Attachment 9, Browns Ferry Unit 1 - Technical Specification Change 467, ANP-2807(NP), Revision 0, Thermal Hydraulic Design Report for ATRIUM-10 Fuel AssembliesML093080160 |
Person / Time |
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Site: |
Browns Ferry ![Tennessee Valley Authority icon.png](/w/images/c/ce/Tennessee_Valley_Authority_icon.png) |
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Issue date: |
06/30/2009 |
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From: |
AREVA NP |
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To: |
Office of Nuclear Reactor Regulation |
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References |
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L32 090715 803, TS 467 ANP-2807(NP), Rev 0 |
Download: ML093080160 (28) |
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Category:Report
MONTHYEARML23192A4472023-07-31031 July 2023 Staff Assessment of Updated Seismic Hazards at TVA Sites Following the NRC Process for the Ongoing Assessment of Natural Hazards Information ML23025A0752023-01-25025 January 2023 American Society of Mechanical Engineers, Section XI, Third 10 Year Inspection Interval, Inservice Inspection, System Pressure Test, Containment Inspection, and Repair and Replacement Programs, Owners Activity Report Cycle . ML22363A3922022-12-28028 December 2022 Cycle 14 Mellla+ Eigenvalue Tracking Data CNL-22-090, Request to Use a Later Edition of the American Society of Mechanical Engineers Operation and Maintenance Code and Alternative Requests for the Fifth Inservice Testing Interval2022-12-12012 December 2022 Request to Use a Later Edition of the American Society of Mechanical Engineers Operation and Maintenance Code and Alternative Requests for the Fifth Inservice Testing Interval CNL-22-100, to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for Browns Ferry Nuclear Plant Units 1, 2, and 3,in Support of Atrium 11 Fuel Use at Browns Ferry (TS-535)2022-12-0909 December 2022 to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for Browns Ferry Nuclear Plant Units 1, 2, and 3,in Support of Atrium 11 Fuel Use at Browns Ferry (TS-535) CNL-22-066, Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for in Support of Atrium 11 Fuel Use Supplement 3, Response to Request for Additional Information2022-07-18018 July 2022 Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for in Support of Atrium 11 Fuel Use Supplement 3, Response to Request for Additional Information ML22154A4042022-06-0303 June 2022 Unit 3 Cycle 20 Mellla+ Eigenvalue Tracking Data CNL-22-057, to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 in Support of Atrium 11 Fuel Use2022-05-27027 May 2022 to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 in Support of Atrium 11 Fuel Use ML21277A1232021-10-0404 October 2021 Submittal of Browns Ferry Unit 2 Reactor Pressure Vessel Vertical Weld Flaw Evaluation ML21246A2942021-09-29029 September 2021 Enclosufinal Ea/Fonsi for Tva'S Initial and Updated Triennial Decommissioning Funding Plans for Browns Ferry Nuclear Plant ISFSIs ML21246A2952021-09-29029 September 2021 Memo to File CNL-20-102, 10 CFR 71.95 Report for 3-60B Casks User2020-12-16016 December 2020 10 CFR 71.95 Report for 3-60B Casks User ML20255A0002020-09-24024 September 2020 Staff Review of Seismic Probabilistic Risk Assessment Associated with Reevaluated Seismic Hazard Implementation of the Near Term Task Force Recommendation 2.1: Seismic ML20112F4852020-05-0606 May 2020 Staff Assessment of Flooding Focused Evaluation CNL-19-074, Extended Power Uprate - Flow Induced Vibration Summary Report2019-09-0404 September 2019 Extended Power Uprate - Flow Induced Vibration Summary Report CNL-19-004, Tennessee Valley Authority, Browns Ferry Nuclear Plant, Unit 2, Completion of Required Action for NRC Order EA-13-109, Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions2019-06-0707 June 2019 Tennessee Valley Authority, Browns Ferry Nuclear Plant, Unit 2, Completion of Required Action for NRC Order EA-13-109, Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions CNL-19-041, Extended Power Uprate - Unit 1 Flow Induced Vibration Summary Report2019-04-16016 April 2019 Extended Power Uprate - Unit 1 Flow Induced Vibration Summary Report CNL-19-032, Proposed Technical Specifications (TS) Change TS-510 - Request for License Amendments - Maximum Extended Load Line Limit Analysis Plus - Supplement 8, Additional Operator Training Information2019-03-13013 March 2019 Proposed Technical Specifications (TS) Change TS-510 - Request for License Amendments - Maximum Extended Load Line Limit Analysis Plus - Supplement 8, Additional Operator Training Information CNL-18-134, Extended Power Uprate - Replacement Steam Dryer Revised Analysis and Limit Curves Report2018-11-29029 November 2018 Extended Power Uprate - Replacement Steam Dryer Revised Analysis and Limit Curves Report ML18283B5472018-10-10010 October 2018 Responding to Letter of 11/18/1977 from E. G. Case to G. Williams, Providing Environmental Qualification Information for Electrical Connectors in Reference of IE Bulletins 77-05 & 77-05A CNL-18-112, Extended Power Uprate - Flow Induced Vibration Summary Report2018-09-13013 September 2018 Extended Power Uprate - Flow Induced Vibration Summary Report CNL-18-060, Completion of Required Action for NRC Order EA-13-109, Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions2018-05-31031 