Letter Sequence Approval |
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MONTHYEARML0425702122004-09-13013 September 2004 RAI, Allowable Value for Reactor Vessel Water Level Project stage: RAI ML0500303612005-01-0606 January 2005 Ltr, Request for Information for Status of Amendment (Tac No. MC1330, MC1427, MC2305, MC3812, MC4070, MC4071, MC4072, MC4161, MC3743, MC3744) Project stage: RAI ML0505900362005-02-16016 February 2005 Enclosure 2, 02/01/2005 Conference Call with TVA Presentation Slides on TS Changes Using Method 3 for Interim Solution Project stage: Other ML0511501912005-04-19019 April 2005 Enclosure 3, Ltr, Licensing Action Status and Interdependencies Project stage: Other ML0514300362005-05-20020 May 2005 Response to Nrc'S Letter on Licensing Action Status and Interdependencies, Project stage: Other ML0520706042005-07-26026 July 2005 Notice of Meeting with Tennessee Valley Authority (TVA) BFN Units 1, 2, and 3, and Sequoyah Units 1 and 2 on Resolution of Instrument Setpoint Concerns Project stage: Meeting ML0522903972005-09-0202 September 2005 TVA Handout Method 3 Issue BFN Proposed Resolution Project stage: Other ML0522901442005-09-0202 September 2005 Summary of Meeting with the Tennessee Valley Authority Regarding Plant-Specific Resolution of Instrument Setpoint Concerns Project stage: Meeting ML0626300962006-09-18018 September 2006 Tech Spec Pages for Amendment 258, Regarding Allowable Value for Reactor Vessel Water Level (TS-434) Project stage: Other ML0621401112006-09-18018 September 2006 License Amendment 258, Regarding Allowable Value for Reactor Vessel Water Level (TS-434) Project stage: Approval 2005-04-19
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Category:Letter
MONTHYEARCNL-24-060, Supplement to Request for Approval of the Tennessee Valley Authority Nuclear Quality Assurance Program Description2024-09-24024 September 2024 Supplement to Request for Approval of the Tennessee Valley Authority Nuclear Quality Assurance Program Description ML24262A1502024-09-24024 September 2024 Requalification Program Inspection - Browns Ferry Nuclear Plant ML24262A0602024-09-23023 September 2024 Summary of August 19, 2024, Meeting with Tennessee Valley Authority Regarding a Proposed Supplement to the Tennessee Valley Authority Nuclear Quality Assurance Plan ML24263A2952024-09-19019 September 2024 Site Emergency Plan Implementing Procedure Revision CNL-24-065, Tennessee Valley Authority – Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-09-18018 September 2024 Tennessee Valley Authority – Central Emergency Control Center Emergency Plan Implementing Procedure Revisions IR 05000260/20240902024-09-17017 September 2024 NRC Inspection Report 05000260/2024090 and Preliminary White Finding and Apparent Violation - 1 CNL-24-062, Cycle 16 Reload Analysis Report2024-09-16016 September 2024 Cycle 16 Reload Analysis Report ML24255A8862024-09-10010 September 2024 Core Operating Limits Report for Cycle 16 Operation, Revision 0 ML24239A3332024-09-0303 September 2024 Full Audit Plan IR 05000259/20244042024-09-0303 September 2024 Cyber Security Inspection Report 05000259/2024404 and 05000260-2024404 and 05000296/2024404-Cover Letter IR 05000259/20240052024-08-26026 August 2024 Updated Inspection Plan for Browns Ferry Nuclear Plant, Units 1, 2 and 3 - Report 05000259/2024005, 05000260/2024005 and 05000296/2024005 ML24225A1682024-08-16016 August 2024 – Notification of Inspection and Request IR 05000259/20244022024-08-0606 August 2024 Security Baseline Inspection Report 05000259/2024402 and 05000260/2024402 and 05000296/2024402 ML24219A0272024-08-0606 August 2024 Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations IR 05000259/20240022024-08-0202 August 2024 Brown Ferry Nuclear Plant – Integrated Inspection Report05000259/2024002 and 05000260/2024002 and 05000296/2024002 ML24199A0012024-07-22022 July 2024 Clarification and Correction to Exemption from Requirement of 10 CFR 37.11(c)(2) ML24172A1342024-07-15015 July 2024 Exemptions from 10 CFR 37.11(C)(2) (EPID L-2023-LLE-0024) - Letter ML24183A4142024-07-10010 July 2024 – License Renewal Regulatory Limited Scope Audit Regarding the Environmental Review of the License Renewal Application (EPID Number: L-2024-SLE-0000) (Docket Numbers: 50-259, 50-260, and 50-296) ML24190A1292024-07-0808 July 2024 Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints ML24185A1512024-07-0303 July 2024 Inoperability of Unit 3 Diesel Generator Due to Relay Failure ML24184A1142024-07-0202 July 2024 Site Emergency Plan Implementing Procedure Revision ML24183A3842024-07-0101 July 2024 Registration of Use of Cask to Store Spent Fuel (MPC-364, -365) ML24179A0282024-06-26026 June 2024 Evaluation of Effects of Out-of-Limits Condition as Described in IWB-3720(a) ML24176A1022024-06-24024 June 2024 Reactor Scram Due to Generator Step-Up Transformer Failure ML24176A1132024-06-23023 June 2024 American Society of Mechanical Engineers, Section XI, Fourth 10 Year Inspection Interval, Inservice Inspection, System Pressure Test, Containment Inspection, and Repair and Replacement Programs, Owner’S Activity Report Cycle 21 Oper ML24175A0042024-06-23023 June 2024 Interim Report of a Deviation or Failure to Comply Associated with a Valve in the Unit 3 High Pressure Coolant Injection System ML24089A1152024-06-21021 June 2024 Transmittal Letter, Environmental Assessments and Findings of No Significant Impact Related to Exemption Requests from 10 CFR 37.11(c)(2) ML24155A0042024-06-18018 June 2024 Proposed Alternative to the Requirements of the ASME Code (Revised Alternative Request 0-ISI-47) ML24158A5312024-06-0606 June 2024 Registration of Use of Cask to Store Spent Fuel (MPC-361, -362, -363) ML24071A0292024-06-0505 June 2024 Subsequent License Renewal Application Enclosure 3 - Proprietary Determination Letter ML24068A2612024-06-0505 June 2024 SLRA Fluence