266 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.943 18.050 20 1.065 1.153 53 36 47 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 47 17 0.950 18.404 21 1.061 1.158 51 38 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.962 18.708 39 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 39 15 0.994 19.257 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.009 19.194 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.137 18.169 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.187 18.740 23 -- -- -- 12 -- -- -- 8 -- -- -- 12 -- -- -- 23 11 1.210 19.028 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.229 19.401 15 -- -- -- -- 12 -- -- -- 12 -- -- -- -- 15 9 1.258 19.845 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.284 20.111 7 -- -- -- -- -- -- -- -- -- 7 7 1.316 20.374 3 -- -- -- -- -- -- -- 3 6 1.365 20.909 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.379* 20.931* 4 1.335 20.158 Control Rod Density: % 5.05 3 1.227 18.409 2 0.944 13.809 k-effective: 0.99958 Bottom 1 0.257 3.975 Void Fraction: 0.470 Core Delta-P: psia 23.761  % AXIAL TILT -17.268 -7.418 Core Plate Delta-P: psia 19.203 AVG BOT 8ft/12ft 1.0942 1.0438 Coolant Temp: Deg-F 548.4 In Channel Flow: Mlb/hr 87.53 Active Channel Flow: Mlb/hr 84.37 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00007

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.266 19 21 38 1.723 0.824 19 29 34 7.84 0.688 30.6 16 31 14 5 10.70 0.759 4.3 19 31 50 4 1.266 19 23 22 1.725 0.823 19 27 32 7.61 0.684 33.0 16 31 10 5 10.57 0.750 4.2 19 49 32 5 1.264 16 29 32 1.733 0.819 19 21 38 7.53 0.679 33.3 16 51 30 5 10.56 0.749 4.2 19 33 10 5 1.264 19 29 16 1.740 0.816 19 23 22 7.66 0.672 30.4 16 33 50 5 10.55 0.749 4.2 20 25 8 5 1.255 16 23 24 1.752 0.810 19 23 36 7.48 0.670 32.7 16 47 30 5 10.54 0.748 4.2 19 31 16 5 1.254 19 29 28 1.752 0.810 19 29 46 7.26 0.667 35.5 17 35 10 5 10.52 0.746 4.2 20 7 26 5 1.253 16 31 48 1.753 0.810 19 25 38 7.35 0.667 34.1 17 51 36 5 10.48 0.743 4.2 19 51 34 5 1.252 19 27 30 1.759 0.807 19 45 30 7.61 0.666 30.1 16 49 34 5 10.48 0.743 4.2 19 33 48 5 1.252 19 45 30 1.761 0.806 19 19 36 7.49 0.652 29.6 16 33 16 5 10.47 0.743 4.2 19 31 8 5 1.247 19 23 26 1.761 0.806 19 19 40 7.34 0.642 30.1 16 21 40 5 10.37 0.735 4.1 19 7 32 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-10 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 16925.7 Exposure: MWd/MTU (GWd) 1820.3 ( 251.84 ) Delta E: MWd/MTU, (GWd) 0.7 ( 0.10 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.180 3.690 13 0.486 0.507 59 34 Inlet Subcooling: Btu/lbm -30.88 24 0.503 10.374 14 0.429 0.520 59 30 Flow: Mlb/hr 90.30 ( 88.10 %) 23 0.648 13.385 15 0.540 0.546 3 18 22 0.741 14.978 16 1.105 1.255 45 28 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.804 16.168 17 0.907 1.233 13 40 59 -- -- -- -- -- -- -- 59 20 0.856 17.061 18 0.997 1.188 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.885 17.520 19 1.151 1.284 47 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.910 18.050 20 1.088 1.206 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.928 18.404 21 1.085 1.226 51 38 43 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 43 16 0.949 18.709 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 0.989 19.258 35 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 35 14 1.009 19.195 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.140 18.170 27 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 27 12 1.192 18.740 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.219 19.029 19 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 19 10 1.241 19.402 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.274 19.846 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.305 20.111 7 -- -- -- -- -- -- -- -- -- 7 7 1.344 20.375 3 -- -- -- -- -- -- -- 3 6 1.404 20.910 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.435* 20.932* 4 1.413 20.159 Control Rod Density: % 4.28 3 1.321 18.410 2 1.029 13.809 k-effective: 0.99957 Bottom 1 0.281 3.976 Void Fraction: 0.495 Core Delta-P: psia 20.773  % AXIAL TILT -21.190 -7.419 Core Plate Delta-P: psia 16.220 AVG BOT 8ft/12ft 1.1208 1.0438 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 79.16 Active Channel Flow: Mlb/hr 76.20 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.3 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.284 19 47 28 1.682 0.844 20 51 22 8.50 0.733 28.5 16 49 28 4 11.45 0.812 3.9 21 51 24 4 1.269 19 49 30 1.686 0.842 19 47 28 8.10 0.729 33.2 17 51 26 4 11.44 0.812 4.0 19 49 26 4 1.267 19 13 38 1.688 0.841 19 49 30 8.31 0.725 29.7 16 49 24 4 11.44 0.811 4.3 19 49 32 4 1.267 19 49 26 1.694 0.838 19 11 36 8.11 0.723 32.3 16 51 30 4 11.42 0.810 4.2 19 51 34 4 1.266 19 45 30 1.702 0.834 19 13 38 8.15 0.723 31.6 16 47 30 4 11.42 0.810 4.2 20 7 26 4 1.265 19 15 36 1.717 0.827 19 45 30 8.13 0.719 31.3 16 47 26 4 11.39 0.808 4.0 19 47 34 4 1.255 16 45 28 1.718 0.827 19 15 36 7.78 0.704 33.8 17 13 22 5 11.28 0.800 3.9 20 51 40 4 1.252 16 47 26 1.736 0.818 19 45 22 8.07 0.700 29.1 16 45 28 4 11.22 0.796 3.4 19 49 40 4 1.249 19 15 40 1.738 0.817 19 29 34 7.52 0.692 35.5 17 35 10 5 11.21 0.795 3.0 19 47 24 4 1.248 16 49 34 1.754 0.809 20 21 52 7.62 0.691 34.0 16 11 20 5 11.16 0.791 4.1 19 15 32 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-11 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 17405.4 Exposure: MWd/MTU (GWd) 2300.0 ( 318.20 ) Delta E: MWd/MTU, (GWd) 479.7 ( 66.36 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.185 3.783 13 0.482 0.502 59 34 Inlet Subcooling: Btu/lbm -30.84 24 0.514 10.635 14 0.425 0.515 59 30 Flow: Mlb/hr 90.40 ( 88.20 %) 23 0.661 13.721 15 0.537 0.542 3 18 22 0.753 15.365 16 1.102 1.250 45 28 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.815 16.587 17 0.904 1.229 13 40 59 -- -- -- -- -- -- -- 59 20 0.867 17.507 18 0.994 1.185 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.894 17.980 19 1.155 1.287 47 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.918 18.524 20 1.091 1.211 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.935 18.887 21 1.088 1.231 51 38 43 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 43 16 0.955 19.202 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 0.993 19.770 35 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 35 14 1.012 19.718 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.141 18.681 27 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 27 12 1.192 19.275 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.217 19.575 19 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 19 10 1.239 19.958 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.271 20.417 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.300 20.696 7 -- -- -- -- -- -- -- -- -- 7 7 1.337 20.977 3 -- -- -- -- -- -- -- 3 6 1.393 21.538 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.420* 21.572* 4 1.394 20.789 Control Rod Density: % 4.28 3 1.301 18.998 2 1.014 14.268 k-effective: 0.99943 Bottom 1 0.279 4.108 Void Fraction: 0.493 Core Delta-P: psia 20.780  % AXIAL TILT -20.345 -7.604 Core Plate Delta-P: psia 16.227 AVG BOT 8ft/12ft 1.1157 1.0448 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 79.27 Active Channel Flow: Mlb/hr 76.31 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.3 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.287 19 47 28 1.682 0.844 20 51 22 8.35 0.726 29.4 16 49 28 4 11.30 0.801 5.1 21 51 24 4 1.272 19 49 30 1.686 0.842 19 47 28 7.98 0.724 34.0 17 51 26 4 11.27 0.799 5.4 20 7 26 4 1.270 19 13 38 1.688 0.841 19 49 30 8.18 0.719 30.6 16 49 24 4 11.26 0.799 5.4 19 49 32 4 1.270 19 49 26 1.693 0.839 19 11 36 7.98 0.718 33.1 16 51 30 4 11.26 0.798 5.2 19 49 26 4 1.269 19 45 30 1.699 0.836 19 13 38 8.01 0.716 32.4 16 47 30 4 11.24 0.797 5.4 19 51 34 4 1.269 19 15 36 1.715 0.828 19 15 36 7.99 0.712 32.1 16 47 26 4 11.24 0.797 5.2 19 47 34 4 1.253 19 15 40 1.716 0.827 19 45 30 7.68 0.700 34.6 17 13 22 5 11.15 0.791 5.0 20 51 40 4 1.250 16 45 28 1.733 0.819 19 45 22 7.85 0.693 31.1 16 45 28 5 11.07 0.785 4.2 19 47 24 4 1.247 16 47 26 1.736 0.818 19 29 34 7.44 0.689 36.2 17 35 10 5 11.05 0.784 4.5 19 49 40 4 1.244 16 49 34 1.748 0.813 20 21 52 7.54 0.689 34.8 16 11 20 5 10.97 0.778 5.2 19 15 32 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-12 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 17905.4 Exposure: MWd/MTU (GWd) 2800.0 ( 387.38 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.190 3.882 13 0.478 0.497 59 34 Inlet Subcooling: Btu/lbm -30.62 24 0.525 10.913 14 0.421 0.509 59 30 Flow: Mlb/hr 91.02 ( 88.80 %) 23 0.674 14.078 15 0.533 0.539 3 18 22 0.766 15.775 16 1.100 1.247 45 28 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.828 17.031 17 0.901 1.227 13 40 59 -- -- -- -- -- -- -- 59 20 0.878 17.978 18 0.991 1.182 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.904 18.465 19 1.159 1.292 47 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.927 19.021 20 1.095 1.216 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.942 19.393 21 1.093 1.237 51 38 43 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 43 16 0.961 19.719 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 0.997 20.307 35 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 35 14 1.014 20.265 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.142 19.214 27 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 27 12 1.191 19.831 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.215 20.143 19 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 19 10 1.237 20.537 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.268 21.011 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.296 21.303 7 -- -- -- -- -- -- -- -- -- 7 7 1.331 21.600 3 -- -- -- -- -- -- -- 3 6 1.384 22.187 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.406* 22.233* 4 1.375 21.437 Control Rod Density: % 4.28 3 1.279 19.601 2 0.996 14.738 k-effective: 0.99940 Bottom 1 0.276 4.244 Void Fraction: 0.489 Core Delta-P: psia 20.956  % AXIAL TILT -19.482 -7.763 Core Plate Delta-P: psia 16.402 AVG BOT 8ft/12ft 1.1105 1.0456 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 79.84 Active Channel Flow: Mlb/hr 76.87 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.3 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.292 19 47 28 1.685 0.843 20 51 22 7.83 0.722 35.7 17 51 26 5 11.17 0.792 6.3 21 51 24 4 1.277 19 49 30 1.690 0.840 19 47 28 8.23 0.721 30.2 16 49 28 4 11.14 0.790 6.6 20 7 26 4 1.276 19 49 26 1.691 0.840 19 49 30 7.98 0.715 32.7 16 49 24 5 11.13 0.789 6.6 19 49 32 4 1.276 19 13 38 1.695 0.838 19 11 36 7.87 0.714 33.9 16 51 30 4 11.11 0.788 6.4 19 47 34 4 1.273 19 45 30 1.702 0.834 19 13 38 7.82 0.712 34.4 16 47 30 5 11.10 0.787 6.4 19 49 26 4 1.273 19 15 36 1.715 0.828 19 15 36 7.83 0.709 34.0 16 47 26 5 11.09 0.786 6.6 19 51 34 4 1.258 19 15 40 1.717 0.827 19 45 30 7.60 0.698 35.4 17 13 22 5 11.05 0.784 6.2 20 51 40 4 1.247 16 45 28 1.733 0.819 19 45 22 7.76 0.690 31.9 16 45 28 5 10.96 0.777 5.3 19 47 24 4 1.244 16 47 26 1.742 0.815 19 29 34 7.47 0.688 35.6 16 11 20 5 10.91 0.774 5.7 19 49 40 4 1.242 16 49 34 1.746 0.813 20 21 52 7.37 0.688 37.0 17 35 10 5 10.83 0.768 6.5 19 7 32 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-13 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 18405.3 Exposure: MWd/MTU (GWd) 3300.0 ( 456.55 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.195 3.984 13 0.474 0.492 59 34 Inlet Subcooling: Btu/lbm -30.51 24 0.535 11.197 14 0.418 0.504 59 30 Flow: Mlb/hr 91.33 ( 89.10 %) 23 0.687 14.442 15 0.530 0.535 3 18 22 0.778 16.191 16 1.097 1.243 45 28 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.838 17.480 17 0.897 1.225 13 40 59 -- -- -- -- -- -- -- 59 20 0.887 18.454 18 0.988 1.180 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.912 18.955 19 1.163 1.296 47 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.933 19.523 20 1.098 1.222 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.948 19.903 21 1.098 1.244 51 38 43 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 43 16 0.965 20.239 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.000 20.846 35 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 35 14 1.015 20.812 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.141 19.747 27 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 27 12 1.188 20.387 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.212 20.710 19 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 19 10 1.234 21.115 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.265 21.603 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.292 21.909 7 -- -- -- -- -- -- -- -- -- 7 7 1.325 22.222 3 -- -- -- -- -- -- -- 3 6 1.376 22.833 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.395* 22.889* 4 1.361 22.077 Control Rod Density: % 4.28 3 1.262 20.195 2 0.983 15.201 k-effective: 0.99926 Bottom 1 0.274 4.380 Void Fraction: 0.486 Core Delta-P: psia 21.035  % AXIAL TILT -18.752 -7.891 Core Plate Delta-P: psia 16.482 AVG BOT 8ft/12ft 1.1059 1.0462 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 80.13 Active Channel Flow: Mlb/hr 77.16 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.296 19 47 28 1.684 0.843 20 51 22 7.77 0.722 36.6 17 51 26 5 11.09 0.787 7.5 21 51 24 4 1.282 19 49 30 1.691 0.840 19 47 28 8.14 0.718 31.1 16 49 28 4 11.04 0.783 7.8 20 7 26 4 1.281 19 49 26 1.691 0.840 19 49 30 7.93 0.716 33.6 16 49 24 5 11.03 0.782 7.8 19 49 32 4 1.281 19 13 38 1.695 0.838 19 11 36 7.73 0.713 35.8 16 51 30 5 11.01 0.781 7.6 19 47 34 4 1.277 19 15 36 1.703 0.834 19 13 38 7.75 0.711 35.2 16 47 30 5 11.00 0.780 7.4 19 49 26 4 1.277 19 45 30 1.713 0.829 19 15 36 7.76 0.709 34.8 16 47 26 5 10.98 0.779 7.4 20 51 40 4 1.262 19 15 40 1.716 0.828 19 45 30 7.53 0.697 36.2 17 13 22 5 10.97 0.778 7.7 19 9 34 4 1.246 19 51 34 1.730 0.821 19 45 22 7.43 0.689 36.4 16 11 20 5 10.87 0.771 6.5 19 47 24 4 1.245 19 13 42 1.743 0.815 20 21 52 7.68 0.688 32.7 16 45 28 5 10.80 0.766 6.8 19 49 40 4 1.245 19 11 40 1.745 0.814 19 29 34 7.31 0.687 37.8 17 35 10 5 10.75 0.762 7.6 19 7 32 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-14 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 18905.3 Exposure: MWd/MTU (GWd) 3800.0 ( 525.73 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.199 4.089 13 0.470 0.487 59 34 Inlet Subcooling: Btu/lbm -30.29 24 0.546 11.486 14 0.414 0.498 59 30 Flow: Mlb/hr 91.94 ( 89.70 %) 23 0.699 14.813 15 0.526 0.532 3 18 22 0.789 16.614 16 1.094 1.240 45 28 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.848 17.935 17 0.894 1.222 13 40 59 -- -- -- -- -- -- -- 59 20 0.896 18.935 18 0.985 1.177 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.919 19.449 19 1.167 1.301 47 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.940 20.029 20 1.102 1.228 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.953 20.415 21 1.103 1.250 51 38 43 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 43 16 0.969 20.761 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.002 21.386 35 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 35 14 1.015 21.360 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.139 20.280 27 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 27 12 1.186 20.941 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.209 21.276 19 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 19 10 1.231 21.691 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.262 22.194 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.289 22.512 7 -- -- -- -- -- -- -- -- -- 7 7 1.320 22.840 3 -- -- -- -- -- -- -- 3 6 1.369 23.475 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.386* 23.539* 4 1.347 22.710 Control Rod Density: % 4.28 3 1.246 20.782 2 0.969 15.658 k-effective: 0.99917 Bottom 1 0.271 4.514 Void Fraction: 0.483 Core Delta-P: psia 21.217  % AXIAL TILT -18.063 -7.993 Core Plate Delta-P: psia 16.664 AVG BOT 8ft/12ft 1.1015 1.0467 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 80.69 Active Channel Flow: Mlb/hr 77.71 Total Bypass Flow (%): 12.2 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.301 19 47 28 1.685 0.843 20 51 22 7.72 0.723 37.4 17 51 26 5 11.02 0.782 8.7 21 51 24 4 1.287 19 49 30 1.694 0.838 19 49 30 7.94 0.717 33.5 16 49 28 5 10.97 0.778 9.0 20 7 26 4 1.287 19 49 26 1.695 0.838 19 47 28 7.87 0.717 34.4 16 49 24 5 10.95 0.776 9.0 19 49 32 4 1.286 19 13 38 1.697 0.837 19 11 36 7.68 0.714 36.6 16 51 30 5 10.93 0.775 8.6 19 49 26 4 1.281 19 15 36 1.706 0.832 19 13 38 7.68 0.710 36.0 16 47 30 5 10.93 0.775 8.6 20 51 40 5 1.281 19 45 30 1.714 0.829 19 15 36 7.70 0.709 35.6 16 47 26 5 10.92 0.775 8.8 19 47 34 4 1.267 19 15 40 1.717 0.827 19 45 30 7.47 0.697 37.0 17 13 22 5 10.90 0.773 8.9 19 9 34 4 1.251 19 51 34 1.731 0.820 19 45 22 7.39 0.691 37.2 16 11 20 5 10.80 0.766 7.7 19 47 24 4 1.251 19 11 40 1.744 0.814 20 21 52 7.62 0.688 33.5 16 45 28 5 10.71 0.760 8.1 19 49 40 5 1.251 19 13 42 1.751 0.811 19 29 34 7.25 0.687 38.6 17 35 10 5 10.67 0.757 8.8 19 7 32 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-15 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 19405.2 Exposure: MWd/MTU (GWd) 4300.0 ( 594.90 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.204 4.195 13 0.466 0.482 59 34 Inlet Subcooling: Btu/lbm -30.18 24 0.555 11.781 14 0.410 0.493 59 30 Flow: Mlb/hr 92.25 ( 90.00 %) 23 0.710 15.190 15 0.523 0.528 3 18 22 0.799 17.043 16 1.091 1.236 45 28 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.857 18.395 17 0.891 1.220 13 40 59 -- -- -- -- -- -- -- 59 20 0.903 19.420 18 0.981 1.175 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.925 19.947 19 1.172 1.306 47 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.945 20.537 20 1.105 1.234 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.956 20.930 21 1.108 1.257 51 38 43 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 43 16 0.971 21.284 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.004 21.928 35 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 35 14 1.015 21.908 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.137 20.811 27 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 27 12 1.182 21.494 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.205 21.840 19 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 19 10 1.227 22.266 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.258 22.783 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.285 23.114 7 -- -- -- -- -- -- -- -- -- 7 7 1.316 23.457 3 -- -- -- -- -- -- -- 3 6 1.364 24.114 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.378* 24.185* 4 1.338 23.338 Control Rod Density: % 4.28 3 1.235 21.363 2 0.960 16.109 k-effective: 0.99904 Bottom 1 0.270 4.647 Void Fraction: 0.481 Core Delta-P: psia 21.302  % AXIAL TILT -17.485 -8.074 Core Plate Delta-P: psia 16.749 AVG BOT 8ft/12ft 1.0976 1.0471 Coolant Temp: Deg-F 548.1 In Channel Flow: Mlb/hr 80.98 Active Channel Flow: Mlb/hr 78.00 Total Bypass Flow (%): 12.2 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00007

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.306 19 47 28 1.685 0.843 20 51 22 7.69 0.726 38.2 17 51 26 5 10.98 0.779 9.8 21 51 24 4 1.292 19 49 26 1.696 0.837 19 49 30 7.83 0.719 35.3 16 49 24 5 10.92 0.774 10.2 20 7 26 4 1.292 19 49 30 1.697 0.837 19 47 28 7.90 0.718 34.3 16 49 28 5 10.91 0.773 9.8 20 51 40 5 1.291 19 13 38 1.697 0.837 19 11 36 7.63 0.715 37.4 16 51 30 5 10.89 0.773 10.1 19 49 32 4 1.285 19 15 36 1.708 0.832 19 13 38 7.63 0.711 36.8 16 47 30 5 10.89 0.772 9.7 19 49 26 4 1.285 19 45 30 1.713 0.829 19 15 36 7.64 0.709 36.4 16 47 26 5 10.86 0.770 10.0 19 47 34 4 1.271 19 15 40 1.716 0.827 19 45 30 7.42 0.698 37.8 17 13 22 5 10.86 0.770 10.1 19 9 34 4 1.258 19 11 40 1.729 0.821 19 45 22 7.36 0.693 38.0 16 11 20 5 10.75 0.762 8.8 19 47 24 4 1.257 19 51 34 1.744 0.814 20 21 52 7.21 0.689 39.4 17 35 10 5 10.64 0.755 9.2 19 49 40 5 1.257 21 51 38 1.753 0.810 19 11 40 7.55 0.687 34.4 16 45 28 5 10.63 0.754 9.6 20 9 20 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-16 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 19668.2 Exposure: MWd/MTU (GWd) 4563.0 ( 631.29 ) Delta E: MWd/MTU, (GWd) 263.0 ( 36.39 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.206 4.252 13 0.464 0.479 59 34 Inlet Subcooling: Btu/lbm -30.11 24 0.560 11.938 14 0.408 0.490 59 30 Flow: Mlb/hr 92.45 ( 90.20 %) 23 0.716 15.391 15 0.521 0.526 3 18 22 0.804 17.270 16 1.090 1.234 45 28 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.861 18.639 17 0.889 1.219 13 40 59 -- -- -- -- -- -- -- 59 20 0.907 19.677 18 0.979 1.173 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.928 20.210 19 1.174 1.308 47 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.947 20.805 20 1.107 1.237 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.958 21.202 21 1.110 1.260 51 38 43 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 43 16 0.972 21.560 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.004 22.213 35 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 35 14 1.014 22.196 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.135 21.090 27 -- -- -- -- -- 6 -- -- -- 6 -- -- -- -- -- 27 12 1.179 21.784 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.202 22.136 19 -- -- -- -- -- 12 -- 14 -- 12 -- -- -- -- -- 19 10 1.225 22.567 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.256 23.093 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.283 23.430 7 -- -- -- -- -- -- -- -- -- 7 7 1.314 23.781 3 -- -- -- -- -- -- -- 3 6 1.362 24.450 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.376* 24.524* 4 1.335 23.667 Control Rod Density: % 4.28 3 1.231 21.666 2 0.956 16.345 k-effective: 0.99899 Bottom 1 0.269 4.717 Void Fraction: 0.479 Core Delta-P: psia 21.363  % AXIAL TILT -17.226 -8.109 Core Plate Delta-P: psia 16.810 AVG BOT 8ft/12ft 1.0958 1.0472 Coolant Temp: Deg-F 548.1 In Channel Flow: Mlb/hr 81.17 Active Channel Flow: Mlb/hr 78.18 Total Bypass Flow (%): 12.2 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.308 19 47 28 1.685 0.842 20 51 22 7.67 0.728 38.6 17 51 26 5 10.97 0.778 10.5 21 51 24 4 1.295 19 49 26 1.697 0.837 19 49 30 7.82 0.721 35.7 16 49 24 5 10.90 0.773 10.4 20 51 40 5 1.295 19 49 30 1.698 0.836 19 11 36 7.87 0.719 34.8 16 49 28 5 10.90 0.773 10.8 20 7 26 4 1.293 19 13 38 1.699 0.836 19 47 28 7.62 0.717 37.9 16 51 30 5 10.87 0.771 10.3 19 49 26 4 1.288 19 15 36 1.709 0.831 19 13 38 7.60 0.711 37.3 16 47 30 5 10.87 0.771 10.8 19 49 32 4 1.287 19 45 30 1.713 0.829 19 15 36 7.62 0.710 36.8 16 47 26 5 10.84 0.769 10.7 19 9 34 4 1.274 19 15 40 1.716 0.827 19 45 30 7.40 0.699 38.2 17 13 22 5 10.84 0.768 10.6 19 47 34 4 1.261 19 11 40 1.729 0.821 19 45 22 7.35 0.695 38.4 16 11 20 5 10.73 0.761 9.6 19 47 24 5 1.260 19 51 34 1.744 0.814 20 21 52 7.20 0.690 39.8 17 35 10 5 10.63 0.754 10.2 20 9 20 5 1.260 21 51 38 1.753 0.810 19 11 40 7.53 0.688 34.8 16 45 28 5 10.62 0.753 9.8 19 49 40 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-17 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 19669.0 Exposure: MWd/MTU (GWd) 4563.7 ( 631.39 ) Delta E: MWd/MTU, (GWd) 0.7 ( 0.10 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.187 4.252 13 0.457 0.467 59 34 Inlet Subcooling: Btu/lbm -30.11 24 0.509 11.938 14 0.402 0.477 59 30 Flow: Mlb/hr 92.45 ( 90.20 %) 23 0.650 15.391 15 0.515 0.518 3 18 22 0.728 17.271 16 1.081 1.264 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.777 18.640 17 0.879 1.184 13 40 59 -- -- -- -- -- -- -- 59 20 0.823 19.678 18 1.074 1.202 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.855 20.210 19 1.182 1.317 29 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.895 20.806 20 1.098 1.210 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.927 21.203 21 1.096 1.224 51 38 43 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 43 16 0.959 21.561 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.004 22.213 35 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 35 14 1.025 22.197 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.155 21.091 27 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 27 12 1.206 21.785 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.234 22.136 19 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 19 10 1.260 22.568 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.294 23.093 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.324 23.431 7 -- -- -- -- -- -- -- -- -- 7 7 1.358 23.782 3 -- -- -- -- -- -- -- 3 6 1.411 24.451 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.431* 24.525* 4 1.397 23.668 Control Rod Density: % 3.78 3 1.295 21.667 2 1.009 16.345 k-effective: 0.99914 Bottom 1 0.285 4.717 Void Fraction: 0.489 Core Delta-P: psia 21.520  % AXIAL TILT -21.652 -8.109 Core Plate Delta-P: psia 16.968 AVG BOT 8ft/12ft 1.1273 1.0472 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 81.08 Active Channel Flow: Mlb/hr 78.07 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.3 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.317 19 29 28 1.656 0.858 19 29 28 7.51 0.712 38.6 17 51 26 5 10.77 0.764 10.2 20 39 52 5 1.316 19 27 30 1.657 0.857 19 27 32 7.40 0.709 39.8 17 35 10 5 10.77 0.764 10.4 20 51 40 5 1.284 19 25 28 1.726 0.823 19 25 28 7.66 0.706 35.7 16 49 24 5 10.73 0.761 10.4 21 51 24 5 1.283 19 27 26 1.731 0.820 16 31 32 7.66 0.704 35.5 16 37 12 5 10.73 0.761 10.1 21 37 10 4 1.264 19 17 42 1.732 0.820 19 27 26 7.44 0.696 37.3 16 31 10 5 10.68 0.758 10.2 20 17 50 5 1.264 16 29 32 1.749 0.812 20 51 22 7.40 0.696 37.9 16 51 30 5 10.65 0.755 10.7 20 7 26 5 1.264 19 19 44 1.753 0.810 20 39 52 7.35 0.696 38.4 16 11 20 5 10.65 0.755 10.0 20 19 52 5 1.256 19 17 46 1.753 0.810 19 23 30 7.38 0.695 38.0 16 41 12 5 10.65 0.755 10.5 20 25 8 5 1.256 19 15 44 1.763 0.806 19 29 38 7.33 0.694 38.4 16 17 48 5 10.64 0.755 10.3 20 49 18 5 1.243 19 13 42 1.776 0.800 19 17 46 7.35 0.693 38.0 17 39 14 5 10.63 0.754 10.2 20 51 20 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-18 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 20105.2 Exposure: MWd/MTU (GWd) 5000.0 ( 691.75 ) Delta E: MWd/MTU, (GWd) 436.3 ( 60.36 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.191 4.340 13 0.454 0.463 59 34 Inlet Subcooling: Btu/lbm -29.89 24 0.517 12.178 14 0.399 0.473 59 30 Flow: Mlb/hr 93.07 ( 90.80 %) 23 0.659 15.697 15 0.512 0.515 3 18 22 0.737 17.616 16 1.078 1.256 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.785 19.007 17 0.876 1.183 13 40 59 -- -- -- -- -- -- -- 59 20 0.829 20.067 18 1.070 1.199 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.861 20.614 19 1.186 1.317 29 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.899 21.228 20 1.101 1.216 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.931 21.640 21 1.101 1.231 51 38 43 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 43 16 0.962 22.013 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.006 22.686 35 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 35 14 1.025 22.679 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.153 21.562 27 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 27 12 1.203 22.276 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.230 22.639 19 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 19 10 1.257 23.082 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.291 23.621 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.321 23.971 7 -- -- -- -- -- -- -- -- -- 7 7 1.355 24.335 3 -- -- -- -- -- -- -- 3 6 1.407 25.026 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.426* 25.108* 4 1.389 24.237 Control Rod Density: % 3.78 3 1.285 22.193 2 0.999 16.755 k-effective: 0.99908 Bottom 1 0.283 4.839 Void Fraction: 0.486 Core Delta-P: psia 21.711  % AXIAL TILT -21.140 -8.254 Core Plate Delta-P: psia 17.159 AVG BOT 8ft/12ft 1.1240 1.0481 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 81.64 Active Channel Flow: Mlb/hr 78.62 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.3 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.317 19 29 28 1.664 0.853 19 29 34 7.49 0.715 39.3 17 51 26 5 10.77 0.764 11.4 20 51 40 5 1.316 19 27 30 1.665 0.853 19 27 32 7.36 0.711 40.5 17 35 10 5 10.75 0.763 11.2 20 39 52 5 1.284 19 25 28 1.733 0.819 19 25 28 7.64 0.709 36.4 16 49 24 5 10.72 0.761 11.4 21 51 24 5 1.283 19 27 26 1.739 0.817 19 27 26 7.62 0.706 36.2 16 37 12 5 10.70 0.759 11.1 21 37 10 5 1.268 19 17 42 1.744 0.814 16 31 32 7.34 0.699 39.1 16 11 20 5 10.64 0.755 11.2 20 17 50 5 1.267 19 19 44 1.750 0.811 20 51 22 7.37 0.699 38.5 16 51 30 5 10.64 0.755 11.7 20 7 26 5 1.261 19 15 44 1.758 0.808 20 21 10 7.35 0.697 38.7 16 41 12 5 10.64 0.754 11.0 20 19 52 5 1.261 19 17 46 1.759 0.807 19 23 30 7.40 0.697 38.0 16 31 10 5 10.63 0.754 11.2 20 51 20 5 1.256 16 29 32 1.769 0.803 19 29 38 7.30 0.695 39.1 16 17 48 5 10.61 0.753 11.5 20 25 8 5 1.248 19 13 42 1.779 0.798 19 17 46 7.29 0.695 39.1 16 13 18 5 10.61 0.752 11.3 20 49 18 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-19 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 20605.1 Exposure: MWd/MTU (GWd) 5500.0 ( 760.92 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.194 4.441 13 0.450 0.458 59 34 Inlet Subcooling: Btu/lbm -29.93 24 0.525 12.457 14 0.396 0.467 59 30 Flow: Mlb/hr 92.97 ( 90.70 %) 23 0.668 16.052 15 0.508 0.511 3 18 22 0.744 18.015 16 1.075 1.249 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.791 19.432 17 0.873 1.181 13 40 59 -- -- -- -- -- -- -- 59 20 0.834 20.516 18 1.065 1.196 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.864 21.080 19 1.190 1.319 29 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.902 21.714 20 1.105 1.222 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.933 22.143 21 1.107 1.238 51 38 43 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 43 16 0.963 22.532 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.005 23.229 35 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 35 14 1.023 23.232 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.149 22.100 27 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 27 12 1.197 22.837 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.225 23.212 19 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 19 10 1.252 23.668 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.287 24.224 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.317 24.588 7 -- -- -- -- -- -- -- -- -- 7 7 1.352 24.968 3 -- -- -- -- -- -- -- 3 6 1.405 25.684 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.424* 25.775* 4 1.386 24.886 Control Rod Density: % 3.78 3 1.281 22.793 2 0.995 17.222 k-effective: 0.99895 Bottom 1 0.283 4.978 Void Fraction: 0.485 Core Delta-P: psia 21.668  % AXIAL TILT -20.786 -8.402 Core Plate Delta-P: psia 17.116 AVG BOT 8ft/12ft 1.1214 1.0489 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 81.56 Active Channel Flow: Mlb/hr 78.54 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.3 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.319 19 29 28 1.664 0.854 19 29 34 7.48 0.720 40.1 17 51 26 5 10.78 0.765 12.6 20 51 40 5 1.318 19 27 30 1.665 0.853 19 27 32 7.35 0.715 41.2 17 35 10 5 10.76 0.763 12.3 20 39 52 5 1.286 19 25 28 1.732 0.820 19 25 28 7.62 0.713 37.2 16 49 24 5 10.73 0.761 12.6 21 51 24 5 1.286 19 27 26 1.737 0.817 19 27 36 7.61 0.710 37.0 16 37 12 5 10.69 0.758 12.3 21 37 10 5 1.273 19 17 42 1.747 0.813 20 51 22 7.33 0.704 39.8 16 11 20 5 10.65 0.755 12.2 20 19 52 5 1.271 19 19 44 1.751 0.811 16 31 32 7.35 0.702 39.5 16 41 12 5 10.65 0.755 12.4 20 51 20 5 1.266 19 15 44 1.755 0.809 20 21 10 7.36 0.702 39.3 16 51 30 5 10.63 0.754 12.9 20 7 26 5 1.266 19 17 46 1.758 0.808 19 23 30 7.38 0.701 38.8 16 31 10 5 10.62 0.753 12.3 20 17 50 5 1.254 19 13 42 1.767 0.804 19 29 38 7.28 0.699 39.9 16 17 48 5 10.60 0.752 12.7 20 25 8 5 1.252 19 19 48 1.777 0.799 19 17 46 7.27 0.698 39.9 16 13 18 5 10.59 0.751 12.5 20 49 18 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-20 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 21105.1 Exposure: MWd/MTU (GWd) 6000.0 ( 830.10 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.197 4.545 13 0.446 0.453 59 34 Inlet Subcooling: Btu/lbm -29.93 24 0.531 12.739 14 0.392 0.462 59 30 Flow: Mlb/hr 92.97 ( 90.70 %) 23 0.675 16.411 15 0.505 0.508 3 18 22 0.750 18.418 16 1.072 1.242 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.795 19.860 17 0.869 1.178 13 40 59 -- -- -- -- -- -- -- 59 20 0.838 20.967 18 1.060 1.193 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.867 21.547 19 1.195 1.322 29 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.904 22.202 20 1.109 1.229 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.934 22.647 21 1.112 1.246 51 38 43 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 43 16 0.963 23.052 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.004 23.771 35 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 35 14 1.021 23.784 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.144 22.635 27 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 27 12 1.191 23.395 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.218 23.783 19 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 19 10 1.246 24.253 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.282 24.825 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.314 25.203 7 -- -- -- -- -- -- -- -- -- 7 7 1.350 25.601 3 -- -- -- -- -- -- -- 3 6 1.405 26.341 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.425* 26.441* 4 1.387 25.534 Control Rod Density: % 3.78 3 1.281 23.393 2 0.994 17.687 k-effective: 0.99888 Bottom 1 0.283 5.117 Void Fraction: 0.484 Core Delta-P: psia 21.662  % AXIAL TILT -20.529 -8.536 Core Plate Delta-P: psia 17.110 AVG BOT 8ft/12ft 1.1194 1.0496 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 81.56 Active Channel Flow: Mlb/hr 78.54 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.322 19 29 28 1.665 0.853 19 29 34 7.48 0.726 40.9 17 51 26 5 10.81 0.767 13.7 20 51 40 5 1.320 19 27 30 1.666 0.852 19 27 32 7.35 0.721 42.0 17 35 10 5 10.79 0.765 13.5 20 39 52 5 1.289 19 27 26 1.732 0.820 19 25 28 7.62 0.718 38.0 16 49 24 5 10.75 0.762 13.7 21 51 24 4 1.289 19 25 28 1.737 0.817 19 27 36 7.60 0.715 37.8 16 37 12 5 10.71 0.760 13.4 21 37 10 5 1.277 19 17 42 1.746 0.813 20 51 22 7.34 0.710 40.6 16 11 20 5 10.68 0.758 13.3 20 19 52 5 1.276 19 19 44 1.753 0.810 20 21 10 7.35 0.709 40.2 16 41 12 5 10.68 0.757 13.5 20 51 20 5 1.272 19 15 44 1.758 0.808 16 31 32 7.35 0.707 40.1 16 51 30 5 10.64 0.755 14.1 20 7 26 4 1.272 19 17 46 1.758 0.808 19 23 30 7.37 0.705 39.6 16 31 10 5 10.62 0.753 13.5 20 17 50 5 1.260 19 13 42 1.767 0.804 19 29 38 7.27 0.703 40.6 16 17 48 5 10.61 0.752 13.8 20 25 8 5 1.258 19 19 48 1.776 0.800 19 17 46 7.26 0.703 40.7 16 13 18 5 10.61 0.752 12.9 19 39 50 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-21 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 21605.1 Exposure: MWd/MTU (GWd) 6500.0 ( 899.27 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.200 4.650 13 0.442 0.448 59 34 Inlet Subcooling: Btu/lbm -30.00 24 0.536 13.025 14 0.388 0.457 59 30 Flow: Mlb/hr 92.76 ( 90.50 %) 23 0.681 16.774 15 0.501 0.504 3 18 22 0.755 18.824 16 1.070 1.235 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.799 20.290 17 0.865 1.176 13 40 59 -- -- -- -- -- -- -- 59 20 0.840 21.420 18 1.056 1.190 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.869 22.015 19 1.199 1.324 29 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.905 22.690 20 1.113 1.235 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.934 23.151 21 1.117 1.253 51 38 43 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 43 16 0.962 23.571 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.002 24.313 35 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 35 14 1.017 24.334 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.138 23.168 27 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 27 12 1.183 23.949 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.210 24.351 19 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 19 10 1.240 24.834 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.277 25.423 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.311 25.817 7 -- -- -- -- -- -- -- -- -- 7 7 1.349 26.232 3 -- -- -- -- -- -- -- 3 6 1.406 26.998 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.428* 27.108* 4 1.392 26.184 Control Rod Density: % 3.78 3 1.286 23.993 2 0.996 18.152 k-effective: 0.99882 Bottom 1 0.284 5.256 Void Fraction: 0.484 Core Delta-P: psia 21.593  % AXIAL TILT -20.383 -8.659 Core Plate Delta-P: psia 17.041 AVG BOT 8ft/12ft 1.1178 1.0503 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 81.38 Active Channel Flow: Mlb/hr 78.37 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.324 19 29 28 1.664 0.853 19 29 34 7.49 0.733 41.7 17 51 26 5 10.89 0.772 14.8 20 51 40 4 1.323 19 27 30 1.666 0.852 19 27 32 7.36 0.728 42.8 17 35 10 5 10.87 0.771 14.5 20 39 52 4 1.293 19 27 26 1.731 0.820 19 25 28 7.62 0.724 38.8 16 49 24 5 10.81 0.766 14.9 21 51 24 4 1.292 19 25 28 1.735 0.818 19 27 36 7.61 0.722 38.6 16 37 12 5 10.77 0.764 14.5 21 37 10 4 1.282 19 17 42 1.744 0.814 20 51 22 7.36 0.717 41.4 16 11 20 5 10.73 0.761 14.5 20 19 52 5 1.281 19 19 44 1.751 0.811 20 21 10 7.37 0.716 41.0 16 41 12 5 10.73 0.761 14.7 20 51 20 5 1.278 19 15 44 1.757 0.808 19 23 30 7.35 0.713 40.9 16 51 30 5 10.69 0.758 15.3 20 7 26 4 1.277 19 17 46 1.764 0.805 16 31 32 7.37 0.711 40.4 16 31 10 5 10.66 0.756 14.1 19 39 50 5 1.266 19 13 42 1.765 0.804 19 29 38 7.28 0.709 41.4 16 17 48 5 10.66 0.756 14.3 19 49 40 5 1.264 19 19 48 1.774 0.801 19 17 46 7.27 0.709 41.5 16 13 18 5 10.65 0.756 15.0 20 25 8 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-22 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 21855.0 Exposure: MWd/MTU (GWd) 6750.0 ( 933.86 ) Delta E: MWd/MTU, (GWd) 250.0 ( 34.59 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.201 4.703 13 0.440 0.445 59 34 Inlet Subcooling: Btu/lbm -30.14 24 0.538 13.169 14 0.386 0.454 59 30 Flow: Mlb/hr 92.35 ( 90.10 %) 23 0.683 16.957 15 0.499 0.502 3 18 22 0.756 19.028 16 1.068 1.231 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.799 20.506 17 0.864 1.174 13 40 59 -- -- -- -- -- -- -- 59 20 0.840 21.646 18 1.053 1.189 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.868 22.249 19 1.201 1.326 29 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.904 22.934 20 1.115 1.239 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.933 23.403 21 1.120 1.257 51 38 43 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 43 16 0.960 23.831 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.000 24.583 35 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 35 14 1.014 24.608 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.134 23.433 27 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 27 12 1.178 24.225 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.206 24.633 19 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 19 10 1.236 25.124 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.274 25.722 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.309 26.123 7 -- -- -- -- -- -- -- -- -- 7 7 1.349 26.548 3 -- -- -- -- -- -- -- 3 6 1.408 27.327 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.433* 27.443* 4 1.398 26.511 Control Rod Density: % 3.78 3 1.293 24.295 2 1.002 18.386 k-effective: 0.99874 Bottom 1 0.286 5.326 Void Fraction: 0.485 Core Delta-P: psia 21.460  % AXIAL TILT -20.437 -8.717 Core Plate Delta-P: psia 16.908 AVG BOT 8ft/12ft 1.1178 1.0506 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 81.01 Active Channel Flow: Mlb/hr 78.01 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.3 (of total core flow) Source Convergence 0.00006