May 2018 Completion of Required Action for NRC Order EA-13-109, Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions ML18079B1402018-02-23023 February 2018 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 ML17222A3282017-09-0505 September 2017 Flood Hazard Mitigation Strategies Assessment ML17170A0732017-06-15015 June 2017 Report Pursuant to 10 CFR 71.95 (a)(3) and (B) - Failure to Follow Conditions of TN-RAM Packaging Certificate of Compliance No. 9233 ML17114A3712017-04-20020 April 2017 Errata for BWRVIP-271NP: BWR Vessel and Internals Project, Testing and Evaluation of the Browns Ferry, Unit 2, 120 Degree Capsule ML17033B1642017-02-0202 February 2017 American Society of Mechanical Engineers Section XI, Inservice Inspection, System Pressure Test, Containment Inservice Inspection, and Repair and Replacement - Cycle 11 Operation Programs ML17024A0362016-12-31031 December 2016 Operating Data Report for 2016 CNL-16-169, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 35, Consolidated Power Uprate Safety Analysis Report Revision2016-10-28028 October 2016 Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 35, Consolidated Power Uprate Safety Analysis Report Revision ML16196A0882016-08-0505 August 2016 Staff Assessment of Response to 10 CFR 50.54(f) Information Request - Flood Causing Mechanism Reevaluation ML16146A0182016-05-25025 May 2016 Special Report 296/2016-001 for Inoperable Post Accident Monitoring (PAM) Instrumentation ML16028A2952016-01-29029 January 2016 10 CFR 71.95 Notification Associated with the Failure to Observe Certificate of Compliance Condition of the 8-120B Secondary Lid Test Port Configuration ML16027A0592016-01-27027 January 2016 Snubbers Added to Lnservice Testing Program ML15356A6542015-12-22022 December 2015 Submittal of 10 CFR 50.46 30-Day Report CNL-15-056, Application to Modify Technical Specification 2.1.1.2, Reactor Core Minimum Critical Power Ratio Safety Limits (TS-506)2015-09-25025 September 2015 Application to Modify Technical Specification 2.1.1.2, Reactor Core Minimum Critical Power Ratio Safety Limits (TS-506) NL-15-169, Non-Proprietary - Safety Analysis Report for Browns Ferry, Units 1, 2 and 3, Extended Power Uprate, Attachment 72015-09-21021 September 2015 Non-Proprietary - Safety Analysis Report for Browns Ferry, Units 1, 2 and 3, Extended Power Uprate, Attachment 7 ML15282A2392015-09-21021 September 2015 Flow Induced Vibration Analysis and Monitoring Program ML15282A2402015-09-21021 September 2015 Startup Test Plan NL-15-169, Browns Ferry Units 1, 2 and 3, Flow Induced Vibration Analysis and Monitoring Program2015-09-21021 September 2015 Browns Ferry Units 1, 2 and 3, Flow Induced Vibration Analysis and Monitoring Program ML15282A1812015-09-21021 September 2015 Non-Proprietary - Safety Analysis Report for Browns Ferry, Units 1, 2 and 3, Extended Power Uprate, Attachment 7 NL-15-169, Browns Ferry, Units 1, 2, and 3, Startup Test Plan2015-09-21021 September 2015 Browns Ferry, Units 1, 2, and 3, Startup Test Plan ML15254A5432015-09-11011 September 2015 Submittal of 10 CFR 72.48 Changes, Tests, and Experiments, Biennial Summary Report Associated with the Independent Spent Fuel Storage Installation NL-15-169, ANP-3377NP, Browns Ferry Units 1, 2 and 3 LOCA Break Spectrum Analysis Atrium 10XM Fuel (Epu).2015-08-31031 August 2015 ANP-3377NP, Browns Ferry Units 1, 2 and 3 LOCA Break Spectrum Analysis Atrium 10XM Fuel (Epu). ML15282A2362015-08-31031 August 2015 NEDO-33824, Revision 0, Engineering Report, Browns Ferry Replacement Steam Dryer Stress Analysis. ML15282A1822015-08-31031 August 2015 ANP-3403NP, Revision 2, Fuel Uprate Safety Analysis Report for Browns Ferry Units 1, 2, and 3, Attachment 9 NL-15-169, NEDO-33824, Revision 0, Engineering Report, Browns Ferry Replacement Steam Dryer Stress Analysis.2015-08-31031 August 2015 NEDO-33824, Revision 0, Engineering Report, Browns Ferry Replacement Steam Dryer Stress Analysis. ML15282A1842015-08-31031 August 2015 ANP-3377NP, Browns Ferry Units 1, 2 and 3 LOCA Break Spectrum Analysis Atrium 10XM Fuel (Epu). ML15282A1852015-08-31031 August 2015 ANP-3378NP, Browns Ferry Units 1, 2 and 3, LOCA-ECCS Analysis MAPLHGR Limits for Atrium 10XM Fuel (Epu). NL-15-169, ANP-3378NP, Browns Ferry Units 1, 2 and 3, LOCA-ECCS Analysis MAPLHGR Limits for Atrium 10XM Fuel (Epu).2015-08-31031 August 2015 ANP-3378NP, Browns Ferry Units 1, 2 and 3, LOCA-ECCS Analysis MAPLHGR Limits for Atrium 10XM Fuel (Epu). NL-15-169, ANP-3403NP, Revision 2, Fuel Uprate Safety Analysis Report for Browns Ferry Units 1, 2, and 3, Attachment 92015-08-31031 August 2015 ANP-3403NP, Revision 2, Fuel Uprate Safety Analysis Report for Browns Ferry Units 1, 2, and 3, Attachment 9 2023-07-31
[Table view] Category:Technical