Methodology Report - Proprietary Determination Letter IR 05000259/20244032024-05-22022 May 2024 – Security Baseline Report 05000259/2024403 and 05000260/2024403 and 05000296/2024403 ML24141A2462024-05-20020 May 2024 High Pressure Coolant Injection Inoperable Due to Rupture Disc Failure and Resulting System Isolation ML24141A0482024-05-17017 May 2024 EN 56958_1 Ametek Solidstate Controls, Inc ML24136A0702024-05-15015 May 2024 2023 Annual Radiological Environmental Operating Report IR 05000259/20240012024-05-14014 May 2024 Integrated Inspection Report 05000259/2024001, 05000260/2024001, and 05000296/2024001 CNL-24-040, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-05-0808 May 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML24123A2012024-05-0202 May 2024 NRC Cybersecurity Baseline Inspection (NRC Inspection Report 05000259/2024404, 05000260-2024404, 05000296/2024404) and Request for Information ML24122A6852024-05-0101 May 2024 2023 Annual Radioactive Effluent Release Report and Offsite Dose Calculation Manual CNL-24-036, – 10 CFR 50.46 Annual Report2024-04-25025 April 2024 – 10 CFR 50.46 Annual Report ML24116A2522024-04-25025 April 2024 Site Emergency Plan Implementing Procedure Revision ML24115A1652024-04-24024 April 2024 Breaker Trip Automatically Started an Emergency Diesel Generator CNL-24-037, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 422024-04-22022 April 2024 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 42 05000296/LER-2024-001, Primary Containment Isolation Valve Inoperable Due to Incorrect Motor Operated Valve Setup2024-04-22022 April 2024 Primary Containment Isolation Valve Inoperable Due to Incorrect Motor Operated Valve Setup ML24087A2302024-04-18018 April 2024 Exemption from Select Requirements of 10 CFR Part 73, Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting ML24108A1832024-04-17017 April 2024 Secondary Containment Isolation Valve Inoperable Due to Mechanical Failure CNL-24-033, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-04-17017 April 2024 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions CNL-24-024, Hydrologic Engineering Center River Analysis System Project Milestone Status Update2024-04-17017 April 2024 Hydrologic Engineering Center River Analysis System Project Milestone Status Update ML24106A2632024-04-16016 April 2024 Ltr to State Recognized Tribes-Hamilton, Cher-O-Creek Intra Tribal Indians ML24106A2572024-04-16016 April 2024 Ltr to Federally Recognized Tribes-Yahola, Kialegee Tribal Town 2024-09-03
[Table view] Category:License-Operating (New/Renewal/Amendments) DKT 50
MONTHYEARML24043A0462024-02-16016 February 2024 Summary of Meeting with TVA to Discuss Use of Risk-Informed Process Evaluation (RIPE) for a Proposed LAR Regarding Elimination of Limiting Condition for Operation Actions for Rod Worth Minimizers 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 ML23257A1232023-09-22022 September 2023 Administrative Changes to Technical Specification Pages Issued for License Amendment Nos. 332, 355, and 315 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 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 ML23116A2472023-05-23023 May 2023 Issuance of Amendment Nos. 330, 353, and 313 Regarding Adoption of TSTF-478, Revision 2 for Combustible Gas Control ML23101A1102023-05-16016 May 2023 Issuance of Amendment Nos. 329, 352, and 312 Regarding Removal of Site Acreage Description from Technical Specifications ML23073A2902023-05-0202 May 2023 Issuance of Amendment Nos. 328, 351, and 311 Adoption of TSTF-505, Revision 2, for Risk-Informed Completion Times and TSTF-439, Revision 2, to Eliminate Second Completion Times ML22348A0052023-01-25025 January 2023 Issuance of Amendment Nos. 326, 349, and 309; 363 and 35; 159 and 67 Regarding Adoption of TSTF-554, Revise Reactor Coolant Leakage Requirements ML22349A6472023-01-20020 January 2023 Issuance of Amendment Nos. 325, 348, and 308; 362 and 356; and 158 and 66 Regarding Adoption of TSTF-529, Rev. 4, Clarify Use and Application Rules ML22348A0662023-01-13013 January 2023 Issuance of Amendment Nos. 325, 348, & 308 Regarding Application of Advanced Framatome Methodologies, & Adoption of TSTF Traveler TSTF-564-A, Rev. 2, in Support of Atrium 11 Fuel Use (EPID L-2021-LLA-0132) - Nonproprietary ML22271A9142022-12-0707 December 2022 Issuance of Amendment Nos. 324, 347, and 307; 360 and 354; 157 and 65 Regarding a Revision to the Emergency Action Level Scheme ML22327A0182022-12-0101 December 2022 Correction of Error Incurred During Issuance of License Amendment No. 322 ML22273A1032022-11-22022 November 2022 Issuance of Amendment Nos. 323, 346, and 306 Regarding Chilled Water Cross-Tie Modification ML22220A2602022-11-21021 November 2022 Issuance of Amendment Nos. 322, 345, and 305 Regarding Adoption of TSTF Traveler TSTF-205-A, Rev. 3, and TSTF-563-A ML22138A3252022-06-24024 June 2022 Issuance of Amendment Nos. 321, 344, and 304 Regarding Spent Fuel Pool Criticality Safety Analysis ML22020A2282022-03-16016 March 2022 Issuance of Amendment Nos. 320, 343, and 303 Regarding the Adoption of Approved Technical Specification Task Force Traveler TSTF-568, Revision 2 ML21285A0682021-10-28028 October 2021 Issuance of Amendment Nos. 319, 342, and 302 Regarding the Adoption of Technical Specification Task Force Traveler TSTF-582, Revision 2 ML21214A1392021-08-30030 August 2021 Issuance of Amendment Nos. 318, 341, and 301 Regarding Changes to Technical Specification 3.8.6, Battery Cell Parameters ML21173A1772021-07-27027 July 2021 Issuance of Amendment Nos. 317, 340, and 300 Regarding Adoption of Title 10 of the Code of Federal Register Section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components ML21075A0762021-04-30030 April 2021 Issuance of Amendment Nos. 316, 339, and 299 Regarding the Incorporation of the Tornado Missile Risk Evaluator Into the