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MONTHYEARML1013401882010-06-0808 June 2010 Request for Withholding Information from Public Disclosure for Browns Ferry Nuclear Plant, Unit 1, Areva Fuel Transition, Areva Fuel Mechanical Design Report Project stage: Withholding Request Acceptance ML1013802442010-06-0808 June 2010 Request for Withholding Information from Public Disclosure for Browns Ferry Nuclear Plant, Unit 1, Areva Fuel Transition, Thermal-Hydraulic Design Report Project stage: Withholding Request Acceptance ML1025202302010-09-24024 September 2010 Request for Withholding Information from Public Disclosure for Browns Ferry Nuclear Plant, Unit 1, Areva Fuel Transition, GE14 Fuel Thermal Mechanical Information Project stage: Withholding Request Acceptance ML1025203012010-11-0404 November 2010 Letter Request for Withholding Information from Public Disclosure for Browns Ferry Nuclear Plant, Unit 1, Areva Fuel Transition, Unit 1 Cycle 9 Reload Safety Analysis for 105% Original Licensed Thermal Power Project stage: Withholding Request Acceptance ML1101805852011-01-24024 January 2011 Request for Additonal Information Regarding Amendment Request to Transition to Areva Fuel Project stage: Other ML1105900542011-02-23023 February 2011 Response to NRC Request for Additional Information Regarding Amendment Request to Transition to Areva Fuel Project stage: Response to RAI ML11095A0732011-03-28028 March 2011 Request for Additional Information Regarding a Request to Transition to Areva Fuel Project stage: RAI ML11136A1872011-04-26026 April 2011 NRR E-mail Capture - Audit Results Summary Report Supporting TAC ME3775 - BF1 Areva Fuel Transition and Related BF2/3 Request Project stage: Other ML11137A1992011-05-12012 May 2011 Response to NRC Request for Additional Information Regarding to Transition to Areva Fuel Project stage: Response to RAI ML11159A0412011-06-0707 June 2011 NRR E-mail Capture - Audit Results Summary Report Supporting TAC ME3775 - BF1 Areva Fuel Transition and Related BF2/3 Request Project stage: Other ML11234A0042011-08-23023 August 2011 Request for Additional Information Regarding Technical Specification TS-473, Areva Fuel Transition Project stage: RAI ML11286A1072011-10-0707 October 2011 Response to NRC Request for Additional Information Regarding Amendment Request to Transition to Areva Fuel Project stage: Response to RAI ML12018A3762012-01-23023 January 2012 Request for Withholding Information from Public Disclosure for Areva Fuel Transition Documents (TAC No. ME3775)(TS-473) Project stage: Withholding Request Acceptance ML12083A0012012-04-0404 April 2012 Request for Withholding Information from Public Disclosure for Areva Fuel Transition Documents (TAC No. ME3775)(TS-473) Project stage: Withholding Request Acceptance ML1208300592012-04-0404 April 2012 Request for Withholding Information from Public Disclosure for Areva Fuel Transition Documents (TAC No. ME3775) (TS-473) Project stage: Withholding Request Acceptance ML12083A0032012-04-0404 April 2012 Request for Withholding Information from Public Disclosure for Areva Fuel Transition Documents (TAC No. ME3775)(TS-473) Project stage: Withholding Request Acceptance ML12083A0142012-04-0404 April 2012 Request for Withholding Information from Public Disclosure for Areva Fuel Transition Documents (GEH-Harrison) (TAC No. ME3775) (TS-473) Project stage: Withholding Request Acceptance ML12100A1032012-04-11011 April 2012 Audit Report Regarding Tennessee Valley Authority Browns Ferry, Unit 1 Areva Fuel Transition Emergency Core Cooling System Evaluation Model Application (TAC