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.326 19 29 28 1.662 0.854 19 29 34 7.51 0.737 42.1 17 51 26 5 10.96 0.777 15.4 20 51 40 4 1.324 19 27 30 1.664 0.853 19 27 32 7.38 0.733 43.2 17 35 10 5 10.93 0.775 15.1 20 39 52 4 1.294 19 27 26 1.728 0.822 19 25 28 7.64 0.729 39.2 16 49 24 5 10.87 0.771 15.5 21 51 24 4 1.294 19 25 28 1.733 0.820 19 27 36 7.62 0.726 39.0 16 37 12 5 10.83 0.768 15.1 21 37 10 4 1.285 19 17 42 1.741 0.815 20 51 22 7.38 0.722 41.8 16 11 20 5 10.78 0.764 15.1 20 19 52 5 1.283 19 19 44 1.748 0.813 20 21 10 7.39 0.721 41.4 16 41 12 5 10.78 0.764 15.2 20 51 20 5 1.280 19 15 44 1.755 0.809 19 23 30 7.36 0.717 41.3 16 51 30 5 10.74 0.762 15.8 20 7 26 4 1.280 19 17 46 1.763 0.805 19 29 38 7.38 0.715 40.7 16 31 10 5 10.73 0.761 14.5 19 39 50 4 1.269 19 13 42 1.766 0.804 16 31 32 7.29 0.713 41.8 16 17 48 5 10.72 0.760 14.7 19 49 40 4 1.267 19 19 48 1.770 0.802 19 17 46 7.28 0.713 41.9 16 13 18 5 10.70 0.759 15.5 20 25 8 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-23 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 22215.5 Exposure: MWd/MTU (GWd) 7110.5 ( 983.73 ) Delta E: MWd/MTU, (GWd) 360.5 ( 49.87 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.202 4.780 13 0.437 0.442 59 34 Inlet Subcooling: Btu/lbm -30.29 24 0.541 13.377 14 0.384 0.451 9 52 Flow: Mlb/hr 91.94 ( 89.70 %) 23 0.686 17.221 15 0.497 0.499 3 18 22 0.757 19.322 16 1.066 1.226 29 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.800 20.817 17 0.861 1.172 13 40 59 -- -- -- -- -- -- -- 59 20 0.840 21.973 18 1.050 1.186 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.868 22.587 19 1.205 1.328 29 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.903 23.285 20 1.118 1.243 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.931 23.765 21 1.124 1.262 51 38 43 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 43 16 0.958 24.204 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 0.997 24.971 35 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 35 14 1.010 25.001 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.127 23.814 27 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 27 12 1.171 24.621 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.199 25.038 19 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 19 10 1.230 25.539 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.270 26.151 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.307 26.564 7 -- -- -- -- -- -- -- -- -- 7 7 1.349 27.003 3 -- -- -- -- -- -- -- 3 6 1.411 27.802 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.438* 27.927* 4 1.407 26.984 Control Rod Density: % 3.78 3 1.301 24.732 2 1.008 18.725 k-effective: 0.99869 Bottom 1 0.288 5.428 Void Fraction: 0.485 Core Delta-P: psia 21.328  % AXIAL TILT -20.466 -8.801 Core Plate Delta-P: psia 16.776 AVG BOT 8ft/12ft 1.1173 1.0511 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 80.64 Active Channel Flow: Mlb/hr 77.65 Total Bypass Flow (%): 12.3 (of total core flow) Total Water Rod Flow (%): 3.3 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.328 19 29 28 1.660 0.855 19 29 34 7.53 0.744 42.7 17 51 26 5 11.05 0.783 16.2 20 51 40 4 1.327 19 27 30 1.662 0.854 19 27 32 7.40 0.740 43.8 17 35 10 5 11.01 0.781 15.9 20 39 52 4 1.297 19 27 26 1.725 0.823 19 25 28 7.66 0.735 39.8 16 49 24 5 10.95 0.776 16.3 21 51 24 4 1.296 19 25 28 1.730 0.821 19 27 36 7.65 0.732 39.6 16 37 12 5 10.90 0.773 15.9 21 37 10 4 1.288 19 17 42 1.738 0.817 20 51 22 7.41 0.729 42.4 16 11 20 5 10.83 0.768 15.9 20 19 52 5 1.287 19 19 44 1.744 0.814 20 39 52 7.42 0.728 42.0 16 41 12 5 10.83 0.768 16.1 20 51 20 5 1.285 19 15 44 1.753 0.810 19 23 30 7.38 0.723 41.8 16 51 30 5 10.81 0.767 15.3 19 39 50 4 1.284 19 17 46 1.760 0.807 19 29 38 7.40 0.721 41.3 16 31 10 5 10.81 0.767 16.7 20 7 26 4 1.273 19 13 42 1.767 0.804 19 17 46 7.31 0.719 42.4 16 17 48 5 10.81 0.767 15.6 19 49 40 4 1.272 19 19 48 1.769 0.803 16 31 32 7.30 0.719 42.4 16 13 18 5 10.77 0.764 16.4 20 25 8 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-24 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 22216.3 Exposure: MWd/MTU (GWd) 7111.2 ( 983.83 ) Delta E: MWd/MTU, (GWd) 0.7 ( 0.10 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.214 4.780 13 0.427 0.437 47 56 Inlet Subcooling: Btu/lbm -28.38 24 0.576 13.377 14 0.373 0.448 9 52 Flow: Mlb/hr 97.68 ( 95.30 %) 23 0.733 17.221 15 0.480 0.499 43 58 22 0.814 19.323 16 1.081 1.268 21 40 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.868 20.817 17 0.834 1.166 39 14 59 -- -- -- -- -- -- -- 59 20 0.902 21.974 18 1.108 1.249 19 42 55 -- -- -- -- -- -- -- -- -- 55 19 0.914 22.588 19 1.235 1.373 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.937 23.286 20 1.061 1.237 17 50 47 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 47 17 0.956 23.766 21 1.043 1.225 37 10 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.974 24.205 39 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 39 15 1.007 24.972 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.014 25.002 31 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 31 13 1.128 23.815 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.169 24.622 23 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 23 11 1.193 25.039 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.220 25.540 15 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 15 9 1.255 26.152 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.286 26.565 7 -- -- -- -- -- -- -- -- -- 7 7 1.320 27.004 3 -- -- -- -- -- -- -- 3 6 1.371 27.803 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.384* 27.928* 4 1.336 26.985 Control Rod Density: % 4.59 3 1.221 24.733 2 0.940 18.725 k-effective: 0.99849 Bottom 1 0.268 5.428 Void Fraction: 0.465 Core Delta-P: psia 23.155  % AXIAL TILT -16.988 -8.801 Core Plate Delta-P: psia 18.599 AVG BOT 8ft/12ft 1.0946 1.0511 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 85.87 Active Channel Flow: Mlb/hr 82.77 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.373 19 21 38 1.652 0.860 19 19 36 7.36 0.713 40.8 16 17 30 5 10.62 0.753 13.8 19 17 34 5 1.372 19 19 40 1.656 0.858 19 21 38 7.33 0.712 41.1 16 47 30 5 10.57 0.750 13.4 19 41 26 5 1.368 19 19 36 1.656 0.858 19 19 40 7.53 0.712 38.3 16 45 28 5 10.57 0.749 15.9 20 19 52 5 1.364 19 17 34 1.672 0.849 19 43 28 7.47 0.711 39.0 16 17 36 5 10.55 0.748 14.9 19 15 32 5 1.358 19 23 40 1.677 0.847 19 21 42 7.49 0.708 38.3 16 19 38 5 10.51 0.745 15.7 20 43 12 5 1.354 19 17 38 1.678 0.846 19 23 22 7.21 0.707 42.0 16 41 12 5 10.50 0.744 14.3 19 41 40 5 1.354 19 19 32 1.681 0.845 19 45 30 7.55 0.707 37.4 16 21 40 5 10.49 0.744 14.0 19 39 24 5 1.352 19 21 42 1.686 0.842 19 19 32 7.18 0.706 42.4 16 17 48 5 10.49 0.744 14.0 19 17 24 5 1.351 19 15 32 1.696 0.837 19 17 38 7.29 0.706 40.6 16 17 40 5 10.48 0.743 13.8 19 39 42 5 1.343 19 17 42 1.704 0.833 19 21 34 7.45 0.702 38.1 16 19 34 5 10.47 0.743 15.9 19 47 34 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-25 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 22605.1 Exposure: MWd/MTU (GWd) 7500.0 (1037.62 ) Delta E: MWd/MTU, (GWd) 388.8 ( 53.79 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.216 4.868 13 0.424 0.434 47 56 Inlet Subcooling: Btu/lbm -28.60 24 0.579 13.618 14 0.371 0.446 9 52 Flow: Mlb/hr 96.96 ( 94.60 %) 23 0.736 17.527 15 0.478 0.496 43 58 22 0.816 19.665 16 1.078 1.263 21 40 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.869 21.182 17 0.832 1.164 39 14 59 -- -- -- -- -- -- -- 59 20 0.902 22.352 18 1.104 1.245 19 42 55 -- -- -- -- -- -- -- -- -- 55 19 0.913 22.971 19 1.239 1.375 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.936 23.679 20 1.065 1.242 17 12 47 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 47 17 0.954 24.167 21 1.047 1.232 37 10 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.972 24.613 39 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 39 15 1.003 25.393 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.009 25.427 31 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 31 13 1.121 24.224 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.161 25.045 23 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 23 11 1.185 25.471 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.213 25.983 15 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 15 9 1.250 26.607 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.283 27.032 7 -- -- -- -- -- -- -- -- -- 7 7 1.320 27.484 3 -- -- -- -- -- -- -- 3 6 1.375 28.303 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.391* 28.433* 4 1.346 27.472 Control Rod Density: % 4.59 3 1.232 25.179 2 0.947 19.069 k-effective: 0.99841 Bottom 1 0.271 5.532 Void Fraction: 0.466 Core Delta-P: psia 22.913  % AXIAL TILT -17.040 -8.827 Core Plate Delta-P: psia 18.357 AVG BOT 8ft/12ft 1.0942 1.0512 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 85.23 Active Channel Flow: Mlb/hr 82.14 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.375 19 21 38 1.649 0.861 19 19 36 7.32 0.716 41.7 16 47 30 5 10.64 0.755 16.8 20 19 52 5 1.375 19 19 40 1.651 0.860 19 19 40 7.25 0.716 42.6 16 41 12 5 10.61 0.752 14.7 19 17 34 5 1.369 19 19 36 1.652 0.860 19 21 38 7.33 0.715 41.4 16 17 30 5 10.58 0.750 16.5 20 43 12 5 1.365 19 17 34 1.670 0.851 19 43 28 7.51 0.714 38.9 16 45 28 5 10.56 0.749 14.3 19 41 26 5 1.360 19 23 40 1.673 0.849 19 21 42 7.20 0.714 43.0 16 17 48 5 10.54 0.748 15.7 19 15 32 5 1.357 19 17 38 1.674 0.848 19 23 22 7.44 0.712 39.6 16 17 36 5 10.51 0.745 16.3 19 17 16 5 1.355 19 19 32 1.677 0.847 19 45 30 7.47 0.710 38.9 16 19 38 5 10.51 0.745 16.3 19 39 50 5 1.355 19 21 42 1.682 0.844 19 19 32 7.29 0.709 41.3 16 17 40 5 10.50 0.745 16.7 20 39 52 5 1.353 19 15 32 1.691 0.840 19 17 38 7.53 0.709 38.0 16 21 40 5 10.49 0.744 15.2 19 41 40 5 1.347 19 17 42 1.701 0.835 19 21 34 7.44 0.709 39.1 16 15 46 5 10.49 0.744 14.7 19 39 42 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-26 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 23105.0 Exposure: MWd/MTU (GWd) 8000.0 (1106.80 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.218 4.983 13 0.420 0.431 47 56 Inlet Subcooling: Btu/lbm -28.77 24 0.583 13.929 14 0.367 0.442 9 52 Flow: Mlb/hr 96.45 ( 94.10 %) 23 0.740 17.922 15 0.474 0.493 43 58 22 0.817 20.106 16 1.075 1.259 21 40 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.869 21.650 17 0.828 1.162 39 14 59 -- -- -- -- -- -- -- 59 20 0.901 22.839 18 1.099 1.241 19 42 55 -- -- -- -- -- -- -- -- -- 55 19 0.911 23.463 19 1.244 1.380 19 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.933 24.183 20 1.069 1.249 17 50 47 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 47 17 0.950 24.680 21 1.053 1.240 37 10 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.967 25.136 39 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 39 15 0.997 25.933 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.002 25.969 31 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 31 13 1.111 24.745 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.149 25.584 23 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 23 11 1.173 26.021 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.204 26.548 15 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 15 9 1.244 27.191 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.280 27.632 7 -- -- -- -- -- -- -- -- -- 7 7 1.322 28.102 3 -- -- -- -- -- -- -- 3 6 1.381 28.947 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.403* 29.086* 4 1.362 28.106 Control Rod Density: % 4.59 3 1.248 25.759 2 0.959 19.514 k-effective: 0.99842 Bottom 1 0.275 5.666 Void Fraction: 0.468 Core Delta-P: psia 22.748  % AXIAL TILT -17.205 -8.863 Core Plate Delta-P: psia 18.192 AVG BOT 8ft/12ft 1.0942 1.0513 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 84.76 Active Channel Flow: Mlb/hr 81.68 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.380 19 19 40 1.646 0.863 19 19 36 7.31 0.727 43.4 16 41 12 5 10.76 0.763 17.9 20 19 52 5 1.379 19 21 38 1.647 0.862 19 19 40 7.25 0.724 43.7 16 17 48 5 10.68 0.757 17.7 20 43 12 5 1.373 19 19 36 1.649 0.861 19 21 38 7.35 0.724 42.4 16 47 30 5 10.64 0.755 15.8 19 17 34 5 1.370 19 17 34 1.667 0.852 19 43 28 7.53 0.722 39.7 16 45 28 5 10.62 0.753 17.4 19 39 50 5 1.365 19 23 40 1.670 0.850 19 21 42 7.34 0.721 42.2 16 17 30 5 10.61 0.753 17.9 20 39 52 4 1.362 19 17 38 1.671 0.850 19 23 22 7.48 0.719 39.9 16 15 46 5 10.59 0.751 15.4 19 41 26 5 1.359 19 19 32 1.673 0.849 19 45 30 7.44 0.718 40.4 16 17 36 5 10.59 0.751 16.9 19 15 32 5 1.359 19 21 42 1.679 0.846 19 19 32 7.31 0.717 42.0 16 17 40 5 10.58 0.750 17.4 19 17 16 5 1.359 19 15 32 1.687 0.842 19 17 38 7.47 0.716 39.7 16 19 38 5 10.55 0.748 17.9 19 47 34 5 1.352 19 17 42 1.698 0.836 19 21 34 7.53 0.715 38.8 16 21 40 5 10.54 0.747 15.8 19 39 42 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-27 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 23605.0 Exposure: MWd/MTU (GWd) 8500.0 (1175.97 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.220 5.099 13 0.416 0.428 47 56 Inlet Subcooling: Btu/lbm -29.00 24 0.585 14.241 14 0.363 0.439 9 52 Flow: Mlb/hr 95.74 ( 93.40 %) 23 0.743 18.319 15 0.470 0.489 43 58 22 0.818 20.547 16 1.072 1.255 21 40 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.868 22.119 17 0.823 1.160 39 14 59 -- -- -- -- -- -- -- 59 20 0.899 23.324 18 1.094 1.237 19 42 55 -- -- -- -- -- -- -- -- -- 55 19 0.908 23.954 19 1.249 1.386 19 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.930 24.686 20 1.072 1.256 17 50 47 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 47 17 0.946 25.192 21 1.058 1.248 37 10 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.962 25.656 39 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 39 15 0.991 26.470 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.994 26.508 31 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 31 13 1.100 25.261 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.137 26.118 23 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 23 11 1.161 26.567 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.194 27.109 15 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 15 9 1.236 27.771 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.277 28.230 7 -- -- -- -- -- -- -- -- -- 7 7 1.323 28.720 3 -- -- -- -- -- -- -- 3 6 1.389 29.595 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.417* 29.746* 4 1.381 28.748 Control Rod Density: % 4.59 3 1.269 26.348 2 0.974 19.967 k-effective: 0.99847 Bottom 1 0.279 5.802 Void Fraction: 0.469 Core Delta-P: psia 22.517  % AXIAL TILT -17.468 -8.901 Core Plate Delta-P: psia 17.962 AVG BOT 8ft/12ft 1.0946 1.0513 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 84.11 Active Channel Flow: Mlb/hr 81.05 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.386 19 19 40 1.641 0.866 19 19 36 7.39 0.741 44.1 16 41 12 5 10.89 0.773 19.1 20 19 52 5 1.385 19 21 38 1.641 0.865 19 19 40 7.30 0.735 44.5 16 17 48 5 10.80 0.767 19.0 20 39 52 4 1.378 19 19 36 1.643 0.864 19 21 38 7.38 0.733 43.2 16 47 30 5 10.80 0.766 18.8 20 43 12 5 1.375 19 17 34 1.662 0.855 19 43 28 7.56 0.730 40.5 16 45 28 5 10.80 0.766 18.4 19 39 50 4 1.370 19 23 40 1.664 0.853 19 21 42 7.54 0.730 40.7 16 15 46 5 10.70 0.759 17.3 19 17 34 5 1.367 19 17 38 1.665 0.853 19 23 22 7.36 0.729 43.0 16 17 30 5 7.97 0.758 48.6 16 41 50 5 1.365 19 19 32 1.666 0.852 19 45 30 7.46 0.726 41.2 16 17 36 5 10.66 0.756 18.6 19 15 48 4 1.365 19 21 42 1.673 0.849 19 19 32 7.33 0.725 42.8 16 17 40 5 10.66 0.756 18.6 19 17 16 5 1.364 19 15 32 1.681 0.845 19 17 38 7.47 0.725 40.8 16 41 16 5 10.65 0.755 18.0 19 15 32 5 1.358 19 17 42 1.694 0.838 19 17 42 7.50 0.724 40.5 16 19 38 5 10.65 0.755 16.5 19 41 26 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-28 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 24104.9 Exposure: MWd/MTU (GWd) 9000.0 (1245.15 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.221 5.216 13 0.412 0.425 47 56 Inlet Subcooling: Btu/lbm -29.51 24 0.586 14.555 14 0.359 0.436 9 52 Flow: Mlb/hr 94.20 ( 91.90 %) 23 0.743 18.717 15 0.466 0.485 17 4 22 0.816 20.987 16 1.068 1.251 21 40 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.864 22.586 17 0.819 1.157 39 14 59 -- -- -- -- -- -- -- 59 20 0.894 23.808 18 1.089 1.233 19 42 55 -- -- -- -- -- -- -- -- -- 55 19 0.903 24.443 19 1.255 1.391 19 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.924 25.186 20 1.076 1.262 17 50 47 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 47 17 0.939 25.700 21 1.063 1.255 37 10 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.955 26.174 39 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 39 15 0.982 27.002 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.984 27.041 31 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 31 13 1.087 25.772 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.122 26.646 23 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 23 11 1.148 27.107 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.183 27.665 15 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 15 9 1.229 28.347 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.273 28.826 7 -- -- -- -- -- -- -- -- -- 7 7 1.326 29.340 3 -- -- -- -- -- -- -- 3 6 1.398 30.247 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.436* 30.414* 4 1.407 29.400 Control Rod Density: % 4.59 3 1.297 26.948 2 0.997 20.427 k-effective: 0.99841 Bottom 1 0.286 5.941 Void Fraction: 0.473 Core Delta-P: psia 22.017  % AXIAL TILT -17.951 -8.946 Core Plate Delta-P: psia 17.462 AVG BOT 8ft/12ft 1.0963 1.0515 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 82.72 Active Channel Flow: Mlb/hr 79.69 Total Bypass Flow (%): 12.2 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.391 19 19 40 1.628 0.872 19 19 40 7.47 0.755 44.9 16 41 12 5 11.14 0.798 20.0 20 19 52 4 1.390 19 21 38 1.629 0.872 19 19 36 7.38 0.749 45.3 16 17 48 5 11.01 0.790 20.2 20 39 52 4 1.383 19 19 36 1.630 0.871 19 21 38 7.43 0.744 44.0 16 47 30 5 11.03 0.787 19.6 19 39 50 4 1.380 19 17 34 1.649 0.861 19 43 28 7.61 0.743 41.5 16 15 46 5 11.02 0.787 19.7 20 43 12 4 1.375 19 23 40 1.652 0.859 19 23 22 7.60 0.740 41.3 16 45 28 5 10.87 0.777 19.7 19 15 48 4 1.373 19 17 38 1.652 0.859 19 45 30 7.40 0.739 43.8 16 17 30 5 10.78 0.775 20.5 19 49 32 4 1.371 19 15 32 1.653 0.859 19 21 42 7.54 0.736 41.6 16 19 46 5 10.85 0.775 19.7 19 17 16 4 1.370 19 19 32 1.660 0.855 19 19 32 7.38 0.736 43.6 16 17 40 5 8.05 0.773 49.4 16 41 50 5 1.370 19 21 42 1.668 0.851 19 17 38 7.50 0.736 42.0 16 17 36 5 10.78 0.773 20.1 19 47 34 4 1.364 19 17 42 1.678 0.846 19 17 42 7.54 0.734 41.3 16 19 38 5 10.85 0.772 19.4 19 41 48 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-29 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 24355.0 Exposure: MWd/MTU (GWd) 9250.0 (1279.73 ) Delta E: MWd/MTU, (GWd) 250.0 ( 34.59 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.221 5.274 13 0.410 0.423 47 56 Inlet Subcooling: Btu/lbm -29.75 24 0.586 14.712 14 0.357 0.434 9 52 Flow: Mlb/hr 93.48 ( 91.20 %) 23 0.743 18.916 15 0.464 0.483 17 4 22 0.814 21.207 16 1.067 1.249 21 40 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.862 22.819 17 0.817 1.156 39 14 59 -- -- -- -- -- -- -- 59 20 0.891 24.049 18 1.086 1.231 19 42 55 -- -- -- -- -- -- -- -- -- 55 19 0.900 24.686 19 1.257 1.394 19 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.920 25.434 20 1.078 1.266 17 50 47 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 47 17 0.935 25.953 21 1.065 1.259 37 10 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.950 26.431 39 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 39 15 0.977 27.266 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.978 27.306 31 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 31 13 1.080 26.025 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.115 26.907 23 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 23 11 1.141 27.374 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.177 27.941 15 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 15 9 1.225 28.634 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.272 29.124 7 -- -- -- -- -- -- -- -- -- 7 7 1.328 29.650 3 -- -- -- -- -- -- -- 3 6 1.404 30.575 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.446* 30.751* 4 1.422 29.731 Control Rod Density: % 4.59 3 1.314 27.253 2 1.009 20.662 k-effective: 0.99840 Bottom 1 0.289 6.011 Void Fraction: 0.475 Core Delta-P: psia 21.789  % AXIAL TILT -18.254 -8.972 Core Plate Delta-P: psia 17.235 AVG BOT 8ft/12ft 1.0974 1.0515 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 82.07 Active Channel Flow: Mlb/hr 79.06 Total Bypass Flow (%): 12.2 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.394 19 19 40 1.621 0.876 19 19 40 7.52 0.763 45.3 16 41 12 5 11.27 0.811 20.6 20 19 52 4 1.393 19 21 38 1.623 0.875 19 19 36 7.42 0.756 45.7 16 17 48 5 11.12 0.802 20.8 20 39 52 4 1.386 19 19 36 1.624 0.874 19 21 38 7.46 0.750 44.4 16 47 30 5 11.16 0.800 20.2 19 39 50 4 1.383 19 17 34 1.643 0.864 19 43 28 7.65 0.750 41.9 16 15 46 5 11.14 0.800 20.3 20 43 12 4 1.378 19 23 40 1.646 0.863 19 45 30 7.63 0.746 41.7 16 45 28 5 10.98 0.789 20.3 19 15 48 4 1.376 19 17 38 1.647 0.862 19 23 22 9.56 0.745 16.6 20 39 10 4 10.96 0.786 20.3 19 17 16 4 1.374 19 15 32 1.647 0.862 19 21 42 7.43 0.745 44.2 16 17 30 5 10.87 0.785 21.1 19 49 32 4 1.373 19 19 32 1.654 0.858 19 19 32 7.57 0.743 42.0 16 19 46 5 10.96 0.785 20.0 19 41 48 4 1.373 19 21 42 1.662 0.854 19 17 38 7.41 0.742 44.0 16 17 40 5 10.87 0.783 20.7 19 47 34 4 1.367 19 17 42 1.671 0.850 19 17 42 7.52 0.741 42.4 16 17 36 5 8.10 0.781 49.9 16 41 50 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-30 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 24762.8 Exposure: MWd/MTU (GWd) 9657.9 (1336.17 ) Delta E: MWd/MTU, (GWd) 407.9 ( 56.44 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.221 5.369 13 0.406 0.421 47 56 Inlet Subcooling: Btu/lbm -30.14 24 0.586 14.967 14 0.354 0.431 9 52 Flow: Mlb/hr 92.35 ( 90.10 %) 23 0.742 19.240 15 0.461 0.480 17 4 22 0.811 21.565 16 1.064 1.246 21 40 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.857 23.197 17 0.813 1.154 39 14 59 -- -- -- -- -- -- -- 59 20 0.886 24.440 18 1.082 1.228 19 42 55 -- -- -- -- -- -- -- -- -- 55 19 0.894 25.081 19 1.262 1.400 19 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.914 25.838 20 1.081 1.272 17 50 47 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 47 17 0.929 26.364 21 1.069 1.264 37 10 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.943 26.847 39 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 39 15 0.969 27.695 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.969 27.734 31 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 31 13 1.068 26.434 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.103 27.330 23 -- -- 8 -- -- -- -- -- -- -- -- -- 8 -- -- 23 11 1.130 27.807 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.169 28.389 15 -- -- -- -- -- 10 -- 10 -- -- -- -- -- 15 9 1.219 29.100 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.270 29.609 7 -- -- -- -- -- -- -- -- -- 7 7 1.331 30.158 3 -- -- -- -- -- -- -- 3 6 1.413 31.112 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.463* 31.306* 4 1.445 30.278 Control Rod Density: % 4.59 3 1.340 27.760 2 1.030 21.051 k-effective: 0.99842 Bottom 1 0.296 6.129 Void Fraction: 0.478 Core Delta-P: psia 21.434  % AXIAL TILT -18.749 -9.019 Core Plate Delta-P: psia 16.879 AVG BOT 8ft/12ft 1.0993 1.0516 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 81.05 Active Channel Flow: Mlb/hr 78.06 Total Bypass Flow (%): 12.2 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.400 19 19 40 1.611 0.881 19 19 40 7.59 0.776 46.0 16 41 12 5 11.47 0.833 21.6 20 19 52 4 1.398 19 21 38 1.613 0.880 19 19 36 7.49 0.769 46.3 16 17 48 5 11.36 0.821 21.1 19 39 50 4 1.390 19 19 36 1.615 0.879 19 21 38 9.78 0.767 17.4 20 39 10 4 11.32 0.821 21.5 20 39 52 4 1.388 19 17 34 1.633 0.869 19 43 28 9.75 0.762 17.0 20 41 10 4 11.34 0.821 21.3 20 43 12 4 1.382 19 23 40 1.635 0.869 19 45 30 7.72 0.762 42.6 16 15 46 5 11.17 0.809 21.3 19 15 48 4 1.381 19 17 38 1.637 0.867 19 23 22 7.51 0.760 45.1 16 47 30 5 11.14 0.806 21.2 19 17 16 4 1.379 19 15 32 1.638 0.867 19 21 42 7.68 0.756 42.4 16 45 28 5 11.14 0.804 20.9 19 41 48 4 1.377 19 19 32 1.644 0.864 19 19 32 7.47 0.755 44.8 16 17 30 5 11.01 0.802 22.0 19 49 32 4 1.377 19 21 42 1.651 0.860 19 17 38 9.65 0.755 17.0 20 17 50 4 11.01 0.800 21.7 19 47 34 4 1.373 19 17 42 1.659 0.856 19 17 42 7.63 0.753 42.7 16 19 46 5 11.04 0.796 20.8 20 13 48 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-31 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 24763.5 Exposure: MWd/MTU (GWd) 9658.7 (1336.27 ) Delta E: MWd/MTU, (GWd) 0.7 ( 0.10 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.250 5.370 13 0.419 0.424 59 34 Inlet Subcooling: Btu/lbm -28.09 24 0.664 14.968 14 0.367 0.432 59 30 Flow: Mlb/hr 98.60 ( 96.20 %) 23 0.839 19.240 15 0.477 0.479 3 18 22 0.908 21.566 16 1.062 1.198 31 14 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.947 23.198 17 0.818 1.144 51 36 59 -- -- -- -- -- -- -- 59 20 0.960 24.441 18 1.043 1.158 21 26 55 -- -- -- -- -- -- -- -- -- 55 19 0.942 25.082 19 1.230 1.330 49 32 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.942 25.839 20 1.127 1.238 39 52 47 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 47 17 0.940 26.364 21 1.145 1.268 37 10 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.940 26.848 39 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 39 15 0.956 27.696 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.947 27.735 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.039 26.435 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.070 27.331 23 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 23 11 1.095 27.808 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.133 28.389 15 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 15 9 1.183 29.101 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.234 29.610 7 -- -- -- -- -- -- -- -- -- 7 7 1.295 30.159 3 -- -- -- -- -- -- -- 3 6 1.376 31.113 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.419* 31.307* 4 1.391 30.279 Control Rod Density: % 5.23 3 1.277 27.761 2 0.974 21.052 k-effective: 0.99847 Bottom 1 0.278 6.129 Void Fraction: 0.462 Core Delta-P: psia 23.402  % AXIAL TILT -14.562 -9.019 Core Plate Delta-P: psia 18.844 AVG BOT 8ft/12ft 1.0671 1.0516 Coolant Temp: Deg-F 548.3 In Channel Flow: Mlb/hr 86.74 Active Channel Flow: Mlb/hr 83.63 Total Bypass Flow (%): 12.0 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.330 19 49 32 1.727 0.822 19 21 38 7.49 0.763 45.7 16 51 30 5 11.15 0.813 22.1 19 9 34 4 1.329 19 15 32 1.730 0.821 19 27 32 7.40 0.759 46.2 17 51 26 5 11.18 0.811 21.6 19 33 10 4 1.328 19 31 12 1.733 0.820 19 29 34 7.30 0.759 47.6 17 35 10 5 11.08 0.803 21.4 19 31 8 4 1.328 19 21 38 1.733 0.819 19 23 22 7.50 0.758 44.9 16 31 52 5 11.04 0.802 21.7 19 7 32 4 1.327 19 29 16 1.734 0.819 19 45 30 9.65 0.756 17.3 19 51 28 4 10.96 0.798 21.9 20 25 8 4 1.326 19 23 40 1.744 0.814 19 29 46 9.59 0.749 17.0 19 33 10 4 10.94 0.798 22.1 19 49 32 4 1.314 19 47 28 1.744 0.814 19 49 30 9.52 0.745 17.1 21 53 34 4 10.94 0.796 21.8 20 7 26 4 1.308 19 27 14 1.745 0.814 19 23 36 9.47 0.743 17.5 19 49 32 4 11.00 0.792 20.7 21 33 8 4 1.304 19 19 40 1.746 0.813 19 25 38 9.50 0.742 17.0 21 33 54 4 10.95 0.789 20.7 21 7 34 4 1.303 19 25 24 1.750 0.811 19 31 12 7.50 0.741 42.8 16 49 28 5 10.92 0.789 21.1 19 31 50 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-32 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 25104.8 Exposure: MWd/MTU (GWd) 10000.0 (1383.49 ) Delta E: MWd/MTU, (GWd) 341.3 ( 47.23 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.250 5.460 13 0.416 0.421 59 34 Inlet Subcooling: Btu/lbm -28.60 24 0.664 15.210 14 0.364 0.430 9 52 Flow: Mlb/hr 96.96 ( 94.60 %) 23 0.837 19.547 15 0.474 0.476 3 18 22 0.905 21.900 16 1.060 1.194 31 14 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.943 23.546 17 0.814 1.141 51 36 59 -- -- -- -- -- -- -- 59 20 0.955 24.794 18 1.040 1.156 21 26 55 -- -- -- -- -- -- -- -- -- 55 19 0.938 25.428 19 1.234 1.334 15 32 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.937 26.185 20 1.129 1.242 39 52 47 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 47 17 0.935 26.710 21 1.148 1.270 37 10 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.935 27.194 39 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 39 15 0.949 28.046 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.940 28.083 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.030 26.765 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.060 27.670 23 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 23 11 1.085 28.155 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.125 28.750 15 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 15 9 1.178 29.478 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.232 30.004 7 -- -- -- -- -- -- -- -- -- 7 7 1.298 30.573 3 -- -- -- -- -- -- -- 3 6 1.384 31.554 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.433* 31.762* 4 1.411 30.726 Control Rod Density: % 5.23 3 1.300 28.172 2 0.992 21.366 k-effective: 0.99845 Bottom 1 0.283 6.223 Void Fraction: 0.466 Core Delta-P: psia 22.854  % AXIAL TILT -14.987 -9.006 Core Plate Delta-P: psia 18.297 AVG BOT 8ft/12ft 1.0688 1.0514 Coolant Temp: Deg-F 548.3 In Channel Flow: Mlb/hr 85.26 Active Channel Flow: Mlb/hr 82.18 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.334 19 15 32 1.712 0.830 19 21 38 9.77 0.770 18.0 19 51 28 4 11.22 0.824 22.9 19 9 34 4 1.333 19 49 32 1.718 0.826 19 23 22 7.50 0.769 46.3 16 51 30 5 11.26 0.823 22.4 19 33 10 4 1.332 19 21 38 1.719 0.826 19 45 30 7.31 0.765 48.1 17 35 10 5 11.17 0.815 22.2 19 31 8 4 1.331 19 31 12 1.721 0.825 19 27 32 7.42 0.765 46.8 17 51 26 5 11.13 0.814 22.5 19 7 32 4 1.330 19 23 40 1.722 0.825 19 29 34 7.51 0.764 45.5 16 31 10 5 11.01 0.808 22.9 19 49 32 4 1.330 19 29 16 1.731 0.820 19 49 30 9.72 0.764 17.7 19 33 10 4 11.02 0.808 22.7 20 25 8 4 1.318 19 47 28 1.732 0.820 19 25 24 9.60 0.757 18.2 19 49 32 4 11.02 0.807 22.6 20 7 26 4 1.311 19 27 14 1.732 0.820 19 23 26 9.58 0.754 17.8 21 53 34 4 11.05 0.801 21.5 21 33 8 4 1.309 19 19 40 1.732 0.820 19 29 16 9.56 0.752 17.6 21 33 54 4 11.00 0.801 21.9 19 31 50 4 1.307 19 25 24 1.738 0.817 19 31 12 9.48 0.748 18.1 20 39 52 4 11.01 0.798 21.5 21 7 34 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-33 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 25604.8 Exposure: MWd/MTU (GWd) 10500.0 (1452.67 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.251 5.592 13 0.412 0.415 59 34 Inlet Subcooling: Btu/lbm -29.10 24 0.663 15.566 14 0.360 0.427 9 52 Flow: Mlb/hr 95.43 ( 93.10 %) 23 0.836 19.994 15 0.471 0.472 3 18 22 0.901 22.387 16 1.056 1.190 31 14 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.938 24.054 17 0.810 1.137 51 36 59 -- -- -- -- -- -- -- 59 20 0.950 25.308 18 1.035 1.152 21 26 55 -- -- -- -- -- -- -- -- -- 55 19 0.932 25.932 19 1.239 1.339 15 32 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.931 26.689 20 1.133 1.248 39 52 47 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 47 17 0.929 27.212 21 1.152 1.274 37 10 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.928 27.696 39 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 39 15 0.942 28.557 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.932 28.588 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.019 27.243 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.049 28.163 23 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 23 11 1.075 28.660 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.117 29.274 15 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 15 9 1.173 30.028 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.231 30.580 7 -- -- -- -- -- -- -- -- -- 7 7 1.301 31.180 3 -- -- -- -- -- -- -- 3 6 1.392 32.203 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.447* 32.436* 4 1.433 31.391 Control Rod Density: % 5.23 3 1.327 28.787 2 1.014 21.835 k-effective: 0.99848 Bottom 1 0.290 6.364 Void Fraction: 0.470 Core Delta-P: psia 22.352  % AXIAL TILT -15.491 -8.997 Core Plate Delta-P: psia 17.795 AVG BOT 8ft/12ft 1.0709 1.0510 Coolant Temp: Deg-F 548.3 In Channel Flow: Mlb/hr 83.87 Active Channel Flow: Mlb/hr 80.82 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.339 19 15 32 1.699 0.836 19 21 38 9.84 0.782 19.1 19 51 28 4 11.28 0.834 23.6 19 33 10 4 1.338 19 21 38 1.704 0.833 19 45 30 9.82 0.778 18.7 19 33 10 4 11.22 0.833 24.1 19 9 34 4 1.338 19 49 32 1.705 0.833 19 23 22 7.48 0.773 47.1 16 51 30 5 11.20 0.826 23.4 19 31 8 4 1.336 19 31 12 1.716 0.827 19 29 34 7.62 0.771 45.1 16 31 52 4 11.14 0.824 23.6 19 7 32 4 1.336 19 23 40 1.717 0.827 19 27 32 7.31 0.771 48.9 17 35 10 5 11.02 0.817 23.9 20 25 8 4 1.335 19 29 16 1.718 0.827 19 29 16 7.41 0.771 47.6 17 51 26 5 11.04 0.817 23.7 20 7 26 4 1.323 19 47 28 1.720 0.826 19 25 24 9.67 0.769 19.2 19 49 32 4 11.00 0.816 24.1 19 49 32 4 1.316 19 27 14 1.720 0.825 19 49 30 9.63 0.766 19.1 20 39 52 4 11.05 0.813 23.1 19 31 50 4 1.316 19 19 40 1.720 0.825 19 23 26 9.69 0.765 18.2 19 31 50 4 10.97 0.810 23.5 20 19 52 4 1.312 19 25 24 1.726 0.823 19 31 12 9.62 0.763 18.7 20 41 52 4 10.95 0.810 23.7 20 39 52 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-34 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 26104.8 Exposure: MWd/MTU (GWd) 11000.0 (1521.84 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.252 5.725 13 0.409 0.410 59 34 Inlet Subcooling: Btu/lbm -29.51 24 0.665 15.921 14 0.357 0.425 9 52 Flow: Mlb/hr 94.20 ( 91.90 %) 23 0.837 20.442 15 0.467 0.469 3 18 22 0.901 22.873 16 1.053 1.187 31 14 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.937 24.560 17 0.806 1.134 51 36 59 -- -- -- -- -- -- -- 59 20 0.948 25.820 18 1.031 1.148 21 26 55 -- -- -- -- -- -- -- -- -- 55 19 0.931 26.435 19 1.244 1.344 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.930 27.191 20 1.137 1.253 21 10 47 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 47 17 0.928 27.713 21 1.155 1.276 37 10 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.927 28.197 39 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 39 15 0.940 29.065 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.928 29.090 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.014 27.718 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.042 28.651 23 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 23 11 1.069 29.161 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.113 29.796 15 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 15 9 1.170 30.576 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.231 31.156 7 -- -- -- -- -- -- -- -- -- 7 7 1.302 31.789 3 -- -- -- -- -- -- -- 3 6 1.394 32.855 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.449* 33.113* 4 1.438 32.063 Control Rod Density: % 5.23 3 1.338 29.410 2 1.024 22.312 k-effective: 0.99862 Bottom 1 0.292 6.507 Void Fraction: 0.472 Core Delta-P: psia 21.942  % AXIAL TILT -15.589 -8.994 Core Plate Delta-P: psia 17.386 AVG BOT 8ft/12ft 1.0711 1.0506 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 82.76 Active Channel Flow: Mlb/hr 79.75 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.344 19 21 38 1.687 0.842 19 21 38 9.77 0.781 19.7 19 33 10 4 11.16 0.833 24.8 19 33 10 4 1.344 19 15 32 1.692 0.839 19 45 30 9.73 0.781 20.1 19 51 28 4 11.04 0.828 25.3 19 9 34 4 1.342 19 23 40 1.693 0.839 19 23 22 9.67 0.773 19.7 20 41 52 4 11.06 0.825 24.6 19 31 8 4 1.342 19 29 16 1.703 0.834 19 29 16 9.69 0.772 19.3 19 31 50 4 10.97 0.819 24.8 19 7 32 4 1.341 19 31 12 1.707 0.832 19 25 24 9.61 0.771 20.1 20 39 52 4 10.94 0.817 24.7 20 19 52 4 1.340 19 49 32 1.709 0.831 19 23 26 7.23 0.769 49.7 17 35 10 5 10.96 0.814 24.2 19 31 50 4 1.327 19 47 34 1.710 0.831 19 29 34 7.33 0.769 48.4 17 51 26 5 10.89 0.814 24.9 20 7 26 4 1.322 19 27 14 1.711 0.830 19 27 32 7.52 0.769 46.0 16 31 10 4 10.87 0.814 25.1 20 25 8 4 1.321 19 19 40 1.711 0.830 19 49 30 7.37 0.768 47.8 16 51 30 5 10.95 0.811 23.9 20 13 48 4 1.318 19 25 24 1.714 0.829 19 31 12 9.56 0.767 20.2 19 49 32 4 10.81 0.811 25.2 19 49 32 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-35 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 26604.7 Exposure: MWd/MTU (GWd) 11500.0 (1591.02 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.254 5.859 13 0.405 0.407 5 48 Inlet Subcooling: Btu/lbm -29.89 24 0.669 16.278 14 0.354 0.423 9 52 Flow: Mlb/hr 93.07 ( 90.80 %) 23 0.841 20.891 15 0.464 0.466 3 18 22 0.904 23.360 16 1.050 1.183 31 14 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.940 25.066 17 0.803 1.130 51 36 59 -- -- -- -- -- -- -- 59 20 0.952 26.332 18 1.027 1.145 21 26 55 -- -- -- -- -- -- -- -- -- 55 19 0.934 26.938 19 1.249 1.350 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.934 27.694 20 1.140 1.255 21 10 47 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 47 17 0.932 28.215 21 1.157 1.277 51 38 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.932 28.698 39 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 39 15 0.945 29.573 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.932 29.591 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.015 28.192 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.043 29.138 23 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 23 11 1.070 29.661 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.115 30.317 15 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 15 9 1.172 31.124 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.232 31.731 7 -- -- -- -- -- -- -- -- -- 7 7 1.300 32.398 3 -- -- -- -- -- -- -- 3 6 1.387 33.505 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.437* 33.788* 4 1.424 32.732 Control Rod Density: % 5.23 3 1.328 30.034 2 1.018 22.789 k-effective: 0.99878 Bottom 1 0.291 6.650 Void Fraction: 0.472 Core Delta-P: psia 21.552  % AXIAL TILT -15.213 -8.989 Core Plate Delta-P: psia 16.996 AVG BOT 8ft/12ft 1.0693 1.0503 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 81.76 Active Channel Flow: Mlb/hr 78.77 Total Bypass Flow (%): 12.2 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.350 19 21 38 1.673 0.849 19 21 38 9.54 0.769 20.8 19 33 10 4 10.85 0.819 25.9 19 33 10 4 1.349 19 23 40 1.679 0.846 19 23 22 9.54 0.769 20.7 20 41 52 4 10.82 0.810 25.0 20 9 20 4 1.348 19 29 16 1.681 0.845 19 45 30 9.57 0.766 19.8 20 9 20 4 10.68 0.809 26.4 19 9 34 4 1.347 19 15 32 1.688 0.841 19 29 16 9.45 0.764 21.1 19 51 34 4 10.75 0.809 25.7 19 31 8 4 1.344 19 31 12 1.694 0.838 19 25 24 9.52 0.764 20.3 19 31 50 4 10.80 0.809 25.0 20 13 48 4 1.340 19 49 32 1.695 0.838 19 23 26 9.42 0.762 21.2 20 39 10 4 10.71 0.808 25.9 20 19 52 4 1.329 19 47 34 1.701 0.835 19 29 34 7.08 0.760 50.4 17 35 10 5 10.77 0.806 24.9 19 31 50 4 1.328 19 27 14 1.702 0.834 19 31 12 7.27 0.759 47.9 16 31 10 5 10.61 0.801 26.0 19 7 32 4 1.327 19 19 40 1.702 0.834 19 49 30 7.17 0.759 49.2 17 51 36 5 10.72 0.801 24.8 19 15 48 4 1.325 19 25 24 1.703 0.834 19 27 32 9.43 0.757 20.2 19 15 48 4 10.70 0.799 24.8 20 17 12 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-36 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 27104.7 Exposure: MWd/MTU (GWd) 12000.0 (1660.19 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.256 5.994 13 0.402 0.405 5 48 Inlet Subcooling: Btu/lbm -30.29 24 0.674 16.637 14 0.351 0.421 9 52 Flow: Mlb/hr 91.94 ( 89.70 %) 23 0.847 21.343 15 0.461 0.463 3 18 22 0.910 23.850 16 1.047 1.180 31 14 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.948 25.576 17 0.800 1.126 51 36 59 -- -- -- -- -- -- -- 59 20 0.960 26.848 18 1.024 1.143 21 26 55 -- -- -- -- -- -- -- -- -- 55 19 0.944 27.445 19 1.254 1.356 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.945 28.201 20 1.143 1.257 21 10 47 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 47 17 0.944 28.721 21 1.159 1.279 51 38 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.944 29.205 39 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 39 15 0.957 30.086 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.943 30.097 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.025 28.669 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.052 29.628 23 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 23 11 1.078 30.163 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.122 30.840 15 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 15 9 1.178 31.674 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.233 32.308 7 -- -- -- -- -- -- -- -- -- 7 7 1.294 33.004 3 -- -- -- -- -- -- -- 3 6 1.370 34.150 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.408* 34.454* 4 1.390 33.390 Control Rod Density: % 5.23 3 1.297 30.648 2 0.995 23.260 k-effective: 0.99892 Bottom 1 0.284 6.792 Void Fraction: 0.471 Core Delta-P: psia 21.146  % AXIAL TILT -14.293 -8.971 Core Plate Delta-P: psia 16.590 AVG BOT 8ft/12ft 1.0651 1.0499 Coolant Temp: Deg-F 548.1 In Channel Flow: Mlb/hr 80.76 Active Channel Flow: Mlb/hr 77.81 Total Bypass Flow (%): 12.2 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.356 19 21 38 1.657 0.857 19 23 40 9.37 0.756 20.9 20 9 20 4 10.49 0.794 26.2 20 9 20 4 1.355 19 23 22 1.658 0.856 19 21 38 9.24 0.751 21.7 20 41 52 4 10.38 0.792 27.1 19 33 10 4 1.355 19 29 16 1.669 0.851 19 45 30 9.31 0.746 20.1 20 51 40 4 10.45 0.790 26.2 20 13 48 4 1.349 19 15 32 1.671 0.850 19 29 16 9.15 0.744 21.7 19 33 10 4 10.40 0.786 26.1 19 31 50 4 1.346 19 31 12 1.674 0.848 19 25 24 9.18 0.744 21.3 19 31 50 4 10.38 0.786 26.3 20 19 10 4 1.340 19 49 32 1.680 0.845 19 23 26 6.85 0.742 51.2 17 35 10 5 10.39 0.785 26.0 19 15 48 4 1.333 19 27 14 1.682 0.844 19 31 34 9.08 0.741 22.2 20 39 10 4 10.40 0.783 25.7 19 47 16 4 1.331 19 25 24 1.689 0.841 19 31 12 7.04 0.741 48.7 16 31 10 5 10.44 0.783 25.4 20 11 44 4 1.331 19 19 40 1.689 0.841 19 19 36 6.94 0.741 49.9 17 51 36 5 10.60 0.783 23.5 20 51 40 4 1.330 19 47 34 1.692 0.839 19 49 30 9.15 0.741 21.2 19 15 48 4 10.27 0.782 26.8 19 31 8 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-37 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 27310.1 Exposure: MWd/MTU (GWd) 12205.4 (1688.61 ) Delta E: MWd/MTU, (GWd) 205.4 ( 28.41 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.257 6.050 13 0.401 0.404 5 48 Inlet Subcooling: Btu/lbm -30.43 24 0.677 16.786 14 0.350 0.420 9 52 Flow: Mlb/hr 91.53 ( 89.30 %) 23 0.850 21.529 15 0.460 0.462 3 18 22 0.914 24.052 16 1.046 1.178 31 14 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.952 25.786 17 0.799 1.125 51 36 59 -- -- -- -- -- -- -- 59 20 0.966 27.062 18 1.023 1.142 21 26 55 -- -- -- -- -- -- -- -- -- 55 19 0.950 27.655 19 1.256 1.358 21 38 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 0.952 28.411 20 1.143 1.258 21 10 47 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 47 17 0.951 28.931 21 1.160 1.280 51 38 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 0.951 29.415 39 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 39 15 0.964 30.299 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 0.949 30.307 31 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 31 13 1.031 28.866 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.057 29.830 23 -- -- -- 10 -- -- -- 8 -- -- -- 10 -- -- -- 23 11 1.083 30.370 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.127 31.056 15 -- -- -- -- 10 -- -- -- 10 -- -- -- -- 15 9 1.181 31.900 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.234 32.545 7 -- -- -- -- -- -- -- -- -- 7 7 1.290 33.253 3 -- -- -- -- -- -- -- 3 6 1.360 34.412 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.393* 34.723* 4 1.371 33.656 Control Rod Density: % 5.23 3 1.278 30.895 2 0.981 23.450 k-effective: 0.99901 Bottom 1 0.280 6.849 Void Fraction: 0.470 Core Delta-P: psia 20.992  % AXIAL TILT -13.770 -8.959 Core Plate Delta-P: psia 16.437 AVG BOT 8ft/12ft 1.0628 1.0497 Coolant Temp: Deg-F 548.0 In Channel Flow: Mlb/hr 80.41 Active Channel Flow: Mlb/hr 77.47 Total Bypass Flow (%): 12.2 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.358 19 21 38 1.646 0.862 19 23 40 9.24 0.748 21.3 20 9 42 4 10.32 0.784 26.6 20 9 20 4 1.357 19 23 22 1.650 0.860 19 39 24 9.08 0.741 22.1 20 41 52 4 10.26 0.779 26.6 20 13 48 4 1.357 19 29 16 1.664 0.853 19 45 30 9.18 0.739 20.5 20 51 40 4 10.17 0.779 27.5 19 33 10 4 1.350 19 15 32 1.664 0.853 19 25 24 6.75 0.733 51.5 17 35 10 5 10.22 0.775 26.5 19 31 50 4 1.346 19 31 12 1.664 0.853 19 29 16 9.02 0.733 21.7 19 31 50 4 10.19 0.775 26.7 20 19 10 4 1.340 19 49 32 1.670 0.850 19 23 26 8.97 0.732 22.1 19 33 10 4 10.22 0.774 26.4 19 15 48 4 1.335 19 27 14 1.673 0.849 19 31 34 9.04 0.732 21.3 20 13 48 4 10.27 0.774 25.8 20 11 44 4 1.334 19 25 24 1.683 0.844 19 19 36 6.93 0.732 49.0 16 31 10 5 10.23 0.774 26.2 19 47 16 4 1.333 19 19 40 1.683 0.844 19 31 12 6.83 0.731 50.2 17 51 36 5 10.43 0.773 23.9 20 51 40 4 1.331 19 23 26 1.688 0.841 19 49 30 9.11 0.731 20.2 19 9 18 4 10.16 0.769 26.3 20 17 12 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-38 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 27310.8 Exposure: MWd/MTU (GWd) 12206.1 (1688.71 ) Delta E: MWd/MTU, (GWd) 0.7 ( 0.10 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.240 6.050 13 0.402 0.417 47 56 Inlet Subcooling: Btu/lbm -29.51 24 0.633 16.786 14 0.353 0.433 9 52 Flow: Mlb/hr 94.20 ( 91.90 %) 23 0.796 21.530 15 0.470 0.475 17 4 22 0.859 24.053 16 1.028 1.176 21 30 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.899 25.787 17 0.831 1.159 21 14 59 -- -- -- -- -- -- -- 59 20 0.919 27.062 18 1.018 1.146 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.927 27.656 19 1.245 1.352 21 12 51 -- -- -- -- -- 16 -- -- -- -- -- 51 18 0.946 28.412 20 1.170 1.325 21 10 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.955 28.932 21 1.172 1.325 37 10 43 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 43 16 0.958 29.415 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 0.974 30.300 35 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 35 14 0.962 30.308 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.048 28.867 27 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 27 12 1.076 29.831 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.103 30.371 19 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 19 10 1.148 31.057 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.204 31.901 11 -- -- -- -- -- 16 -- -- -- -- -- 11 8 1.258 32.546 7 -- -- -- -- -- -- -- -- -- 7 7 1.314 33.253 3 -- -- -- -- -- -- -- 3 6 1.385 34.413 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.417* 34.724* 4 1.394 33.657 Control Rod Density: % 5.41 3 1.300 30.896 2 0.999 23.451 k-effective: 0.99888 Bottom 1 0.286 6.849 Void Fraction: 0.469 Core Delta-P: psia 21.957  % AXIAL TILT -15.923 -8.959 Core Plate Delta-P: psia 17.400 AVG BOT 8ft/12ft 1.0799 1.0497 Coolant Temp: Deg-F 548.2 In Channel Flow: Mlb/hr 82.76 Active Channel Flow: Mlb/hr 79.74 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.352 19 21 12 1.679 0.846 20 21 10 9.82 0.795 21.3 20 9 20 4 11.12 0.844 26.6 20 9 20 4 1.351 19 15 44 1.722 0.825 19 39 50 9.64 0.787 22.1 20 19 10 4 10.88 0.833 27.4 20 19 52 4 1.344 19 17 16 1.730 0.821 19 29 46 9.68 0.786 21.7 20 11 18 4 11.09 0.832 25.3 19 49 40 4 1.341 19 21 28 1.731 0.820 19 21 34 9.73 0.782 20.5 20 9 22 4 10.96 0.830 26.4 20 11 44 4 1.341 19 19 14 1.734 0.819 19 15 44 9.58 0.781 22.0 20 17 12 4 10.87 0.825 26.6 20 13 48 4 1.339 19 13 42 1.735 0.819 19 27 44 9.61 0.780 21.5 19 13 16 4 10.84 0.823 26.7 19 39 50 4 1.337 19 29 16 1.737 0.818 20 17 50 7.34 0.780 49.6 16 11 20 5 10.96 0.823 25.4 20 9 40 4 1.333 19 17 42 1.741 0.815 20 19 52 9.58 0.777 21.6 19 15 14 4 10.80 0.822 26.9 20 43 12 4 1.333 19 23 26 1.741 0.815 19 33 14 7.29 0.776 49.8 16 41 50 5 10.86 0.821 26.2 19 47 16 4 1.328 19 29 20 1.744 0.814 19 37 14 9.46 0.774 22.5 20 39 10 4 10.82 0.820 26.4 19 15 48 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-39 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 27604.7 Exposure: MWd/MTU (GWd) 12500.0 (1729.37 ) Delta E: MWd/MTU, (GWd) 293.9 ( 40.66 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.242 6.125 13 0.401 0.415 47 56 Inlet Subcooling: Btu/lbm -29.65 24 0.638 16.986 14 0.352 0.432 9 52 Flow: Mlb/hr 93.79 ( 91.50 %) 23 0.802 21.781 15 0.469 0.473 17 4 22 0.866 24.326 16 1.026 1.175 21 30 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.907 26.073 17 0.829 1.158 21 14 59 -- -- -- -- -- -- -- 59 20 0.928 27.355 18 1.016 1.144 21 26 55 -- -- -- -- -- -- -- -- -- 55 19 0.937 27.951 19 1.248 1.354 21 12 51 -- -- -- -- -- 16 -- -- -- -- -- 51 18 0.957 28.714 20 1.171 1.326 21 10 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.967 29.237 21 1.173 1.326 37 10 43 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 43 16 0.971 29.721 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 0.987 30.611 35 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 35 14 0.974 30.615 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.059 29.156 27 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 27 12 1.086 30.128 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.113 30.675 19 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 19 10 1.157 31.374 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.210 32.233 11 -- -- -- -- -- 16 -- -- -- -- -- 11 8 1.258 32.892 7 -- -- -- -- -- -- -- -- -- 7 7 1.307 33.614 3 -- -- -- -- -- -- -- 3 6 1.367 34.791 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.389* 35.110* 4 1.360 34.035 Control Rod Density: % 5.41 3 1.267 31.249 2 0.973 23.722 k-effective: 0.99901 Bottom 1 0.278 6.930 Void Fraction: 0.467 Core Delta-P: psia 21.785  % AXIAL TILT -14.992 -8.957 Core Plate Delta-P: psia 17.229 AVG BOT 8ft/12ft 1.0758 1.0496 Coolant Temp: Deg-F 548.1 In Channel Flow: Mlb/hr 82.41 Active Channel Flow: Mlb/hr 79.41 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.354 19 21 12 1.672 0.849 20 21 10 9.56 0.778 21.9 20 9 20 4 10.78 0.824 27.3 20 9 20 4 1.351 19 15 44 1.713 0.829 19 39 50 9.40 0.768 22.3 20 11 18 4 10.78 0.814 26.0 19 49 40 4 1.345 19 21 28 1.716 0.828 19 31 16 9.35 0.767 22.7 20 19 10 4 10.52 0.810 28.1 20 41 52 4 1.344 19 17 16 1.722 0.825 19 21 34 9.48 0.767 21.1 20 9 22 4 10.62 0.810 27.0 20 11 44 4 1.342 19 19 14 1.723 0.824 19 27 44 7.17 0.765 50.0 16 11 20 5 10.66 0.805 26.1 20 9 40 4 1.341 19 29 16 1.728 0.822 20 17 50 9.29 0.761 22.6 20 17 12 4 10.52 0.804 27.2 20 13 48 4 1.340 19 13 42 1.728 0.822 19 15 44 7.11 0.760 50.2 16 41 50 5 10.53 0.801 26.8 19 47 16 4 1.336 19 23 26 1.730 0.821 19 33 14 9.32 0.760 22.1 19 13 16 4 10.48 0.801 27.3 19 39 50 4 1.334 19 17 42 1.731 0.820 19 31 42 9.39 0.758 20.8 19 9 18 4 10.44 0.800 27.6 20 43 12 4 1.334 19 29 20 1.732 0.820 19 37 14 9.28 0.757 22.2 19 15 14 4 10.48 0.799 27.1 19 15 48 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-40 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 28104.6 Exposure: MWd/MTU (GWd) 13000.0 (1798.54 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.247 6.255 13 0.398 0.413 47 56 Inlet Subcooling: Btu/lbm -29.93 24 0.649 17.330 14 0.350 0.430 9 52 Flow: Mlb/hr 92.97 ( 90.70 %) 23 0.816 22.214 15 0.466 0.470 17 4 22 0.881 24.797 16 1.024 1.172 21 30 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.925 26.567 17 0.827 1.156 21 14 59 -- -- -- -- -- -- -- 59 20 0.948 27.861 18 1.013 1.142 21 26 55 -- -- -- -- -- -- -- -- -- 55 19 0.959 28.463 19 1.252 1.356 21 12 51 -- -- -- -- -- 16 -- -- -- -- -- 51 18 0.981 29.237 20 1.172 1.327 21 10 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 0.993 29.766 21 1.174 1.326 37 10 43 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 43 16 0.997 30.252 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.014 31.151 35 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 35 14 0.998 31.147 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.083 29.657 27 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 27 12 1.108 30.640 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.132 31.200 19 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 19 10 1.173 31.919 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.219 32.801 11 -- -- -- -- -- 16 -- -- -- -- -- 11 8 1.257 33.480 7 -- -- -- -- -- -- -- -- -- 7 7 1.290 34.221 3 -- -- -- -- -- -- -- 3 6 1.329 35.422 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.331* 35.746* 4 1.290 34.655 Control Rod Density: % 5.41 3 1.197 31.825 2 0.920 24.165 k-effective: 0.99919 Bottom 1 0.263 7.063 Void Fraction: 0.462 Core Delta-P: psia 21.440  % AXIAL TILT -12.969 -8.926 Core Plate Delta-P: psia 16.883 AVG BOT 8ft/12ft 1.0666 1.0493 Coolant Temp: Deg-F 548.0 In Channel Flow: Mlb/hr 81.71 Active Channel Flow: Mlb/hr 78.75 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.356 19 21 12 1.657 0.857 20 21 10 9.04 0.742 22.9 20 9 42 4 10.12 0.781 28.4 20 9 20 4 1.352 19 15 44 1.684 0.843 19 31 16 6.84 0.737 50.8 16 11 20 5 10.16 0.775 27.1 19 49 40 4 1.352 19 21 28 1.696 0.837 19 31 42 8.98 0.732 22.0 20 9 40 4 9.89 0.770 29.3 20 41 52 5 1.347 19 29 16 1.697 0.837 19 39 50 6.77 0.731 50.9 16 41 50 5 9.97 0.769 28.4 20 11 44 5 1.345 19 17 16 1.700 0.835 19 39 34 8.86 0.730 23.2 20 11 18 4 10.06 0.767 27.2 20 9 40 4 1.343 19 19 14 1.702 0.834 19 27 44 8.80 0.728 23.7 20 19 10 4 9.89 0.764 28.4 19 39 50 5 1.342 19 29 20 1.709 0.831 20 17 50 8.90 0.724 21.8 19 9 18 4 9.83 0.762 28.8 20 43 12 5 1.342 19 23 26 1.712 0.829 19 33 14 8.73 0.721 23.6 20 17 12 5 9.91 0.761 27.9 19 47 16 4 1.341 19 13 42 1.712 0.829 19 37 14 8.76 0.720 23.0 19 13 16 4 9.86 0.761 28.3 20 13 14 4 1.337 19 27 18 1.713 0.829 19 15 44 6.65 0.719 51.1 16 13 18 5 9.85 0.758 28.1 19 15 48 4