MONTHYEARML23192A4472023-07-31031 July 2023 Staff Assessment of Updated Seismic Hazards at TVA Sites Following the NRC Process for the Ongoing Assessment of Natural Hazards Information ML22363A3922022-12-28028 December 2022 Cycle 14 Mellla+ Eigenvalue Tracking Data CNL-22-100, to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for Browns Ferry Nuclear Plant Units 1, 2, and 3,in Support of Atrium 11 Fuel Use at Browns Ferry (TS-535)2022-12-0909 December 2022 to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for Browns Ferry Nuclear Plant Units 1, 2, and 3,in Support of Atrium 11 Fuel Use at Browns Ferry (TS-535) CNL-22-066, Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for in Support of Atrium 11 Fuel Use Supplement 3, Response to Request for Additional Information2022-07-18018 July 2022 Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for in Support of Atrium 11 Fuel Use Supplement 3, Response to Request for Additional Information ML22154A4042022-06-0303 June 2022 Unit 3 Cycle 20 Mellla+ Eigenvalue Tracking Data CNL-22-057, to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 in Support of Atrium 11 Fuel Use2022-05-27027 May 2022 to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 in Support of Atrium 11 Fuel Use ML21277A1232021-10-0404 October 2021 Submittal of Browns Ferry Unit 2 Reactor Pressure Vessel Vertical Weld Flaw Evaluation CNL-20-102, 10 CFR 71.95 Report for 3-60B Casks User2020-12-16016 December 2020 10 CFR 71.95 Report for 3-60B Casks User ML20255A0002020-09-24024 September 2020 Staff Review of Seismic Probabilistic Risk Assessment Associated with Reevaluated Seismic Hazard Implementation of the Near Term Task Force Recommendation 2.1: Seismic ML20112F4852020-05-0606 May 2020 Staff Assessment of Flooding Focused Evaluation CNL-19-074, Extended Power Uprate - Flow Induced Vibration Summary Report2019-09-0404 September 2019 Extended Power Uprate - Flow Induced Vibration Summary Report CNL-19-004, Tennessee Valley Authority, Browns Ferry Nuclear Plant, Unit 2, Completion of Required Action for NRC Order EA-13-109, Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions2019-06-0707 June 2019 Tennessee Valley Authority, Browns Ferry Nuclear Plant, Unit 2, Completion of Required Action for NRC Order EA-13-109, Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions CNL-19-041, Extended Power Uprate - Unit 1 Flow Induced Vibration Summary Report2019-04-16016 April 2019 Extended Power Uprate - Unit 1 Flow Induced Vibration Summary Report CNL-19-032, Proposed Technical Specifications (TS) Change TS-510 - Request for License Amendments - Maximum Extended Load Line Limit Analysis Plus - Supplement 8, Additional Operator Training Information2019-03-13013 March 2019 Proposed Technical Specifications (TS) Change TS-510 - Request for License Amendments - Maximum Extended Load Line Limit Analysis Plus - Supplement 8, Additional Operator Training Information CNL-18-134, Extended Power Uprate - Replacement Steam Dryer Revised Analysis and Limit Curves Report2018-11-29029 November 2018 Extended Power Uprate - Replacement Steam Dryer Revised Analysis and Limit Curves Report ML18283B5472018-10-10010 October 2018 Responding to Letter of 11/18/1977 from E. G. Case to G. Williams, Providing Environmental Qualification Information for Electrical Connectors in Reference of IE Bulletins 77-05 & 77-05A CNL-18-112, Extended Power Uprate - Flow Induced Vibration Summary Report2018-09-13013 September 2018 Extended Power Uprate - Flow Induced Vibration Summary Report ML18079B1402018-02-23023 February 2018 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 ML17222A3282017-09-0505 September 2017 Flood Hazard Mitigation Strategies Assessment ML17114A3712017-04-20020 April 2017 Errata for BWRVIP-271NP: BWR Vessel and Internals Project, Testing and Evaluation of the Browns Ferry, Unit 2, 120 Degree Capsule CNL-16-169, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 35, Consolidated Power Uprate Safety Analysis Report Revision2016-10-28028 October 2016 Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 35, Consolidated Power Uprate Safety Analysis Report Revision ML16196A0882016-08-0505 August 2016 Staff Assessment of Response to 10 CFR 50.54(f) Information Request - Flood Causing Mechanism Reevaluation ML16027A0592016-01-27027 January 2016 Snubbers Added to Lnservice Testing Program CNL-15-056, Application to Modify Technical Specification 2.1.1.2, Reactor Core Minimum Critical Power Ratio Safety Limits (TS-506)2015-09-25025 September 2015 Application to Modify Technical Specification 2.1.1.2, Reactor Core Minimum Critical Power Ratio Safety Limits (TS-506) ML15282A2402015-09-21021 September 2015 Startup Test Plan ML15282A2392015-09-21021 September 2015 Flow Induced Vibration Analysis and Monitoring Program NL-15-169, Non-Proprietary - Safety Analysis Report for Browns Ferry, Units 1, 2 and 3, Extended Power Uprate, Attachment 72015-09-21021 September 2015 Non-Proprietary - Safety Analysis Report for Browns Ferry, Units 1, 2 and 3, Extended Power Uprate, Attachment 7 NL-15-169, Browns Ferry, Units 1, 2, and 3, Startup Test Plan2015-09-21021 September 2015 Browns Ferry, Units 1, 2, and 3, Startup Test Plan ML15282A1812015-09-21021 September 2015 Non-Proprietary - Safety Analysis Report for Browns Ferry, Units 1, 2 and 3, Extended Power Uprate, Attachment 7 NL-15-169, Browns Ferry Units 1, 2 and 3, Flow Induced Vibration Analysis and Monitoring Program2015-09-21021 September 2015 Browns Ferry Units 1, 2 and 3, Flow Induced Vibration Analysis and Monitoring Program ML15282A1822015-08-31031 August 2015 ANP-3403NP, Revision 2, Fuel Uprate Safety Analysis Report for Browns Ferry Units 1, 2, and 3, Attachment 9 NL-15-169, ANP-3378NP, Browns Ferry Units 1, 2 and 3, LOCA-ECCS Analysis MAPLHGR Limits for Atrium 10XM Fuel (Epu).2015-08-31031 August 2015 ANP-3378NP, Browns Ferry Units 1, 2 and 3, LOCA-ECCS Analysis MAPLHGR Limits for Atrium 10XM Fuel (Epu). NL-15-169, ANP-3377NP, Browns Ferry Units 1, 2 and 3 LOCA Break