No. ME3775) (Np) Project stage: Other ML12108A2482012-04-17017 April 2012 Request for Withholding Information from Public Disclosure for Areva Fuel Transition Documents (TAC No. ME3775)(TS-473) Project stage: Withholding Request Acceptance ML12114A0042012-04-18018 April 2012 Supplement to License Amendment Request to Transition to Areva Fuel Project stage: Supplement ML12129A1492012-07-0303 July 2012 Issuance of Amendment Regarding the Transition to Areva Fuel (TAC No. ME3775) (TS-473) Project stage: Approval ML13010A0162013-01-0404 January 2013 CFR 50.46 30-Day Report Project stage: Other 2011-05-12
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Category:Letter
MONTHYEARML24032A4762024-02-0101 February 2024 Final Report of a Part 21 Evaluation Associated with Starter Contactors for the BFN Unit 1 High Pressure Coolant Injection Suppression Pool Inboard Suction Valve ML24023A2802024-01-23023 January 2024 Final Report of a Deviation or Failure to Comply Associated with a Relay in the Reactor Core Isolation Cooling Condensate Pump CNL-24-017, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions2024-01-17017 January 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions ML24016A3042024-01-16016 January 2024 Final Report of a Part 21 Evaluation Associated with Check Valve 0-CKV-023-0565 for D1 Residual Heat Removal Service Water Pump CNL-23-071, Response to Request for Additional Information Regarding Browns Ferry Nuclear Plant, Units 1, 2, and 3, Inservice Inspection and Augmented Program Interval Revised Request for Alternative 0-ISI-472024-01-11011 January 2024 Response to Request for Additional Information Regarding Browns Ferry Nuclear Plant, Units 1, 2, and 3, Inservice Inspection and Augmented Program Interval Revised Request for Alternative 0-ISI-47 ML24022A1732024-01-0303 January 2024 Receipt and Availability of the Subsequent License Renewal Application ML23319A1992024-01-0303 January 2024 Issuance of Amendment Nos. 333, 356, and 316 Regarding the Technical Specification Surveillance Requirements 3.4.3.2 and 3.5.1.11 Regarding Safety Relief Valves ML23355A2062023-12-21021 December 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23348A3942023-12-14014 December 2023 Interim Part 21 Report of a Potential Deviation or Failure to Comply Associated with Starter Contactors for the High Pressure Coolant Injection Suppression Pool Inboard Suction Valve IR 05000259/20230102023-12-11011 December 2023 Commercial Grade Dedication Inspection Report 05000259/2023010 and 05000260/2023010 and 05000296/2023010 ML23335A0722023-12-0101 December 2023 Interim Report of a Deviation or Failure to Comply Associated with a Relay in the Unit 2 Reactor Core Isolation Cooling Condensate Pump ML23334A2492023-11-30030 November 2023 Site Emergency Plan Implementing Procedure Revision CNL-23-070, Submittal of Fifth 10-Year Interval Inservice Testing Program Plan2023-11-29029 November 2023 Submittal of Fifth 10-Year Interval Inservice Testing Program Plan ML23331A2532023-11-27027 November 2023 Summary Report for 10 CFR 50.9 Evaluations, Technical Specifications Bases Changes, Technical Requirement Manual Changes, and NRC Commitment Revisions CNL-23-067, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-11-27027 November 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML23325A1102023-11-21021 November 2023 Anchor Darling Double Disc Gate Valve Commitment Revision ML23320A2542023-11-16016 November 2023 Interim Part 21 Report of a Deviation or Failure to Comply Associated with Check Valve 0-CKV-023-0565 for D1 Residual Heat Removal Service Water Pump