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-41 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 28604.5 Exposure: MWd/MTU (GWd) 13500.0 (1867.72 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.252 6.386 13 0.396 0.411 47 56 Inlet Subcooling: Btu/lbm -30.11 24 0.661 17.681 14 0.348 0.428 9 52 Flow: Mlb/hr 92.45 ( 90.20 %) 23 0.830 22.655 15 0.464 0.468 17 4 22 0.899 25.278 16 1.022 1.169 21 30 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.946 27.072 17 0.825 1.154 21 14 59 -- -- -- -- -- -- -- 59 20 0.972 28.379 18 1.010 1.139 21 26 55 -- -- -- -- -- -- -- -- -- 55 19 0.986 28.988 19 1.255 1.359 21 12 51 -- -- -- -- -- 16 -- -- -- -- -- 51 18 1.010 29.774 20 1.174 1.330 21 10 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.023 30.310 21 1.176 1.328 23 10 43 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 43 16 1.029 30.799 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.045 31.706 35 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 35 14 1.028 31.694 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.112 30.169 27 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 27 12 1.135 31.164 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.155 31.734 19 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 19 10 1.191 32.472 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.229 33.374 11 -- -- -- -- -- 16 -- -- -- -- -- 11 8 1.253 34.067 7 -- -- -- -- -- -- -- -- -- 7 7 1.267 34.820 3 -- -- -- -- -- -- -- 3 6 1.284* 36.033 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.264 36.353* 4 1.209 35.239 Control Rod Density: % 5.41 3 1.116 32.366 2 0.859 24.581 k-effective: 0.99940 Bottom 1 0.245 7.188 Void Fraction: 0.455 Core Delta-P: psia 21.188  % AXIAL TILT -10.576 -8.858 Core Plate Delta-P: psia 16.631 AVG BOT 8ft/12ft 1.0559 1.0489 Coolant Temp: Deg-F 547.8 In Channel Flow: Mlb/hr 81.30 Active Channel Flow: Mlb/hr 78.36 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00006