Licensing Basis ML21041A4892021-04-0808 April 2021 Issuance of Amendment Nos. 315, 338, and 298 Regarding the Adoption of Technical Specifications Task Force Traveler, TSTF-425, Revision 3 ML20268A0822021-01-12012 January 2021 Issuance of Amendment Nos. 314, 337, and 297; 351 and 345; 140 and 46 Regarding Changes to the Technical Specifications ML20282A3452020-11-19019 November 2020 Issuance of Amendment Nos. 313, 336, 296, 350, 344, 138, and 44 Revise Emergency Plan On-Shift Emergency Medical Technician & Onsite Ambulance Requirements ML20190A1052020-08-11011 August 2020 Correction to Amendment No. 319 Regarding Revisions to Technical Specification 3.3.6.1, Primary Containment Isolation Instrumentation ML20085G8962020-06-26026 June 2020 Issuance of Amendment Nos. 312, 335, and 295 Regarding Request to Revise Emergency Plan Staff Augmentation Times ML19294A0112019-12-26026 December 2019 Issuance of Amendment Nos. 311, 334, and 294 Adopt Technical Specifications Task Force Traveler, TSTF-542, Revision 2 ML18277A1102019-08-27027 August 2019 Units, 1 & 2 Issuance of Amendment Nos. 309, 332, 292, 345, 339, 128, and 31 Regarding Unbalanced Voltage Protection ML19198A0012019-08-13013 August 2019 Issuance of Amendment Nos. 308, 331, and 291 to Extend Implementation Due Date for Modifications 102 and 106 Related to NFPA 805 (Eid L-2019-LLA-0140) ML19037A1372019-04-0202 April 2019 Revisions to Modifications 85, 102 and 106 Related to National Fire Protection Association 805 Performance-based Standard for Fire Protection of Light Water Reactor Electric Generating Plants ML18241A3192018-10-0909 October 2018 Issuance of Amendment Nos. 306, 329, and 289 to Revise Approved NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants-Revision ML18251A0032018-09-27027 September 2018 Issuance of Amendment Nos. 305, 328, and 288 to Revise Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program (CAC Nos. MG0113, MG0114, and MG0115; EPID L-2017-LLA-0292) ML18138A4522018-05-29029 May 2018 Bf, Units 1, 2, and 3; Sequoyah, Units 1 and 2; Watts Bar, Units 1 and 2 - Correction to an Omitted Reference for License Amendment Regarding Request to Upgrade (CAC Nos. MF9054, MF9055, MF9056, MF9057, MF9058, MF9059, and MF9060, EPID L-20 ML17289A0322017-12-22022 December 2017 Issuance of Amendments Regarding Request to Upgrade Emergency Action Level Scheme (CAC Nos. MF9054-60; EPID L2017-LLA-0160) ML17215A2432017-10-0202 October 2017 Browns Ferry Nuclear Plant, Units 1, 2, and 3; Watts Bar Nuclear Plant, Units 1 and 2 - Issuance of Amendments to Change Technical Specifications to Adopt Technical Specifications Task Force Traveler-522 (CAC No. MF9562-MF9566) ML17032A1202017-08-14014 August 2017 Non-Proprietary Issuance of Amendments Regarding Extended Power Uprate ML17034A3602017-03-27027 March 2017 Issuance of Amendment Nos. 298, 322, 282, and 338 and 331 - Revise Technical Specification 5.3, Unit Staff Qualifications to Replace References to Rg 1.8, Rev. 2 with TVA Nuclear Quality ... ML17052A1362017-03-16016 March 2017 Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Issuance of Amendment Nos. 297, 321, and 281 to Change the Completion Date of Cyber Security Plan Implementation Milestone 8 (CAC Nos. MF8125, MF8126, and MF8127) ML16330A1582017-01-17017 January 2017 Issuance of Amendments Regarding Revisions to Technical Specification 4.3.1.2, Fuel Storage Criticality ML16028A4142016-04-26026 April 2016 Issuance of Amendment to Revise Technical Specifications Related to Cycle 18 Safety Limit Minimum Critical Power Ratio ML15317A4782016-02-0909 February 2016 Issuance of Amendment to Revise Technical Specifications Related to Cycle 18 Safety Limit Minimum Critical Power Ratio ML15344A3212016-01-0707 January 2016 Issuance of Amendment Regarding Modification of Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits ML15321A4722015-12-23023 December 2015 Issuance of Amendments Regarding Revision to Table 3.3.6.1-1, Primary Containment Isolation Instrumentation ML15287A2132015-12-16016 December 2015 Issuance of Amendments Regarding Technical Specification Changes to Reactor Core Safety Limits ML15324A2472015-12-14014 December 2015 Issuance of Amendment to Adopt TSTF-501, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Valves to Licensee Control ML15287A3712015-12-0404 December 2015 Issuance of Amendments for the Adoption of Technical Specifications Task Force Standard Technical Specifications Change Traveler TSTF-535 (CNL-15-029) ML15212A7962015-10-28028 October 2015 Issuance of Amendments to Transition to Fire Protection Program NFPA-805 ML15251A5402015-09-29029 September 2015 Issuance of Amendment Regarding Control Rod Scram Time Testing Frequency Per TSTF-460, Revision 0 ML15065A0492015-06-0202 June 2015 Issuance of Amendment Revising Pressure and Temperature Limit Curves ML13254A0812013-09-13013 September 2013 Issuance of Amendment Under Exigent Circumstances to Remove Notes from Technical Specifications 3.4.9-1 and 3.4.9-2 2024-02-16
[Table view] Category:Safety Evaluation