IR 05000259/20230032023-11-13013 November 2023 Integrated Inspection Report 05000259/2023003, 05000260/2023003 and 05000296/2023003 IR 05000259/20230402023-11-0202 November 2023 Supplemental Inspection Supplemental Report 05000259 2023040 and Follow-Up Assessment Letter ML23292A2532023-10-18018 October 2023 BFN 2024-301, Corporate Notification Letter (210-day Ltr) ML23282A0022023-10-0606 October 2023 Interim Part 21 Report of a Deviation or Failure to Comply Associated with Check Valve 0-CKV-023-0565 for D1 Residual Heat Removal Service Water Pump ML23278A0122023-10-0505 October 2023 Updated Final Safety Analysis Report, Amendment 30 ML23271A1702023-09-28028 September 2023 Site Emergency Plan Implementing Procedure Revision ML23270A0702023-09-26026 September 2023 SLRA Pre-Application Meeting Summary 09-13-2023 ML23257A1232023-09-22022 September 2023 Administrative Changes to Technical Specification Pages Issued for License Amendment Nos. 332, 355, and 315 CNL-23-061, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision2023-09-20020 September 2023 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision ML23263B1042023-09-20020 September 2023 Special Report 260/2023-001 for Inoperable Post Accident Monitoring (PAM) Instrumentation ML23205A2132023-09-0808 September 2023 Issuance of Amendment Nos. 332, 355, and 315 Regarding the Revision of Technical Specifications to Adopt TSTF-566-A and TSTF-580-A, Rev. 1 CNL-23-057, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 642023-09-0505 September 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 64 IR 05000259/20230052023-08-29029 August 2023 Updated Inspection Plan for Browns Ferry Nuclear Plant, Units 1, 2 and 3 - Report 05000259/2023005, 05000260/2023005 and 05000296/2023005 ML23233A0432023-08-18018 August 2023 Enforcement Action EA-22-122 Inspection Readiness Notification ML23219A1542023-08-17017 August 2023 Request to Use Later Edition of ASME Code for Operation and Maintenance and Alternative Requests BFN-IST-01 Through 05 for the Fifth 10-Year Interval Inservice Testing Program ML23228A1642023-08-16016 August 2023 Site Emergency Plan Implementing Procedure Revision ML23228A0202023-08-15015 August 2023 (BFN) Unit 1 - Special Report 259/2023-001 for Inoperable Post Accident Monitoring (PAM) Instrumentation IR 05000259/20230022023-08-10010 August 2023 Integrated Inspection Report 05000259/2023002, 05000260/2023002, 05000296/2023002 and 07200052/2023001 ML23192A4472023-07-31031 July 2023 Staff Assessment of Updated Seismic Hazards at TVA Sites Following the NRC Process for the Ongoing Assessment of Natural Hazards Information CNL-23-055, Tennessee Valley Authority - Emergency Preparedness Department Procedure Revision. Includes EPDP-3, Revision 3, Emergency Plan Exercises and Preparedness Drills2023-07-25025 July 2023 Tennessee Valley Authority - Emergency Preparedness Department Procedure Revision. Includes EPDP-3, Revision 3, Emergency Plan Exercises and Preparedness Drills ML23171A8862023-07-24024 July 2023 Issuance of Amend. Nos. 331, 354, and 314; 365 and 359 Regarding Adoption of TSTF-541-A, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position ML23201A2182023-07-20020 July 2023 Registration of Use of Cask to Store Spent Fuel (MPC-298 and -299) ML23159A2552023-07-20020 July 2023 Proposed Alternative to the Requirements of the ASME Code Regarding Volumetric Inspection of Standby Liquid Control Nozzles ML23199A3072023-07-18018 July 2023 Site Emergency Plan Implementing Procedure Revision CNL-23-053, Tennessee Valley Authority - Radiological Emergency Plan Revisions2023-07-18018 July 2023 Tennessee Valley Authority - Radiological Emergency Plan Revisions IR 05000259/20233012023-07-18018 July 2023 NRC Operator License Examination Report Nos. 05000259/2023301, 05000260/2023301, and 05000296/2023301 2024-02-01