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.359 19 21 12 1.643 0.865 20 21 10 6.57 0.707 50.8 16 49 42 6 9.53 0.745 29.7 20 51 42 5 1.356 19 21 28 1.655 0.858 19 31 16 6.55 0.706 50.9 16 41 50 6 9.58 0.740 28.4 19 49 40 5 1.352 19 15 44 1.665 0.853 19 31 42 8.47 0.702 23.9 20 9 42 5 9.34 0.734 30.3 20 41 52 5 1.351 19 29 16 1.677 0.847 19 39 34 6.44 0.696 51.1 16 13 18 6 9.42 0.734 29.4 20 11 44 5 1.348 19 29 20 1.680 0.845 19 39 50 8.37 0.695 24.3 20 11 18 5 9.63 0.730 26.5 20 51 22 5 1.346 19 17 16 1.685 0.843 19 27 44 8.45 0.695 23.1 20 9 40 5 9.37 0.730 29.4 19 39 50 5 1.346 19 19 14 1.687 0.842 19 23 26 6.40 0.693 51.2 16 17 48 6 7.04 0.728 55.5 16 49 42 6 1.344 19 23 26 1.690 0.840 20 17 50 8.31 0.692 24.6 20 19 10 5 7.03 0.727 55.6 16 41 50 6 1.343 19 27 18 1.693 0.839 19 37 14 6.60 0.691 48.2 16 15 16 6 9.36 0.727 29.0 19 47 16 5 1.343 19 13 42 1.696 0.837 19 33 14 8.27 0.689 24.6 20 17 12 5 9.28 0.726 29.8 20 43 12 5

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-42 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 29104.6 Exposure: MWd/MTU (GWd) 14000.0 (1936.89 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.257 6.521 13 0.394 0.409 47 56 Inlet Subcooling: Btu/lbm -30.25 24 0.674 18.038 14 0.346 0.427 9 52 Flow: Mlb/hr 92.05 ( 89.80 %) 23 0.847 23.104 15 0.462 0.466 17 4 22 0.920 25.769 16 1.020 1.164 21 32 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.970 27.589 17 0.823 1.153 21 14 59 -- -- -- -- -- -- -- 59 20 0.999 28.911 18 1.006 1.135 21 36 55 -- -- -- -- -- -- -- -- -- 55 19 1.016 29.528 19 1.257 1.362 21 12 51 -- -- -- -- -- 16 -- -- -- -- -- 51 18 1.043 30.328 20 1.177 1.333 21 10 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.058 30.871 21 1.179 1.329 23 10 43 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 43 16 1.064 31.364 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.080 32.280 35 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 35 14 1.060 32.257 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.144 30.696 27 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 27 12 1.164 31.701 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.179 32.279 19 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 19 10 1.209 33.033 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.236 33.950 11 -- -- -- -- -- 16 -- -- -- -- -- 11 8 1.244* 34.651 7 -- -- -- -- -- -- -- -- -- 7 7 1.239 35.406 3 -- -- -- -- -- -- -- 3 6 1.232 36.622 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.192 36.927* 4 1.123 35.785 Control Rod Density: % 5.41 3 1.029 32.868 2 0.794 24.968 k-effective: 0.99956 Bottom 1 0.227 7.304 Void Fraction: 0.447 Core Delta-P: psia 20.963  % AXIAL TILT -7.876 -8.748 Core Plate Delta-P: psia 16.407 AVG BOT 8ft/12ft 1.0436 1.0482 Coolant Temp: Deg-F 547.6 In Channel Flow: Mlb/hr 80.98 Active Channel Flow: Mlb/hr 78.07 Total Bypass Flow (%): 12.0 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.362 19 21 12 1.626 0.873 19 31 16 6.31 0.685 51.5 16 49 42 6 7.58 0.722 48.7 16 31 10 18 1.357 19 21 28 1.627 0.873 20 21 10 6.28 0.683 51.5 16 41 50 6 9.06 0.707 29.6 20 51 42 6 1.354 19 29 16 1.636 0.868 19 31 42 6.82 0.679 43.5 16 29 52 18 6.76 0.705 56.2 16 49 42 6 1.353 19 15 44 1.652 0.859 19 39 34 6.16 0.671 51.8 16 13 18 6 9.03 0.704 29.4 19 49 40 5 1.352 19 29 20 1.658 0.857 19 39 50 8.02 0.669 24.8 20 9 42 5 6.74 0.703 56.3 16 41 50 6 1.348 19 19 14 1.663 0.854 19 23 26 6.13 0.668 51.9 16 17 48 6 8.97 0.703 30.0 20 41 52 6 1.348 19 23 14 1.665 0.853 19 33 18 6.32 0.666 48.8 16 15 16 6 8.98 0.699 29.3 20 11 44 6 1.348 19 17 16 1.669 0.851 19 23 32 7.99 0.662 24.0 20 9 40 5 9.00 0.699 29.1 19 39 50 6 1.348 19 27 18 1.669 0.851 20 17 50 7.89 0.660 25.2 20 11 18 5 8.91 0.696 29.6 20 43 12 6 1.346 19 23 30 1.671 0.850 19 33 48 7.86 0.659 25.4 20 19 10 5 6.61 0.695 56.8 16 47 18 6

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-43 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 29604.5 Exposure: MWd/MTU (GWd) 14500.0 (2006.07 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.263 6.658 13 0.392 0.408 47 56 Inlet Subcooling: Btu/lbm -30.22 24 0.689 18.403 14 0.344 0.427 9 52 Flow: Mlb/hr 92.15 ( 89.90 %) 23 0.866 23.562 15 0.461 0.465 17 4 22 0.944 26.272 16 1.018 1.161 23 12 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.998 28.120 17 0.822 1.153 21 14 59 -- -- -- -- -- -- -- 59 20 1.031 29.459 18 1.002 1.132 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.050 30.086 19 1.259 1.366 21 12 51 -- -- -- -- -- 16 -- -- -- -- -- 51 18 1.080 30.901 20 1.181 1.337 21 10 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.096 31.452 21 1.182 1.331 23 10 43 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 43 16 1.101 31.948 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.117 32.873 35 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 35 14 1.093 32.838 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.177 31.238 27 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 27 12 1.193 32.252 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.202 32.836 19 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 19 10 1.224 33.602 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.240* 34.529 11 -- -- -- -- -- 16 -- -- -- -- -- 11 8 1.231 35.230 7 -- -- -- -- -- -- -- -- -- 7 7 1.205 35.977 3 -- -- -- -- -- -- -- 3 6 1.176 37.185 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.116 37.467* 4 1.035 36.290 Control Rod Density: % 5.41 3 0.940 33.328 2 0.727 25.323 k-effective: 0.99970 Bottom 1 0.208 7.411 Void Fraction: 0.438 Core Delta-P: psia 20.902  % AXIAL TILT -4.905 -8.592 Core Plate Delta-P: psia 16.346 AVG BOT 8ft/12ft 1.0298 1.0474 Coolant Temp: Deg-F 547.4 In Channel Flow: Mlb/hr 81.13 Active Channel Flow: Mlb/hr 78.24 Total Bypass Flow (%): 12.0 (of total core flow) Total Water Rod Flow (%): 3.1 (of total core flow) Source Convergence 0.00006

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.366 19 21 12 1.603 0.886 19 31 16 7.02 0.704 44.3 16 29 52 18 7.79 0.749 49.5 16 31 10 18 1.357 19 21 28 1.614 0.880 19 31 42 6.81 0.680 43.7 16 29 14 18 7.36 0.710 49.9 16 31 48 18 1.356 19 29 16 1.616 0.879 20 21 10 8.56 0.676 18.4 19 29 50 18 9.73 0.708 21.9 19 31 50 18 1.355 19 15 44 1.631 0.871 19 39 50 6.35 0.671 49.0 16 41 12 8 7.26 0.703 50.1 16 33 50 18 1.354 19 31 42 1.633 0.869 19 39 34 6.66 0.667 44.0 16 27 50 18 7.15 0.702 51.3 17 39 48 15 1.353 19 23 14 1.644 0.864 19 33 48 6.31 0.667 49.1 16 49 20 8 6.89 0.695 53.6 16 41 50 8 1.352 19 19 14 1.645 0.863 19 25 16 6.34 0.665 48.3 16 17 48 9 7.06 0.693 51.4 17 25 10 15 1.350 19 27 18 1.645 0.863 19 23 26 6.34 0.664 48.3 16 13 44 9 6.75 0.692 54.9 16 17 48 9 1.350 19 17 16 1.646 0.863 19 33 18 6.58 0.663 44.7 16 37 12 15 7.06 0.691 51.0 16 23 50 15 1.348 19 13 42 1.646 0.862 19 23 48 6.43 0.661 46.6 16 15 46 9 6.63 0.690 56.0 16 47 18 8

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-44 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 29857.3 Exposure: MWd/MTU (GWd) 14752.8 (2041.05 ) Delta E: MWd/MTU, (GWd) 252.8 ( 34.98 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.266 6.729 13 0.391 0.407 47 56 Inlet Subcooling: Btu/lbm -30.11 24 0.697 18.590 14 0.344 0.426 9 52 Flow: Mlb/hr 92.45 ( 90.20 %) 23 0.876 23.798 15 0.460 0.464 17 4 22 0.957 26.531 16 1.017 1.160 23 12 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.014 28.394 17 0.822 1.153 21 14 59 -- -- -- -- -- -- -- 59 20 1.048 29.742 18 1.000 1.131 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.068 30.375 19 1.260 1.369 21 12 51 -- -- -- -- -- 16 -- -- -- -- -- 51 18 1.098 31.198 20 1.183 1.339 21 10 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.115 31.754 21 1.184 1.332 23 10 43 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 43 16 1.120 32.251 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.134 33.180 35 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 35 14 1.109 33.138 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.192 31.518 27 -- -- -- 8 -- -- -- 0 -- -- -- 8 -- -- -- 27 12 1.206 32.535 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.212 33.121 19 -- -- -- -- -- 8 -- -- -- 8 -- -- -- -- -- 19 10 1.230 33.892 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.239* 34.822 11 -- -- -- -- -- 16 -- -- -- -- -- 11 8 1.223 35.520 7 -- -- -- -- -- -- -- -- -- 7 7 1.186 36.260 3 -- -- -- -- -- -- -- 3 6 1.147 37.460 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.078 37.727* 4 0.992 36.529 Control Rod Density: % 5.41 3 0.899 33.546 2 0.695 25.491 k-effective: 0.99978 Bottom 1 0.199 7.462 Void Fraction: 0.433 Core Delta-P: psia 20.956  % AXIAL TILT -3.375 -8.497 Core Plate Delta-P: psia 16.399 AVG BOT 8ft/12ft 1.0225 1.0469 Coolant Temp: Deg-F 547.4 In Channel Flow: Mlb/hr 81.43 Active Channel Flow: Mlb/hr 78.54 Total Bypass Flow (%): 11.9 (of total core flow) Total Water Rod Flow (%): 3.1 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.369 19 21 12 1.594 0.891 19 31 16 7.12 0.717 44.6 16 29 52 18 7.90 0.763 49.9 16 31 10 18 1.356 19 29 16 1.606 0.884 19 31 42 6.91 0.692 44.1 16 29 14 18 7.46 0.724 50.3 16 31 48 18 1.356 19 15 44 1.612 0.881 20 21 10 8.72 0.692 18.8 19 29 50 18 9.88 0.722 22.4 19 31 50 18 1.355 19 21 28 1.619 0.877 19 39 50 6.76 0.680 44.4 16 27 50 18 7.37 0.717 50.5 16 33 50 18 1.355 19 23 14 1.627 0.873 19 39 34 6.68 0.676 45.1 16 37 12 15 7.26 0.716 51.7 17 39 48 15 1.354 19 31 42 1.633 0.870 19 33 48 6.42 0.672 48.2 16 41 12 9 7.16 0.707 51.8 17 25 10 15 1.353 19 19 14 1.634 0.869 19 25 16 6.63 0.671 45.1 17 21 48 15 7.17 0.704 51.4 16 23 50 15 1.351 19 27 18 1.635 0.869 19 23 48 6.55 0.668 45.7 16 29 44 15 7.18 0.701 50.9 16 41 50 15 1.351 19 17 16 1.637 0.867 19 33 18 6.52 0.668 46.2 16 15 42 15 6.99 0.698 52.8 16 15 42 15 1.349 19 13 42 1.639 0.866 19 25 50 6.34 0.666 48.6 16 13 44 9 7.04 0.696 52.0 17 47 40 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-45 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 29858.0 Exposure: MWd/MTU (GWd) 14753.6 (2041.15 ) Delta E: MWd/MTU, (GWd) 0.7 ( 0.10 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.215 6.729 13 0.391 0.403 5 48 Inlet Subcooling: Btu/lbm -31.22 24 0.562 18.591 14 0.342 0.421 9 52 Flow: Mlb/hr 89.38 ( 87.20 %) 23 0.712 23.798 15 0.456 0.459 3 18 22 0.791 26.532 16 1.018 1.158 11 38 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.861 28.395 17 0.823 1.153 39 14 59 -- -- -- -- -- -- -- 59 20 0.928 29.743 18 0.979 1.145 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 0.979 30.375 19 1.260 1.365 11 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.035 31.199 20 1.186 1.334 9 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.078 31.755 21 1.195 1.335 51 38 43 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 43 16 1.111 32.252 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.147 33.181 35 -- -- -- -- 0 -- -- -- -- -- 0 -- -- -- -- 35 14 1.139 33.139 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.236 31.519 27 -- -- -- -- 0 -- -- -- -- -- 0 -- -- -- -- 27 12 1.258 32.536 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.270 33.122 19 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 19 10 1.291 33.893 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.303* 34.823 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.290 35.521 7 -- -- -- -- -- -- -- -- -- 7 7 1.259 36.261 3 -- -- -- -- -- -- -- 3 6 1.229 37.461 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.173 37.727* 4 1.100 36.530 Control Rod Density: % 4.32 3 1.013 33.546 2 0.792 25.492 k-effective: 1.00005 Bottom 1 0.228 7.462 Void Fraction: 0.456 Core Delta-P: psia 20.216  % AXIAL TILT -11.144 -8.496 Core Plate Delta-P: psia 15.663 AVG BOT 8ft/12ft 1.0797 1.0469 Coolant Temp: Deg-F 547.5 In Channel Flow: Mlb/hr 78.52 Active Channel Flow: Mlb/hr 75.66 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.365 19 11 40 1.618 0.878 19 33 32 6.60 0.691 48.2 16 41 12 9 7.20 0.718 52.7 16 41 50 9 1.364 19 17 16 1.618 0.878 19 17 16 6.54 0.688 48.7 16 17 48 9 7.05 0.714 54.0 16 49 42 8 1.362 19 15 44 1.621 0.876 19 15 18 6.48 0.688 49.4 16 49 20 8 7.19 0.713 52.2 17 39 48 15 1.361 19 21 12 1.622 0.876 19 31 34 6.52 0.686 48.6 16 13 44 9 6.92 0.712 55.2 16 17 48 9 1.360 19 19 14 1.623 0.875 19 41 18 6.64 0.685 46.9 16 37 12 9 7.19 0.712 52.0 17 47 40 15 1.358 19 13 42 1.625 0.874 19 41 48 6.48 0.685 49.0 17 35 10 9 7.00 0.710 54.0 16 15 16 9 1.358 19 31 34 1.625 0.874 19 39 50 6.58 0.684 47.6 17 39 14 9 6.90 0.709 55.1 16 47 18 9 1.358 19 27 30 1.626 0.873 19 17 42 6.61 0.683 46.9 16 15 46 9 7.09 0.708 52.8 16 15 42 15 1.355 19 19 44 1.627 0.873 19 47 42 6.65 0.681 46.2 16 49 24 9 6.98 0.707 53.9 16 23 50 9 1.351 19 17 42 1.632 0.870 20 51 40 6.55 0.681 47.5 17 13 40 9 6.91 0.702 54.2 16 49 24 9