MONTHYEARML23319A1992024-01-0303 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 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 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 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 ML23116A2472023-05-23023 May 2023 Issuance of Amendment Nos. 330, 353, and 313 Regarding Adoption of TSTF-478, Revision 2 for Combustible Gas Control ML23101A1102023-05-16016 May 2023 Issuance of Amendment Nos. 329, 352, and 312 Regarding Removal of Site Acreage Description from Technical Specifications ML23073A2902023-05-0202 May 2023 Issuance of Amendment Nos. 328, 351, and 311 Adoption of TSTF-505, Revision 2, for Risk-Informed Completion Times and TSTF-439, Revision 2, to Eliminate Second Completion Times ML23054A2902023-03-13013 March 2023 Request for Relief from the Requirements of the ASME Boiler and Pressure Vessel Code Regarding Weld Examination Coverage ML23048A3042023-03-0808 March 2023 Tennessee Valley Authority - Request for Relief from Requirements of ASME Boiler and Pressure Vessel Code Regarding Weld Examination Coverage (EPID L-2022-LLR-0045,-0046,-0047) ML22348A0052023-01-25025 January 2023 Issuance of Amendment Nos. 326, 349, and 309; 363 and 35; 159 and 67 Regarding Adoption of TSTF-554, Revise Reactor Coolant Leakage Requirements ML22349A6472023-01-20020 January 2023 Issuance of Amendment Nos. 325, 348, and 308; 362 and 356; and 158 and 66 Regarding Adoption of TSTF-529, Rev. 4, Clarify Use and Application Rules ML22348A0662023-01-13013 January 2023 Issuance of Amendment Nos. 325, 348, & 308 Regarding Application of Advanced Framatome Methodologies, & Adoption of TSTF Traveler TSTF-564-A, Rev. 2, in Support of Atrium 11 Fuel Use (EPID L-2021-LLA-0132) - Nonproprietary ML22271A9142022-12-0707 December 2022 Issuance of Amendment Nos. 324, 347, and 307; 360 and 354; 157 and 65 Regarding a Revision to the Emergency Action Level Scheme ML22273A1032022-11-22022 November 2022 Issuance of Amendment Nos. 323, 346, and 306 Regarding Chilled Water Cross-Tie Modification ML22220A2602022-11-21021 November 2022 Issuance of Amendment Nos. 322, 345, and 305 Regarding Adoption of TSTF Traveler TSTF-205-A, Rev. 3, and TSTF-563-A ML22298A2852022-11-14014 November 2022 Request for Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, Paragraph IWB-2420(B) and Use of Code Case N-526 ML22138A3252022-06-24024 June 2022 Issuance of Amendment Nos. 321, 344, and 304 Regarding Spent Fuel Pool Criticality Safety Analysis ML22084A0012022-04-0505 April 2022 Clinch River Nuclear Site; Sequoyah Nuclear Plant, Units 1 and 2; Watts Bar Nuclear Plant, Units 1 and 2, Review of Quality Assurance Plan Changes ML22020A2282022-03-16016 March 2022 Issuance of Amendment Nos. 320, 343, and 303 Regarding the Adoption of Approved Technical Specification Task Force Traveler TSTF-568, Revision 2 ML22010A1952022-01-12012 January 2022 Request to Use Later Edition of the American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants ML21285A0682021-10-28028 October 2021 Issuance of Amendment Nos. 319, 342, and 302 Regarding the Adoption of Technical Specification Task Force Traveler TSTF-582, Revision 2 ML21214A1392021-08-30030 August 2021 Issuance of Amendment Nos. 318, 341, and 301 Regarding Changes to Technical Specification 3.8.6, Battery Cell Parameters ML21173A1772021-07-27027 July 2021 Issuance of Amendment Nos. 317, 340, and 300 Regarding Adoption of Title 10 of the Code of Federal Register Section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components ML21075A0762021-04-30030 April 2021 Issuance of Amendment Nos. 316, 339, and 299 Regarding the Incorporation of the Tornado Missile Risk Evaluator Into the Licensing Basis ML21041A4892021-04-0808 April 2021 Issuance of Amendment Nos. 315, 338, and 298 Regarding the Adoption of Technical Specifications Task Force Traveler, TSTF-425, Revision 3 ML20268A0822021-01-12012 January 2021 Issuance of Amendment Nos. 314, 337, and 297; 351 and 345; 140 and 46 Regarding Changes to the Technical Specifications ML20282A3452020-11-19019 November 2020 Issuance of Amendment Nos. 313, 336, 296, 350, 344, 138, and 44 Revise Emergency Plan On-Shift Emergency Medical Technician & Onsite Ambulance Requirements ML20253A1812020-09-23023 September 2020 Proposed Alternative to the Requirements of the American Society of Mechanical Engineers Code ML20085G8962020-06-26026 June 2020 Issuance of Amendment Nos. 312, 335, and 295 Regarding Request to Revise Emergency Plan Staff Augmentation Times ML19329E3192020-01-0202 January 2020 SE Implementation of Hardened Containment Vents Capable of Operation Under Severe Accident Conditions Related to Order EA-13-109 (CAC Nos. MF4540, MF4541 and MF4542; EPID No. L-2014-JLD-0044) ML19294A0112019-12-26026 December 2019 Issuance of Amendment Nos. 311, 334, and 294 Adopt Technical Specifications Task Force Traveler, TSTF-542, Revision 2 ML18277A1102019-08-27027 August 2019 Units, 1 & 2 Issuance of Amendment Nos. 309, 332, 292, 345, 339, 128, and 31 Regarding Unbalanced Voltage Protection ML19198A0012019-08-13013 August 2019 Issuance of Amendment Nos. 308, 331, and 291 to Extend Implementation Due Date for Modifications 102 and 106 Related to NFPA 805 (Eid L-2019-LLA-0140) ML19037A1372019-04-0202 April 2019 Revisions to Modifications 85, 102 and 106 Related to National Fire Protection Association 805 Performance-based Standard for Fire Protection of Light Water Reactor Electric Generating Plants ML19010A2742019-03-18018 March 2019 Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report- TVA Overall Basin Probable Maximum Precipitation and Local Intense Precipitation, Calculation, CDQ0000002016000041 Tennessee Valley Authority ML19016A2152019-02-28028 February 2019 Relief from the Requirements of American Association of Mechanical Engineers Code Section XI Inservice Inspection Program, Request for an Alternative ISI-46 ML18323A1722018-12-31031 December 2018 Relief Request No. 1 ISI 29 Regarding Second 10-Year Inservice Inspection Interval Regarding Examination Coverage for Certain Pressure Retaining Welds ML18323A0262018-12-0404 December 2018 Relief Request No. 1-ISI-28 Regarding Third 10-Year Inservice Inspection Interval Regarding Examination Coverage for Certain Pressure Retaining Reactor Vessel Welds ML18241A3192018-10-0909 October 2018 Issuance of Amendment Nos. 306, 329, and 289 to Revise Approved NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants-Revision ML18251A0032018-09-27027 September 2018 Issuance of Amendment Nos. 305, 328, and 288 to Revise Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program (CAC Nos. MG0113, MG0114, and MG0115; EPID L-2017-LLA-0292) ML18236A3312018-09-24024 