[Table view] Category:Licensee 30-Day Written Event Report
MONTHYEARML17200A0842017-07-19019 July 2017 Special Report 259/2017-001 for Inoperable Post Accident Monitoring (PAM) Instrumentation ML16340A2072016-12-0505 December 2016 Transmittal of 10 CFR 50.46 30-Day Report ML16197A2872016-07-18018 July 2016 Voluntary Reporting of High Groundwater Tritium Levels ML15022A0142015-01-16016 January 2015 Special Report 260/2015-001 for Inoperable Post Accident Monitoring (PAM) Instrumentation ML14364A0252014-12-23023 December 2014 Offsite Dose Calculation Manual (ODCM) Special Report ML14175B3902014-06-20020 June 2014 CFR 50.46 30-Day Report Regarding Changes and Errors to the Calculated Peak Cladding Temperature for Emergency Core Cooling System Evaluation Model ML13010A0162013-01-0404 January 2013 CFR 50.46 30-Day Report ML12153A0532012-05-30030 May 2012 CFR 50.46 30-Day Report for Browns Ferry Nuclear Plant, Unit 3 ML1004801402010-02-12012 February 2010 Day Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes ML1004700522010-02-0808 February 2010 Submittal of 30-Day Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes 2017-07-19
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Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 January 4, 2013 10 CFR 50.4 10 CFR 50.46 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Unit 1 Facility Operating License No DPR-33 NRC Docket No 50-259
Subject:
10 CFR 50.46 30-Day Report for Browns Ferry Nuclear Plant, Unit 1
Reference:
- 1. TVA Letter to NRC, "10 CFR 50.46 Annual Report for Browns Ferry Nuclear Plant, Unit 1," dated April 30, 2012 2. NRC Letter, "Browns Ferry Nuclear Plant, Unit 1 -Issuance of Amendments Regarding the Transition to AREVA Fuel (TAC No.ME3775) (TS-473)," April 27, 2012 The purpose of this letter is to provide a 30-day report as required by Title 10 of the Code of Federal Regulations (10 CFR) 50.46 of significant changes in the Emergency Core Cooling System (ECCS) evaluation model for Browns Ferry Nuclear Plant (BFN), Unit 1. In accordance with 10 CFR 50.46, "Acceptance Criteria for ECCS for Light-Water Nuclear Power Reactors," paragraph (a)(3)(ii), the enclosure to this letter describes the nature and the estimated effect on the limiting ECCS analysis of changes or errors discovered since submittal of the 10 CFR 50.46 Annual Report for BFN, Unit 1 dated April 30, 2012 (Reference 1).During the BFN, Unit 1, Fall 2012 refueling outage, modifications were completed that restored the automatic initiation capability of the BFN, Unit 1, Automatic Depressurization System. As a result, NEDC-32484P Revision 6, "Browns Ferry Nuclear Plant Units 1, 2, and 3: SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," GE Nuclear Energy, February 2005 was re-established as the Loss of Coolant Accident (LOCA) analysis of record for GE14 fuel, with a baseline peak cladding temperature (PCT) of 1760°F for GE14 Printed on recycled paper U.S. Nuclear Regulatory Commission Page 2 January 4, 2013 fuel. The previous baseline PCT reported in the Reference 1 letter was 1920°F for GE14 fuel.AREVA ATRIUM-10 fuel was introduced into BFN, Unit 1 reactor core during the Fall 2012 refueling outage. Reference 2 documents the NRC approval of the LOCA analysis for ATRIUM-10 fuel for BFN, Unit 1 which includes the application of a modified EXEM BWR-2000 LOCA methodology.