Spectrum Analysis Atrium 10XM Fuel (Epu).2015-08-31031 August 2015 ANP-3377NP, Browns Ferry Units 1, 2 and 3 LOCA Break Spectrum Analysis Atrium 10XM Fuel (Epu). NL-15-169, ANP-3403NP, Revision 2, Fuel Uprate Safety Analysis Report for Browns Ferry Units 1, 2, and 3, Attachment 92015-08-31031 August 2015 ANP-3403NP, Revision 2, Fuel Uprate Safety Analysis Report for Browns Ferry Units 1, 2, and 3, Attachment 9 ML15282A2362015-08-31031 August 2015 NEDO-33824, Revision 0, Engineering Report, Browns Ferry Replacement Steam Dryer Stress Analysis. ML15282A1842015-08-31031 August 2015 ANP-3377NP, Browns Ferry Units 1, 2 and 3 LOCA Break Spectrum Analysis Atrium 10XM Fuel (Epu). ML15282A1852015-08-31031 August 2015 ANP-3378NP, Browns Ferry Units 1, 2 and 3, LOCA-ECCS Analysis MAPLHGR Limits for Atrium 10XM Fuel (Epu). NL-15-169, NEDO-33824, Revision 0, Engineering Report, Browns Ferry Replacement Steam Dryer Stress Analysis.2015-08-31031 August 2015 NEDO-33824, Revision 0, Engineering Report, Browns Ferry Replacement Steam Dryer Stress Analysis. CNL-15-143, the Tennessee Valley Authority (TVA) Nuclear Power Group Commercial Grade Dedication Recovery Project - Closure Report2015-07-31031 July 2015 the Tennessee Valley Authority (TVA) Nuclear Power Group Commercial Grade Dedication Recovery Project - Closure Report CNL-15-085, Response to NRC Request for Additional Information Regarding Proposed Technical Specification Change to Modify Technical Proposed Technical Specification Change to Modify Technical2015-06-0303 June 2015 Response to NRC Request for Additional Information Regarding Proposed Technical Specification Change to Modify Technical Proposed Technical Specification Change to Modify Technical CNL-14-208, Response to NRC Request for Additional Information Regarding the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Pl2014-12-17017 December 2014 Response to NRC Request for Additional Information Regarding the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Plan CNL-14-089, (Bfn), Units 1, 2, and 3 - Application to Modify Technical Specification 2.1.1, Reactor Core Safety Limits (BFN-TS-492)2014-12-11011 December 2014 (Bfn), Units 1, 2, and 3 - Application to Modify Technical Specification 2.1.1, Reactor Core Safety Limits (BFN-TS-492) ML14176A9612014-06-24024 June 2014 Submittal of Non-Proprietary BWROG Technical Product, BWROGTP-11-006 - ECCS Containment Walkdown Procedure, Rev 1 (January 2011), as Formally Requested During the Public Meeting Held on April 30, 2014 CNL-14-038, Tennessee Valley Authority'S Seismic Hazard & Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Recommendation 2.1 of Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Acc2014-03-31031 March 2014 Tennessee Valley Authority'S Seismic Hazard & Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Recommendation 2.1 of Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accid ML14077A0952014-01-30030 January 2014 BWROG-TP-14-001, Rev. 0, Containment Accident Pressure Committee (344) Task 1 - Cfd Report and Combined Npshr Uncertainty for Browns Ferry/ Peach Bottom Cvic RHR Pumps, Attachment 8 ML14077A0902013-12-31031 December 2013 BWROG-TP-13-021, Rev. 0, Containment Accident Pressure Committee (344) Task 4 - Operation in Maximum Erosion Rate Zone (Cvic Pump), Attachment 11 ML13225A5412013-12-19019 December 2013 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML13338A6342013-12-18018 December 2013 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for Browns Ferry Nuclear Plant, Units 1, 2, and 3, TAC Nos.: MF0902, MF0903, and MF0904 ML13276A0642013-09-30030 September 2013 ANP-3248NP, Revision 1, Areva RAI Responses for Browns Ferry Atrium 10XM Fuel Transition, Enclosure 3 ML13276A0662013-08-31031 August 2013 ANP-3153(NP), Revision 0, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for Atrium Tm 1OXM Fuel, Enclosure 8 2023-07-31
[Table view] Category:Technical Specifications
MONTHYEARML22348A0052023-01-25025 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 CNL-22-039, Application to Revise Technical Specifications to Adopt TSTF-554-A, Revision 1, Revise Reactor Coolant Leakage Requirements (BFN TS-537) (SQN-21-05) (WBN-TS-21-04)2022-07-13013 July 2022 Application to Revise Technical Specifications to Adopt TSTF-554-A, Revision 1, Revise Reactor Coolant Leakage Requirements (BFN TS-537) (SQN-21-05) (WBN-TS-21-04) 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 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 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 CNL-20-003, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (BFN-TS-516)2020-03-27027 March 2020 Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (BFN-TS-516) ML18277A1102019-08-27027 August 2019 Units, 1 & 2 Issuance of Amendment Nos. 309, 332, 292, 345, 339, 128, and 31 Regarding Unbalanced Voltage Protection ML18283A8972018-10-10010 October 2018 Enclosure 1: Proposed Changes to Browns Ferry Nuclear Plant Unit 1 Technical Specifications, Attachments 1, 2, & 3 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) 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) CNL-17-015, Transmittal of Response to Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 36, Transmission System Update -Safety Aspects2017-01-20020 January 2017 Transmittal of Response to Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 36, Transmission System