September 2018 Safety Evaluation Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation Related to Orders EA-12-049 and EA-12-051 ML18197A3072018-08-30030 August 2018 Issuance of Amendments Regarding Request to Change Technical Specification 3.3.1 and Surveillance Requirement 3.2.4 ML18171A3372018-07-10010 July 2018 Issuance of Amendment to Revise License Condition 2.C(18)(a)3 ML18138A4522018-05-29029 May 2018 Bf, Units 1, 2, and 3; Sequoyah, Units 1 and 2; Watts Bar, Units 1 and 2 - Correction to an Omitted Reference for License Amendment Regarding Request to Upgrade (CAC Nos. MF9054, MF9055, MF9056, MF9057, MF9058, MF9059, and MF9060, EPID L-20 ML17289A0322017-12-22022 December 2017 Issuance of Amendments Regarding Request to Upgrade Emergency Action Level Scheme (CAC Nos. MF9054-60; EPID L2017-LLA-0160) ML17261A0362017-10-0606 October 2017 Request for Request 3-ISI-28 Regarding Third 10-Year Inservice Inspection Interval Examination Coverage of Inside Radius and Nozzle-to-Vessel Welds ML17215A2432017-10-0202 October 2017 Browns Ferry Nuclear Plant, Units 1, 2, and 3; Watts Bar Nuclear Plant, Units 1 and 2 - Issuance of Amendments to Change Technical Specifications to Adopt Technical Specifications Task Force Traveler-522 (CAC No. MF9562-MF9566) ML17032A1202017-08-14014 August 2017 Non-Proprietary Issuance of Amendments Regarding Extended Power Uprate ML17145A5522017-08-11011 August 2017 Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Alternative Request IST-RR-1 for the Fourth 10-Year Inservice Testing Interval (CAC Nos. MF9087, MF9088, and MF9089) ML17034A3602017-03-27027 March 2017 Issuance of Amendment Nos. 298, 322, 282, and 338 and 331 - Revise Technical Specification 5.3, Unit Staff Qualifications to Replace References to Rg 1.8, Rev. 2 with TVA Nuclear Quality ... 2024-01-03
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September 18, 2006Mr. Karl W. SingerChief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801
SUBJECT:
BROWNS FERRY NUCLEAR PLANT, UNIT 1 - ISSUANCE OF AMENDMENTREGARDING ALLOWABLE VALUE FOR REACTOR VESSEL WATER LEVEL (TAC NO. MC2305) (TS-434)
Dear Mr. Singer:
The Commission has issued the enclosed Amendment No. 258 to Renewed Facility OperatingLicense No. DPR-33 for the Browns Ferry Nuclear Plant, Unit 1. This amendment is in response to your application dated March 9, 2004, as supplemented by letters dated November 15, 2004, and March 7, 2006. The amendment reduces the Allowable Value usedfor Reactor Vessel Water Level - Low, Level 3, for several instrument functions.A copy of the Safety Evaluation is also enclosed. Notice of Issuance will be included in theCommission's biweekly Federal Register notice.Sincerely,/RA/
Margaret H. Chernoff, Project ManagerPlant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-259
Enclosures:
- 1. Amendment No. 258 to DPR-33
- 2. Safety Evaluationcc w/encls: See next page
ML062140111* "NLO" [No Legal Objection] wW/comments NRR-058OFFICE LPL2-2/PMLPL2-2/PMLPL2-2/LASBWB/BCEICB/BCOGC *LPL2-2/BCNAMECPatelMChernoffBClaytonGCranstonAHoweDRothLRaghavanDATE 9/8/069/11/069/8/068/14/068/10/069/7/069/11/06 TENNESSEE VALLEY AUTHORITYDOCKET NO. 50-259BROWNS FERRY NUCLEAR PLANT, UNIT 1AMENDMENT TO RENEWED FACILITY OPERATING LICENSEAmendment No. 258 Renewed License No. DPR-331.The Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by Tennessee Valley Authority (the licensee)dated March 9, 2004, as supplemented by letters dated November 15, 2004, andMarch 7, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by thisamendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with theCommission's regulations;D.The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public; andE.The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied. 2.Accordingly, the license is amended by changes to the Technical Specifications asindicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-33 is hereby amended to read as follows:(2)Technical SpecificationsThe Technical Specifications contained in Appendices A and B, as revised throughAmendment No. 258, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.3.This license amendment is effective as of its date of issuance and shall be implementedwithin 60 days from the date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION L. Raghavan, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical SpecificationsDate of Issuance: September 18, 2006 ATTACHMENT TO LICENSE AMENDMENT NO. 258RENEWED FACILITY OPERATING LICENSE NO. DPR-33DOCKET NO. 50-259Replace Page 3 of Renewed Operating License DPR-33 with the attached Page 3.Replace the following pages of the Appendix A Technical Specifications with the attached revisedpages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. REMOVE INSERT3.3-73.3-7 3.3-463.3-46 3.3-473.3-47 3.3-583.3-58 3.3-603.3-60 3.3-643.3-64 3.3-693.3-69 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 258TO RENEWED FACILITY OPERATING LICENSE NO. DPR-33TENNESSEE VALLEY AUTHORITYBROWNS FERRY NUCLEAR PLANT, UNIT 1DOCKET NO. 50-25
91.0 INTRODUCTION
By letter dated March 9, 2004, as supplemented by letters dated November 15, 2004, andMarch 7, 2006, the Tennessee Valley Authority (the licensee) requested an amendment toRenewed Facility Operating License No. DPR-33 for the Browns Ferry Nuclear Plant (BFN),Unit 1. The proposed amendment would reduce the Allowable Value (AV) specified for Reactor Vessel Water Level - Low, Level 3, for several instrument functions, in the Technical Specifications (TSs). The supplements dated November 15, 2004, and March 7, 2006, provided additional informationthat clarified the application, did not expand the scope of the application as originally noticed, anddid not change the staff's original proposed no significant hazards consideration determination aspublished in the Federal Register on April 13, 2004 (69 FR 19575).