This analysis is established as the LOCA analysis of record for ATRIUM-10 fuel with a baseline PCT of 1926°F for ATRIUM-10 fuel.In accordance with the BFN, Unit 1, Renewed Operating License and Technical Specifications, the new baseline PCT values described above became effective on December 5, 2012. The 160'F change in the baseline PCT for GE14 fuel meets the criteria of 10 CFR 50.46 (a)(3)(i) as a significant change. As such, in accordance with 10 CFR 50.46 (a)(3)(ii), this 30-day report is required to be submitted by January 4, 2013.There are no new regulatory commitments in this letter. Please direct questions concerning this issue to Tom Hess at (423) 751-3487.Respellly, JSShea President, Nuclear Licensing
Enclosure:
10 CFR 50.46 30-Day Report for Browns Ferry Nuclear Plant, Unit 1 cc (w/Enclosure):
NRC Regional Administrator-Region II NRC Senior Resident Inspector
-Browns Ferry Nuclear Plant ENCLOSURE 10 CFR 50.46 30-DAY REPORT FOR BROWNS FERRY NUCLEAR PLANT, UNIT I The Browns Ferry Nuclear Plant (BFN), Unit 1, reactor core contains both the ATRIUM-1 0 and GE14 fuel designs. This report establishes new baseline peak cladding temperature (PCT)values for both fuel types, as described below.ATRIUM-10 Fuel Evaluation AREVA ATRIUM-10 fuel was introduced into the BFN, Unit 1 reactor core during the Fall 2012 refueling outage. References 1 and 2 constitute the Loss of Coolant Accident (LOCA) analysis of record for ATRIUM-1 0 fuel in the BFN, Unit 1, reactor core. These analyses were reviewed by the NRC and approved for application to BFN, Unit 1, in Reference
- 3. The baseline PCT for ATRIUM-10 fuel is 1926 'F.On June 28, 2012, AREVA notified the Tennessee Valley Authority (TVA) of a change to their evaluation of thermal conductivity degradation over the approved burnup range. When older generation codes, like AREVA's RODEX2 were approved, experimental data was not available to support explicit modeling of thermal conductivity degradation with fuel burnup. However, in recent evaluations of this phenomenon, it appears that the use of the RODEX2 code (which provides inputs to RELAX and HUXY in the LOCA analysis methodology) results in conservatively high temperatures at low burnup (less than 15 Giga-Watt Day per Metric Ton Uranium), but underpredicts pellet temperatures at higher exposures.
For BFN, Unit 1, the current analysis (Reference
- 2) shows that the limiting PCT occurs at beginning of life (BOL). As discussed in Reference 4, the effects of thermal conductivity degradation at higher burnups result in a zero degree change in the limiting PCT, which occurs at BOL. Therefore, there is no change in the reported PCT due to thermal conductivity degradation for BFN, Unit 1.Table 1 details the accumulated PCT impact due to errors and changes in the ATRIUM-10 LOCA analyses since the References 1 and 2 analyses of record.Table 1: Cumulative Effect of PCT Changes -BFN, Unit I (ATRIUM-10)
Baseline PCT 1926 OF Thermal Conductivity Degradation (Reference
- 4) + 0 OF Accumulated changes since baseline analysis 0 OF New licensing PCT 1926 OF Absolute value of accumulated changes 0 OF Enclosure
-1 of 4 ENCLOSURE 10 CFR 50.46 30-DAY REPORT FOR BROWNS FERRY NUCLEAR PLANT, UNIT 1 GE14 Fuel Evaluation During the BFN, Unit 1, Fall 2012 refueling outage, modifications were completed that restored the automatic initiation capability of the BFN, Unit 1, Automatic Depressurization System. As a result, Reference 5 was re-established as the LOCA analysis of record for GE14 fuel, with a baseline PCT of 1760 OF. The applicability of this analysis to the as-modified plant configuration was confirmed by GE-Hitachi in Reference
- 6. Reference 5 provides PCT results for both Extended Power Uprate (EPU) and Current Licensed Thermal Power (CLTP) conditions.
The TVA has elected to use the CLTP results for 10 CFR 50.46 reporting, since EPU has not been approved for BFN, Unit 1, and all GE14 fuel is scheduled to be discharged from the reactor core prior to the planned EPU implementation date.In addition, GE-Hitachi has provided three 10 CFR 50.46 error reports that are applicable to the limiting BFN LOCA analysis for GE14 fuel.On July 20, 2011, GE-Hitachi issued 10 CFR 50.46 Notification Letter 2011-02 (Reference 7), which notified TVA of a database error that affected input coefficients used to direct the deposition of gamma radiation energy produced by fuel when determining whether this energy would heat the fuel rod, cladding, channel, or control rod structure materials.
The input caused the heat deposited in the fuel channel (post scram) to be over predicted and the corresponding heat to the fuel to be under predicted.
This effect was determined to be non-conservative.
The error only applies to 10x10 fuel and increased the PCT by 25 °F. As discussed in Reference 6, this error is applicable to the current BFN LOCA analysis for GE14 fuel.On July 20, 2011, GE-Hitachi also issued 10 CFR 50.46 Notification Letter 2011-03 (Reference 8), which notified TVA of an updated formulation for gamma heat deposition in the channel wall for 9x9 and 1 0x1 0 fuel assemblies.