Update -Safety Aspects NL-17-015, Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Transmittal of Response to Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 36, Transmission System Update -Saf2017-01-20020 January 2017 Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Transmittal of Response to Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 36, Transmission System Update -Safe 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 CNL-15-056, Application to Modify Technical Specification 2.1.1.2, Reactor Core Minimum Critical Power Ratio Safety Limits (TS-506)2015-09-25025 September 2015 Application to Modify Technical Specification 2.1.1.2, Reactor Core Minimum Critical Power Ratio Safety Limits (TS-506) CNL-15-073, Application to Modify the Technical Specifications by Adding New Specification TS 3.3.8.3, Emergency Core Cooling System Preferred Pump Logic, Common Accident Signal (CAS) Logic, and Unit Priority Re-Trip Logic, And.2015-09-16016 September 2015 Application to Modify the Technical Specifications by Adding New Specification TS 3.3.8.3, Emergency Core Cooling System Preferred Pump Logic, Common Accident Signal (CAS) Logic, and Unit Priority Re-Trip Logic, And. ML15065A0492015-06-0202 June 2015 Issuance of Amendment Revising Pressure and Temperature Limit Curves CNL-15-070, Withdrawal of Proposed Technical Specification Change to Revise the Leakage Rate Through MSIVs - TS-4852015-05-29029 May 2015 Withdrawal of Proposed Technical Specification Change to Revise the Leakage Rate Through MSIVs - TS-485 CNL-15-019, License Amendment Request for the Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-460-A, Revision 0, Control Rod Scram Time Testing Frequency (TS-501)2015-03-0909 March 2015 License Amendment Request for the Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-460-A, Revision 0, Control Rod Scram Time Testing Frequency (TS-501) CNL-15-029, License Amendment Request for the Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-535, Revision 0, Revise Shutdown Margin Definition to Address Advanced Fuel Designs (TS-502)2015-03-0909 March 2015 License Amendment Request for the Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-535, Revision 0, Revise Shutdown Margin Definition to Address Advanced Fuel Designs (TS-502) CNL-15-033, License Amendment Request Under Exigent Circumstances for the Change to Add a Reference to the ATRIUM-10 Xm NRC Safety Evaluation Approval in Technical Specification 5.6.5.b in Support of ATRIUM-10 Xm Fuel Use2015-02-12012 February 2015 License Amendment Request Under Exigent Circumstances for the Change to Add a Reference to the ATRIUM-10 Xm NRC Safety Evaluation Approval in Technical Specification 5.6.5.b in Support of ATRIUM-10 Xm Fuel Use CNL-14-156, Technical Specifications (TS) Changes TS-431 and TS-418 - Extended Power Uprate (EPU) - Withdrawal of Requests and Update to EPU Plans and Schedules2014-09-18018 September 2014 Technical Specifications (TS) Changes TS-431 and TS-418 - Extended Power Uprate (EPU) - Withdrawal of Requests and Update to EPU Plans and Schedules NL-14-081, Browns Ferry, Units, 1, 2 & 3, Revised Pages to Technical Specification Change TS-478-Addition of Analytical Methodologies to Technical Specification 5.6.5 and Revision of Tech Spec 2.1.1.2 in Support of ATRIUM-10 Xm Fuel Use at Browns Fer2014-05-16016 May 2014 Browns Ferry, Units, 1, 2 & 3, Revised Pages to Technical Specification Change TS-478-Addition of Analytical Methodologies to Technical Specification 5.6.5 and Revision of Tech Spec 2.1.1.2 in Support of ATRIUM-10 Xm Fuel Use at Browns Ferr CNL-14-081, Units, 1, 2 & 3, Revised Pages to Technical Specification Change TS-478-Addition of Analytical Methodologies to Technical Specification 5.6.5 and Revision of Tech Spec 2.1.1.2 in Support of ATRIUM-10 Xm Fuel Use at Browns Ferry2014-05-16016 May 2014 Units, 1, 2 & 3, Revised Pages to Technical Specification Change TS-478-Addition of Analytical Methodologies to Technical Specification 5.6.5 and Revision of Tech Spec 2.1.1.2 in Support of ATRIUM-10 Xm Fuel Use at Browns Ferry NL-13-148, Browns Ferry, Unit 1, Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature Limits (BFN TS 484)2013-12-18018 December 2013 Browns Ferry, Unit 1, Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature Limits (BFN TS 484) CNL-13-148, Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature Limits (BFN TS 484)2013-12-18018 December 2013 Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature Limits (BFN TS 484) CNL-13-126, Proposed Technical Specification Change to Revise the Leakage Rate Through MSIVs-TS-4852013-11-22022 November 2013 Proposed Technical Specification Change to Revise the Leakage Rate Through MSIVs-TS-485 ML13199A2212013-08-30030 August 2013 Issuance of Amendments Re. Deletion of References to Section XI of the ASME Code and Incorporate References to the ASME OM Code and Allow Application of 25% Extension of Surveillance Intervals ML13092A3932013-03-27027 March 2013 License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) (Technical Specification Change TS-480) ML13070A3072013-02-28028 February 2013 Technical Specification Change TS-478 - Addition of Analytical Methodologies to Technical Specification 5.6.5 for Browns Ferry 1, 2, & 3, & Revision of Technical Specification 2.1.1.2 for Browns Ferry Unit 2, in Support of ATRIUM-10 Xm Fuel ML11189A2172012-07-30030 