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-46 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 30204.5 Exposure: MWd/MTU (GWd) 15100.0 (2089.08 ) Delta E: MWd/MTU, (GWd) 346.4 ( 47.93 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.221 6.809 13 0.389 0.403 5 48 Inlet Subcooling: Btu/lbm -30.66 24 0.577 18.802 14 0.341 0.421 9 52 Flow: Mlb/hr 90.92 ( 88.70 %) 23 0.731 24.066 15 0.455 0.458 3 18 22 0.814 26.832 16 1.016 1.158 11 38 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.887 28.722 17 0.823 1.153 39 14 59 -- -- -- -- -- -- -- 59 20 0.956 30.095 18 0.976 1.144 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.008 30.747 19 1.260 1.369 11 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.066 31.591 20 1.190 1.339 9 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.109 32.164 21 1.198 1.337 51 38 43 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 43 16 1.142 32.673 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.176 33.615 35 -- -- -- -- 0 -- -- -- -- -- 0 -- -- -- -- 35 14 1.165 33.570 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.262 31.923 27 -- -- -- -- 0 -- -- -- -- -- 0 -- -- -- -- 27 12 1.280 32.946 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.286 33.536 19 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 19 10 1.300 34.313 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.301* 35.245 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.275 35.937 7 -- -- -- -- -- -- -- -- -- 7 7 1.230 36.665 3 -- -- -- -- -- -- -- 3 6 1.184 37.852 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.113 38.098* 4 1.030 36.875 Control Rod Density: % 4.32 3 0.941 33.863 2 0.736 25.740 k-effective: 1.00011 Bottom 1 0.212 7.537 Void Fraction: 0.446 Core Delta-P: psia 20.634  % AXIAL TILT -8.576 -8.435 Core Plate Delta-P: psia 16.081 AVG BOT 8ft/12ft 1.0673 1.0468 Coolant Temp: Deg-F 547.4 In Channel Flow: Mlb/hr 79.95 Active Channel Flow: Mlb/hr 77.06 Total Bypass Flow (%): 12.1 (of total core flow) Total Water Rod Flow (%): 3.2 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.369 19 11 40 1.616 0.885 19 17 16 6.79 0.699 46.7 16 15 42 15 7.38 0.736 52.7 17 39 48 15 1.365 19 17 16 1.619 0.883 19 15 18 6.81 0.696 46.0 16 19 16 15 7.38 0.735 52.5 17 47 40 15 1.364 19 21 12 1.620 0.883 19 39 50 6.61 0.695 48.7 16 41 12 9 7.26 0.729 53.3 16 15 42 15 1.363 19 15 44 1.621 0.882 19 41 48 6.83 0.695 45.6 17 21 14 15 7.20 0.723 53.2 16 41 50 9 1.362 19 19 14 1.622 0.882 19 41 18 6.83 0.695 45.4 17 13 40 15 7.26 0.722 52.4 16 41 16 15 1.361 19 13 42 1.623 0.881 19 47 42 6.57 0.692 48.7 16 49 20 9 7.26 0.721 52.3 16 15 16 15 1.355 19 19 44 1.623 0.881 19 33 32 6.54 0.691 49.2 16 17 48 9 7.16 0.719 53.3 16 49 42 9 1.354 19 31 34 1.626 0.880 19 31 34 6.52 0.689 49.1 16 13 44 9 7.20 0.719 52.8 16 17 40 15 1.353 19 27 30 1.626 0.880 19 17 42 6.64 0.689 47.4 16 37 12 9 7.23 0.718 52.3 17 25 10 15 1.350 19 17 42 1.631 0.877 19 49 40 6.48 0.688 49.5 17 35 10 9 7.30 0.717 51.3 17 9 26 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-47 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 30704.5 Exposure: MWd/MTU (GWd) 15600.0 (2158.25 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.230 6.928 13 0.388 0.403 5 48 Inlet Subcooling: Btu/lbm -29.82 24 0.602 19.117 14 0.341 0.421 9 52 Flow: Mlb/hr 93.28 ( 91.00 %) 23 0.763 24.466 15 0.455 0.458 3 18 22 0.852 27.281 16 1.014 1.159 49 38 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.928 29.212 17 0.824 1.154 39 14 59 -- -- -- -- -- -- -- 59 20 1.001 30.624 18 0.970 1.142 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.054 31.303 19 1.259 1.376 11 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.112 32.179 20 1.197 1.347 9 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.154 32.774 21 1.203 1.341 51 38 43 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 43 16 1.183 33.300 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.215 34.260 35 -- -- -- -- 0 -- -- -- -- -- 0 -- -- -- -- 35 14 1.198 34.207 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.294 32.520 27 -- -- -- -- 0 -- -- -- -- -- 0 -- -- -- -- 27 12 1.307 33.550 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.304 34.141 19 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 19 10 1.307* 34.922 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.294 35.852 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.250 36.528 7 -- -- -- -- -- -- -- -- -- 7 7 1.184 37.229 3 -- -- -- -- -- -- -- 3 6 1.117 38.390 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.028 38.599* 4 0.932 37.334 Control Rod Density: % 4.32 3 0.842 34.280 2 0.659 26.066 k-effective: 1.00014 Bottom 1 0.190 7.635 Void Fraction: 0.431 Core Delta-P: psia 21.287  % AXIAL TILT -4.725 -8.295 Core Plate Delta-P: psia 16.734 AVG BOT 8ft/12ft 1.0478 1.0464 Coolant Temp: Deg-F 547.3 In Channel Flow: Mlb/hr 82.13 Active Channel Flow: Mlb/hr 79.21 Total Bypass Flow (%): 11.9 (of total core flow) Total Water Rod Flow (%): 3.1 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.376 19 11 40 1.612 0.887 19 39 50 6.99 0.725 47.4 16 15 42 15 7.63 0.769 53.5 17 39 48 15 1.368 19 21 12 1.614 0.886 19 17 16 7.05 0.723 46.3 17 21 14 15 7.63 0.768 53.3 17 47 40 15 1.366 19 17 16 1.616 0.885 19 41 48 7.06 0.723 46.2 17 13 40 15 7.47 0.758 54.1 16 15 42 15 1.365 19 19 14 1.617 0.885 19 15 18 7.01 0.722 46.7 16 19 16 15 7.51 0.752 53.0 16 15 16 15 1.365 19 15 44 1.618 0.884 19 47 42 6.94 0.716 46.8 16 15 46 15 7.49 0.752 53.1 16 41 16 15 1.364 19 13 42 1.619 0.883 19 41 18 6.94 0.716 46.8 16 17 14 15 7.57 0.751 52.2 16 41 50 15 1.353 19 19 44 1.620 0.883 19 49 40 7.03 0.715 45.5 16 49 24 15 7.47 0.749 53.1 17 25 10 15 1.350 19 39 16 1.624 0.881 19 17 42 6.98 0.715 46.3 16 37 12 15 7.55 0.748 52.1 17 9 26 15 1.348 19 23 48 1.624 0.880 20 9 40 6.93 0.714 46.8 16 13 44 15 7.54 0.748 52.1 16 49 42 15 1.347 19 17 42 1.626 0.879 20 39 10 6.92 0.711 46.6 16 19 12 15 7.48 0.746 52.6 16 17 48 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-48 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 31204.5 Exposure: MWd/MTU (GWd) 16100.0 (2227.43 ) Delta E: MWd/MTU, (GWd) 500.0 ( 69.17 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.240 7.053 13 0.387 0.403 5 48 Inlet Subcooling: Btu/lbm -28.54 24 0.631 19.447 14 0.340 0.422 9 52 Flow: Mlb/hr 97.17 ( 94.80 %) 23 0.799 24.884 15 0.455 0.458 3 18 22 0.893 27.752 16 1.011 1.160 49 38 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 0.973 29.725 17 0.826 1.155 13 40 59 -- -- -- -- -- -- -- 59 20 1.047 31.176 18 0.964 1.140 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.100 31.884 19 1.258 1.384 11 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.157 32.791 20 1.206 1.357 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.195 33.408 21 1.209 1.347 51 38 43 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 43 16 1.220 33.948 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.247 34.924 35 -- -- -- -- 0 -- -- -- -- -- 0 -- -- -- -- 35 14 1.225 34.861 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.321 33.131 27 -- -- -- -- 0 -- -- -- -- -- 0 -- -- -- -- 27 12 1.327* 34.166 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.316 34.753 19 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 19 10 1.307 35.534 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.281 36.454 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.222 37.106 7 -- -- -- -- -- -- -- -- -- 7 7 1.137 37.772 3 -- -- -- -- -- -- -- 3 6 1.052 38.897 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.948 39.061* 4 0.844 37.749 Control Rod Density: % 4.32 3 0.755 34.654 2 0.592 26.359 k-effective: 1.00011 Bottom 1 0.171 7.724 Void Fraction: 0.415 Core Delta-P: psia 22.472  % AXIAL TILT -0.950 -8.099 Core Plate Delta-P: psia 17.918 AVG BOT 8ft/12ft 1.0277 1.0457 Coolant Temp: Deg-F 547.2 In Channel Flow: Mlb/hr 85.70 Active Channel Flow: Mlb/hr 82.71 Total Bypass Flow (%): 11.8 (of total core flow) Total Water Rod Flow (%): 3.1 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.384 19 11 40 1.616 0.885 19 39 50 7.25 0.749 46.9 17 13 40 15 7.84 0.798 54.3 17 39 48 15 1.373 19 21 12 1.617 0.884 19 49 40 7.23 0.748 47.1 17 21 14 15 7.85 0.798 54.1 17 47 40 15 1.368 19 13 42 1.618 0.884 19 17 16 7.15 0.748 48.2 16 15 42 15 7.65 0.784 54.9 16 15 42 15 1.368 19 19 14 1.619 0.883 19 47 42 7.17 0.745 47.4 16 19 16 15 7.81 0.782 53.0 16 41 50 15 1.368 19 17 16 1.619 0.883 19 15 18 7.14 0.742 47.6 16 15 46 15 7.72 0.781 53.8 16 15 16 15 1.366 19 15 44 1.620 0.883 19 41 48 7.24 0.742 46.3 16 49 24 15 7.80 0.780 52.9 16 49 42 15 1.357 20 51 40 1.623 0.881 20 9 40 7.13 0.742 47.6 16 17 14 15 7.78 0.779 52.9 17 9 26 15 1.351 19 39 16 1.628 0.879 19 41 18 7.13 0.741 47.5 16 13 44 15 7.68 0.778 53.9 17 25 10 15 1.351 19 23 48 1.630 0.877 20 39 10 7.16 0.740 47.0 16 37 12 15 7.68 0.778 53.9 16 41 16 15 1.351 19 19 44 1.633 0.876 19 17 42 7.13 0.739 47.3 16 19 50 15 7.72 0.777 53.4 16 17 48 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-49 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 31604.5 Exposure: MWd/MTU (GWd) 16500.0 (2282.77 ) Delta E: MWd/MTU, (GWd) 400.0 ( 55.34 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.250 7.156 13 0.387 0.403 5 48 Inlet Subcooling: Btu/lbm -27.08 24 0.658 19.723 14 0.339 0.422 51 52 Flow: Mlb/hr 101.99 ( 99.50 %) 23 0.834 25.233 15 0.455 0.458 3 18 22 0.932 28.146 16 1.008 1.163 49 38 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.014 30.154 17 0.827 1.156 13 40 59 -- -- -- -- -- -- -- 59 20 1.089 31.637 18 0.958 1.138 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.139 32.368 19 1.256 1.391 11 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.194 33.299 20 1.214 1.367 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.228 33.931 21 1.216 1.352 51 38 43 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 43 16 1.247 34.481 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.270 35.467 35 -- -- -- -- 0 -- -- -- -- -- 0 -- -- -- -- 35 14 1.243 35.394 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.338 33.628 27 -- -- -- -- 0 -- -- -- -- -- 0 -- -- -- -- 27 12 1.340* 34.664 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.321 35.246 19 -- -- -- -- -- -- 0 -- 0 -- -- -- -- -- -- 19 10 1.304 36.022 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.267 36.931 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.194 37.558 7 -- -- -- -- -- -- -- -- -- 7 7 1.095 38.189 3 -- -- -- -- -- -- -- 3 6 0.998 39.281 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.884 39.403* 4 0.776 38.052 Control Rod Density: % 4.32 3 0.689 34.924 2 0.541 26.571 k-effective: 1.00014 Bottom 1 0.156 7.788 Void Fraction: 0.400 Core Delta-P: psia 24.029  % AXIAL TILT 2.255 -7.902 Core Plate Delta-P: psia 19.474 AVG BOT 8ft/12ft 1.0098 1.0449 Coolant Temp: Deg-F 547.2 In Channel Flow: Mlb/hr 90.10 Active Channel Flow: Mlb/hr 87.02 Total Bypass Flow (%): 11.7 (of total core flow) Total Water Rod Flow (%): 3.0 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.391 19 11 40 1.618 0.884 19 49 40 7.39 0.768 47.6 17 13 40 15 8.02 0.821 54.8 17 47 40 15 1.378 19 39 50 1.624 0.881 19 47 42 7.36 0.767 47.7 17 21 48 15 8.00 0.821 55.0 17 39 48 15 1.372 19 13 42 1.626 0.879 20 9 40 7.25 0.764 48.8 16 15 42 15 7.99 0.806 53.6 16 41 50 15 1.371 19 19 14 1.626 0.879 19 39 50 7.39 0.762 46.9 16 49 24 15 7.99 0.806 53.6 16 49 42 15 1.369 19 17 16 1.626 0.879 19 17 16 7.28 0.762 48.2 16 15 46 15 7.88 0.804 54.5 16 15 16 15 1.368 19 15 44 1.626 0.879 19 15 18 7.28 0.762 48.2 16 17 48 15 7.96 0.803 53.6 17 9 26 15 1.367 20 51 40 1.627 0.879 19 41 48 7.28 0.762 48.1 16 13 44 15 7.77 0.803 55.5 16 15 42 15 1.352 19 37 14 1.640 0.872 20 39 10 7.28 0.761 48.1 16 19 16 15 7.86 0.801 54.4 16 17 48 15 1.352 21 51 38 1.642 0.871 19 41 18 7.29 0.761 47.8 16 49 20 15 7.84 0.800 54.5 17 25 10 15 1.352 20 39 52 1.647 0.868 19 17 42 7.29 0.761 47.9 16 41 12 15 7.86 0.799 54.2 16 47 18 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-50 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 31605.5 Exposure: MWd/MTU (GWd) 16501.0 (2282.91 ) Delta E: MWd/MTU, (GWd) 1.0 ( 0.14 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.278 7.157 13 0.383 0.399 5 48 Inlet Subcooling: Btu/lbm -29.86 24 0.730 19.724 14 0.337 0.418 51 52 Flow: Mlb/hr 93.17 ( 90.90 %) 23 0.916 25.234 15 0.450 0.454 3 18 22 1.008 28.147 16 1.015 1.152 49 38 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.064 30.155 17 0.817 1.148 13 40 59 -- -- -- -- -- -- -- 59 20 1.119 31.639 18 0.969 1.131 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.154 32.369 19 1.265 1.373 11 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.195 33.300 20 1.194 1.345 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.218 33.933 21 1.197 1.333 51 38 43 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 43 16 1.228 34.482 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.243 35.469 35 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 35 14 1.212 35.395 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.301* 33.629 27 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 27 12 1.301 34.665 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.281 35.247 19 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 19 10 1.265 36.024 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.232 36.932 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.165 37.559 7 -- -- -- -- -- -- -- -- -- 7 7 1.073 38.190 3 -- -- -- -- -- -- -- 3 6 0.983 39.281 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.876 39.404* 4 0.772 38.053 Control Rod Density: % 3.78 3 0.689 34.925 2 0.542 26.571 k-effective: 1.00009 Bottom 1 0.156 7.789 Void Fraction: 0.408 Core Delta-P: psia 20.970  % AXIAL TILT 4.352 -7.901 Core Plate Delta-P: psia 16.418 AVG BOT 8ft/12ft 0.9901 1.0449 Coolant Temp: Deg-F 546.8 In Channel Flow: Mlb/hr 82.20 Active Channel Flow: Mlb/hr 79.34 Total Bypass Flow (%): 11.8 (of total core flow) Total Water Rod Flow (%): 3.1 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.373 19 11 40 1.570 0.911 19 49 40 7.21 0.750 47.6 17 13 40 15 7.82 0.800 54.8 17 47 40 15 1.360 19 39 50 1.575 0.908 19 47 42 7.19 0.749 47.7 17 21 48 15 7.79 0.800 55.0 17 39 48 15 1.356 19 13 42 1.578 0.906 19 17 16 7.09 0.747 48.8 16 15 42 15 7.78 0.785 53.6 16 49 42 15 1.355 19 19 14 1.578 0.906 19 41 48 7.21 0.744 46.9 16 49 24 15 7.78 0.785 53.7 16 41 50 15 1.354 19 17 16 1.578 0.906 19 15 18 7.12 0.744 48.1 16 19 16 15 7.59 0.784 55.5 16 15 42 15 1.353 19 15 44 1.580 0.905 19 41 18 7.10 0.743 48.2 16 15 46 15 7.68 0.783 54.5 16 15 16 15 1.348 19 39 16 1.580 0.905 19 39 50 7.10 0.743 48.2 16 17 48 15 7.75 0.782 53.6 17 9 26 15 1.346 19 37 14 1.583 0.904 20 9 40 7.10 0.743 48.1 16 13 44 15 7.68 0.780 54.1 16 17 48 15 1.346 19 47 24 1.585 0.902 19 17 42 7.11 0.741 47.8 16 49 20 15 7.64 0.780 54.5 17 25 10 15 1.345 20 51 40 1.592 0.898 19 39 16 7.10 0.741 47.9 16 41 12 15 7.68 0.778 54.0 16 47 18 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-51 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 32104.5 Exposure: MWd/MTU (GWd) 17000.0 (2351.94 ) Delta E: MWd/MTU, (GWd) 499.0 ( 69.04 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.292 7.307 13 0.382 0.398 5 48 Inlet Subcooling: Btu/lbm -27.29 24 0.769 20.124 14 0.335 0.418 51 52 Flow: Mlb/hr 101.27 ( 98.80 %) 23 0.965 25.737 15 0.450 0.454 3 18 22 1.063 28.704 16 1.012 1.155 49 24 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.120 30.743 17 0.820 1.150 47 22 59 -- -- -- -- -- -- -- 59 20 1.173 32.256 18 0.962 1.129 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.203 33.003 19 1.263 1.383 49 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.239 33.955 20 1.205 1.359 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.254 34.598 21 1.205 1.341 51 38 43 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 43 16 1.257 35.151 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.266 36.144 35 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 35 14 1.230 36.053 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.319* 34.240 27 -- -- -- -- 6 -- -- -- -- -- 6 -- -- -- -- 27 12 1.314 35.275 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.286 35.846 19 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 19 10 1.260 36.613 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.214 37.503 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.130 38.095 7 -- -- -- -- -- -- -- -- -- 7 7 1.021 38.679 3 -- -- -- -- -- -- -- 3 6 0.914 39.724 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.796 39.794* 4 0.689 38.394 Control Rod Density: % 3.78 3 0.609 35.228 2 0.480 26.810 k-effective: 1.00012 Bottom 1 0.139 7.861 Void Fraction: 0.386 Core Delta-P: psia 23.569  % AXIAL TILT 8.356 -7.573 Core Plate Delta-P: psia 19.016 AVG BOT 8ft/12ft 0.9666 1.0433 Coolant Temp: Deg-F 546.9 In Channel Flow: Mlb/hr 89.56 Active Channel Flow: Mlb/hr 86.54 Total Bypass Flow (%): 11.6 (of total core flow) Total Water Rod Flow (%): 3.0 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.383 19 49 40 1.587 0.901 19 49 40 7.36 0.771 48.3 17 47 22 15 8.05 0.828 55.2 17 47 40 15 1.366 19 39 50 1.599 0.894 20 9 40 7.33 0.769 48.4 17 39 14 15 7.97 0.826 55.8 17 39 48 15 1.361 19 13 42 1.599 0.894 19 47 42 7.30 0.768 48.6 16 49 20 15 8.01 0.816 54.4 16 49 42 15 1.359 20 51 40 1.604 0.892 19 39 50 7.37 0.767 47.7 16 49 24 15 7.98 0.814 54.5 16 41 50 15 1.358 19 41 48 1.605 0.891 19 41 48 7.26 0.766 48.9 16 13 44 15 7.96 0.811 54.4 17 9 26 15 1.356 19 17 46 1.605 0.891 19 15 18 7.26 0.766 48.9 16 15 46 15 7.85 0.809 55.3 16 15 16 15 1.355 19 15 44 1.606 0.890 19 17 16 7.26 0.766 49.0 16 17 48 15 7.86 0.809 55.2 16 17 48 15 1.349 19 47 24 1.616 0.885 19 41 18 7.28 0.766 48.7 16 41 12 15 7.87 0.808 55.0 16 47 18 15 1.348 19 37 14 1.618 0.884 19 47 24 7.20 0.765 49.5 16 15 42 15 7.82 0.806 55.4 17 25 10 15 1.347 19 39 16 1.620 0.883 20 39 10 7.23 0.762 48.8 16 19 16 15 7.73 0.806 56.2 16 15 42 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-52 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 32105.4 Exposure: MWd/MTU (GWd) 17001.0 (2352.08 ) Delta E: MWd/MTU, (GWd) 1.0 ( 0.14 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.298 7.307 13 0.380 0.396 5 48 Inlet Subcooling: Btu/lbm -28.77 24 0.782 20.125 14 0.334 0.415 51 52 Flow: Mlb/hr 96.45 ( 94.10 %) 23 0.984 25.738 15 0.448 0.452 3 18 22 1.090 28.705 16 1.015 1.150 49 24 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.159 30.744 17 0.815 1.146 47 22 59 -- -- -- -- -- -- -- 59 20 1.198 32.257 18 0.967 1.126 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.216 33.005 19 1.267 1.374 49 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.244 33.957 20 1.195 1.348 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.253 34.600 21 1.196 1.332 51 38 43 -- -- -- -- -- -- 8 -- 8 -- -- -- -- -- -- 43 16 1.250 35.153 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.255 36.146 35 -- -- -- -- 8 -- -- -- -- -- 8 -- -- -- -- 35 14 1.216 36.054 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.302* 34.241 27 -- -- -- -- 8 -- -- -- -- -- 8 -- -- -- -- 27 12 1.295 35.276 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.267 35.847 19 -- -- -- -- -- -- 8 -- 8 -- -- -- -- -- -- 19 10 1.242 36.614 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.197 37.504 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.117 38.096 7 -- -- -- -- -- -- -- -- -- 7 7 1.012 38.680 3 -- -- -- -- -- -- -- 3 6 0.909 39.725 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.794 39.795* 4 0.689 38.395 Control Rod Density: % 3.60 3 0.611 35.228 2 0.481 26.810 k-effective: 1.00011 Bottom 1 0.139 7.861 Void Fraction: 0.391 Core Delta-P: psia 21.904  % AXIAL TILT 9.248 -7.572 Core Plate Delta-P: psia 17.352 AVG BOT 8ft/12ft 0.9580 1.0433 Coolant Temp: Deg-F 546.7 In Channel Flow: Mlb/hr 85.23 Active Channel Flow: Mlb/hr 82.33 Total Bypass Flow (%): 11.6 (of total core flow) Total Water Rod Flow (%): 3.0 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.374 19 49 40 1.563 0.915 19 49 40 7.28 0.763 48.3 17 47 22 15 7.95 0.818 55.2 17 47 40 15 1.357 19 39 50 1.575 0.908 19 47 42 7.25 0.761 48.4 17 39 14 15 7.87 0.816 55.8 17 39 48 15 1.353 19 13 42 1.577 0.907 20 9 40 7.29 0.759 47.7 16 49 24 15 7.90 0.805 54.4 16 49 42 15 1.350 19 41 48 1.580 0.905 19 39 50 7.21 0.758 48.6 16 49 20 15 7.88 0.803 54.5 16 41 50 15 1.348 19 17 46 1.580 0.905 19 41 48 7.13 0.757 49.5 16 15 42 15 7.86 0.801 54.4 17 9 26 15 1.348 20 51 40 1.581 0.905 19 15 18 7.18 0.757 48.9 16 13 44 15 7.75 0.799 55.3 16 15 16 15 1.348 19 15 44 1.582 0.904 19 17 16 7.17 0.757 48.9 16 15 46 15 7.75 0.798 55.2 16 17 48 15 1.347 19 47 24 1.583 0.903 19 41 18 7.17 0.757 49.0 16 17 48 15 7.76 0.797 55.1 16 47 18 15 1.345 19 39 16 1.583 0.903 19 47 24 7.19 0.756 48.7 16 41 12 15 7.65 0.797 56.2 16 15 42 15 1.345 19 37 14 1.588 0.900 19 17 42 7.16 0.754 48.8 16 19 16 15 7.72 0.796 55.4 17 25 10 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-53 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 32404.6 Exposure: MWd/MTU (GWd) 17300.3 (2393.49 ) Delta E: MWd/MTU, (GWd) 299.3 ( 41.41 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.306 7.403 13 0.379 0.396 5 48 Inlet Subcooling: Btu/lbm -26.97 24 0.807 20.379 14 0.333 0.416 51 52 Flow: Mlb/hr 102.40 ( 99.90 %) 23 1.016 26.058 15 0.447 0.451 3 18 22 1.126 29.063 16 1.013 1.152 49 24 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.195 31.124 17 0.816 1.147 47 22 59 -- -- -- -- -- -- -- 59 20 1.232 32.650 18 0.962 1.125 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.246 33.402 19 1.266 1.380 49 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.269 34.362 20 1.203 1.357 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.273 35.008 21 1.201 1.337 51 38 43 -- -- -- -- -- -- 8 -- 8 -- -- -- -- -- -- 43 16 1.265 35.559 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.266 36.553 35 -- -- -- -- 8 -- -- -- -- -- 8 -- -- -- -- 35 14 1.224 36.448 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.310* 34.607 27 -- -- -- -- 8 -- -- -- -- -- 8 -- -- -- -- 27 12 1.300 35.639 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.267 36.201 19 -- -- -- -- -- -- 8 -- 8 -- -- -- -- -- -- 19 10 1.236 36.961 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.184 37.838 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.094 38.405 7 -- -- -- -- -- -- -- -- -- 7 7 0.979 38.959 3 -- -- -- -- -- -- -- 3 6 0.868 39.974 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.749 40.011* 4 0.643 38.581 Control Rod Density: % 3.60 3 0.568 35.393 2 0.447 26.940 k-effective: 1.00011 Bottom 1 0.129 7.900 Void Fraction: 0.376 Core Delta-P: psia 23.854  % AXIAL TILT 11.653 -7.342 Core Plate Delta-P: psia 19.301 AVG BOT 8ft/12ft 0.9433 1.0421 Coolant Temp: Deg-F 546.8 In Channel Flow: Mlb/hr 90.63 Active Channel Flow: Mlb/hr 87.61 Total Bypass Flow (%): 11.5 (of total core flow) Total Water Rod Flow (%): 3.0 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.380 19 49 40 1.576 0.907 19 49 40 7.35 0.775 48.8 17 47 22 15 8.05 0.834 55.7 17 47 40 15 1.361 19 39 50 1.590 0.900 20 9 40 7.32 0.773 49.1 16 49 20 15 7.96 0.831 56.3 17 39 48 15 1.357 20 51 40 1.593 0.898 19 47 42 7.32 0.772 48.9 17 39 14 15 8.02 0.823 54.9 16 49 42 15 1.356 19 13 42 1.597 0.896 19 39 50 7.37 0.771 48.2 16 49 24 15 7.99 0.819 55.0 16 41 50 15 1.353 19 41 48 1.599 0.894 19 41 48 7.28 0.770 49.1 16 41 50 15 7.97 0.818 54.9 17 9 26 15 1.349 19 17 46 1.600 0.894 19 15 18 7.26 0.770 49.4 16 13 44 15 7.88 0.814 55.5 16 47 18 15 1.349 19 15 44 1.601 0.893 19 17 16 7.25 0.769 49.4 16 15 46 15 7.86 0.814 55.7 16 17 48 15 1.348 19 47 24 1.602 0.893 19 47 24 7.25 0.769 49.4 16 17 48 15 7.85 0.814 55.8 16 15 16 15 1.345 19 37 14 1.608 0.889 19 41 18 7.18 0.767 50.0 16 15 42 15 7.82 0.811 55.8 17 25 10 15 1.345 19 39 16 1.610 0.888 19 15 40 7.21 0.764 49.2 16 41 16 15 7.71 0.809 56.7 16 15 42 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-54 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 32405.3 Exposure: MWd/MTU (GWd) 17301.0 (2393.59 ) Delta E: MWd/MTU, (GWd) 0.7 ( 0.10 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.292 7.403 13 0.377 0.393 5 48 Inlet Subcooling: Btu/lbm -28.93 24 0.766 20.380 14 0.331 0.412 51 52 Flow: Mlb/hr 95.94 ( 93.60 %) 23 0.964 26.059 15 0.444 0.448 3 18 22 1.071 29.064 16 1.026 1.157 49 24 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.141 31.125 17 0.812 1.157 47 22 59 -- -- -- -- -- -- -- 59 20 1.199 32.650 18 0.895 1.129 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.223 33.403 19 1.271 1.379 49 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.254 34.363 20 1.193 1.348 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.265 35.009 21 1.193 1.332 51 38 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.262 35.560 39 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 39 15 1.266 36.554 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.226 36.449 31 -- -- -- -- 10 -- -- -- -- -- 10 -- -- -- -- 31 13 1.312* 34.607 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.302 35.640 23 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 23 11 1.270 36.202 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.240 36.962 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.192 37.838 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.110 38.406 7 -- -- -- -- -- -- -- -- -- 7 7 1.005 38.959 3 -- -- -- -- -- -- -- 3 6 0.905 39.975 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.794 40.012* 4 0.693 38.582 Control Rod Density: % 2.75 3 0.618 35.394 2 0.489 26.940 k-effective: 1.00005 Bottom 1 0.142 7.901 Void Fraction: 0.391 Core Delta-P: psia 21.748  % AXIAL TILT 9.159 -7.341 Core Plate Delta-P: psia 17.199 AVG BOT 8ft/12ft 0.9607 1.0421 Coolant Temp: Deg-F 546.7 In Channel Flow: Mlb/hr 84.76 Active Channel Flow: Mlb/hr 81.87 Total Bypass Flow (%): 11.6 (of total core flow) Total Water Rod Flow (%): 3.0 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.379 19 49 40 1.532 0.933 19 47 24 7.42 0.782 48.8 17 47 22 15 8.06 0.834 55.7 17 47 40 15 1.376 19 37 14 1.539 0.929 19 15 40 7.39 0.780 48.9 17 39 14 15 7.96 0.831 56.3 17 39 48 15 1.375 19 47 24 1.541 0.928 19 49 40 7.39 0.774 48.2 16 49 24 15 7.97 0.817 54.9 16 49 42 15 1.371 19 39 16 1.542 0.927 19 23 48 7.23 0.772 50.0 16 15 42 15 7.78 0.816 56.7 16 15 42 15 1.366 19 15 40 1.546 0.925 19 39 16 7.29 0.769 48.9 16 37 12 15 7.93 0.814 55.0 16 41 50 15 1.362 19 39 50 1.559 0.917 19 47 42 7.27 0.768 49.1 16 49 20 15 7.90 0.812 55.1 16 49 24 15 1.357 19 47 20 1.562 0.916 19 41 18 7.25 0.768 49.2 16 41 16 15 7.91 0.811 54.9 17 9 26 15 1.357 19 33 48 1.563 0.915 19 39 50 7.21 0.765 49.4 16 13 44 15 7.86 0.811 55.4 16 23 50 15 1.356 19 35 16 1.564 0.914 20 9 40 7.23 0.765 49.2 16 41 12 15 7.81 0.809 55.7 16 41 16 15 1.353 19 41 14 1.564 0.914 19 17 42 7.20 0.764 49.4 16 17 48 15 7.79 0.808 55.9 17 25 10 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-55 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 32694.2 Exposure: MWd/MTU (GWd) 17590.0 (2433.57 ) Delta E: MWd/MTU, (GWd) 289.0 ( 39.98 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.300 7.494 13 0.376 0.393 5 48 Inlet Subcooling: Btu/lbm -26.94 24 0.791 20.621 14 0.330 0.412 51 52 Flow: Mlb/hr 102.50 (100.00 %) 23 0.996 26.362 15 0.444 0.448 3 18 22 1.107 29.404 16 1.025 1.159 49 24 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.177 31.486 17 0.814 1.158 47 22 59 -- -- -- -- -- -- -- 59 20 1.232 33.030 18 0.891 1.127 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.252 33.789 19 1.270 1.385 49 40 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.279 34.758 20 1.200 1.357 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.284 35.406 21 1.199 1.337 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.275 35.955 39 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 39 15 1.276 36.950 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.233 36.832 31 -- -- -- -- 10 -- -- -- -- -- 10 -- -- -- -- 31 13 1.319* 34.963 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.306 35.992 23 -- -- -- -- -- -- 6 -- 6 -- -- -- -- -- -- 23 11 1.270 36.545 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.235 37.297 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.179 38.159 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.088 38.703 7 -- -- -- -- -- -- -- -- -- 7 7 0.974 39.227 3 -- -- -- -- -- -- -- 3 6 0.866 40.214 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.750 40.220* 4 0.648 38.763 Control Rod Density: % 2.75 3 0.575 35.555 2 0.456 27.068 k-effective: 1.00004 Bottom 1 0.132 7.940 Void Fraction: 0.376 Core Delta-P: psia 23.906  % AXIAL TILT 11.502 -7.124 Core Plate Delta-P: psia 19.357 AVG BOT 8ft/12ft 0.9462 1.0410 Coolant Temp: Deg-F 546.8 In Channel Flow: Mlb/hr 90.72 Active Channel Flow: Mlb/hr 87.69 Total Bypass Flow (%): 11.5 (of total core flow) Total Water Rod Flow (%): 3.0 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.385 19 49 40 1.558 0.918 19 47 24 7.48 0.792 49.2 17 47 22 15 8.14 0.848 56.2 17 47 40 15 1.375 19 47 24 1.560 0.917 19 49 40 7.44 0.789 49.4 17 39 14 15 8.04 0.844 56.8 17 39 48 15 1.375 19 37 14 1.567 0.912 19 15 40 7.46 0.785 48.6 16 49 24 15 8.08 0.833 55.4 16 49 42 15 1.370 19 39 16 1.573 0.909 19 23 48 7.36 0.781 49.5 16 49 20 15 8.03 0.829 55.5 16 41 50 15 1.365 19 39 50 1.577 0.907 19 39 16 7.27 0.780 50.4 16 15 42 15 8.45 0.828 51.2 16 17 32 20 1.365 19 15 40 1.581 0.904 20 9 40 7.35 0.779 49.3 16 37 12 15 8.01 0.827 55.4 17 9 26 15 1.360 19 47 20 1.582 0.904 19 47 42 7.31 0.777 49.6 16 41 50 15 7.84 0.827 57.2 16 15 42 15 1.357 20 51 40 1.586 0.901 19 39 50 7.29 0.776 49.8 16 13 44 15 7.99 0.826 55.6 16 49 24 15 1.355 19 41 14 1.592 0.898 19 41 48 7.30 0.776 49.7 16 41 16 15 7.93 0.823 55.9 16 23 50 15 1.354 19 33 48 1.593 0.898 19 41 18 7.27 0.775 49.9 16 17 48 15 7.88 0.822 56.3 17 25 10 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-56 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 32695.2 Exposure: MWd/MTU (GWd) 17591.0 (2433.71 ) Delta E: MWd/MTU, (GWd) 1.0 ( 0.14 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.290 7.494 13 0.370 0.386 5 48 Inlet Subcooling: Btu/lbm -29.41 24 0.761 20.622 14 0.325 0.404 7 50 Flow: Mlb/hr 94.50 ( 92.20 %) 23 0.960 26.363 15 0.436 0.441 3 18 22 1.074 29.405 16 1.032 1.154 45 28 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.152 31.488 17 0.799 1.148 47 22 59 -- -- -- -- -- -- -- 59 20 1.207 33.031 18 0.906 1.110 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.234 33.791 19 1.284 1.378 43 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.266 34.760 20 1.175 1.334 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.275 35.407 21 1.175 1.320 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.270 35.957 39 -- -- -- -- -- -- 8 -- 8 -- -- -- -- -- -- 39 15 1.271 36.952 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.230 36.834 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.315* 34.964 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.302 35.993 23 -- -- -- -- -- -- 8 -- 8 -- -- -- -- -- -- 23 11 1.266 36.546 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.233 37.298 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.181 38.160 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.097 38.704 7 -- -- -- -- -- -- -- -- -- 7 7 0.993 39.228 3 -- -- -- -- -- -- -- 3 6 0.895 40.215 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.787 40.221* 4 0.689 38.764 Control Rod Density: % 1.80 3 0.617 35.555 2 0.492 27.069 k-effective: 1.00008 Bottom 1 0.143 7.940 Void Fraction: 0.392 Core Delta-P: psia 21.271  % AXIAL TILT 9.690 -7.123 Core Plate Delta-P: psia 16.725 AVG BOT 8ft/12ft 0.9580 1.0410 Coolant Temp: Deg-F 546.6 In Channel Flow: Mlb/hr 83.47 Active Channel Flow: Mlb/hr 80.61 Total Bypass Flow (%): 11.7 (of total core flow) Total Water Rod Flow (%): 3.0 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.378 19 43 28 1.515 0.944 19 15 26 7.40 0.784 49.2 17 47 22 15 8.02 0.835 56.2 17 47 40 15 1.376 19 47 24 1.516 0.943 19 47 24 7.33 0.778 49.4 16 47 26 15 7.82 0.821 56.8 17 39 48 15 1.375 19 45 26 1.523 0.939 19 17 34 7.39 0.778 48.6 16 49 24 15 7.92 0.819 55.6 16 49 24 15 1.366 19 49 40 1.531 0.934 19 47 34 7.23 0.773 50.2 16 15 38 15 7.86 0.818 56.2 16 47 26 15 1.362 19 47 28 1.531 0.934 19 15 40 7.47 0.770 46.8 16 45 28 17 7.91 0.816 55.4 16 49 42 15 1.360 19 15 40 1.533 0.933 19 15 32 7.17 0.769 50.4 16 15 42 15 7.73 0.816 57.2 16 15 42 15 1.355 19 45 30 1.539 0.929 19 49 40 7.25 0.769 49.4 17 39 14 15 7.75 0.814 56.9 16 15 24 15 1.352 19 41 26 1.540 0.929 19 41 32 7.23 0.768 49.6 16 47 30 17 7.89 0.814 55.4 17 9 26 15 1.350 19 41 32 1.550 0.923 19 41 26 7.20 0.765 49.5 16 49 20 15 7.71 0.812 57.1 16 49 34 15 1.348 19 37 14 1.555 0.920 19 17 24 7.13 0.764 50.4 16 49 34 15 7.68 0.812 57.3 16 47 32 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-57 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 32979.4 Exposure: MWd/MTU (GWd) 17875.0 (2473.00 ) Delta E: MWd/MTU, (GWd) 284.0 ( 39.29 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.300 7.583 13 0.369 0.386 5 48 Inlet Subcooling: Btu/lbm -27.11 24 0.789 20.857 14 0.324 0.404 53 50 Flow: Mlb/hr 101.89 ( 99.40 %) 23 0.995 26.660 15 0.435 0.441 3 18 22 1.112 29.740 16 1.030 1.151 49 24 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.189 31.846 17 0.801 1.149 47 22 59 -- -- -- -- -- -- -- 59 20 1.242 33.406 18 0.902 1.109 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.264 34.173 19 1.282 1.376 47 24 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.290 35.151 20 1.183 1.342 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.294 35.801 21 1.181 1.324 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.282 36.348 39 -- -- -- -- -- -- 8 -- 8 -- -- -- -- -- -- 39 15 1.280 37.343 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.236 37.212 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.321* 35.314 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.306 36.339 23 -- -- -- -- -- -- 8 -- 8 -- -- -- -- -- -- 23 11 1.266 36.882 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.228 37.625 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.169 38.472 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.076 38.993 7 -- -- -- -- -- -- -- -- -- 7 7 0.962 39.488 3 -- -- -- -- -- -- -- 3 6 0.855 40.447* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.741 40.424 4 0.642 38.941 Control Rod Density: % 1.80 3 0.572 35.713 2 0.456 27.195 k-effective: 1.00011 Bottom 1 0.133 7.978 Void Fraction: 0.375 Core Delta-P: psia 23.686  % AXIAL TILT 12.105 -6.909 Core Plate Delta-P: psia 19.139 AVG BOT 8ft/12ft 0.9429 1.0400 Coolant Temp: Deg-F 546.8 In Channel Flow: Mlb/hr 90.17 Active Channel Flow: Mlb/hr 87.15 Total Bypass Flow (%): 11.5 (of total core flow) Total Water Rod Flow (%): 3.0 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.376 19 47 24 1.551 0.922 19 47 24 7.45 0.793 49.7 17 47 22 15 8.10 0.849 56.6 17 47 40 15 1.371 19 49 40 1.554 0.920 19 15 26 7.46 0.788 49.1 16 49 24 15 7.91 0.835 57.3 17 39 48 15 1.371 19 43 28 1.564 0.914 19 17 34 7.54 0.787 47.9 16 47 26 17 8.02 0.832 55.9 16 49 42 15 1.371 19 45 26 1.564 0.914 19 49 40 7.45 0.783 48.6 16 45 24 17 8.00 0.832 56.0 16 49 24 15 1.359 19 47 28 1.567 0.913 19 47 34 7.53 0.780 47.2 16 45 28 17 7.99 0.829 55.9 17 9 26 15 1.358 19 45 22 1.567 0.913 19 15 40 7.30 0.779 50.0 16 47 30 17 7.91 0.829 56.6 16 47 26 15 1.350 19 45 30 1.573 0.909 19 15 32 7.31 0.779 49.8 17 39 14 15 7.79 0.827 57.6 16 15 42 15 1.348 19 37 14 1.583 0.903 19 41 32 7.37 0.779 49.0 16 15 42 17 8.02 0.825 55.2 16 15 24 17 1.348 19 49 26 1.589 0.900 19 49 26 7.29 0.778 50.0 16 49 20 15 7.96 0.823 55.6 16 49 34 17 1.345 19 39 16 1.589 0.900 20 9 40 7.32 0.774 49.1 16 49 34 17 7.94 0.823 55.8 16 47 32 17

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-58 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 32980.3 Exposure: MWd/MTU (GWd) 17876.0 (2473.14 ) Delta E: MWd/MTU, (GWd) 1.0 ( 0.14 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.300 7.583 13 0.365 0.381 5 48 Inlet Subcooling: Btu/lbm -29.17 24 0.786 20.858 14 0.320 0.399 53 50 Flow: Mlb/hr 95.22 ( 92.90 %) 23 0.993 26.661 15 0.431 0.435 3 18 22 1.116 29.741 16 1.033 1.141 45 28 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.205 31.848 17 0.791 1.135 47 22 59 -- -- -- -- -- -- -- 59 20 1.273 33.408 18 0.929 1.102 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.302 34.175 19 1.291 1.368 43 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.324* 35.153 20 1.165 1.320 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.312 35.802 21 1.164 1.303 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.288 36.349 39 -- -- -- -- -- -- 14 -- 14 -- -- -- -- -- -- 39 15 1.277 37.344 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.227 37.213 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.306 35.315 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.287 36.340 23 -- -- -- -- -- -- 14 -- 14 -- -- -- -- -- -- 23 11 1.246 36.883 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.207 37.626 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.151 38.473 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.061 38.994 7 -- -- -- -- -- -- -- -- -- 7 7 0.951 39.488 3 -- -- -- -- -- -- -- 3 6 0.848 40.448* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.737 40.425 4 0.641 38.941 Control Rod Density: % 1.53 3 0.572 35.714 2 0.456 27.195 k-effective: 1.00009 Bottom 1 0.133 7.978 Void Fraction: 0.382 Core Delta-P: psia 21.414  % AXIAL TILT 13.142 -6.908 Core Plate Delta-P: psia 16.868 AVG BOT 8ft/12ft 0.9345 1.0400 Coolant Temp: Deg-F 546.4 In Channel Flow: Mlb/hr 84.17 Active Channel Flow: Mlb/hr 81.31 Total Bypass Flow (%): 11.6 (of total core flow) Total Water Rod Flow (%): 3.0 (of total core flow) Source Convergence 0.00007

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.368 19 43 28 1.520 0.941 19 17 34 8.10 0.832 46.4 18 33 38 19 8.34 0.837 53.2 16 25 26 19 1.360 19 45 26 1.525 0.937 19 15 26 8.08 0.819 45.1 18 25 22 19 7.97 0.835 56.6 17 47 40 15 1.359 19 47 24 1.528 0.936 19 47 24 9.74 0.806 23.9 19 25 24 19 8.26 0.833 53.5 16 31 40 20 1.351 19 41 26 1.532 0.933 19 41 32 9.74 0.802 23.3 19 33 40 19 8.19 0.831 54.0 16 23 24 19 1.351 19 49 40 1.535 0.931 19 41 26 7.60 0.788 47.4 16 31 40 20 7.84 0.828 57.3 17 39 48 15 1.347 19 41 32 1.541 0.928 19 15 40 7.61 0.787 47.0 16 25 26 19 10.77 0.827 27.8 19 25 24 19 1.346 19 45 22 1.544 0.926 19 47 34 7.53 0.782 47.6 16 23 24 19 8.25 0.825 52.9 16 31 26 20 1.344 19 47 28 1.545 0.925 19 15 32 7.35 0.782 49.7 17 47 22 15 8.00 0.823 55.2 16 15 24 17 1.343 19 37 14 1.550 0.923 19 49 40 7.58 0.782 46.8 16 31 36 20 7.95 0.822 55.6 16 15 42 17 1.343 19 39 16 1.551 0.922 19 39 16 7.44 0.782 48.6 16 45 24 17 10.76 0.820 27.1 19 33 40 19

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-59 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 33204.3 Exposure: MWd/MTU (GWd) 18100.0 (2504.13 ) Delta E: MWd/MTU, (GWd) 224.0 ( 30.99 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.307 7.655 13 0.364 0.381 5 48 Inlet Subcooling: Btu/lbm -27.26 24 0.807 21.049 14 0.320 0.400 53 50 Flow: Mlb/hr 101.37 ( 98.90 %) 23 1.020 26.903 15 0.431 0.435 3 18 22 1.146 30.015 16 1.032 1.137 45 28 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.233 32.142 17 0.793 1.136 47 22 59 -- -- -- -- -- -- -- 59 20 1.299 33.719 18 0.925 1.101 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.324 34.492 19 1.289 1.363 43 28 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.342* 35.475 20 1.172 1.327 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.325 36.121 21 1.169 1.307 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.297 36.662 39 -- -- -- -- -- -- 14 -- 14 -- -- -- -- -- -- 39 15 1.283 37.653 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.231 37.510 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.310 35.589 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.290 36.610 23 -- -- -- -- -- -- 14 -- 14 -- -- -- -- -- -- 23 11 1.246 37.144 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.203 37.878 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.141 38.713 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.044 39.214 7 -- -- -- -- -- -- -- -- -- 7 7 0.927 39.685 3 -- -- -- -- -- -- -- 3 6 0.818 40.623* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.704 40.576 4 0.607 39.072 Control Rod Density: % 1.53 3 0.540 35.830 2 0.431 27.288 k-effective: 1.00005 Bottom 1 0.125 8.007 Void Fraction: 0.369 Core Delta-P: psia 23.428  % AXIAL TILT 14.931 -6.720 Core Plate Delta-P: psia 18.882 AVG BOT 8ft/12ft 0.9232 1.0390 Coolant Temp: Deg-F 546.6 In Channel Flow: Mlb/hr 89.75 Active Channel Flow: Mlb/hr 86.76 Total Bypass Flow (%): 11.5 (of total core flow) Total Water Rod Flow (%): 2.9 (of total core flow) Source Convergence 0.00006

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.363 19 43 28 1.553 0.921 19 17 34 8.23 0.840 45.8 18 33 38 20 8.51 0.849 52.7 16 25 26 20 1.359 19 47 24 1.557 0.919 19 15 26 8.13 0.827 45.5 18 25 22 19 8.03 0.846 57.0 17 47 40 15 1.357 19 45 26 1.557 0.919 19 47 24 9.80 0.815 24.3 19 25 24 19 8.35 0.845 53.9 16 31 40 20 1.355 19 49 40 1.568 0.912 19 41 32 9.81 0.811 23.8 19 33 40 19 8.32 0.842 53.9 16 23 24 20 1.345 19 41 26 1.569 0.911 19 41 26 7.66 0.798 47.7 16 31 40 20 7.90 0.838 57.7 17 39 48 15 1.345 19 45 22 1.570 0.911 19 15 40 7.74 0.796 46.5 16 25 26 20 10.86 0.837 28.3 19 25 24 19 1.343 19 37 14 1.570 0.911 19 49 40 7.61 0.794 47.8 17 47 22 17 8.32 0.836 53.2 16 31 26 20 1.342 19 47 28 1.573 0.909 19 47 34 7.57 0.791 47.9 16 23 24 19 8.07 0.834 55.6 16 15 24 17 1.342 19 39 16 1.578 0.906 19 15 32 7.49 0.790 48.9 16 45 24 17 8.03 0.834 56.0 16 15 42 17 1.340 19 41 32 1.581 0.905 19 39 16 7.63 0.790 47.1 16 31 36 20 8.13 0.832 54.8 16 47 26 17

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-60 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 33205.2 Exposure: MWd/MTU (GWd) 18101.0 (2504.26 ) Delta E: MWd/MTU, (GWd) 1.0 ( 0.14 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.284 7.655 13 0.356 0.372 5 48 Inlet Subcooling: Btu/lbm -29.54 24 0.743 21.050 14 0.313 0.390 53 50 Flow: Mlb/hr 94.10 ( 91.80 %) 23 0.939 26.904 15 0.422 0.426 3 18 22 1.058 30.016 16 1.034 1.137 35 18 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.143 32.144 17 0.775 1.110 47 22 59 -- -- -- -- -- -- -- 59 20 1.212 33.720 18 0.984 1.100 21 36 55 -- -- -- -- -- -- -- -- -- 55 19 1.247 34.494 19 1.306 1.365 33 18 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.281 35.477 20 1.144 1.293 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.289 36.123 21 1.141 1.273 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.282 36.663 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.281 37.655 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.239 37.511 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.325* 35.590 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.308 36.611 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.267 37.146 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.227 37.880 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.171 38.714 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.085 39.215 7 -- -- -- -- -- -- -- -- -- 7 7 0.980 39.686 3 -- -- -- -- -- -- -- 3 6 0.885 40.623* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.783 40.577 4 0.693 39.073 Control Rod Density: % 0.00 3 0.628 35.831 2 0.504 27.288 k-effective: 1.00004 Bottom 1 0.148 8.007 Void Fraction: 0.392 Core Delta-P: psia 21.159  % AXIAL TILT 9.799 -6.719 Core Plate Delta-P: psia 16.617 AVG BOT 8ft/12ft 0.9592 1.0390 Coolant Temp: Deg-F 546.5 In Channel Flow: Mlb/hr 83.09 Active Channel Flow: Mlb/hr 80.24 Total Bypass Flow (%): 11.7 (of total core flow) Total Water Rod Flow (%): 3.0 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.365 19 33 18 1.538 0.930 19 33 18 7.18 0.767 50.1 17 47 22 15 7.78 0.819 57.0 17 47 40 15 1.362 19 25 42 1.539 0.929 19 25 42 7.15 0.764 50.2 17 39 14 15 7.69 0.816 57.7 17 39 48 15 1.356 19 37 18 1.544 0.926 19 23 18 7.21 0.761 48.9 16 45 24 17 7.54 0.804 58.0 16 15 42 15 1.352 19 31 42 1.549 0.923 19 17 34 7.26 0.760 48.2 16 47 26 17 7.56 0.803 57.7 16 15 24 15 1.347 19 43 28 1.549 0.923 19 41 26 7.15 0.759 49.4 16 49 24 15 7.69 0.802 56.3 16 49 42 15 1.346 19 41 26 1.551 0.922 19 31 42 7.11 0.758 49.9 16 21 18 15 7.68 0.802 56.4 16 49 24 15 1.346 19 35 16 1.557 0.918 19 39 34 7.08 0.758 50.3 16 37 16 15 7.61 0.801 57.0 16 47 26 15 1.339 19 39 34 1.557 0.918 19 41 32 7.23 0.758 48.3 16 43 26 17 7.60 0.800 57.0 16 41 16 15 1.339 19 21 42 1.558 0.918 19 23 32 7.28 0.757 47.6 16 45 28 17 7.56 0.799 57.4 16 17 40 15 1.337 19 33 40 1.558 0.918 19 25 16 7.13 0.756 49.4 16 15 42 17 7.59 0.799 57.0 16 23 16 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-61 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 33504.2 Exposure: MWd/MTU (GWd) 18400.0 (2545.63 ) Delta E: MWd/MTU, (GWd) 299.0 ( 41.37 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.294 7.747 13 0.355 0.372 5 48 Inlet Subcooling: Btu/lbm -26.65 24 0.775 21.293 14 0.312 0.391 53 50 Flow: Mlb/hr 103.53 (101.00 %) 23 0.978 27.210 15 0.421 0.426 3 18 22 1.100 30.364 16 1.031 1.130 35 18 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.184 32.519 17 0.778 1.113 47 22 59 -- -- -- -- -- -- -- 59 20 1.249 34.117 18 0.977 1.091 21 36 55 -- -- -- -- -- -- -- -- -- 55 19 1.277 34.901 19 1.303 1.357 33 18 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.305 35.894 20 1.153 1.304 51 40 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.308 36.542 21 1.149 1.282 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.294 37.079 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.290 38.070 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.244 37.912 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.331* 35.961 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.312 36.977 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.267 37.500 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.222 38.222 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.159 39.040 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.063 39.516 7 -- -- -- -- -- -- -- -- -- 7 7 0.948 39.956 3 -- -- -- -- -- -- -- 3 6 0.843 40.865* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.735 40.789 4 0.643 39.259 Control Rod Density: % 0.00 3 0.579 36.000 2 0.465 27.424 k-effective: 1.00005 Bottom 1 0.137 8.049 Void Fraction: 0.372 Core Delta-P: psia 24.275  % AXIAL TILT 12.368 -6.499 Core Plate Delta-P: psia 19.734 AVG BOT 8ft/12ft 0.9428 1.0379 Coolant Temp: Deg-F 546.8 In Channel Flow: Mlb/hr 91.64 Active Channel Flow: Mlb/hr 88.58 Total Bypass Flow (%): 11.5 (of total core flow) Total Water Rod Flow (%): 2.9 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.357 19 33 18 1.598 0.895 19 33 18 7.45 0.780 48.3 17 47 22 17 7.88 0.835 57.5 17 47 40 15 1.353 19 25 42 1.600 0.894 19 17 34 7.20 0.774 50.6 17 39 14 15 7.78 0.831 58.1 17 39 48 15 1.351 19 37 18 1.600 0.894 19 25 42 7.35 0.774 48.7 16 47 26 17 7.82 0.820 56.7 16 49 42 15 1.342 19 35 16 1.600 0.894 19 23 18 7.30 0.774 49.4 16 45 24 17 7.84 0.820 56.5 16 15 42 17 1.341 19 43 28 1.603 0.892 19 41 26 7.41 0.773 47.9 16 49 24 17 8.01 0.818 54.7 16 49 24 17 1.340 19 31 42 1.606 0.890 19 15 26 7.23 0.770 49.8 16 15 42 17 7.94 0.818 55.2 16 47 26 17 1.339 19 41 26 1.606 0.890 19 47 24 7.35 0.768 48.1 16 45 28 17 7.87 0.818 56.0 16 15 24 17 1.332 19 21 42 1.612 0.887 19 49 40 7.29 0.767 48.8 16 43 26 17 7.76 0.815 56.8 17 9 26 15 1.332 19 47 24 1.612 0.887 19 25 16 7.22 0.766 49.4 16 17 40 17 7.76 0.814 56.8 16 41 50 15 1.331 19 45 26 1.613 0.887 19 41 32 7.10 0.765 50.7 16 37 16 15 7.86 0.813 55.7 16 17 40 17