2.0 REGULATORY EVALUATION
The proposed TS changes will lower the current Reactor Vessel Water Level - Low, Level 3 AV forseveral instrumentation functions. The affected functions are reactor protection system (RPS)actuation, Emergency Core Cooling Systems (ECCS) (including automatic depressurization system reactor vessel level confirmatory signal), primary containment isolation (including reactorwater cleanup (RWCU) system and shutdown cooling system isolation), secondary containmentisolation, and control room emergency ventilation system (CREVS) initiation. The regulationsapplicable to these instrumentation functions include Title 10 to the Code of Federal Regulations(10 CFR), Section 50.36, Technical Specifications; Section 50.46, Acceptance Criteria for ECCSsfor Light-Water Nuclear Power Reactors, and Appendix K, ECCS Evaluation Models; Appendix A, General Design Criterion 19; Part 20, Standards for protection against radiation; and Part 100,Reactor site criteria.
3.0 TECHNICAL EVALUATION
3.1 BackgroundDuring
operation, significant changes in reactor vessel water level can occur due to pressuretransients that cause shrinking or swelling of the steam within the coolant system, or due toexcessive rates of addition or removal of coolant from the reactor vessel, such as might result from a feedwater pump trip. There is a 23-inch difference in elevation between the normal reactor water level (561 inches) and the current reactor trip (scram) initiation level (Level 3, 538 inches).
Process control systems are designed so the reactor can automatically recover from manytransients, such as a trip of a feedwater system pump, which might cause a significant change inthe water level. However, in some cases, with this narrow water level range, reactor scrams may result that would have been avoidable if plant control systems or operators had slightly more timeto take control. In addition to tripping the reactor, a drop in vessel level to Level 3 initiates primary and secondary containment isolations, CREVS operation and arming of the Automatic Depressurization System (ADS). The proposed change will provide additional operating range between the normal reactor vesselwater level and the level used as the setpoint for initiation of the above functions. The increased range should provide additional time for operators or plant control systems to automaticallyrespond to recoverable transients such as feedwater system malfunctions and, thus, may avertunnecessary reactor scrams. This change should similarly reduce the likelihood of initiation of the other aforementioned system actuations, without increasi ng the consequences of events that relyupon these functions. 3.2 Description of ChangeThe proposed TS changes will lower the current Reactor Vessel Water Level - Low, Level 3 AVfrom 538 inches above vessel zero to 528 inches above vessel zero for the following functions:- Table 3.3.1.1-1, Reactor Protection System Instrumentation, Function 4, PrimaryContainment Isolation - Reactor Vessel Water Level - Low, Level 3;- Table 3.3.6.1-1, Primary Containment Isolation Instrumentation, Function 2.a, PrimaryContainment Isolation Instrumentation - Reactor Vessel Water Level - Low, Level 3,- Table 3.3.6.1-1, Primary Containment Isolation Instrumentation, Function 5.h, ReactorWater Cleanup (RWCU) System Isolation - Reactor Vessel Water Level - Low, Level 3,- Table 3.3.6.1-1, Primary Containment Isolation Instrumentation, Function 6.b, ShutdownCooling System Isolation, Reactor Vessel Water Level - Low, Level 3,- Table 3.3.6.2-1, Secondary Containment Isolation Instrumentation, Function 1, ReactorVessel Water Level - Low, Level 3,- Table 3.3.7.1-1, Control Room Emergency Ventilation System Instrumentation, Function 1,Reactor Vessel Water Level - Low, Level 3.
Additionally, the proposed TS changes will lower the current Reactor Vessel Water Level - Low,Level 3 AV from 544 inches above vessel zero to 528 inches above vessel zero for the followingfunctions:- Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation, Function 4.d, ADS TripSystem A - Reactor Vessel Water Level - Low, Level 3 (Confirmatory),- Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation, Function 5.d, ADS TripSystem B - Reactor Vessel Water Level - Low, Level 3 (Confirmatory).The proposed changes to Reactor vessel water level - Low, Level 3 instrumentation setpoint AVswere calculated for each function using staff approved setpoint methodology. The staff's review of this methodology and the instrument operability determination was documented in a SafetyEvaluation dated September 14, 2006.For each analyzed accident/event, the effect of the change in initiation of these protective safetyfunctions is discussed below:
3.3 Evaluation
of Changes3.3.1 RPS Actuation
To determine the effects of the change in reactor vessel water level - low, Level 3 scram AV, the licensee evaluated (a) abnormal operational occurrences (AOO), (b) loss-of-coolant accidents (LOCA), (c) anticipated transients without scram (ATWS), (d) Appendix R events (fires), and (e) events involving potential radiological releases. The results of the licensee's analysis and U.S. Nuclear Regulatory Commission (NRC) staff review are summarized below.