July 2012 Issuance of Amendments Regarding Request to Add Technical Specification 3.7.3, Control Room Emergency Ventilation System, Action to Address Two Crev Subsystems Inoperable ML12114A0042012-04-18018 April 2012 Supplement to License Amendment Request to Transition to Areva Fuel ML1011601532010-04-16016 April 2010 Supplemental Information for Technical Specification Change TS-473 - Areva Fuel Transition Amendment Request ML0923700452009-11-19019 November 2009 Issuance of Amendments Regarding Technical Specification Improvement to Adopt Technical Specification Task Force (TSTF) TSTF-476, Revision 1 ML0931001212009-10-31031 October 2009 Attachment 19, Browns Ferry Unit 1 -Technical Specification Change 467, ANP-2638NP, Revision 2, Applicability of Areva Np BWR Methods to Extend Power Uprate Conditions ML0931402652009-10-31031 October 2009 Attachment 13, Browns Ferry Unit 1 - Technical Specification Change 467, ANP-2864(NP), Revision 2, Reload Safety Analysis Report 2023-01-25
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ATTACHMENT 9 Browns Ferry Nuclear Plant (BFN)
Unit 1 Technical Specifications (TS) Change 467 Revision of Technical Specifications to allow utilization of AREVA NP fuel and associated analysis methodologies Thermal Hydraulic Design Report Attached is the non proprietary version of the thermal hydraulic design report for 120%
OLTP conditions.
L32 090715 803 ANP-2807(NP)
Revision 0 Browns esign Unit 1 FerryReport T
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f~rAF~l -,,,juFuel Assemblies June 2009 A
AAREVA
AREVA NP Inc.
ANP-2807(NP)
Revision 0 Browns Ferry Unit I Thermal-Hydraulic Design Report for ATRIUM TM -10 Fuel Assemblies
AREVA NP Inc.
ANP-2807(NP)
Revision 0 Copyright © 2009 AREVA NP Inc.
All Rights Reserved paj
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUMTM-10 Fuel Assemblies Page i Nature of Changes Item Page Description and Justification
- 1. All This is the initial issue.
AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page ii Contents 1 .0 Intro d u ctio n .................................................................................................................... 1-1 2.0 S um m ary and C onclusions .............................................................................................. 2-1 3.0 Thermal-Hydraulic Design Evaluation ............................................................................ 3-1 3.1 Hydraulic C haracterization ................................................................................. 3-2 3.2 Hydraulic Com patibility ....................................................................................... 3-3 3.3 Therm al Margin Perform ance ............................................................................. 3-4 3 .4 Ro d B o w ............................................................................................................. 3 -5 3 .5 B ypass F low ....................................................................................................... 3-5 3 .6 Sta b ility ............................................................................................................... 3 -5 4.0 References ..... .................................................... 4-1 Tables 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM-10 Fuel As se m b ly ........................................................................................................................ 3 -7 3.2 Comparative Description of Browns Ferry Unit 1 ATRIUM-10 and GE14 Fuel .............. 3-9 3.3 Hydraulic Characterization Comparison Between Browns Ferry Unit 1 ATRIUM-10 and GE14 Fuel Assemblies ...................................................................... 3-10 3.4 Browns Ferry Unit 1 Thermal-Hydraulic Design Conditions ......................................... 3-11 3.5 Browns Ferry Unit 1 Transition Core Thermal-Hydraulic Results at Rated C onditions (100% P / 100% F) ....................................................................................... 3-12 3.6 Browns Ferry Unit 1 Transition Core Thermal-Hydraulic Results at Off-Rated C onditions (54.3% P / 37.3% F) ..................................................................................... 3-13 3.7 Browns Ferry Unit 1 Thermal-Hydraulic Results at Rated Conditions (100%P /
100%F) for Transition to-ATRIUM-1 0 Fuel ................................................................... 3-14 3.8 Browns Ferry Unit 1 Thermal-Hydraulic Results at Off-Rated Conditions (54.3%P / 37.3%F) for Transition to ATRIUM-1 0 Fuel ................................................. 3-15 Figures 3 .1 Axial P ow er S hapes ..................................................................................................... 3-16 3.2 Transition Core: Hydraulic Demand Curves 100%P/100%F ........................................ 3-17 3.3 Transition Core: Hydraulic Demand Curves 54.3%P/37.3%F ...................................... 3-18 AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUMTM-10 Fuel Assemblies Page iii Nomenclature AOO anticipated operational occurrence ASME American Society of Mechanical Engineers BWR boiling water reactor CHF critical heat flux CPR critical power ratio CRDA control rod drop accident LOCA loss-of-coolant accident LTP lower tie plate MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio NRC Nuclear Regulatory Commission, U.S.