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-62 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 33629.3 Exposure: MWd/MTU (GWd) 18525.0 (2562.92 ) Delta E: MWd/MTU, (GWd) 125.0 ( 17.29 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1050.0 Top 25 0.299 7.786 13 0.355 0.372 5 48 Inlet Subcooling: Btu/lbm -25.55 24 0.789 21.397 14 0.311 0.391 53 50 Flow: Mlb/hr 107.62 (105.00 %) 23 0.996 27.343 15 0.421 0.425 3 18 22 1.120 30.514 16 1.031 1.128 35 18 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.203 32.680 17 0.778 1.113 47 22 59 -- -- -- -- -- -- -- 59 20 1.267 34.287 18 0.975 1.088 19 42 55 -- -- -- -- -- -- -- -- -- 55 19 1.292 35.074 19 1.302 1.354 33 18 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.317 36.071 20 1.157 1.308 51 22 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.316 36.719 21 1.152 1.284 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.300 37.253 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.293 38.244 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.246 38.080 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.333* 36.116 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.313 37.130 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.266 37.647 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.220 38.365 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.152 39.175 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.052 39.639 7 -- -- -- -- -- -- -- -- -- 7 7 0.932 40.066 3 -- -- -- -- -- -- -- 3 6 0.824 40.962* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.713 40.873 4 0.621 39.333 Control Rod Density: % 0.00 3 0.557 36.066 2 0.448 27.477 k-effective: 1.00018 Bottom 1 0.131 8.065 Void Fraction: 0.364 Core Delta-P: psia 25.698  % AXIAL TILT 13.552 -6.401 Core Plate Delta-P: psia 21.157 AVG BOT 8ft/12ft 0.9353 1.0374 Coolant Temp: Deg-F 546.9 In Channel Flow: Mlb/hr 95.36 Active Channel Flow: Mlb/hr 92.22 Total Bypass Flow (%): 11.4 (of total core flow) Total Water Rod Flow (%): 2.9 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.354 19 33 18 1.615 0.885 19 17 34 7.50 0.787 48.5 17 47 22 17 8.18 0.842 55.2 17 47 40 17 1.350 19 25 42 1.616 0.885 19 33 18 7.47 0.781 48.1 16 49 24 17 7.81 0.837 58.4 17 39 48 15 1.349 19 37 18 1.618 0.884 19 23 18 7.39 0.780 48.9 16 47 26 17 7.89 0.827 56.7 16 15 42 17 1.340 19 35 16 1.618 0.884 19 25 42 7.33 0.779 49.6 16 45 24 17 7.87 0.827 56.9 16 49 42 15 1.339 19 43 28 1.619 0.883 19 47 24 7.22 0.778 50.8 17 39 14 15 8.07 0.827 54.9 16 49 24 17 1.337 19 31 42 1.619 0.883 19 41 26 7.27 0.776 50.0 16 15 42 17 8.00 0.826 55.5 16 47 26 17 1.336 19 41 26 1.619 0.883 19 49 40 7.44 0.774 47.6 16 45 28 18 7.91 0.825 56.3 16 15 24 17 1.332 19 49 22 1.620 0.883 19 15 26 7.31 0.772 49.0 16 43 26 17 7.81 0.821 57.0 17 9 26 15 1.332 19 47 24 1.629 0.878 19 25 16 7.20 0.771 50.2 16 47 30 18 7.80 0.821 57.0 16 41 50 15 1.330 19 39 16 1.630 0.877 19 41 32 7.25 0.771 49.6 16 17 40 17 7.90 0.820 55.9 16 17 40 17

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-63 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 33644.2 Exposure: MWd/MTU (GWd) 18540.0 (2565.00 ) Delta E: MWd/MTU, (GWd) 15.0 ( 2.08 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1044.7 Top 25 0.285 7.791 13 0.359 0.376 5 48 Inlet Subcooling: Btu/lbm -36.39 24 0.752 21.410 14 0.314 0.395 53 50 Flow: Mlb/hr 94.92 ( 92.60 %) 23 0.949 27.358 15 0.430 0.435 3 18 22 1.069 30.532 16 1.030 1.123 35 18 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.150 32.699 17 0.787 1.109 47 22 59 -- -- -- -- -- -- -- 59 20 1.213 34.307 18 0.975 1.083 19 42 55 -- -- -- -- -- -- -- -- -- 55 19 1.241 35.095 19 1.295 1.345 33 18 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.269 36.092 20 1.160 1.304 51 22 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.273 36.739 21 1.155 1.280 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.263 37.274 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.264 38.264 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.227 38.100 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.325* 36.135 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.319 37.149 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.285 37.665 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.254 38.382 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.201 39.192 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.110 39.655 7 -- -- -- -- -- -- -- -- -- 7 7 0.994 40.079 3 -- -- -- -- -- -- -- 3 6 0.884 40.974* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.767 40.884 4 0.670 39.342 Control Rod Density: % 0.00 3 0.604 36.074 2 0.486 27.484 k-effective: 1.00004 Bottom 1 0.142 8.067 Void Fraction: 0.363 Core Delta-P: psia 20.965  % AXIAL TILT 9.461 -6.390 Core Plate Delta-P: psia 16.391 AVG BOT 8ft/12ft 0.9588 1.0374 Coolant Temp: Deg-F 544.1 In Channel Flow: Mlb/hr 84.04 Active Channel Flow: Mlb/hr 81.24 Total Bypass Flow (%): 11.5 (of total core flow) Total Water Rod Flow (%): 2.9 (of total core flow) Source Convergence 0.00006

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.345 19 33 18 1.598 0.895 19 33 18 7.07 0.761 50.7 17 47 22 15 7.70 0.819 57.8 17 47 40 15 1.341 19 25 42 1.601 0.893 19 23 18 7.02 0.757 50.8 17 39 14 15 7.60 0.814 58.4 17 39 48 15 1.340 19 37 18 1.602 0.893 19 25 42 7.06 0.754 50.1 16 49 24 15 7.66 0.806 57.0 16 49 42 15 1.332 19 35 16 1.603 0.892 19 17 34 7.11 0.751 48.9 16 47 26 17 7.62 0.802 57.0 17 9 26 15 1.331 19 43 28 1.604 0.891 19 49 40 7.05 0.750 49.6 16 45 24 17 7.44 0.801 58.7 16 15 42 15 1.328 19 31 42 1.606 0.891 19 47 24 6.95 0.750 51.0 16 49 20 15 7.59 0.800 57.1 16 49 24 15 1.327 19 41 26 1.606 0.890 19 41 26 6.86 0.748 51.9 16 15 42 15 7.60 0.800 57.0 16 41 50 15 1.326 19 49 22 1.608 0.889 19 15 26 6.91 0.746 51.0 16 37 16 15 7.50 0.797 57.7 16 47 26 15 1.325 19 47 24 1.612 0.887 19 25 16 6.92 0.745 50.8 16 37 12 15 7.42 0.796 58.4 16 15 24 15 1.323 19 39 16 1.617 0.884 19 39 16 6.89 0.745 51.1 16 41 16 15 7.49 0.795 57.7 16 41 16 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-64 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 34044.2 Exposure: MWd/MTU (GWd) 18940.0 (2620.34 ) Delta E: MWd/MTU, (GWd) 400.0 ( 55.34 ) Axial Profile Edit Radial Power Power: MWt 3952.0 (100.00 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1044.7 Top 25 0.301 7.915 13 0.357 0.376 5 48 Inlet Subcooling: Btu/lbm -31.83 24 0.797 21.741 14 0.313 0.395 53 50 Flow: Mlb/hr 107.62 (105.00 %) 23 1.007 27.777 15 0.428 0.433 3 18 22 1.131 31.006 16 1.027 1.117 35 18 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.210 33.209 17 0.789 1.111 47 22 59 -- -- -- -- -- -- -- 59 20 1.268 34.842 18 0.968 1.079 19 42 55 -- -- -- -- -- -- -- -- -- 55 19 1.286 35.640 19 1.293 1.338 33 18 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.306 36.647 20 1.171 1.317 51 22 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.301 37.295 21 1.164 1.289 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.283 37.824 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.279 38.813 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.238 38.632 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.337* 36.632 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.327 37.643 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.288 38.146 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.248 38.850 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.180 39.637 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.072 40.063 7 -- -- -- -- -- -- -- -- -- 7 7 0.938 40.441 3 -- -- -- -- -- -- -- 3 6 0.815 41.292* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.694 41.157 4 0.599 39.580 Control Rod Density: % 0.00 3 0.537 36.288 2 0.432 27.656 k-effective: 0.99989 Bottom 1 0.127 8.120 Void Fraction: 0.338 Core Delta-P: psia 25.165  % AXIAL TILT 13.305 -6.101 Core Plate Delta-P: psia 20.593 AVG BOT 8ft/12ft 0.9349 1.0359 Coolant Temp: Deg-F 544.5 In Channel Flow: Mlb/hr 95.58 Active Channel Flow: Mlb/hr 92.52 Total Bypass Flow (%): 11.2 (of total core flow) Total Water Rod Flow (%): 2.8 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.338 19 33 18 1.639 0.873 19 49 40 7.38 0.780 49.1 17 47 22 17 7.85 0.842 58.4 17 47 40 15 1.335 19 37 18 1.654 0.864 19 47 24 7.37 0.776 48.7 16 49 24 17 7.74 0.836 59.0 17 39 48 15 1.334 19 49 22 1.658 0.862 20 9 40 7.12 0.773 51.5 17 39 14 15 7.83 0.831 57.6 16 49 42 15 1.332 19 25 42 1.662 0.861 19 17 34 7.33 0.771 48.7 16 47 26 18 7.78 0.826 57.6 17 9 26 15 1.328 19 35 16 1.662 0.860 19 15 26 7.08 0.770 51.6 16 49 20 15 7.98 0.824 55.5 16 49 24 17 1.325 19 47 24 1.664 0.859 19 33 18 7.20 0.769 50.0 16 15 42 18 7.77 0.824 57.6 16 41 50 15 1.324 19 43 28 1.665 0.859 19 23 18 7.18 0.769 50.2 16 45 24 17 7.79 0.823 57.3 16 15 42 17 1.322 19 37 14 1.668 0.858 19 41 26 7.02 0.764 51.6 16 41 50 15 7.88 0.821 56.1 16 47 26 17 1.321 19 39 16 1.669 0.857 19 25 42 7.08 0.763 50.9 16 47 30 18 7.65 0.818 58.2 16 47 18 15 1.319 19 41 26 1.669 0.857 19 15 40 7.29 0.763 48.2 16 45 28 18 7.77 0.817 56.9 16 15 24 17

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-65 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 34074.3 Exposure: MWd/MTU (GWd) 18970.0 (2624.49 ) Delta E: MWd/MTU, (GWd) 30.0 ( 4.15 ) Axial Profile Edit Radial Power Power: MWt 3926.7 ( 99.36 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1044.2 Top 25 0.301 7.924 13 0.357 0.376 5 48 Inlet Subcooling: Btu/lbm -31.60 24 0.800 21.767 14 0.313 0.395 53 50 Flow: Mlb/hr 107.62 (105.00 %) 23 1.011 27.809 15 0.428 0.433 3 18 22 1.135 31.043 16 1.027 1.117 35 18 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.214 33.248 17 0.789 1.111 47 22 59 -- -- -- -- -- -- -- 59 20 1.271 34.883 18 0.968 1.078 19 42 55 -- -- -- -- -- -- -- -- -- 55 19 1.289 35.682 19 1.292 1.338 33 18 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.308 36.690 20 1.172 1.318 51 22 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.303 37.337 21 1.164 1.289 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.284 37.865 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.280 38.855 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.238 38.672 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.337* 36.670 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.327 37.680 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.288 38.183 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.247 38.885 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.179 39.671 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.069 40.093 7 -- -- -- -- -- -- -- -- -- 7 7 0.935 40.467 3 -- -- -- -- -- -- -- 3 6 0.811 41.315* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.690 41.177 4 0.595 39.597 Control Rod Density: % 0.00 3 0.533 36.303 2 0.429 27.668 k-effective: 1.00000 Bottom 1 0.126 8.124 Void Fraction: 0.336 Core Delta-P: psia 25.131  % AXIAL TILT 13.538 -6.077 Core Plate Delta-P: psia 20.560 AVG BOT 8ft/12ft 0.9333 1.0358 Coolant Temp: Deg-F 544.5 In Channel Flow: Mlb/hr 95.60 Active Channel Flow: Mlb/hr 92.55 Total Bypass Flow (%): 11.2 (of total core flow) Total Water Rod Flow (%): 2.8 (of total core flow) Source Convergence 0.00009

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.338 19 33 18 1.647 0.868 19 49 40 7.34 0.777 49.2 17 47 22 17 7.80 0.837 58.5 17 47 40 15 1.335 19 49 22 1.663 0.860 19 47 24 7.33 0.772 48.7 16 49 24 17 7.69 0.832 59.1 17 39 48 15 1.334 19 37 18 1.666 0.858 20 9 40 7.08 0.768 51.5 17 39 14 15 7.79 0.826 57.7 16 49 42 15 1.332 19 25 42 1.671 0.856 19 17 34 7.29 0.768 48.8 16 47 26 18 7.74 0.822 57.7 17 9 26 15 1.327 19 35 16 1.672 0.855 19 15 26 7.23 0.766 49.3 16 49 20 17 7.93 0.820 55.6 16 49 24 17 1.326 19 47 24 1.674 0.854 19 33 18 7.17 0.766 50.0 16 15 42 18 7.72 0.820 57.7 16 41 50 15 1.324 19 43 28 1.675 0.854 19 23 18 7.20 0.765 49.5 16 45 24 18 7.74 0.819 57.4 16 15 42 17 1.322 19 37 14 1.677 0.853 19 41 26 7.04 0.760 50.9 16 47 30 18 7.84 0.817 56.2 16 47 26 17 1.321 19 39 16 1.679 0.852 19 25 42 6.98 0.760 51.7 16 41 50 15 7.61 0.813 58.2 16 47 18 15 1.319 19 41 26 1.679 0.852 19 15 40 7.15 0.759 49.5 16 47 18 17 7.73 0.813 57.0 16 15 24 17

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-66 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 34315.5 Exposure: MWd/MTU (GWd) 19211.3 (2657.87 ) Delta E: MWd/MTU, (GWd) 241.3 ( 33.38 ) Axial Profile Edit Radial Power Power: MWt 3703.0 ( 93.70 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1039.8 Top 25 0.311 8.003 13 0.356 0.376 5 48 Inlet Subcooling: Btu/lbm -29.58 24 0.830 21.977 14 0.312 0.396 53 50 Flow: Mlb/hr 107.62 (105.00 %) 23 1.048 28.075 15 0.428 0.432 3 18 22 1.175 31.344 16 1.026 1.116 49 24 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.251 33.569 17 0.790 1.111 47 22 59 -- -- -- -- -- -- -- 59 20 1.305 35.219 18 0.964 1.077 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.316 36.021 19 1.291 1.339 49 22 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.329 37.033 20 1.178 1.325 51 22 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.318 37.678 21 1.169 1.294 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.293 38.201 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.285 39.189 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.240 38.995 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.338* 36.971 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.327 37.979 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.283 38.473 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.235 39.166 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.159 39.934 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.041 40.331 7 -- -- -- -- -- -- -- -- -- 7 7 0.901 40.674 3 -- -- -- -- -- -- -- 3 6 0.774 41.494* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.654 41.329 4 0.561 39.727 Control Rod Density: % 0.00 3 0.503 36.420 2 0.406 27.762 k-effective: 1.00012 Bottom 1 0.119 8.153 Void Fraction: 0.323 Core Delta-P: psia 24.829  % AXIAL TILT 15.796 -5.883 Core Plate Delta-P: psia 20.263 AVG BOT 8ft/12ft 0.9185 1.0348 Coolant Temp: Deg-F 544.2 In Channel Flow: Mlb/hr 95.77 Active Channel Flow: Mlb/hr 92.78 Total Bypass Flow (%): 11.0 (of total core flow) Total Water Rod Flow (%): 2.8 (of total core flow) Source Convergence 0.00007

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.339 19 49 22 1.728 0.828 19 49 40 7.09 0.746 48.6 17 47 22 18 7.71 0.805 56.3 17 47 40 17 1.334 19 33 18 1.745 0.820 20 9 40 7.08 0.742 48.4 16 49 24 18 7.31 0.794 59.5 17 39 48 15 1.331 19 37 18 1.754 0.815 19 47 24 6.97 0.737 49.1 16 47 26 18 7.68 0.793 55.5 16 49 42 17 1.327 19 35 42 1.766 0.810 19 15 26 6.93 0.737 49.6 16 49 20 17 7.68 0.790 55.2 16 49 24 18 1.325 19 47 24 1.768 0.809 19 39 50 6.86 0.736 50.4 16 15 42 18 7.48 0.788 57.0 16 15 42 18 1.325 20 51 22 1.768 0.809 19 17 34 6.88 0.734 49.9 16 45 24 18 7.63 0.788 55.4 17 9 26 17 1.325 19 35 16 1.772 0.807 19 15 40 6.90 0.734 49.7 17 39 48 17 7.60 0.786 55.7 16 47 26 18 1.322 19 37 14 1.772 0.807 19 33 18 6.94 0.731 48.7 16 13 44 18 7.35 0.785 58.1 16 41 50 15 1.320 19 43 28 1.772 0.807 19 23 18 6.86 0.730 49.6 16 15 46 18 7.45 0.783 56.9 16 49 34 18 1.319 19 39 16 1.773 0.807 19 23 48 6.80 0.730 50.4 16 49 34 18 7.46 0.781 56.6 16 15 24 18

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-67 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 34598.2 Exposure: MWd/MTU (GWd) 19494.1 (2697.00 ) Delta E: MWd/MTU, (GWd) 282.8 ( 39.13 ) Axial Profile Edit Radial Power Power: MWt 3474.6 ( 87.92 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1035.4 Top 25 0.323 8.098 13 0.356 0.376 5 48 Inlet Subcooling: Btu/lbm -27.52 24 0.864 22.234 14 0.312 0.397 53 50 Flow: Mlb/hr 107.62 (105.00 %) 23 1.091 28.399 15 0.428 0.432 3 18 22 1.220 31.709 16 1.024 1.117 49 24 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.293 33.957 17 0.792 1.112 47 22 59 -- -- -- -- -- -- -- 59 20 1.341 35.623 18 0.959 1.076 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.344 36.427 19 1.288 1.344 49 22 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.351* 37.442 20 1.186 1.332 51 22 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.332 38.083 21 1.175 1.299 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.300 38.597 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.288 39.581 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.239 39.373 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.337 37.325 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.323 38.330 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.275 38.811 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.219 39.490 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.134 40.238 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.009 40.602 7 -- -- -- -- -- -- -- -- -- 7 7 0.863 40.908 3 -- -- -- -- -- -- -- 3 6 0.736 41.694* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.618 41.497 4 0.529 39.871 Control Rod Density: % 0.00 3 0.474 36.549 2 0.383 27.867 k-effective: 1.00014 Bottom 1 0.113 8.185 Void Fraction: 0.309 Core Delta-P: psia 24.523  % AXIAL TILT 18.219 -5.640 Core Plate Delta-P: psia 19.963 AVG BOT 8ft/12ft 0.9021 1.0335 Coolant Temp: Deg-F 543.9 In Channel Flow: Mlb/hr 95.95 Active Channel Flow: Mlb/hr 93.02 Total Bypass Flow (%): 10.8 (of total core flow) Total Water Rod Flow (%): 2.7 (of total core flow) Source Convergence 0.00008

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.344 19 49 22 1.820 0.786 19 49 40 6.76 0.715 49.1 17 47 22 18 7.47 0.774 55.8 17 47 40 18 1.332 20 51 22 1.836 0.779 20 9 40 6.76 0.713 48.8 16 49 24 18 7.46 0.763 54.8 16 49 42 18 1.329 19 33 18 1.859 0.769 19 47 24 6.72 0.710 49.0 16 49 20 18 7.27 0.760 56.5 17 39 48 18 1.326 19 37 18 1.864 0.767 19 39 50 6.63 0.705 49.6 16 47 26 18 7.34 0.760 55.7 16 49 24 18 1.325 19 47 24 1.866 0.766 20 9 42 6.53 0.704 50.8 16 15 42 18 7.42 0.759 54.6 17 9 26 18 1.321 19 35 16 1.876 0.762 19 15 26 6.65 0.703 49.1 16 47 18 18 7.14 0.756 57.5 16 15 42 18 1.321 19 37 14 1.878 0.761 19 23 48 6.62 0.702 49.3 17 39 48 18 7.25 0.755 56.2 16 47 26 18 1.321 19 35 42 1.879 0.761 19 15 40 6.57 0.702 50.1 16 15 46 18 7.32 0.753 55.2 16 47 18 18 1.319 19 39 50 1.880 0.761 19 39 16 6.54 0.701 50.4 16 45 24 18 7.31 0.753 55.2 16 41 50 18 1.318 19 39 16 1.881 0.760 19 47 42 6.68 0.701 48.4 17 51 26 18 7.20 0.752 56.4 16 15 16 18

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-68 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 34915.3 Exposure: MWd/MTU (GWd) 19811.1 (2740.86 ) Delta E: MWd/MTU, (GWd) 317.0 ( 43.86 ) Axial Profile Edit Radial Power Power: MWt 3211.8 ( 81.27 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1030.3 Top 25 0.337 8.208 13 0.356 0.377 5 48 Inlet Subcooling: Btu/lbm -25.17 24 0.905 22.534 14 0.312 0.398 53 50 Flow: Mlb/hr 107.62 (105.00 %) 23 1.142 28.778 15 0.427 0.432 3 18 22 1.274 32.136 16 1.021 1.119 49 24 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.342 34.408 17 0.794 1.112 47 22 59 -- -- -- -- -- -- -- 59 20 1.384* 36.089 18 0.954 1.075 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.377 36.892 19 1.286 1.349 49 22 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.376 37.909 20 1.194 1.342 51 22 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.348 38.541 21 1.182 1.305 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.308 39.043 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.290 40.023 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.237 39.796 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.334 37.720 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.317 38.721 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.261 39.186 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.196 39.848 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.102 40.569 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 0.970 40.896 7 -- -- -- -- -- -- -- -- -- 7 7 0.820 41.157 3 -- -- -- -- -- -- -- 3 6 0.693 41.906* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.580 41.674 4 0.496 40.024 Control Rod Density: % 0.00 3 0.444 36.685 2 0.360 27.977 k-effective: 1.00019 Bottom 1 0.106 8.219 Void Fraction: 0.293 Core Delta-P: psia 24.180  % AXIAL TILT 21.025 -5.348 Core Plate Delta-P: psia 19.626 AVG BOT 8ft/12ft 0.8827 1.0319 Coolant Temp: Deg-F 543.6 In Channel Flow: Mlb/hr 96.15 Active Channel Flow: Mlb/hr 93.30 Total Bypass Flow (%): 10.7 (of total core flow) Total Water Rod Flow (%): 2.6 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.349 19 49 22 1.944 0.735 19 49 40 6.36 0.677 49.6 17 47 22 18 7.05 0.736 56.3 17 47 40 18 1.342 20 51 22 1.957 0.731 20 9 40 6.38 0.676 49.4 16 49 24 18 7.07 0.729 55.3 16 49 42 18 1.327 20 51 20 1.984 0.721 20 9 42 6.36 0.676 49.5 16 49 20 18 7.03 0.723 55.2 17 9 26 18 1.324 19 47 24 1.992 0.718 19 39 50 6.28 0.668 49.6 16 47 18 18 6.87 0.723 57.0 17 39 48 18 1.324 19 39 50 1.998 0.716 19 47 24 6.20 0.667 50.6 16 15 46 18 6.93 0.722 56.2 16 49 24 18 1.323 19 33 18 2.007 0.713 20 39 10 6.24 0.666 49.9 16 41 50 18 6.94 0.719 55.7 16 47 18 18 1.321 19 37 18 2.008 0.712 20 49 18 6.14 0.666 51.3 16 15 42 18 6.93 0.718 55.8 16 41 50 18 1.320 19 37 14 2.013 0.711 20 41 10 6.31 0.665 48.9 17 51 26 18 6.72 0.717 58.1 16 15 42 18 1.317 19 35 16 2.015 0.710 19 9 18 6.22 0.665 50.1 16 47 26 18 6.80 0.716 57.0 16 15 16 18 1.316 20 49 18 2.015 0.710 19 47 42 6.24 0.664 49.8 17 39 48 18 6.82 0.715 56.7 16 47 26 18

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page A-69 AREVA Inc.

Cycle: 21 Core Average Exposure: MWd/MTU 35515.2 Exposure: MWd/MTU (GWd) 20411.0 (2823.85 ) Delta E: MWd/MTU, (GWd) 599.9 ( 82.99 ) Axial Profile Edit Radial Power Power: MWt 2719.0 ( 68.80 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1020.8 Top 25 0.366 8.431 13 0.355 0.378 5 48 Inlet Subcooling: Btu/lbm -20.80 24 0.996 23.144 14 0.312 0.401 53 50 Flow: Mlb/hr 107.62 (105.00 %) 23 1.254 29.547 15 0.428 0.432 3 18 22 1.391 32.998 16 1.017 1.127 49 20 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.448 35.311 17 0.799 1.112 47 22 59 -- -- -- -- -- -- -- 59 20 1.474* 37.014 18 0.943 1.072 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.444 37.806 19 1.280 1.360 49 22 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.424 38.815 20 1.212 1.360 51 22 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.376 39.423 21 1.197 1.317 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.318 39.893 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.287 40.857 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.225 40.593 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.318 38.463 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.292 39.452 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.220 39.882 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.138 40.503 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.028 41.167 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 0.886 41.416 7 -- -- -- -- -- -- -- -- -- 7 7 0.737 41.594 3 -- -- -- -- -- -- -- 3 6 0.616 42.273* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.513 41.981 4 0.439 40.286 Control Rod Density: % 0.00 3 0.394 36.920 2 0.322 28.168 k-effective: 0.99958 Bottom 1 0.095 8.278 Void Fraction: 0.261 Core Delta-P: psia 23.556  % AXIAL TILT 26.699 -4.738 Core Plate Delta-P: psia 19.011 AVG BOT 8ft/12ft 0.8420 1.0286 Coolant Temp: Deg-F 543.1 In Channel Flow: Mlb/hr 96.54 Active Channel Flow: Mlb/hr 93.83 Total Bypass Flow (%): 10.3 (of total core flow) Total Water Rod Flow (%): 2.5 (of total core flow) Source Convergence 0.00010

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.360 20 51 22 2.232 0.641 19 49 40 5.87 0.610 47.5 16 49 20 20 6.23 0.659 57.4 17 47 40 18 1.360 19 49 22 2.247 0.636 20 9 40 5.72 0.606 49.3 16 15 46 20 6.55 0.657 53.2 16 49 42 20 1.350 20 51 20 2.267 0.631 20 9 42 5.83 0.605 47.4 16 47 18 20 6.50 0.652 53.1 17 9 26 20 1.334 20 49 18 2.275 0.628 19 39 50 5.70 0.604 49.3 16 49 24 20 6.31 0.652 55.5 16 15 16 20 1.334 19 39 50 2.289 0.625 19 9 18 5.77 0.602 47.9 16 41 50 20 6.49 0.652 53.2 16 47 18 20 1.331 20 41 10 2.289 0.625 19 47 16 5.70 0.602 49.0 17 47 22 20 6.43 0.649 53.6 16 41 50 20 1.329 20 39 10 2.291 0.624 20 49 18 5.78 0.600 47.5 16 17 48 20 6.06 0.647 58.1 17 39 48 18 1.324 20 43 50 2.293 0.624 19 15 48 5.80 0.600 47.0 17 51 26 20 6.43 0.647 53.3 16 17 48 20 1.322 19 47 24 2.294 0.623 20 39 10 5.52 0.596 51.1 16 45 20 20 6.24 0.646 55.6 16 49 24 20 1.320 19 47 16 2.294 0.623 20 41 10 5.57 0.596 50.2 16 49 34 20 6.15 0.644 56.6 16 49 34 20

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page B-1 AREVA Inc.

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page B-2 AREVA Inc. Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page B-3 AREVA Inc. Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page B-4 AREVA Inc. Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page C-1 AREVA Inc. Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page C-2 AREVA Inc. Fuel Core Fuel Core Fuel Core Fuel Core ID Coord. ID Coord. ID Coord. ID Coord. ______________ ______________ ______________ ______________

21A001 29-54 21A039 45-44 21A077 37-36 21A115 49-30 21A002 31-54 21A040 51-44 21A078 41-36 21A116 53-30 21A003 17-52 21A041 13-42 21A079 45-36 21A117 9-28 21A004 27-52 21A042 17-42 21A080 49-36 21A118 13-28 21A005 33-52 21A043 21-42 21A081 9-34 21A119 17-28 21A006 43-52 21A044 25-42 21A082 13-34 21A120 21-28 21A007 21-50 21A045 29-42 21A083 17-34 21A121 25-28 21A008 25-50 21A046 31-42 21A084 21-34 21A122 29-28 21A009 29-50 21A047 35-42 21A085 25-34 21A123 31-28 21A010 31-50 21A048 39-42 21A086 29-34 21A124 35-28 21A011 35-50 21A049 43-42 21A087 31-34 21A125 39-28 21A012 39-50 21A050 47-42 21A088 35-34 21A126 43-28 21A013 15-48 21A051 11-40 21A089 39-34 21A127 47-28 21A014 19-48 21A052 15-40 21A090 43-34 21A128 51-28 21A015 23-48 21A053 19-40 21A091 47-34 21A129 11-26 21A016 27-48 21A054 23-40 21A092 51-34 21A130 15-26 21A017 33-48 21A055 27-40 21A093 7-32 21A131 19-26 21A018 37-48 21A056 33-40 21A094 11-32 21A132 23-26 21A019 41-48 21A057 37-40 21A095 15-32 21A133 27-26 21A020 45-48 21A058 41-40 21A096 19-32 21A134 33-26 21A021 13-46 21A059 45-40 21A097 23-32 21A135 37-26 21A022 17-46 21A060 49-40 21A098 27-32 21A136 41-26 21A023 21-46 21A061 13-38 21A099 33-32 21A137 45-26 21A024 25-46 21A062 17-38 21A100 37-32 21A138 49-26 21A025 29-46 21A063 21-38 21A101 41-32 21A139 13-24 21A026 31-46 21A064 25-38 21A102 45-32 21A140 17-24 21A027 35-46 21A065 29-38 21A103 49-32 21A141 21-24 21A028 39-46 21A066 31-38 21A104 53-32 21A142 25-24 21A029 43-46 21A067 35-38 21A105 7-30 21A143 29-24 21A030 47-46 21A068 39-38 21A106 11-30 21A144 31-24 21A031 9-44 21A069 43-38 21A107 15-30 21A145 35-24 21A032 15-44 21A070 47-38 21A108 19-30 21A146 39-24 21A033 19-44 21A071 11-36 21A109 23-30 21A147 43-24 21A034 23-44 21A072 15-36 21A110 27-30 21A148 47-24 21A035 27-44 21A073 19-36 21A111 33-30 21A149 11-22 21A036 33-44 21A074 23-36 21A112 37-30 21A150 15-22 21A037 37-44 21A075 27-36 21A113 41-30 21A151 19-22 21A038 41-44 21A076 33-36 21A114 45-30 21A152 23-22 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page C-3 AREVA Inc. (Continued) Fuel Core Fuel Core Fuel Core Fuel Core ID Coord. ID Coord. ID Coord. ID Coord. ______________ ______________ ______________ ______________ 21A153 27-22 21A167 43-20 21A181 21-16 21A195 41-14 21A154 33-22 21A168 47-20 21A182 25-16 21A196 45-14 21A155 37-22 21A169 9-18 21A183 29-16 21A197 21-12 21A156 41-22 21A170 15-18 21A184 31-16 21A198 25-12 21A157 45-22 21A171 19-18 21A185 35-16 21A199 29-12 21A158 49-22 21A172 23-18 21A186 39-16 21A200 31-12 21A159 13-20 21A173 27-18 21A187 43-16 21A201 35-12 21A160 17-20 21A174 33-18 21A188 47-16 21A202 39-12 21A161 21-20 21A175 37-18 21A189 15-14 21A203 17-10 21A162 25-20 21A176 41-18 21A190 19-14 21A204 27-10 21A163 29-20 21A177 45-18 21A191 23-14 21A205 33-10 21A164 31-20 21A178 51-18 21A192 27-14 21A206 43-10 21A165 35-20 21A179 13-16 21A193 33-14 21A207 29- 8 21A166 39-20 21A180 17-16 21A194 37-14 21A208 31- 8 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page C-4 AREVA Inc. (Continued) Fuel Core Fuel Core Fuel Core Fuel Core ID Coord. ID Coord. ID Coord. ID Coord. ______________ ______________ ______________ ______________

21B001 19-56 21B024 13-48 21B047 5-28 21B070 49-14 21B002 23-56 21B025 47-48 21B048 55-28 21B071 13-12 21B003 27-56 21B026 49-48 21B049 7-26 21B072 17-12 21B004 33-56 21B027 9-46 21B050 53-26 21B073 43-12 21B005 37-56 21B028 51-46 21B051 5-24 21B074 47-12 21B006 41-56 21B029 7-44 21B052 7-24 21B075 15-10 21B007 17-54 21B030 11-44 21B053 53-24 21B076 19-10 21B008 23-54 21B031 49-44 21B054 55-24 21B077 21-10 21B009 25-54 21B032 53-44 21B055 9-22 21B078 39-10 21B010 35-54 21B033 5-42 21B056 51-22 21B079 41-10 21B011 37-54 21B034 9-42 21B057 5-20 21B080 45-10 21B012 43-54 21B035 51-42 21B058 9-20 21B081 17- 8 21B013 15-52 21B036 55-42 21B059 51-20 21B082 23- 8 21B014 19-52 21B037 9-40 21B060 55-20 21B083 25- 8 21B015 21-52 21B038 51-40 21B061 7-18 21B084 35- 8 21B016 39-52 21B039 5-38 21B062 11-18 21B085 37- 8 21B017 41-52 21B040 7-38 21B063 49-18 21B086 43- 8 21B018 45-52 21B041 53-38 21B064 53-18 21B087 19- 6 21B019 13-50 21B042 55-38 21B065 9-16 21B088 23- 6 21B020 17-50 21B043 7-36 21B066 51-16 21B089 27- 6 21B021 43-50 21B044 53-36 21B067 11-14 21B090 33- 6 21B022 47-50 21B045 5-34 21B068 13-14 21B091 37- 6 21B023 11-48 21B046 55-34 21B069 47-14 21B092 41- 6 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page C-5 AREVA Inc. (Continued) Fuel Core Fuel Core Fuel Core Fuel Core ID Coord. ID Coord. ID Coord. ID Coord. ______________ ______________ ______________ ______________

21C001 19-54 21C009 9-48 21C017 7-28 21C025 13-10 21C002 27-54 21C010 51-48 21C018 53-28 21C026 23-10 21C003 33-54 21C011 7-42 21C019 9-24 21C027 37-10 21C004 41-54 21C012 53-42 21C020 51-24 21C028 47-10 21C005 13-52 21C013 9-38 21C021 7-20 21C029 19- 8 21C006 23-52 21C014 51-38 21C022 53-20 21C030 27- 8 21C007 37-52 21C015 7-34 21C023 9-14 21C031 33- 8 21C008 47-52 21C016 53-34 21C024 51-14 21C032 41- 8