3.3.1.2 Abnormal Operational Occurrences
The licensee utilized a screening process to examine the AOO in the Updated Final SafetyAnalysis Report (UFSAR) for BFN Unit 1 to determine if a Level 3 RPS actuation is credited for mitigation of the event. The licensee found that a Total Loss of Feedwater event is the only AOOfor which a Level 3 water level initiated scram occurs. For this event, a Reactor Core Isolation Cooling and High Pressure Core Injection systems initiation subsequently occurs at a lower level(Level 2) and adequately maintains core coverage. Thus, the licensee concluded that no unacceptable safety consequences would occur for any AOO if the Level 3 AV setpoint is reduced.However, in a letter dated August 16, 2004, General Electric Company (GE) submitted a lettercontaining a Part 21 60-Day Interim Notification: Narrow Range Water Level Instrument Level 3 Trip (ADAMS Accession No. ML042720291). In the letter, GE stated that a conservative evaluation by GE Nuclear Energy has determined that water level instruments may indicate high by as much as 8 inches should the reactor water level drop below the dryer seal skirt. As part of a review of a separate license amendment request by the licensee, the NRC staff requested TVA todescribe TVA's evaluation of the safety concern and explain, in detail, how the issue was resolved for BFN Unit 1.In a letter dated March 7, 2006 (ADAMS Accession No. ML060720248), TVA responded to theNRC staff's request. TVA confirmed that during a loss-of-feedwater transient event, the dryer skirt will be exposed. TVA calculated the combined potential error to be 8.11 inches. TVA confirmedthat the instrument would continue to function with sufficient margin from the analytical limit andthe trip setting, without affecting safety margin.
Based on the review of the information provided by licensee, the staff determined that the licenseehas adequately addressed the concerns identified in the GE letter described above and the change in the Level 3 AV setpoint for RPS will not result in significant impact on the AOO for BFNUnit 1.
3.3.1.3 LOCA
The licensee's analysis determined that for a large break LOCA, a reactor scram is initiated by high drywell pressure prior to the time that reactor vessel level decreases to Level 3 setpoint. Due to this action, the licensee concluded that the change in reactor vessel Level 3 trip would have no effect on large break LOCA consequences. For a small break LOCA, the licensee concluded that the reduced Analytical Water Level limit at the time of scram initiation slightly decreases the calculated peak cladding temperature (PCT). The reduction in the PCT is related to the earlier initiation of ADS on low level 1 signal due to the lower initial water level.
The licensee's analysis also encompassed a review of the potential effects on containment dynamic loads, safety/relief valve discharge loads, and suppression pool response for a design-basis LOCA. The analysis indicates that because a scram would be initiated as a result of high drywell pressure, prior to Level 3, the Level 3 AV setpoint change would have no effect on these responses.Based on the review of the information provided by the licensee, the NRC staff agrees with thelicensee's conclusion above.3.3.1.4 ATWSThe licensee stated that in an ATWS scenario, no automatic or manual scram occurs. Thus, thechange in Reactor Protection System Level 3 initiation has no effect. The NRC staff agrees withthe licensee's assessment of ATWS event.3.3.1.5 Fire The licensee stated that for Appendix R (fire) events the reactor is manually scrammed and, thus,the Level 3 setpoint has no effect on the consequences. The NRC staff agrees.
3.3.1.6 Radiological Release
The licensee stated that the limiting pipe break for radiological releases outside containment is thedesign-basis main steamline break outside the containment. The licensee's analysis indicates that for the main steamline break event, a scram occurs due to the high steamline flow protective function and, thus, the change in the low water level function will not affect the consequences.
Also, the licensee indicated that the limiting pipe break for radiological releases inside the containment is the design-basis LOCA. The design-basis LOCA assumes that the reactor scramoccurs at time zero due to high drywell pressure with a normal reactor water level. Therefore, reducing the Level 3 RPS AV has no impact on the radiological release analyses inside thecontainment for the design-basis LOCA analyses.
Based on the review of the information provided by the licensee, the NRC staff agrees with thelicensee's conclusion about the impact of the reduced Level 3 AV on potential radioactive release. 3.3.1.7 Conclusion for RPS Level 3 reductionBased on the discussion in previous sections, the NRC staff concludes that the reduced Level 3RPS AV will not have any significant impact on the plant operation. The NRC staff also concl udesthat BFN Unit 1 continues to meet the requirements of 10 CFR 50.36, 10 CFR 50.46, andAppendix K. Therefore, the proposed change is acceptable. 3.3.2 Primary Containment Isolation Including Shutdown Cooling and RWCU System Isolation A protective feature of the BFN Unit 1 is isolation of the primary containment penetrations ifreactor vessel level drops to Level 3 setpoint. This function assures that onsite and offsite dose limits established by 10 CFR Part 20 and 10 CFR Part 100 are not exceeded.The licensee stated that significant radiation releases cannot occur until after the core isuncovered and, with the reduced Level 3 setpoint, containment isolation will still occur well beforecore uncovery; thus, the small delay in primary containment isolation will not affect the ability of the containment isolation valves to perform their intended functions. Also, the licensee stated, thatfor LOCA events inside containment, a high drywell pressure signal will also initiate primarycontainment isolation for all systems affected by Level 3 signal (except RWCU) prior to Level 3water level trip.
The residual heat removal system (RHRS) Primary Containment Isolation function is also requiredto be operable during shutdown cooling operations. During shutdown cooling operations a Level 3 condition will initiate closure of the shutdown cooling isolation valves. This prevents any furtherloss of coolant inventory via the RHRS if RHRS leakage is the reason for the reduction in vessellevel. The licensee stated that the reduction of Level 3 AV will not affect the intended function ofisolation valves since the system will still isolate at a water level far above the top of the core.Therefore, reducing the Level 3 AV has no impact on the ability of the shutdown cooling modeisolation to perform its intended functions.Another primary containment isolation function that occurs on a Low Level 3 signal is isolation ofthe RWCU System. This signal is one of several that initiate an RWCU isolation in the event of loss of reactor coolant due to an RWCU line break. The Level 3 RWCU isolation is not directly analyzed in the UFSAR because the RWCU system line break is bounded by breaks of largersystems (Design-basis LOCA and main steamline break outside the containment).