PLFR part-length fuel rod RPF radial peaking factor UTP upper tie plate AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page 1-1 1.0 Introduction The results of Browns Ferry Unit 1 thermal-hydraulic analyses are presented to demonstrate that AREVA NP* ATRIUMTM-10t fuel is hydraulically compatible with coresident GE14 fuel. This report also provides the hydraulic characterization of the ATRIUM-10 and coresident GE14 fuel designs for Browns Ferry Unit 1.
The generic thermal-hydraulic design criteria applicable to the design have been reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC) in the topical report ANF-89-98(P)(A) Revision 1 and Supplement 1 (Reference 1). In addition, thermal-hydraulic criteria applicable to the design have also been reviewed and approved by the NRC in the topical report XN-NF-80-19(P)(A) Volume 4 Revision 1 (Reference 2).
- AREVA NP Inc. is an AREVA and Siemens company.
t ATRIUM is a trademark of AREVA NP.
AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page 2-1 2.0 Summary and Conclusions ATRIUM-10 fuel assemblies have been determined to be hydraulically compatible with GE14 fuel coresident in the reactor for the entire range of the licensed power-to-flow operating map.
Detailed calculation results supporting this conclusion are provided in Section 3.2 and Tables 3.4 to 3.8.
The ATRIUM-10 fuel design is geometrically different from the coresident GE14 design, but hydraulically the two designs are compatible. [
Core bypass flow (defined as leakage flow through the lower tie plate (LTP) flow holes, channel seal, core support plate, and LTP-fuel support interface) is not adversely affected by the introduction of the ATRIUM-10 fuel design. Analyses at rated conditions show core bypass flow varying between [ ] of rated flow for transition core configurations ranging from a full GE14 fuel core to a full ATRIUM-10 core, respectively.
Analyses demonstrate the thermal-hydraulic design and compatibility criteria discussed in Section 3.0 are satisfied for the Browns Ferry Unit 1 transition core consisting of ATRIUM-10 and GE14 fuel for the expected core power distributions and core power/flow conditions encountered during operation.
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Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page 3-1 3.0 Thermal-Hydraulic Design Evaluation Thermal-hydraulic analyses are performed to verify that design criteria are satisfied and to help establish thermal operating limits with acceptable margins of safety during normal reactor operation and anticipated operational occurrences (AOOs). The design criteria that are applicable to the ATRIUM-1 0 fuel design are described in Reference 1. To the extent possible, these analyses are performed on a generic fuel design basis. However, due to reactor and cycle operating differences, many of the analyses supporting these thermal-hydraulic operating limits are performed on a plant- and cycle-specific basis and are documented in plant- and cycle-specific reports.
The thermal-hydraulic design criteria are summarized below:
Hydraulic compatibility. The hydraulic flow resistance of the reload fuel assemblies shall be sufficiently similar to the existing fuel in the reactor such that there is no significant impact on total core flow or the flow distribution among assemblies in the core. This criterion evaluation is addressed in Sections 3.1 and 3.2.
Thermal margin performance. Fuel assembly geometry, including spacer design and rod-to-rod local power peaking, should minimize the likelihood of boiling transition during normal reactor operation as well as during AQOs. The fuel design should fall within the bounds of the applicable empirically based boiling transition correlation approved for AREVA reload fuel. Within other applicable mechanical, nuclear, and fuel performance constraints, the fuel design should achieve good thermal margin performance. The thermal-hydraulic design impact on steady state thermal margin performance is addressed in Section 3.3. Additional thermal margin performance evaluations dependent on the cycle-specific design are addressed in the reload licensing report.
Fuel centerline temperature. Fuel design and operation shall be such that fuel centerline melting is not projected for normal operation and AQOs. This criterion evaluation is addressed in the mechanical design report.
Rod bow. The anticipated magnitude of fuel rod bowing under irradiation shall be accounted for in establishing thermal margin requirements. This criterion evaluation is addressed in Section 3.4.
Bypass flow. The bypass flow characteristics of the reload fuel assemblies shall not differ significantly from the existing fuel in order to provide adequate flow in the bypass region. This criterion evaluation is addressed in Section 3.5.
Stability. Reactors fueled with new fuel designs must be stable in the approved power and flow operating region. The stability performance of new fuel designs will be equivalent to, or better than, existing (approved) AREVA fuel designs. This criterion evaluation is addressed in Section 3.6. Additional core stability evaluations dependent on the cycle-specific design are addressed in the reload licensing report.
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Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page 3-2 Loss-of-coolant accident (LOCA) analysis. LOCAs are analyzed in accordance with Appendix K modeling requirements using NRC-approved models. The criteria are defined in 10 CFR 50.46. LOCA analysis results are presented in the break spectrum and MAPLHGR reports.
Control rod drop accident (CRDA) analysis. The deposited enthalpy must be less than 280 cal/gm for fuel coolability. This criterion evaluation is addressed in the reload licensing report.
ASME overpressurization analysis. ASME pressure vessel code requirements must be satisfied. This criterion evaluation is addressed in the reload licensing report.