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page D-1 AREVA Inc. Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page D-2 AREVA Inc. 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29

60 40.500 36.650 36.551 36.681 36.580 36.026 35.858 58 37.941 34.342 23.602 22.989 20.754 21.312 21.137 18.992 56 38.538 37.368 23.040 21.894 0.000 22.003 0.000 24.245 0.000 23.152 54 37.746 23.870 19.020 0.000 0.000 23.320 0.000 0.000 0.000 0.000 52 41.157 24.463 0.000 0.000 0.000 0.000 0.000 0.000 24.966 0.000 25.240 50 38.240 37.716 24.146 23.747 0.000 23.814 0.000 24.623 0.000 23.917 0.000 24.044 0.000 48 37.077 23.753 0.000 0.000 0.000 0.000 24.530 0.000 24.833 0.000 24.434 0.000 23.900 46 37.799 22.874 18.854 0.000 24.108 0.000 23.759 0.000 23.855 0.000 25.415 0.000 22.995 0.000 44 40.303 34.324 21.808 0.000 0.000 0.000 24.501 0.000 21.940 0.000 24.839 0.000 24.646 0.000 24.973 42 36.284 23.305 0.000 0.000 0.000 24.744 0.000 24.258 0.000 21.946 0.000 24.746 0.000 24.802 0.000 40 36.653 23.438 21.976 23.477 0.000 0.000 24.701 0.000 24.183 0.000 23.737 0.000 23.521 0.000 24.721 38 36.466 20.581 0.000 0.000 0.000 23.922 0.000 25.442 0.000 24.650 0.000 23.378 0.000 24.766 0.000 36 36.731 21.315 24.257 0.000 24.078 0.000 24.560 0.000 24.923 0.000 23.569 0.000 24.994 0.000 24.493 34 36.014 20.989 0.000 0.000 0.000 23.974 0.000 23.408 0.000 25.037 0.000 23.955 0.000 25.038 0.000 32 35.598 18.960 23.363 0.000 25.537 0.000 25.333 0.000 24.888 0.000 24.290 0.000 25.127 0.000 24.989 30 35.633 18.953 23.149 0.000 25.511 0.000 25.295 0.000 24.884 0.000 24.313 0.000 25.127 0.000 25.040 28 36.233 20.996 0.000 0.000 0.000 23.985 0.000 23.398 0.000 25.027 0.000 23.955 0.000 24.984 0.000 26 36.704 21.324 24.256 0.000 24.078 0.000 24.557 0.000 24.921 0.000 23.574 0.000 24.176 0.000 24.492 24 36.488 20.598 0.000 0.000 0.000 23.905 0.000 25.435 0.000 24.684 0.000 23.373 0.000 24.768 0.000 22 36.597 23.424 21.971 23.479 0.000 0.000 24.700 0.000 24.802 0.000 23.764 0.000 23.520 0.000 24.749 20 36.331 23.314 0.000 0.000 0.000 24.745 0.000 24.231 0.000 21.954 0.000 24.787 0.000 24.801 0.000 18 40.317 34.318 21.804 0.000 0.000 0.000 24.497 0.000 21.955 0.000 24.993 0.000 24.652 0.000 24.972 16 37.798 22.874 18.880 0.000 24.094 0.000 23.750 0.000 23.850 0.000 25.413 0.000 23.000 0.000 14 37.381 23.760 0.000 0.000 0.000 0.000 24.526 0.000 24.833 0.000 24.437 0.000 23.882 12 38.236 37.911 24.148 23.742 0.000 23.819 0.000 24.622 0.000 23.917 0.000 24.057 0.000 10 41.204 24.468 0.000 0.000 0.000 0.000 0.000 0.000 24.970 0.000 25.223 8 37.932 23.868 19.038 0.000 0.000 23.326 0.000 0.000 0.000 0.000 6 38.541 37.068 23.043 21.896 0.000 22.001 0.000 24.247 0.000 23.386 4 37.943 34.314 23.602 22.993 20.750 21.307 21.147 18.993 2 40.503 36.674 36.720 36.703 36.789 36.245 35.844

Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page D-3 AREVA Inc. 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 60 35.888 36.044 36.764 36.709 36.677 36.696 40.496 58 18.980 21.144 21.337 20.774 22.971 23.620 34.331 37.959 56 23.452 0.000 23.539 0.000 21.982 0.000 21.882 23.033 37.082 38.526 54 0.000 0.000 0.000 0.000 23.325 0.000 0.000 19.045 23.897 37.740 52 25.183 0.000 24.970 0.000 0.000 0.000 0.000 0.000 0.000 24.460 41.165 50 0.000 24.056 0.000 23.909 0.000 24.631 0.000 23.806 0.000 23.711 24.141 37.756 38.226 48 23.848 0.000 24.434 0.000 24.812 0.000 24.489 0.000 0.000 0.000 0.000 23.797 37.354 46 0.000 22.990 0.000 25.371 0.000 23.833 0.000 24.095 0.000 24.084 0.000 18.883 22.864 37.840 44 24.972 0.000 24.645 0.000 24.770 0.000 22.192 0.000 24.462 0.000 0.000 0.000 21.794 34.335 40.312 42 0.000 24.759 0.000 24.775 0.000 22.190 0.000 24.204 0.000 24.768 0.000 0.000 0.000 23.336 36.354 40 24.723 0.000 23.541 0.000 24.077 0.000 24.987 0.000 24.681 0.000 0.000 23.509 21.955 23.401 36.703 38 0.000 24.774 0.000 23.457 0.000 24.669 0.000 25.393 0.000 23.901 0.000 0.000 0.000 20.627 36.489 36 24.495 0.000 24.174 0.000 23.612 0.000 24.918 0.000 24.562 0.000 24.080 0.000 23.537 21.373 36.570 34 0.000 24.964 0.000 23.965 0.000 24.984 0.000 23.388 0.000 23.978 0.000 0.000 0.000 20.989 36.045 32 25.019 0.000 25.146 0.000 24.286 0.000 24.893 0.000 25.247 0.000 25.459 0.000 23.128 18.947 35.686 30 24.991 0.000 25.145 0.000 24.252 0.000 24.897 0.000 25.271 0.000 25.490 0.000 23.466 18.942 35.653 28 0.000 25.024 0.000 23.961 0.000 24.968 0.000 23.394 0.000 23.966 0.000 0.000 0.000 20.976 36.228 26 24.495 0.000 24.970 0.000 23.602 0.000 24.919 0.000 24.565 0.000 24.074 0.000 23.539 21.358 36.758 24 0.000 24.780 0.000 23.463 0.000 24.629 0.000 25.403 0.000 23.913 0.000 0.000 0.000 20.612 36.453 22 24.684 0.000 23.554 0.000 24.805 0.000 24.172 0.000 24.682 0.000 0.000 23.504 21.951 23.408 36.698 20 0.000 24.734 0.000 24.731 0.000 22.200 0.000 24.230 0.000 24.765 0.000 0.000 0.000 23.329 36.311 18 24.978 0.000 24.652 0.000 24.107 0.000 22.199 0.000 24.462 0.000 0.000 0.000 21.789 34.330 40.294 16 0.000 22.995 0.000 25.380 0.000 23.857 0.000 24.097 0.000 24.099 0.000 18.853 22.859 37.827 14 23.858 0.000 24.437 0.000 24.810 0.000 24.488 0.000 0.000 0.000 0.000 23.788 37.080 12 0.000 24.039 0.000 23.919 0.000 24.638 0.000 23.814 0.000 23.709 24.136 37.899 38.222 10 25.212 0.000 24.972 0.000 0.000 0.000 0.000 0.000 0.000 24.458 41.198 8 0.000 0.000 0.000 0.000 23.338 0.000 0.000 19.025 23.907 37.926 6 23.149 0.000 23.542 0.000 21.979 0.000 21.877 23.028 37.374 38.524 4 18.974 21.134 21.358 20.778 22.967 23.625 34.337 37.958 2 35.896 36.246 36.739 36.677 36.721 36.676 40.489 (Continued) Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page D-4 AREVA Inc. 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 60 45.876 43.317 44.132 44.707 44.846 44.327 44.326 58 45.334 43.818 36.648 37.104 35.856 36.717 36.700 34.329 56 43.752 45.703 36.234 37.912 19.036 40.501 20.766 43.069 21.323 42.289 54 46.328 37.942 36.683 21.153 22.204 44.082 23.610 23.877 23.528 23.328 52 49.946 38.539 19.009 21.911 23.054 24.471 24.969 24.765 47.309 24.623 46.778 50 43.460 46.295 38.250 41.209 22.020 44.938 24.842 46.867 24.976 46.451 24.954 46.407 24.484 48 45.434 37.811 18.976 21.993 23.761 24.541 46.904 25.027 46.616 23.868 46.192 24.403 46.363 46 45.193 36.051 36.476 21.824 45.136 24.512 46.403 25.416 46.457 23.819 46.385 23.774 44.800 24.625 44 45.667 43.758 37.747 21.007 22.888 24.712 46.842 25.444 44.714 25.237 46.016 23.141 45.219 22.982 46.278 42 42.936 36.295 18.871 21.962 24.159 46.768 24.986 46.949 25.532 44.516 23.904 45.708 22.988 45.373 23.512 40 44.148 37.406 40.309 43.934 24.088 24.178 46.475 24.227 46.852 25.332 46.279 24.801 45.072 24.066 45.556 38 44.459 35.608 20.598 23.321 23.965 45.855 23.892 46.796 24.829 46.901 25.036 45.914 24.703 46.150 23.385 36 44.930 36.664 43.014 23.769 46.489 24.892 46.393 24.102 45.942 23.443 44.909 24.277 47.100 24.703 45.985 34 44.306 36.557 21.332 23.585 24.758 46.487 24.566 45.406 23.415 45.848 23.740 45.121 24.619 46.999 23.985 32 44.098 34.323 42.523 23.494 47.532 25.139 47.690 24.930 46.863 24.242 45.344 23.325 46.426 23.936 46.899 30 44.131 34.319 42.347 23.496 47.510 25.139 47.654 24.928 46.858 24.241 45.369 23.333 46.437 23.947 46.909 28 44.515 36.576 21.341 23.590 24.759 46.495 24.563 45.392 23.406 45.835 23.746 45.144 24.653 46.981 23.997 26 44.937 36.686 43.023 23.776 46.488 24.888 46.383 24.087 45.920 23.428 44.913 24.300 46.514 24.744 45.996 24 44.496 35.642 20.615 23.330 23.964 45.835 23.876 46.764 24.792 46.903 25.025 45.917 24.731 46.175 23.393 22 44.147 37.419 40.323 43.943 24.089 24.171 46.458 24.199 47.246 25.293 46.287 24.800 45.082 24.077 45.585 20 42.999 36.341 18.897 21.972 24.161 46.763 24.976 46.909 25.506 44.497 23.887 45.745 22.991 45.375 23.514 18 45.684 43.809 37.764 21.013 22.889 24.710 46.835 25.436 44.716 25.221 46.198 23.139 45.227 22.987 46.279 16 45.189 36.053 36.498 21.820 45.124 24.508 46.393 25.414 46.448 23.815 46.384 23.780 44.808 24.631 14 45.707 37.810 18.969 21.988 23.757 24.537 46.901 25.029 46.615 23.868 46.196 24.405 46.349 12 43.446 46.465 38.246 41.203 22.017 44.944 24.842 46.868 24.979 46.451 24.953 46.418 24.483 10 49.986 38.542 19.010 21.912 23.056 24.476 24.973 24.767 47.311 24.622 46.763 8 46.501 37.944 36.705 21.162 22.213 44.092 23.610 23.875 23.526 23.333 6 43.758 45.425 36.245 37.933 19.054 40.503 20.762 43.065 21.317 42.492 4 45.339 43.846 36.673 37.106 35.843 36.697 36.700 34.333 2 45.878 43.352 44.269 44.719 45.004 44.513 44.314 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page D-5 AREVA Inc. 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59

60 44.354 44.338 44.999 44.737 44.239 43.378 45.875 58 34.322 36.718 36.756 35.887 37.101 36.694 43.864 45.363 56 42.547 21.351 42.516 20.790 40.496 19.063 37.927 36.246 45.451 43.731 54 23.334 23.550 23.907 23.631 44.097 22.215 21.162 36.711 37.959 46.312 52 46.732 24.632 47.321 24.776 24.975 24.471 23.048 21.901 18.999 38.526 49.951 50 24.486 46.423 24.955 46.446 24.976 46.864 24.824 44.914 22.001 41.169 38.236 46.331 43.451 48 46.319 24.404 46.193 23.862 46.587 25.011 46.843 24.504 23.729 21.974 18.965 37.850 45.671 46 24.625 44.794 23.768 46.336 23.800 46.411 25.376 46.605 24.476 45.099 21.814 36.499 36.033 45.215 44 46.276 22.978 45.208 23.120 45.924 25.184 44.882 25.399 46.780 24.694 22.882 21.010 37.754 43.818 45.677 42 23.513 45.328 22.971 45.709 23.857 44.660 25.459 46.858 24.960 46.775 24.157 21.974 18.903 36.363 43.022 40 45.559 24.068 45.082 24.762 46.468 25.251 47.426 24.177 46.434 24.172 24.094 43.978 40.319 37.411 44.250 38 23.391 46.173 24.708 45.957 24.987 46.861 24.765 46.716 23.875 45.838 23.979 23.356 20.648 35.695 44.513 36 45.999 24.734 46.499 24.277 44.929 23.409 45.907 24.081 46.394 24.900 46.510 23.818 42.473 36.773 44.839 34 23.998 46.957 24.640 45.142 23.735 45.786 23.399 45.382 24.572 46.506 24.786 23.632 21.395 36.605 44.363 32 46.892 23.943 46.448 23.322 45.336 24.235 46.862 24.929 47.622 25.162 47.489 23.530 42.359 34.343 44.202 30 46.899 23.930 46.434 23.312 45.302 24.233 46.868 24.930 47.643 25.161 47.515 23.525 42.641 34.337 44.181 28 23.980 46.979 24.601 45.111 23.721 45.773 23.404 45.392 24.575 46.495 24.783 23.622 21.380 36.588 44.540 26 45.982 24.690 47.057 24.243 44.906 23.416 45.922 24.095 46.399 24.904 46.503 23.808 42.462 36.740 44.961 24 23.380 46.146 24.668 45.931 24.970 46.837 24.799 46.747 23.887 45.849 23.974 23.348 20.633 35.662 44.456 22 45.517 24.049 45.075 24.735 47.086 25.274 46.809 24.201 46.447 24.170 24.087 43.965 40.301 37.391 44.193 20 23.510 45.303 22.967 45.671 23.866 44.692 25.489 46.896 24.965 46.775 24.151 21.959 18.873 36.320 42.963 18 46.283 22.982 45.221 23.140 45.369 25.212 44.902 25.409 46.783 24.695 22.877 20.996 37.723 43.760 45.658 16 24.631 44.802 23.776 46.360 23.824 46.449 25.385 46.610 24.477 45.109 21.808 36.463 36.020 45.205 14 46.329 24.406 46.197 23.872 46.595 25.016 46.843 24.502 23.726 21.970 18.960 37.837 45.413 12 24.486 46.409 24.960 46.459 24.979 46.874 24.821 44.918 21.997 41.161 38.232 46.464 43.445 10 46.760 24.639 47.330 24.781 24.976 24.468 23.043 21.895 18.993 38.525 49.980 8 23.346 23.563 23.916 23.636 44.109 22.208 21.152 36.680 37.958 46.486 6 42.309 21.371 42.529 20.793 40.489 19.042 37.900 36.227 45.713 43.720 4 34.339 36.736 36.787 35.894 37.091 36.674 43.818 45.360 2 44.383 44.561 44.999 44.711 44.296 43.344 45.868 (Continued) Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page D-6 AREVA Inc. 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 60 0.324 0.396 0.455 0.486 0.505 0.517 0.530 58 0.430 0.542 0.732 0.808 0.857 0.883 0.899 0.915 56 0.296 0.473 0.723 0.845 0.934 0.990 1.035 1.028 1.079 1.077 54 0.468 0.744 0.908 0.982 1.026 1.065 1.116 1.142 1.143 1.142 52 0.496 0.741 0.897 1.008 1.027 1.092 1.124 1.143 1.150 1.183 1.185 50 0.298 0.469 0.744 0.894 1.015 1.056 1.120 1.084 1.083 1.117 1.172 1.210 1.235 48 0.477 0.746 0.897 1.015 1.080 1.109 1.120 1.098 0.887 0.910 1.156 1.230 1.258 46 0.433 0.726 0.910 1.008 1.054 1.109 1.146 1.162 1.130 0.915 0.913 1.170 1.242 1.255 44 0.326 0.544 0.847 0.982 1.027 1.119 1.120 1.162 1.180 1.184 1.151 1.169 1.196 1.230 1.222 42 0.398 0.734 0.934 1.026 1.092 1.083 1.098 1.127 1.184 1.213 1.230 1.218 1.230 1.193 1.166 40 0.453 0.805 0.990 1.063 1.123 1.083 0.888 0.916 1.163 1.231 1.253 1.259 1.218 1.172 0.903 38 0.487 0.857 1.034 1.116 1.143 1.117 0.910 0.914 1.169 1.220 1.260 1.264 1.239 1.152 0.921 36 0.503 0.882 1.028 1.142 1.161 1.171 1.151 1.166 1.191 1.229 1.219 1.241 1.223 1.220 1.189 34 0.517 0.899 1.078 1.142 1.181 1.206 1.222 1.231 1.225 1.189 1.173 1.164 1.220 1.232 1.253 32 0.532 0.915 1.073 1.139 1.178 1.226 1.228 1.243 1.217 1.163 0.908 0.922 1.180 1.252 1.282 30 0.532 0.915 1.076 1.139 1.178 1.226 1.228 1.243 1.217 1.163 0.908 0.922 1.181 1.252 1.279 28 0.516 0.900 1.078 1.142 1.181 1.206 1.222 1.231 1.224 1.189 1.174 1.166 1.222 1.234 1.254 26 0.505 0.883 1.029 1.143 1.161 1.170 1.151 1.165 1.190 1.228 1.219 1.242 1.238 1.223 1.190 24 0.488 0.859 1.035 1.117 1.144 1.117 0.909 0.912 1.167 1.217 1.259 1.264 1.241 1.154 0.921 22 0.456 0.807 0.991 1.064 1.124 1.082 0.888 0.915 1.150 1.229 1.252 1.259 1.219 1.173 0.903 20 0.399 0.737 0.936 1.027 1.092 1.082 1.097 1.126 1.183 1.211 1.229 1.217 1.230 1.193 1.166 18 0.326 0.547 0.849 0.983 1.027 1.119 1.120 1.161 1.179 1.183 1.152 1.169 1.196 1.230 1.222 16 0.433 0.726 0.910 1.008 1.054 1.108 1.146 1.162 1.130 0.915 0.914 1.171 1.242 1.255 14 0.475 0.745 0.897 1.015 1.080 1.109 1.120 1.099 0.887 0.911 1.156 1.230 1.259 12 0.297 0.468 0.743 0.894 1.015 1.056 1.120 1.084 1.083 1.117 1.172 1.210 1.235 10 0.495 0.741 0.897 1.008 1.027 1.092 1.124 1.143 1.150 1.183 1.186 8 0.468 0.744 0.908 0.982 1.027 1.065 1.116 1.142 1.143 1.143 6 0.297 0.474 0.723 0.847 0.935 0.990 1.035 1.028 1.078 1.075 4 0.430 0.545 0.733 0.808 0.857 0.882 0.898 0.915 2 0.324 0.397 0.453 0.486 0.502 0.515 0.530 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page D-7 AREVA Inc. 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59

60 0.530 0.516 0.503 0.486 0.454 0.397 0.325 58 0.915 0.899 0.884 0.858 0.809 0.733 0.545 0.431 56 1.074 1.080 1.037 1.036 0.991 0.935 0.847 0.724 0.475 0.296 54 1.143 1.144 1.144 1.118 1.066 1.027 0.982 0.908 0.743 0.468 52 1.186 1.184 1.151 1.144 1.124 1.092 1.027 1.008 0.896 0.740 0.495 50 1.235 1.211 1.172 1.117 1.083 1.083 1.119 1.055 1.014 0.893 0.743 0.469 0.298 48 1.259 1.230 1.156 0.910 0.887 1.097 1.119 1.107 1.078 1.014 0.897 0.745 0.474 46 1.255 1.241 1.170 0.913 0.914 1.128 1.160 1.138 1.106 1.053 1.008 0.910 0.725 0.432 44 1.221 1.230 1.195 1.167 1.149 1.181 1.174 1.159 1.118 1.118 1.027 0.983 0.849 0.546 0.326 42 1.166 1.193 1.229 1.215 1.226 1.206 1.180 1.124 1.096 1.082 1.092 1.027 0.936 0.736 0.399 40 0.903 1.172 1.217 1.257 1.244 1.226 1.149 0.913 0.887 1.082 1.124 1.064 0.992 0.808 0.456 38 0.921 1.153 1.240 1.261 1.256 1.215 1.165 0.912 0.909 1.118 1.145 1.118 1.037 0.860 0.489 36 1.190 1.222 1.237 1.240 1.217 1.227 1.189 1.165 1.151 1.171 1.162 1.145 1.039 0.885 0.507 34 1.254 1.234 1.221 1.165 1.173 1.188 1.224 1.230 1.222 1.207 1.183 1.145 1.082 0.902 0.518 32 1.279 1.252 1.180 0.921 0.907 1.163 1.216 1.243 1.229 1.227 1.180 1.142 1.078 0.918 0.533 30 1.282 1.251 1.179 0.921 0.907 1.163 1.216 1.243 1.229 1.227 1.180 1.141 1.075 0.918 0.534 28 1.253 1.231 1.219 1.163 1.172 1.188 1.224 1.231 1.222 1.207 1.182 1.144 1.081 0.902 0.518 26 1.189 1.219 1.222 1.238 1.216 1.227 1.190 1.166 1.152 1.171 1.162 1.145 1.038 0.884 0.503 24 0.920 1.151 1.237 1.259 1.255 1.216 1.167 0.913 0.910 1.118 1.144 1.118 1.036 0.859 0.488 22 0.903 1.171 1.216 1.255 1.237 1.227 1.161 0.915 0.888 1.082 1.123 1.063 0.991 0.806 0.453 20 1.166 1.192 1.229 1.216 1.227 1.207 1.182 1.125 1.096 1.082 1.091 1.026 0.934 0.734 0.398 18 1.222 1.230 1.195 1.168 1.156 1.182 1.175 1.159 1.119 1.118 1.026 0.982 0.847 0.543 0.326 16 1.255 1.242 1.170 0.914 0.915 1.129 1.160 1.139 1.106 1.052 1.007 0.909 0.725 0.431 14 1.259 1.230 1.156 0.911 0.887 1.098 1.119 1.107 1.078 1.013 0.896 0.745 0.475 12 1.235 1.211 1.173 1.118 1.083 1.084 1.119 1.054 1.013 0.893 0.743 0.469 0.298 10 1.186 1.184 1.152 1.144 1.124 1.092 1.026 1.007 0.896 0.740 0.495 8 1.143 1.145 1.145 1.118 1.066 1.026 0.981 0.907 0.742 0.467 6 1.078 1.082 1.038 1.037 0.991 0.934 0.846 0.723 0.473 0.295 4 0.917 0.902 0.884 0.858 0.809 0.732 0.542 0.430 2 0.531 0.517 0.505 0.487 0.455 0.396 0.324 (Continued) Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page D-8 AREVA Inc. 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 60 0.221 0.278 0.313 0.329 0.337 0.338 0.338 58 0.315 0.416 0.575 0.616 0.657 0.665 0.670 0.655 56 0.226 0.367 0.600 0.744 0.954 0.850 1.025 0.866 1.044 0.863 54 0.389 0.654 0.836 1.095 1.148 0.994 1.193 1.201 1.185 1.189 52 0.389 0.658 0.991 1.138 1.224 1.269 1.282 1.257 1.064 1.264 1.045 50 0.229 0.391 0.661 0.839 1.144 1.030 1.274 1.087 1.309 1.095 1.295 1.076 1.291 48 0.372 0.659 0.996 1.147 1.234 1.272 1.076 1.299 1.108 1.329 1.104 1.318 1.091 46 0.320 0.608 0.845 1.145 1.033 1.274 1.079 1.292 1.091 1.329 1.109 1.340 1.123 1.322 44 0.225 0.422 0.753 1.106 1.235 1.282 1.079 1.292 1.083 1.306 1.108 1.348 1.128 1.354 1.107 42 0.283 0.584 0.966 1.164 1.285 1.096 1.303 1.086 1.299 1.088 1.330 1.110 1.350 1.119 1.336 40 0.316 0.623 0.862 1.009 1.308 1.332 1.112 1.320 1.093 1.303 1.091 1.311 1.086 1.320 1.102 38 0.334 0.666 1.037 1.209 1.283 1.113 1.331 1.101 1.313 1.084 1.305 1.089 1.300 1.064 1.316 36 0.339 0.672 0.874 1.213 1.081 1.302 1.103 1.329 1.113 1.336 1.088 1.307 1.068 1.284 1.079 34 0.341 0.676 1.050 1.191 1.267 1.074 1.311 1.111 1.339 1.110 1.324 1.079 1.287 1.056 1.276 32 0.342 0.660 0.866 1.190 1.036 1.275 1.067 1.305 1.092 1.321 1.107 1.320 1.075 1.276 1.029 30 0.342 0.660 0.867 1.189 1.036 1.275 1.067 1.305 1.092 1.321 1.107 1.320 1.075 1.276 1.027 28 0.341 0.677 1.051 1.191 1.267 1.074 1.311 1.111 1.339 1.110 1.325 1.080 1.288 1.057 1.276 26 0.341 0.673 0.875 1.213 1.081 1.302 1.103 1.329 1.113 1.335 1.088 1.307 1.078 1.285 1.079 24 0.335 0.667 1.037 1.210 1.284 1.113 1.331 1.101 1.312 1.082 1.304 1.090 1.300 1.064 1.316 22 0.319 0.624 0.862 1.009 1.308 1.331 1.112 1.319 1.081 1.301 1.091 1.311 1.085 1.320 1.102 20 0.284 0.585 0.967 1.164 1.285 1.096 1.303 1.086 1.297 1.087 1.329 1.109 1.349 1.119 1.336 18 0.226 0.425 0.754 1.107 1.235 1.282 1.079 1.292 1.083 1.305 1.111 1.347 1.128 1.354 1.107 16 0.320 0.608 0.845 1.145 1.033 1.274 1.079 1.292 1.091 1.328 1.109 1.339 1.122 1.321 14 0.371 0.659 0.995 1.146 1.234 1.272 1.076 1.299 1.108 1.328 1.103 1.317 1.091 12 0.228 0.390 0.661 0.839 1.144 1.030 1.274 1.086 1.309 1.094 1.294 1.075 1.290 10 0.389 0.658 0.991 1.138 1.224 1.269 1.281 1.257 1.063 1.264 1.045 8 0.388 0.654 0.836 1.095 1.148 0.994 1.192 1.201 1.184 1.189 6 0.226 0.368 0.600 0.744 0.955 0.850 1.024 0.865 1.043 0.861 4 0.315 0.419 0.575 0.616 0.657 0.665 0.669 0.654 2 0.221 0.278 0.311 0.328 0.334 0.336 0.338 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page D-9 AREVA Inc. 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59

60 0.338 0.337 0.335 0.329 0.312 0.278 0.221 58 0.655 0.671 0.666 0.658 0.617 0.576 0.419 0.315 56 0.861 1.045 0.871 1.025 0.851 0.956 0.745 0.601 0.368 0.226 54 1.189 1.185 1.202 1.193 0.995 1.149 1.096 0.837 0.654 0.389 52 1.045 1.265 1.064 1.258 1.282 1.270 1.225 1.139 0.991 0.658 0.389 50 1.291 1.076 1.295 1.095 1.310 1.087 1.274 1.030 1.145 0.840 0.661 0.391 0.229 48 1.092 1.318 1.104 1.329 1.109 1.300 1.076 1.272 1.234 1.147 0.996 0.659 0.371 46 1.322 1.123 1.340 1.110 1.329 1.091 1.292 1.074 1.274 1.033 1.146 0.845 0.608 0.320 44 1.107 1.354 1.128 1.348 1.108 1.305 1.081 1.291 1.079 1.282 1.236 1.107 0.755 0.425 0.226 42 1.337 1.119 1.350 1.109 1.329 1.086 1.297 1.086 1.304 1.097 1.286 1.165 0.968 0.586 0.284 40 1.103 1.320 1.086 1.311 1.086 1.301 1.084 1.319 1.113 1.332 1.308 1.009 0.863 0.625 0.318 38 1.317 1.065 1.300 1.089 1.304 1.083 1.312 1.101 1.331 1.114 1.284 1.210 1.039 0.668 0.335 36 1.079 1.285 1.078 1.307 1.088 1.336 1.113 1.329 1.104 1.303 1.082 1.215 0.880 0.675 0.342 34 1.277 1.058 1.288 1.080 1.325 1.111 1.339 1.112 1.312 1.075 1.268 1.192 1.052 0.678 0.342 32 1.028 1.276 1.075 1.320 1.108 1.322 1.092 1.306 1.068 1.276 1.037 1.190 0.869 0.661 0.343 30 1.030 1.276 1.075 1.320 1.108 1.322 1.092 1.306 1.068 1.276 1.037 1.190 0.866 0.661 0.343 28 1.276 1.056 1.286 1.079 1.325 1.111 1.339 1.112 1.312 1.075 1.268 1.192 1.052 0.678 0.342 26 1.079 1.284 1.068 1.306 1.088 1.336 1.113 1.330 1.104 1.303 1.082 1.214 0.880 0.674 0.339 24 1.316 1.064 1.299 1.088 1.304 1.083 1.314 1.102 1.332 1.114 1.284 1.210 1.038 0.667 0.335 22 1.102 1.320 1.085 1.310 1.082 1.302 1.093 1.320 1.113 1.332 1.308 1.009 0.863 0.624 0.316 20 1.336 1.119 1.350 1.109 1.330 1.086 1.298 1.087 1.304 1.097 1.286 1.164 0.967 0.584 0.283 18 1.107 1.354 1.128 1.349 1.113 1.306 1.082 1.292 1.080 1.283 1.235 1.107 0.754 0.423 0.226 16 1.322 1.123 1.340 1.111 1.330 1.092 1.292 1.074 1.274 1.033 1.146 0.845 0.608 0.320 14 1.091 1.318 1.104 1.329 1.109 1.300 1.077 1.272 1.234 1.147 0.996 0.659 0.372 12 1.291 1.076 1.295 1.095 1.310 1.087 1.274 1.030 1.145 0.840 0.662 0.391 0.229 10 1.045 1.265 1.064 1.258 1.282 1.270 1.224 1.139 0.991 0.658 0.389 8 1.189 1.186 1.202 1.193 0.995 1.149 1.096 0.837 0.653 0.388 6 0.864 1.045 0.871 1.025 0.851 0.955 0.744 0.600 0.367 0.226 4 0.655 0.671 0.666 0.658 0.617 0.575 0.417 0.315 2 0.339 0.338 0.336 0.329 0.313 0.278 0.221 (Continued) Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page E-1 AREVA Inc. Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page E-2 AREVA Inc. Cycle: 20 Core Average Exposure: MWd/MTU 34316.0 Exposure: MWd/MTU (GWd) 19211.3 (2657.87 ) Delta E: MWd/MTU, (GWd) 241.3 ( 33.38 ) Axial Profile Edit Radial Power Power: MWt 3703.0 ( 93.70 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1039.8 Top 25 0.313 7.991 10 0.357 0.376 5 48 Inlet Subcooling: Btu/lbm -29.58 24 0.835 21.949 11 0.313 0.396 53 50 Flow: Mlb/hr 107.62 (105.00 %) 23 1.054 28.041 12 0.428 0.433 3 18 22 1.182 31.309 13 1.025 1.117 49 24 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.258 33.538 14 0.791 1.111 47 22 59 -- -- -- -- -- -- -- 59 20 1.311 35.197 15 0.963 1.078 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.320 36.011 16 1.290 1.340 49 22 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.333 37.038 17 1.179 1.326 51 22 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.319 37.695 18 1.170 1.295 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.292 38.225 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.283 39.218 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.237 39.026 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.334* 37.002 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.322 38.009 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.279 38.500 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.230 39.190 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.154 39.954 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 1.037 40.346 7 -- -- -- -- -- -- -- -- -- 7 7 0.897 40.683 3 -- -- -- -- -- -- -- 3 6 0.771 41.496* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.652 41.322 4 0.561 39.710 Control Rod Density: % 0.00 3 0.502 36.392 2 0.406 27.733 k-effective: 1.00011 Bottom 1 0.119 8.143 Void Fraction: 0.323 Core Delta-P: psia 24.819  % AXIAL TILT 16.086 -5.896 Core Plate Delta-P: psia 20.253 AVG BOT 8ft/12ft 0.9160 1.0350 Coolant Temp: Deg-F 544.2 In Channel Flow: Mlb/hr 95.78 Active Channel Flow: Mlb/hr 92.79 Total Bypass Flow (%): 11.0 (of total core flow) Total Water Rod Flow (%): 2.8 (of total core flow) Source Convergence 0.00007

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.340 16 49 22 1.726 0.829 16 49 40 7.11 0.748 48.6 14 47 22 18 7.83 0.806 55.3 14 47 40 18 1.332 16 33 18 1.742 0.821 17 9 40 7.10 0.744 48.4 13 49 24 18 7.55 0.795 57.1 14 39 48 17 1.330 16 37 18 1.752 0.816 16 47 24 6.99 0.739 49.1 13 47 26 18 7.69 0.794 55.5 13 49 42 17 1.326 16 47 24 1.764 0.811 16 15 26 7.03 0.738 48.5 13 49 20 18 7.70 0.792 55.1 13 49 24 18 1.326 17 51 22 1.766 0.810 16 17 34 6.88 0.738 50.4 13 15 42 18 7.51 0.791 57.0 13 15 42 18 1.325 16 25 42 1.767 0.809 16 39 50 6.90 0.736 49.9 13 45 24 18 7.65 0.789 55.4 14 9 26 17 1.323 16 35 16 1.769 0.808 16 15 40 6.90 0.735 49.8 14 39 48 17 7.62 0.788 55.6 13 47 26 18 1.321 16 37 14 1.771 0.808 16 23 18 6.96 0.733 48.7 13 13 44 18 7.47 0.785 56.9 13 49 34 18 1.320 16 43 28 1.771 0.807 16 33 18 6.88 0.732 49.6 13 15 46 18 7.49 0.784 56.6 13 15 24 18 1.319 16 39 16 1.772 0.807 16 23 48 6.82 0.731 50.4 13 49 34 18 7.34 0.784 58.1 13 41 50 15

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page E-3 AREVA Inc.