The licenseestated that the reduced Level 3 setpoint would not impact the capability of the RWCU isolationvalves to perform their intended function. Also, in the event of an RWCU line break, the RWCU system may be isolated earlier as the result of other RWCU system leakage detection functions.Based on the review of the information provided by licensee, the NRC staff agrees that theprimary containment isolation function, including the isolation function for RHRS cold shutdown mode and RWCU isolation function will not have adverse impact from reduced Level 3 AV setpointand associated Part 20 and Part 100 limits will not be exceeded. Therefore, the proposed changefor reduced Level 3 AV setpoints for these systems is acceptable.
3.3.3 Secondary
Containment Isolation and Standby Gas Treatment System (SGTS)
The primary containment system is enclosed by a secondary containment system which, in theevent of an accident, confines gaseous primary containment leakage. This leakage is exhausted from the secondary containment enclosure by an SGTS and discharged to an elevated release point. Like primary containment isolation, operation of the secondary containment system is alsoinitiated upon a vessel Low Level 3 condition.The LOCA provides the most severe potential radiological release to the primary and secondarycontainment and, thus, serves as the bounding design-basis accident in determining the post-accident offsite dose. For LOCA events, the secondary containment and SGTS will actuateon high drywell pressure prior to reaching the Level 3 water level trip. Therefore, a reduced Level 3 AV would have no effect on the LOCA event analysis. For other loss of inventory events, the Level 3 actuation will occur well before any core uncovery, and potential radiological release. Therefore, small change in Level 3 actuation will not affect the secondary containment and SGTS performance. Based on the above discussion, the proposed change in Level 3 AV is acceptable.
3.3.4 CREVS
Actuation The CREVS is designed to provide a radiologically-controlled environment to ensure the habitability of the control room for all plant conditions. In the event of a Level 3 signal, the CREVSis automatically initiated to pressurize the control room with filtered air to minimize the radiological doses to control room personnel. The LOCA provides the most severe potential radiological release to the primary and secondary containment and, thus, serves as the bounding design-basisaccident in determining the control room dose, which must not exceed the criteria of GeneralDesign Criterion 19. For LOCA events, the CREVS will actuate on high drywell pressure prior toreaching the Level 3 water level trip. Therefore, a reduced Level 3 AV would have no effect on the LOCA event analysis. For other loss of inventory events, the Level 3 actuation will occur wellbefore any core uncovery, which could result in potential radiological release. Therefore, a small change in Level 3 actuation will not affect the ability of the CREV system to perform its int endedfunction. Based on the above discussion, the proposed change in Level 3 AV is acceptable.
3.3.5 Automatic
Depressurization System
The proposed TS change lowers ADS confirmatory signal Level 3 AV from 544 inches to 528 inches to maintain consistency with the other Level 3 trip functions. This Level 3 signal is a confirmatory low water level signal for ADS initiation, which serves to prevent unnecessary ADS initiation resulting from spurious Level 1 water level actuations or as a result of a break in the Level 1 instrument line. The intended function of this confirmatory signal will still be successfullyaccomplished even if the Level 3 signal is reduced, since the Level 3 signal will occur well prior toLevel 1. Therefore, reducing the Level 3 AV will not affect the ability of ADS to perform itsintended function and the proposed change is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Alabama State official was notified of theproposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility com ponentlocated within the restricted area as defined in 10 CFR Part 20. The NRC staff has determinedthat the amendment involves no significant increase in the amounts, and no significant change inthe types, of any effluents that may be released offsite, and that there is no significant increase inindividual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and therehas been no public comment on such finding (69 FR 19575). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR51.22(b) no environmental impact statement or environmental assessment need be prepared inconnection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there isreasonable assurance that the health and safety of the public will not be endangered by operationin the proposed manner, (2) such activities will be conducted in compliance with the Commission'sregulations, and (3) the issuance of the amendment will not be inimical to the common defenseand security or to the health and safety of the public.Principal Contributor: Chandu Patel Date: September 18, 2006 Mr. Karl W. SingerBROWNS FERRY NUCLEAR PLANTTennessee Valley Authority
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Mr. Ashok S. Bhatnagar, Senior Vice President Nuclear Operations Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Larry S. Bryant, Vice PresidentNuclear Engineering & Technical Services Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801Brian O'Grady, Site Vice PresidentBrowns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609Mr. Robert J. Beecken, Vice PresidentNuclear Support Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 General CounselTennessee Valley Authority ET 11A 400 West Summit Hill DriveKnoxville, TN 37902Mr. John C. Fornicola, ManagerNuclear Assurance and Licensing Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801Mr. Bruce Aukland, Plant ManagerBrowns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609Mr. Masoud Bajestani, Vice PresidentBrowns Ferry Unit 1 Restart Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609Mr. Robert G. Jones, General ManagerBrowns Ferry Site Operations Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609Mr. Larry S. MellenBrowns Ferry Unit 1 Project Engineer Division of Reactor Projects, Branch 6 U.S. Nuclear Regulatory Commission 61 Forsyth Street, SW.
Suite 23T85 Atlanta, GA 30303-8931 Mr. Glenn W. Morris, Manager Corporate Nuclear Licensing and Industry Affairs Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801Mr. William D. Crouch, M anagerLicensing and Industry Affairs Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609Senior Resident InspectorU.S. Nuclear Regulatory Commission Browns Ferry Nuclear Plant 10833 Shaw Road Athens, AL 35611-6970State Health OfficerAlabama Dept. of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, AL 36130-3017ChairmanLimestone County Commission 310 West Washington Street Athens, AL 35611