An examination of the existing formulation revealed that the contribution of heat from gamma ray absorption by the channel was found to have been minimized.
The method had been simplified such that initially all the energy was assumed to be deposited in the fuel rods prior to the LOCA and then adjusted such that the correct heat deposition was applied after the scram. This modeling was concluded to be potentially non-conservative, as not accounting for this small fraction of total power generation outside the fuel rod would tend to suppress the hot bundle power required to meet the initial operating Average Planar Linear Heat Generation Rate limit. Further, there is a small effect on the initial conditions for the balance of the core, as these are set in relation to the hot bundle condition.
The energy distribution during the pre-scram phase was updated with the appropriate energy distribution.
Since the integral heat deposition is dominated by post-scram energy, the change has only a small impact on the results, increasing PCT by 15 OF. As discussed in Reference 6, this error is applicable to the current BFN LOCA analysis for GE14 fuel.On November 29, 2012, GE Hitachi issued 10 CFR 50.46 Notification Letter 2012-01 (Reference 9), which notified TVA of a change to the Emergency Core Cooling System (ECCS)evaluation model in response to NRC Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation," and addressed inaccuracies in fuel pellet thermal conductivity as a function of exposure.
The PRIME fuel rod thermal-mechanical code addresses these thermal conductivity concerns.
Reference 9 estimates the magnitude of the change in PCT due to the change in fuel Enclosure
-2 of 4 ENCLOSURE 10 CFR 50.46 30-DAY REPORT FOR BROWNS FERRY NUCLEAR PLANT, UNIT I properties in PRIME relative to the existing GESTR model used in Reference
- 5. The most dominant effect of the PRIME fuel properties is in thermal conductivity, which results in a higher fuel stored energy. The impact of PRIME is drawn from stored energy sensitivity results. For BFN, Unit 1, GE14 fuel, the PCT impact of modeling fuel rod mechanical properties with PRIME was determined to be zero degrees.Table 2 details the accumulated PCT impact due to errors and changes in the GE14 LOCA analyses since the Reference 5 analysis of record.Table 2: Cumulative Effect of PCT Changes -BFN, Unit I (GE14)Baseline PCT 1760 OF Input coefficient database error 25 OF Revised gamma heat deposition formulation 15 OF Pellet thermal conductivity degradation 0 OF Accumulated changes since baseline analysis 40 OF New licensing PCT 1800 OF Absolute value of accumulated changes 40 OF Enclosure
-3 of 4 ENCLOSURE 10 CFR 50.46 30-DAY REPORT FOR BROWNS FERRY NUCLEAR PLANT, UNIT I References
- 1. ANP-3015(P)
Revision 0, "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis," AREVA NP Inc., September 2011.2. ANP-3016(P)
Revision 0, "Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM-1 0 Fuel," AREVA NP Inc., December 2011.3. NRC Letter, "Browns Ferry Nuclear Plan, Unit 1 -Issuance of Amendments Regarding the Transition to AREVA Fuel (TAC No. ME3775) (TS-473)," April 27, 2012.4. FAB1 2-2249, "Transmittal of 10 CFR 50.46 PCT Error Reporting for Browns Ferry Units 1, 2, and 3," AREVA NP Inc., June 28, 2012.5. NEDC-32484P Revision 6, "Browns Ferry Nuclear Plant Units 1, 2, and 3: SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," GE Nuclear Energy, February 2005.6. NEDC-32484P Revision 6, Supplement 2 Revision 0, "Browns Ferry Nuclear Plant Unit 1: Supplementary Report Regarding ECCS-LOCA Evaluation Additional Single Failure Evaluation at Current Licensed Thermal Power," GE-Hitachi Nuclear Energy, September 2012.7. GE-Hitachi 10 CFR 50.46 Notification Letter 2011-02, July 20, 2011.8. GE-Hitachi 10 CFR 50.46 Notification Letter 2011-03, July 20, 2011.9. GE-Hitachi 10 CFR 50.46 Notification Letter 2012-01, November 29, 2012.Enclosure
-4 of 4