Cycle: 20 Core Average Exposure: MWd/MTU 34915.9 Exposure: MWd/MTU (GWd) 19811.1 (2740.86 ) Delta E: MWd/MTU, (GWd) 317.0 ( 43.86 ) Axial Profile Edit Radial Power Power: MWt 3211.8 ( 81.27 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1030.3 Top 25 0.339 8.198 10 0.356 0.377 5 48 Inlet Subcooling: Btu/lbm -25.17 24 0.911 22.509 11 0.312 0.399 53 50 Flow: Mlb/hr 107.62 (105.00 %) 23 1.149 28.748 12 0.428 0.433 3 18 22 1.281 32.106 13 1.021 1.119 49 24 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.350 34.382 14 0.795 1.112 47 22 59 -- -- -- -- -- -- -- 59 20 1.391* 36.071 15 0.954 1.075 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.382 36.886 16 1.285 1.350 49 22 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.379 37.916 17 1.195 1.342 51 22 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.349 38.559 18 1.183 1.306 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.307 39.067 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.288 40.051 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.234 39.826 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.329 37.748 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.312 38.748 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.255 39.211 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.191 39.870 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.097 40.586 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 0.965 40.908 7 -- -- -- -- -- -- -- -- -- 7 7 0.817 41.164 3 -- -- -- -- -- -- -- 3 6 0.691 41.906* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.578 41.667 4 0.495 40.006 Control Rod Density: % 0.00 3 0.444 36.657 2 0.360 27.948 k-effective: 1.00019 Bottom 1 0.106 8.210 Void Fraction: 0.292 Core Delta-P: psia 24.172  % AXIAL TILT 21.322 -5.355 Core Plate Delta-P: psia 19.617 AVG BOT 8ft/12ft 0.8801 1.0321 Coolant Temp: Deg-F 543.6 In Channel Flow: Mlb/hr 96.16 Active Channel Flow: Mlb/hr 93.31 Total Bypass Flow (%): 10.7 (of total core flow) Total Water Rod Flow (%): 2.6 (of total core flow) Source Convergence 0.00006

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.350 16 49 22 1.942 0.736 16 49 40 6.38 0.679 49.6 14 47 22 18 7.07 0.738 56.3 14 47 40 18 1.342 17 51 22 1.954 0.732 17 9 40 6.39 0.678 49.3 13 49 24 18 7.09 0.730 55.3 13 49 42 18 1.328 17 51 20 1.981 0.722 17 9 42 6.38 0.677 49.5 13 49 20 18 7.05 0.725 55.2 14 9 26 18 1.325 16 47 24 1.991 0.718 16 39 50 6.29 0.669 49.7 13 13 44 18 6.89 0.725 57.0 14 39 48 18 1.324 16 39 50 1.996 0.716 16 47 24 6.22 0.668 50.5 13 15 46 18 6.95 0.724 56.2 13 49 24 18 1.321 16 33 18 2.005 0.713 17 39 10 6.16 0.667 51.3 13 15 42 18 6.96 0.720 55.7 13 47 18 18 1.320 16 37 18 2.006 0.713 17 49 18 6.25 0.667 49.9 13 41 50 18 6.95 0.720 55.8 13 41 50 18 1.319 16 37 14 2.011 0.711 17 41 10 6.24 0.667 50.1 13 47 26 18 6.74 0.719 58.1 13 15 42 18 1.317 17 49 18 2.012 0.711 16 9 18 6.32 0.667 48.9 14 51 26 18 6.82 0.718 56.9 13 15 16 18 1.316 16 35 16 2.013 0.710 16 47 42 6.25 0.666 49.8 14 39 48 18 6.84 0.717 56.7 13 47 26 18

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3 Browns Ferry EPU (120% OLTP) MELLLA+ Equilibrium Fuel Cycle Design ANP-3544NP Revision 0 Page E-4 AREVA Inc.

Cycle: 20 Core Average Exposure: MWd/MTU 35315.5 Exposure: MWd/MTU (GWd) 20211.0 (2796.19 ) Delta E: MWd/MTU, (GWd) 399.9 ( 55.33 ) Axial Profile Edit Radial Power Power: MWt 2916.6 ( 73.80 %) N(PRA) Power Exposure Zone Avg. Max. IR JR Core Pressure: psia 1024.6 Top 25 0.356 8.345 10 0.356 0.378 5 48 Inlet Subcooling: Btu/lbm -22.54 24 0.964 22.910 11 0.313 0.401 53 50 Flow: Mlb/hr 107.62 (105.00 %) 23 1.214 29.254 12 0.428 0.433 3 18 22 1.349 32.674 13 1.018 1.124 49 20 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 1.411 34.978 14 0.798 1.112 47 22 59 -- -- -- -- -- -- -- 59 20 1.442* 36.682 15 0.947 1.073 17 44 55 -- -- -- -- -- -- -- -- -- 55 19 1.420 37.491 16 1.281 1.356 49 22 51 -- -- -- -- -- -- -- -- -- -- -- 51 18 1.405 38.517 17 1.206 1.353 51 22 47 -- -- -- -- -- -- -- -- -- -- -- -- -- 47 17 1.363 39.144 18 1.192 1.313 51 24 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 16 1.311 39.632 39 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 39 15 1.284 40.606 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 14 1.225 40.357 31 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 31 13 1.319 38.243 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 12 1.297 39.235 23 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 23 11 1.232 39.676 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 10 1.158 40.309 15 -- -- -- -- -- -- -- -- -- -- -- -- -- 15 9 1.055 40.989 11 -- -- -- -- -- -- -- -- -- -- -- 11 8 0.918 41.260 7 -- -- -- -- -- -- -- -- -- 7 7 0.770 41.461 3 -- -- -- -- -- -- -- 3 6 0.647 42.156* 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 0.541 41.876 4 0.464 40.185 Control Rod Density: % 0.00 3 0.417 36.818 2 0.340 28.079 k-effective: 0.99970 Bottom 1 0.100 8.250 Void Fraction: 0.274 Core Delta-P: psia 23.797  % AXIAL TILT 24.519 -4.957 Core Plate Delta-P: psia 19.249 AVG BOT 8ft/12ft 0.8568 1.0299 Coolant Temp: Deg-F 543.3 In Channel Flow: Mlb/hr 96.39 Active Channel Flow: Mlb/hr 93.62 Total Bypass Flow (%): 10.4 (of total core flow) Total Water Rod Flow (%): 2.6 (of total core flow) Source Convergence 0.00007

Top Ten Thermal Limits Summary - Sorted by Margin

Power MCPR APLHGR LHGR Value FT IR JR Value Margin FT IR JR Value Margin Exp. FT IR JR K Value Margin Exp. FT IR JR K 1.356 16 49 22 2.109 0.678 16 49 40 5.94 0.636 50.2 13 49 20 18 6.57 0.692 57.1 14 47 40 18 1.353 17 51 22 2.116 0.676 17 9 40 5.93 0.633 50.0 13 49 24 18 6.61 0.688 56.1 13 49 42 18 1.342 17 51 20 2.138 0.669 17 9 42 5.90 0.632 50.3 14 47 22 18 6.57 0.682 55.9 14 9 26 18 1.329 16 39 50 2.154 0.664 16 39 50 5.99 0.631 48.9 13 15 46 20 6.39 0.680 57.8 14 39 48 18 1.328 17 49 18 2.164 0.661 16 9 18 6.09 0.630 47.1 13 13 44 20 6.49 0.678 56.4 13 47 18 18 1.323 16 47 24 2.167 0.660 17 49 18 5.83 0.626 50.6 13 41 50 18 6.45 0.678 56.9 13 49 24 18 1.322 17 41 10 2.169 0.659 16 47 42 6.04 0.625 47.1 13 17 48 20 6.59 0.678 55.1 13 15 16 20 1.322 17 39 10 2.170 0.659 17 41 10 6.07 0.625 46.6 14 51 26 20 6.48 0.678 56.5 13 41 50 18 1.318 16 37 14 2.172 0.658 17 39 10 5.80 0.624 50.7 13 15 42 20 6.43 0.674 56.6 13 17 48 18 1.317 17 43 50 2.172 0.658 16 47 16 5.85 0.623 49.8 13 49 34 20 6.45 0.672 56.2 13 49 34 20

• LHGR calculated with pin-power reconstruction
• CPR calculated with pin-power reconstruction & CPR limit type 3

0414-12-F04 (Rev. 001, 03/10/2016) ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) March 2017 AREVA Inc. l © 2017 AREVA Inc. ANP-3546NP Revision 0 AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page i Item Section(s) or Page(s) Description and Justification 1 All Initial Issue

AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page ii

1.0 INTRODUCTION

............................................................................................... 1-1 2.0 SUMMARY OF RESULTS ................................................................................. 2-1 3.0 LOCA DESCRIPTION ....................................................................................... 3-1 3.1 Accident Description ............................................................................... 3-1 3.2 Acceptance Criteria ................................................................................ 3-2 4.0 LOCA ANALYSIS DESCRIPTION ..................................................................... 4-1 4.1 Blowdown Analysis ................................................................................. 4-1 4.2 Plant Parameters .................................................................................... 4-2 4.3 Refill / Reflood Analysis .......................................................................... 4-2 4.4 Heatup Analysis ...................................................................................... 4-3 4.5     ................................................................................................. 4-4 4.5.1   ............................................................... 4-4 4.6 ECCS Parameters .................................................................................. 4-5 5.0 BREAK SPECTRUM ANALYSIS DESCRIPTION ............................................. 5-1 5.1 Limiting Single Failure............................................................................. 5-1 5.2 Recirculation Line Breaks ....................................................................... 5-2 5.3 Non-Recirculation Line Breaks ............................................................... 5-3 5.3.1 Main Steam Line Breaks .............................................................. 5-4 5.3.2 Feedwater Line Breaks ................................................................ 5-5 5.3.3 HPCI Line Breaks ......................................................................... 5-5 5.3.4 LPCS Line Breaks ........................................................................ 5-6 5.3.5 LPCI Line Breaks ......................................................................... 5-7 5.3.6 RCIC Line Breaks ........................................................................ 5-7 5.3.7 RWCU Line Breaks ...................................................................... 5-7 5.3.8 Instrument Line Breaks ................................................................ 5-7 6.0 RECIRCULATION LINE BREAK LOCA ANALYSES ......................................... 6-1 6.1 Limiting Break Analysis Results .............................................................. 6-1 6.2 Break Location Analysis Results ............................................................. 6-2 6.3 Break Geometry and Size Analysis Results ............................................ 6-2 6.4 Limiting Single-Failure Analysis Results ................................................. 6-2 6.5 Axial Power Shape Analysis Results ...................................................... 6-2 AREVA Inc. l ANP-3546NP   Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+)  Page iii   7.0 NON-RECIRCULATION LINE LOCA ANALYSIS .............................................. 7-1 7.1 Limiting Non-Recirculation Line Break Results ....................................... 7-1 8.0 SINGLE-LOOP OPERATION LOCA ANALYSIS ............................................... 8-1 8.1 SLO Analysis Modeling Methodology ..................................................... 8-1 8.2 SLO Analysis Results ............................................................................. 8-2 9.0 LONG-TERM COOLABILITY............................................................................. 9-1 

10.0 CONCLUSION

S .............................................................................................. 10-1

11.0 REFERENCES

................................................................................................ 11-1 Appendix A Computer Codes ..................................................................................... A-1 Appendix B Additional Results for SF-BATTlBA TLO Recirculation Line Breaks ....... B-1 Appendix C Additional Results for SF-BATTlBB TLO Recirculation Line Breaks ...... C-1 Appendix D Additional Results for Non-Recirculation Line Breaks ............................ D-1 AREVA Inc. l ANP-3546NP   Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+)  Page iv    4.1 Initial Conditions ............................................................................................. 4-6 4.2 Reactor System Parameters ........................................................................... 4-7 4.3   ATRIUM 10XM Fuel Assembly Parameters .................................................... 4-8 4.4  High-Pressure Coolant Injection Parameters .................................................. 4-9 4.5   Low-Pressure Coolant Injection Parameters ................................................. 4-10 4.6  Low-Pressure Core Spray Parameters ......................................................... 4-11 4.7   Automatic Depressurization System Parameters .......................................... 4-12 4.8   Recirculation Discharge Isolation Valve Parameters .................................... 4-13 5.1   Available ECCS for Recirculation Line Break LOCAs ..................................... 5-8 5.2   Available ECCS for Non-Recirculation Line Break LOCAs ............................. 5-9 6.1   Results for Limiting TLO Recirculation Line Break 0.23 ft2 Split Pump Discharge SF-BATTlBA Top-Peaked Axial .......................................... 6-3 6.2   Event Times for Limiting TLO Recirculation  Line Break 0.23 ft2 Split Pump Discharge SF-BATTlBA Top-Peaked Axial................................... 6-4 6.3   TLO Recirculation Line Break Spectrum Results for SF-BATTlBA ................. 6-5 6.4   Results for TLO Recirculation Line Break 0.60 ft2 Split Pump Discharge SF-BATTlBB Top-Peaked Axial ..................................................... 6-6 6.5  Event Times for TLO Recirculation Line Break 0.60 ft2 Split Pump Discharge SF-BATTlBB Top-Peaked Axial ..................................................... 6-7 6.6   TLO Recirculation Line Break Spectrum Results for SF-BATTlBB ................. 6-8 6.7   Summary of TLO Recirculation Line Break Results Highest PCT Cases ............................................................................................................. 6-9 7.1   Event Times for Limiting Non-Recirculation Line Break 0.4 ft2 Double-Ended Guillotine SF-BATTlBA Top-Peaked Axial .............................. 7-2 7.2  Non-Recirculation Line Break Spectrum Results ............................................ 7-3 8.1   Results for Limiting SLO Recirculation Line Break 1.0 DEG Pump Suction SF-BATTlBA Mid-Peaked Axial [  ] ...................... 8-4 AREVA Inc. l ANP-3546NP   Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+)  Page v   8.2  Event Times for Limiting SLO Recirculation Line Break 1.0 DEG Pump Suction SF-BATTlBA Mid-Peaked Axial [  ] ...................... 8-5 8.3 SLO Recirculation Line Break Spectrum Results for [   ] SF-BATTlBA ...................................................................................... 8-6 8.4 Single- and Two-Loop Operation PCT Summary ............................................ 8-7 AREVA Inc. l ANP-3546NP   Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+)  Page vi    4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model ...................... 4-14 4.2 [  ] ................. 4-15 4.3 RELAX System Blowdown Model ................................................................. 4-16 4.4 RELAX Hot Channel Blowdown Model Top-Peaked Axial ............................ 4-17 4.5 RELAX Hot Channel Blowdown Model Mid-Peaked Axial ............................ 4-18 4.6 ECCS Schematic .......................................................................................... 4-19 4.7 Axial Power Distributions at 102% Power / [  ] ............................ 4-20 4.8 Axial Power Distributions at 102% Power / [  ] .............................. 4-21 4.9 Axial Power Distributions at 102% Power / [  ] .............................. 4-22 4.10 Axial Power Distributions at [  ] Power / [  ] ......................... 4-23 5.1 Steam, Feedwater, and HPCI Lines.............................................................. 5-11 6.1 Limiting Recirculation Line Break Upper Plenum Pressure ........................... 6-10 6.2 Limiting Recirculation Line Break Total Break Flow Rate ............................. 6-10 6.3 Limiting Recirculation Line Break ADS Flow Rate ........................................ 6-11 6.4 Limiting Recirculation Line Break HPCI Flow Rate ....................................... 6-11 6.5 Limiting Recirculation Line Break LPCS Flow Rate ...................................... 6-12 6.6 Limiting Recirculation Line Break Intact Loop LPCI Flow Rate ..................... 6-12 6.7 Limiting Recirculation Line Break Broken Loop LPCI Flow Rate .................. 6-13 6.8 Limiting Recirculation Line Break Upper Downcomer Mixture Level ............................................................................................................. 6-13 6.9  Limiting Recirculation Line Break Middle Downcomer Mixture Level ............................................................................................................. 6-14 6.10 Limiting Recirculation Line Break Lower Downcomer Mixture Level ............................................................................................................. 6-14 6.11 Limiting Recirculation Line Break Intact Loop Discharge Liquid Mass ............................................................................................................. 6-15 6.12 Limiting Recirculation Line Break Upper Plenum Liquid Mass ...................... 6-15 6.13 Limiting Recirculation Line Break Lower Plenum Liquid Mass ...................... 6-16 6.14  Limiting Recirculation Line Break Hot Channel Inlet Flow Rate .................... 6-16 6.15  Limiting Recirculation Line Break Hot Channel Outlet Flow Rate ................. 6-17 AREVA Inc. l ANP-3546NP   Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+)  Page vii   6.16 Limiting Recirculation Line Break Hot Channel Coolant Temperature at the Hot Node at EOB ........................................................... 6-17 6.17 Limiting Recirculation Line Break Hot Channel Quality at the Hot Node at EOB ................................................................................................. 6-18 6.18 Limiting Recirculation Line Break Hot Channel Heat Transfer Coef. at the Hot Node at EOB ....................................................................... 6-18 6.19 Limiting Recirculation Line Break Cladding Temperatures ........................... 6-19 AREVA Inc. l ANP-3546NP   Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+)  Page viii    ADS automatic depressurization system  ADSVOOS ADS valve out of service  ANS American Nuclear Society    BOL beginning of life  BWR boiling water reactor    CFR Code of Federal Regulations  CHF critical heat flux  CMWR core average metal-water reaction    DEG double-ended guillotine  DG diesel generator    ECCS emergency core cooling system  EOB end of blowdown  EPU extended power uprate (defined as 120% OLTP, 3952 MWt)  FFWTR final feedwater temperature reduction FHOOS feedwater heaters out of service  FSAR Final Safety Analysis Report    HPCI high-pressure coolant injection ICF increased core flow LOCA loss-of-coolant accident  LPCI low-pressure coolant injection  LPCS low-pressure core spray    MAPLHGR maximum average planar linear heat generation rate  MCPR minimum critical power ratio MELLLA maximum extended load line limit analysis MSIV main steam isolation valve  MWR metal-water reaction    NRC Nuclear Regulatory Commission, U.S.

OLTP original licensed thermal power (3293 MWt) AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page ix PCT peak cladding temperature RCIC reactor core isolation cooling RDIV recirculation discharge isolation valve RWCU reactor water cleanup SF-ADS single failure of ADS SF-ADSlIL single failure of ADS, initiation logic SF-ADSlSV single failure of ADS, single valve SF-BATT single failure of battery (DC) power SF-BATTlBA single failure of battery (DC) power, board A SF-BATTlBB single failure of battery (DC) power, board B SF-BATTlBC single failure of battery (DC) power, board C SF-DGEN single failure of a diesel generator SF-HPCI single failure of the HPCI system SF-LOCA single failure of opposite unit false LOCA signal SF-LPCI single failure of a LPCI valve SLO single-loop operation TLO two-loop operation AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 1-1 1.0The results of a loss-of-coolant accident (LOCA) break spectrum analysis for Browns Ferry Units 1, 2, and 3 are documented in this report. The purpose of the break spectrum analysis is to identify the parameters that result in the highest calculated peak cladding temperature (PCT) during a postulated LOCA. The LOCA parameters addressed in this report include the following: Break location Break type (double-ended guillotine (DEG) or split) Break size Limiting emergency core cooling system (ECCS) single failure Axial power shape (top- or mid-peaked) The analyses documented in this report were performed with LOCA Evaluation Models developed by AREVA Inc. and approved for reactor licensing analyses by the Nuclear Regulatory Commission, U.S. (NRC). The models and computer codes used by AREVA for LOCA analyses are collectively referred to as the EXEM BWR-2000 Evaluation Model (References 1 - 4). The EXEM BWR-2000 Evaluation Model and NRC approval are documented in Reference 1. A summary description of the LOCA analysis methodology is provided in Section 4.0. The calculations described in this report were performed in conformance with 10 CFR 50 Appendix K requirements and satisfy the event acceptance criteria identified in 10 CFR 50.46. The break spectrum analyses documented in this report were performed for a core composed entirely of ATRIUMŽ 10XM* fuel at beginning-of-life (BOL) conditions. Calculations assumed an initial core power of 102% of 3952 MWt, providing an analysis licensing basis power of 4031 MWt. The 2.0% increase reflects the maximum uncertainty in monitoring reactor power, as per NRC requirements. 3952 MWt corresponds to 120% of the original licensed thermal

• ATRIUM is a trademark of AREVA Inc.

AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 1-2 power (OLTP) and is referred to as extended power uprate (EPU). The limiting assembly in the core was assumed to be at a maximum average planar heat generation rate (MAPLHGR) limit of 13.0 kW/ft. The analyses assumed a generic ATRIUM 10XM neutronic design. Other initial conditions used in the analyses are described in Section 4.0. This report identifies the limiting LOCA break characteristics (location, type, size, single failure, and axial power shape) that will be used in subsequent analyses to determine the LOCA-ECCS MAPLHGR limit versus exposure for ATRIUM 10XM fuel used at Browns Ferry Units 1, 2, and 3. The value of PCT calculated for any given set of break characteristics is dependent on exposure and local fuel rod power peaking. Therefore, heatup analyses are performed to determine the PCT versus exposure for each ATRIUM 10XM nuclear design in the core. The heatup analyses are performed each cycle using the limiting boundary conditions determined in the break spectrum analysis. The maximum PCT versus exposure from the heatup analyses are documented in the MAPLHGR limit report. The operating domain of the power/flow map of Reference 6 is applicable for the ATRIUM 10XM. This report also presents results for single-loop operation (SLO). AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 2-1 2.0Based on analyses presented in this report, the limiting break characteristics are identified below. Location recirculation discharge pipe Type / size split / 0.23 ft2 Single failure battery (DC) power, board A Axial power shape top-peaked Initial state 102% power / The SLO LOCA analyses support operation with an ATRIUM 10XM MAPLHGR multiplier of 0.85 applied to the normal two-loop operation MAPLHGR limit. A more detailed discussion of results is provided in Sections 6.0 - 8.0. The break characteristics identified in this report can be used in subsequent heatup analyses to determine the ATRIUM 10XM MAPLHGR limit appropriate for a full core of ATRIUM 10XM as well as transition cores. AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 2-2 Available ADS valves are presented in Tables 5.1 and 5.2. No additional valves are assumed out-of-service (ADSVOOS). The conclusions of this report are applicable for operation with , FHOOS, FFWTR, and SLO. While the fuel rod temperatures in the limiting plane of the hot channel during a LOCA are dependent on exposure, the factors that determine the limiting break characteristics are primarily associated with the reactor system and are not dependent on fuel-exposure characteristics. Fuel parameters that are dependent on exposure (e.g., stored energy, local peaking) have an insignificant effect on the reactor system response during a LOCA. The limiting break characteristics determined using BOL fuel conditions are applicable for exposed fuel. Fuel exposure effects are addressed in heatup analyses performed to determine or verify MAPLHGR limits versus exposure for each ATRIUM 10XM fuel design. The break spectrum analysis was performed using the NRC-approved AREVA EXEM BWR-2000 LOCA methodology. All SER restrictions and ranges of applicability for the EXEM BWR-2000 methodology were reviewed prior to final documentation of the LOCA analysis to ensure compliance with NRC requirements and methodology limitations. AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 3-1 3.0 3.1The LOCA is described in the Code of Federal Regulations 10 CFR 50.46 as a hypothetical accident that results in a loss of reactor coolant from breaks in reactor coolant pressure boundary piping up to and including a break equivalent in size to a double-ended rupture of the largest pipe in the reactor coolant system. There is not a specifically identified cause that results in the pipe break. However, for the purpose of identifying a design basis accident, the pipe break is postulated to occur inside the primary containment before the first isolation valve. For a boiling water reactor (BWR), a LOCA may occur over a wide spectrum of break locations and sizes. Responses to the break vary significantly over the break spectrum. The largest possible break is a double-ended rupture of a recirculation pipe; however, this is not necessarily the most severe challenge to the emergency core cooling system (ECCS). A double-ended rupture of a main steam line causes the most rapid primary system depressurization, but because of other phenomena, steam line breaks are seldom limiting with respect to the event acceptance criteria of 10 CFR 50.46. Special analysis considerations are required when the break is postulated to occur in a pipe that is used as the injection path for an ECCS (e.g. core spray line). Although these breaks are relatively small, their existence disables the function of an ECCS. In addition to break location dependence, different break sizes in the same pipe produce quite different event responses, and the largest break area is not necessarily the most severe challenge to the event acceptance criteria. Because of these complexities, an analysis covering the full range of break sizes and locations is required. Regardless of the initiating break characteristics, the event response is conveniently separated into three phases: the blowdown phase, the refill phase, and the reflood phase. The relative duration of each phase is strongly dependent upon the break size and location. The last two phases are often combined and will be discussed together in this report. During the blowdown phase of a LOCA, there is a net loss-of-coolant inventory, an increase in fuel cladding temperature due to core flow degradation, and for the larger breaks, the core becomes fully or partially uncovered. There is a rapid decrease in pressure during the AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 3-2 blowdown phase. During the early phase of the depressurization, the exiting coolant provides core cooling. Later in the blowdown, core cooling is provided by lower plenum flashing as the system continues to depressurize. The blowdown phase is defined to end when LPCS reaches rated flow. In the refill phase of a LOCA, the ECCS is functioning and there is a net increase of coolant inventory. During this phase the core sprays provide core cooling and, along with low-pressure and high-pressure coolant injection (LPCI and HPCI), supply liquid to refill the lower portion of the reactor vessel. In general, the core heat transfer to the coolant is less than the fuel decay heat rate and the fuel cladding temperature continues to increase during the refill phase. In the reflood phase, the coolant inventory has increased to the point where the mixture level reenters the core region. During the core reflood phase, cooling is provided above the mixture level by entrained reflood liquid and below the mixture level by pool boiling. Sufficient coolant eventually reaches the core hot node and the fuel cladding temperature decreases. 3.2A LOCA is a potentially limiting event that may place constraints on fuel design, local power peaking, and in some cases, acceptable core power level. During a LOCA, the normal transfer of heat from the fuel to the coolant is disrupted. As the liquid inventory in the reactor decreases, the decay heat and stored energy of the fuel cause a heatup of the undercooled fuel assembly. In order to limit the amount of heat that can contribute to the heatup of the fuel assembly during a LOCA, an operating limit on the MAPLHGR is applied to each fuel assembly in the core. The Code of Federal Regulations prescribes specific acceptance criteria (10 CFR 50.46) for a LOCA event as well as specific requirements and acceptable features for Evaluation Models (10 CFR 50 Appendix K). The conformance of the EXEM BWR-2000 LOCA Evaluation Models to Appendix K is described in Reference 1. The ECCS must be designed such that the plant response to a LOCA meets the following acceptance criteria specified in 10 CFR 50.46: The calculated maximum fuel element cladding temperature shall not exceed 2200°F. The calculated local oxidation of the cladding shall nowhere exceed 0.17 times the local cladding thickness. AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 3-3 The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, except the cladding surrounding the plenum volume, were to react. Calculated changes in core geometry shall be such that the core remains amenable to cooling. After any calculated successful operation of the ECCS, the calculated core temperature shall be maintained for the extended period of time required by the long-lived radioactivity remaining in the core. These criteria are commonly referred to as the peak cladding temperature (PCT) criterion, the local oxidation criterion, the hydrogen generation criterion, the coolable geometry criterion, and the long-term cooling criterion. A MAPLHGR limit is established for the ATRIUM 10XM fuel type to ensure that these criteria are met. LOCA PCT results are provided in Sections 6.0 - 8.0 to determine the limiting LOCA event. LOCA analysis results demonstrating that the PCT, local oxidation, and hydrogen generation criteria are met are provided in the follow-on MAPLHGR report and cycle-specific heatup analyses. Cycle-specific heatup analyses are performed to demonstrate that the MAPLHGR limits versus exposure for the ATRIUM 10XM fuel remains applicable for cycle-specific nuclear designs. Compliance with these three criteria ensures that a coolable geometry is maintained. Long-term coolability criterion is discussed in Section 9.0. AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 4-1 4.0The Evaluation Model used for the break spectrum analysis is the EXEM BWR-2000 LOCA analysis methodology described in Reference 1. The EXEM BWR-2000 methodology employs three major computer codes to evaluate the system and fuel response during all phases of a LOCA. These are the RELAX, HUXY, and RODEX2 computer codes. RELAX is used to calculate the system and hot channel response during the blowdown, refill and reflood phases of the LOCA. The HUXY code is used to perform heatup calculations for the entire LOCA, and calculates the PCT and local clad oxidation at the axial plane of interest. RODEX2 is used to determine fuel parameters (such as stored energy) for input to the other LOCA codes. The code interfaces for the LOCA methodology are illustrated in Figure 4.1. A complete analysis for a given break size starts with the specification of fuel parameters using RODEX2 (Reference 4). RODEX2 is used to determine the initial stored energy for both the blowdown analysis (RELAX system and hot channel) and the heatup analysis (HUXY). This is accomplished by ensuring that the initial stored energy in RELAX and HUXY is the same or higher than that calculated by RODEX2 for the power, exposure, and fuel design being considered. 4.1The RELAX code (Reference 1) is used to calculate the system thermal-hydraulic response during the blowdown phase of the LOCA. For the system blowdown analysis, the core is represented by an average core channel. The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and with decay heating as required by Appendix K of 10 CFR 50. The reactor vessel nodalization for the system blowdown analysis is shown in Figure 4.3. This nodalization is consistent with that used in the topical report submitted to the NRC (Reference 1). The RELAX analysis is performed from the time of the break initiation through the end of blowdown (EOB). The system blowdown calculation provides the upper and lower plenum transient boundary conditions for the hot channel analysis. AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 4-2 Following the system blowdown calculation, another RELAX analysis is performed to analyze the maximum power assembly (hot channel) of the core. The RELAX hot channel calculation is used to calculate hot channel fuel, cladding, and coolant temperatures during the blowdown phase of the LOCA. The RELAX hot channel nodalization is shown in Figure 4.4 for a top-peaked power shape, and in Figure 4.5 for a mid-peaked axial power shape. The hot channel analysis is performed using the system blowdown results to supply the core power and the system boundary conditions at the core inlet and exit. The initial average fuel rod temperature at the limiting plane of the hot channel is conservative relative to the average fuel rod temperature calculated by RODEX2 for operation of the ATRIUM 10XM assembly at the MAPLHGR limit. The heat transfer coefficients and fluid conditions in the hot channel from the RELAX hot channel calculation are used as input to the HUXY heatup analysis. 4.2The LOCA break spectrum analysis is performed using the plant parameters presented in Reference 6. Table 4.1 provides a summary of reactor initial conditions used in the break spectrum analysis. Table 4.2 lists selected reactor system parameters. The break spectrum analysis is performed for a full core of ATRIUM 10XM fuel. Some of the key ATRIUM 10XM fuel parameters used in the break spectrum analysis are summarized in Table 4.3. 4.3The RELAX code is used to compute the system and hot channel hydraulic response during the refill/reflood phase of the LOCA. The RELAX system and RELAX hot channel analyses continue beyond the end of blowdown to analyze system and hot channel responses during the refill and reflood phases. The refill phase is the period when the lower plenum is filling due to ECCS injection. The reflood phase is when some portions of the core and hot assembly are being cooled with ECCS water entering from the lower plenum. The purpose of the RELAX calculations beyond blowdown is to determine the time when the liquid flow via upward entrainment from the bottom of the core becomes high enough at the hot node in the hot assembly to end the temperature increase of the fuel rod cladding. This event time is called the time of hot node reflood. AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 4-3 The time when the core bypass mixture level rises to the elevation of the hot node in the hot assembly is also determined. RELAX provides a prediction of fluid inventory during the ECCS injection period. Allowing for countercurrent flow through the core and bypass, RELAX determines the refill rate of the lower plenum due to ECCS water and the subsequent reflood times for the core, hot assembly, and the core bypass. The RELAX calculations provide HUXY with the time of hot node reflood and the time when the liquid has risen in the bypass to the height of the axial plane of interest (time of bypass reflood). 4.4The HUXY code (Reference 2) is used to perform heatup calculations for the entire LOCA transient and provides PCT and local clad oxidation at the axial plane of interest. The heat generated by metal-water reaction (MWR) is included in the HUXY analysis. HUXY is used to calculate the thermal response of each fuel rod in one axial plane of the hot channel assembly. These calculations consider thermal-mechanical interactions within the fuel rod. The clad swelling and rupture models from NUREG-0630 have been incorporated into HUXY (Reference 3). The HUXY code complies with the 10 CFR 50 Appendix K criteria for LOCA Evaluation Models. HUXY uses the end of blowdown time and the times of core bypass reflood and core reflood at the axial plane of interest from the RELAX analysis. Throughout the calculations, decay power is determined based on the ANS 1971 decay heat curve plus 20% as described in Reference 1. are used in the HUXY analysis. The principal results of a HUXY heatup analysis are the PCT and the percent local oxidation of the fuel cladding, often called the percent maximum local metal water reaction (%MWR). AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 4-4 4.5 4.5.1 AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 4-5 4.6The ECCS configuration is shown in Figure 4.6. Tables 4.4 - 4.8 provide the important ECCS characteristics assumed in the analysis. The ECCS is modeled as fill junctions connected to the appropriate reactor locations: LPCS injects into the upper plenum, HPCI injects into the upper downcomer, and LPCI injects into the recirculation lines. The flow through each ECCS valve is determined based on system pressure and valve position. Flow versus pressure for a fully open valve is obtained by linearly interpolating the pump capacity data provided in Tables 4.4 - 4.6. No credit for ECCS flow is assumed until ECCS pumps reach rated speed. The ADS valves are modeled as a junction connecting the reactor steam line to the suppression pool. The flow through the ADS valves is calculated based on pressure and valve flow characteristics. The valve flow characteristics are determined such that the calculated flow is equal to the rated capacity at the reference pressure shown in Table 4.7. In the AREVA LOCA analysis model, ECCS initiation is assumed to occur when the water level drops to the applicable level setpoint. No credit is assumed for the start of HPCI, LPCS, or LPCI due to high drywell pressure. The recirculation discharge isolation valve (RDIV) parameters are shown in Table 4.8. AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 4-6 Reactor power (% of rated) 102 102 102 Total core flow (% of rated) Reactor power (MWt) 4031 4031 4031 Total core flow (Mlb/hr) Steam flow rate (Mlb/hr) 16.85 16.85 16.85 Steam dome pressure (psia) 1053 1053 1053 Core inlet enthalpy (Btu/lb) ATRIUM 10XM hot assembly MAPLHGR (kW/ft) 13.0 13.0 13.0 13.0 ECCS fluid temperature (°F) 120 120 120 120 Axial power shape Fig. 4.7 Fig. 4.8 Fig. 4.9 Fig. 4.10

• The AREVA calculated heat balance is adjusted to match the 100% power/100% flow values given in the plant parameters document (Reference 6). The model is then rebalanced based on AREVA heat balance calculations to establish these LOCA initial conditions at 102% and of rated thermal power. f AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 4-7 Parameter Value Vessel ID (in) 251 Number of fuel assemblies 764 Recirculation suction pipe area (ft2) 3.507 1.0 DEG suction break area (ft2) 7.013 Recirculation discharge pipe area (ft2) 3.507 1.0 DEG discharge break area (ft2) 7.013 AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 4-8 Parameter Value Fuel rod array 10x10 Number of fuel rods per assembly 79 (full-length rods) 12 (part-length rods) Non-fuel rod type Water channel replaces 9 fuel rods Fuel rod OD (in) 0.4047 Active fuel length (in) (including blankets) 150.0 (full-length rods) 75 (part-length rods) Water channel outside width (in) 1.378 Fuel channel thickness (in) 0.075 (minimum wall) 0.100 (corner) Fuel channel internal width (in) 5.278 AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 4-9 Parameter Value Coolant temperature (°F) 120 Initiating Signals and Setpoints Water level* L2 (448 in) High drywell pressure (psig) 2.6 (not used) Time Delays Time for HPCI pump to reach rated speed and injection valve wide open (sec) 35 Delivered Flow Rate Versus Pressure Vessel to Drywell P (psid) Flow Rate (gpm) 0 0 150 5000 1120 5000 1174 3600
• Relative to vessel zero.

AREVA Inc. l ANP-3546NP Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+) Page 4-10 Parameter Value Reactor pressure permissive for opening valves (psia) 350 Coolant temperature (°F) 120 Initiating Signals and Setpoints Water level* L1 (372.5 in) High drywell pressure (psig) 2.6 (not used) Time DelaysTime for LPCI pumps to reach ADS permissive (max) (sec)f 32 Time for LPCI pumps to reach rated speed (max) (sec)§ 44 LPCI injection valve stroke time (sec) 40 Delivered Flow Rate Versus Pressure Vessel to Drywell P (psid) Flow Rate (gpm) 2 Pumps Into 1 Loop 4 Pumps Into