Seismic/LOCA liftoff. Under accident conditions, the assembly must remain engaged in the fuel support. This criterion evaluation is addressed in the mechanical design report.
A summary of.the thermal-hydraulic design evaluations is given in Table 3.1.
3.1 Hydraulic Characterization Basic geometric parameters for ATRIUM-10 and GEl4 fuel designs are summarized in Table 3.2. Component loss coefficients for the ATRIUM-10 are based on tests and are presented in Table 3.3. These loss coefficients include modifications to the test data reduction process [
] The bare rod friction, ULTRAFLOW TM* spacer, and UTP losses for ATRIUM-10 are based on flow tests. The local losses for the Browns Ferry ATRIUM-10 FUELGUARDTM* LTP are based on pressure drop tests performed at AREVA's Portable Hydraulic Test Facility. [
] The local component (LTP, spacer, and UTP) loss coefficients for the GE14 fuel are based on flow test results.
The primary resistance for the leakage flow through the LTP flow holes is [
3 The resistances for the leakage paths are shown in Table 3.3.
- ULTRAFLOW and FUELGUARD are trademarks of AREVA NP.
AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page 3-3 3.2 Hydraulic Compatibility The thermal-hydraulic analyses were performed in accordance with the AREVA thermal-hydraulic methodology for BWRs. The methodology and constitutive relationships used by AREVA for the calculation of pressure drop in BWR fuel assemblies are presented in Reference 3 and are implemented in the XCOBRA code. The XCOBRA code predicts steady-state thermal-hydraulic performance of the fuel assemblies of BWR cores at various operating conditions and power distributions. XCOBRA received NRC approval in Reference 4.
The NRC reviewed the information provided in Reference 5 regarding inclusion of water rod models in XCOBRA and accepted the inclusion in Reference 6.
Hydraulic compatibility, as it relates to the relative performance of the ATRIUM-10 and GE14 fuel designs, has been evaluated. Detailed analyses were performed for full core GE14 and full core ATRIUM-10 configurations. Analyses for a mixed ATRIUM-10 and GE14 core were also performed to demonstrate that the thermal-hydraulic design criteria are satisfied for a transition core configuration.
The hydraulic compatibility analysis is based on [
Table 3.4 summarizes the input conditions for the analyses. These conditions reflect two of the state points considered in the analyses: 100% power/1 00% flow and 54.3% power/37.3% flow.
Table 3.4 also defines the core loading for the transition core configuration. Input for other core configurations is similar in that core operating conditions remain the same and the same axial power distribution is used. Evaluations were made with the bottom-, middle-, and top-peaked axial power distributions presented in Figure 3.1. Results presented in this report are for the bottom-peaked power distribution. Results for middle- and top-peaked axial power distributions show similar trends.
Table 3.5 and Table 3.6 provide a summary of calculated thermal-hydraulic results using the transition core configuration. Table 3.7 and Table 3.8 provide a summary of results for all core configurations evaluated. Core average results and the differences between ATRIUM-10 and GE14 fuel rated power results are within the range considered compatible, as expected based AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page 3-4 on previous transitions involving GE14 fuel. Similar agreement occurs at lower power levels.
As shown in Table 3.5, [
] Table 3.6 shows that, [
] Differences in assembly flow between the ATRIUM-10 and GEI4 fuel designs as a function of assembly power level are shown in Figure 3.2 and Figure 3.3.
Core pressure drop and core bypass flow fraction are also provided for the configurations evaluated. Based on the reported changes in pressure drop and assembly flow caused by the transition from GE14 to ATRIUM-10, the ATRIUM-10 design is considered hydraulically compatible with the GE14 design since the thermal-hydraulic design criteria are satisfied.
3.3 Thermal Margin Performance Relative thermal margin analyses were performed in accordance with the thermal-hydraulic methodology for AREVA's XCOBRA code. The calculation of the fuel assembly critical power ratio (CPR) (thermal margin performance) is established by means of an empirical correlation based on results of boiling transition test programs. The CPR methodology is the approach used by AREVA to determine the margin to thermal limits for BWRs.
CPR values for ATRIUM-10 and GE14 fuel are calculated with the SPCB critical power correlation (Reference 7). The NRC-approved methodology to demonstrate the acceptability of using the SPCB correlation for computing GE14 fuel CPR is presented in Reference 8.
Assembly design features are incorporated in the CPR calculation through the F-eff term. The F-eff is based on the local power peaking for the nuclear design and on additive constants determined in accordance with approved procedures. The local peaking factors are a function of assembly void fraction and exposure.
For the compatibility evaluation, steady-state analyses evaluated ATRIUM-10 and GE14 assemblies with radial peaking factors (RPFs) between [
AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page 3-5
] Table 3.5 and Table 3.6 show CPR results of the ATRIUM-10 and GE14 fuels.
Table 3.7 and Table 3.8 show similar comparisons of CPR and assembly flow for the various core configurations evaluated. Analysis results indicate ATRIUM-10 fuel will not cause thermal margin problems for the coresident GEl4 fuel.
3.4 Rod Bow The bases for rod bow are discussed in the mechanical design report. Rod bow magnitude is determined during the fuel-specific mechanical design analyses. Rod bow has been measured during post-irradiation examinations of BWR fuel fabricated by AREVA.
3.5 Bypass Flow Total core bypass flow is defined as leakage flow through the LTP flow holes, channel seal, core support plate, and LTP-fuel support interface. Table 3.7 shows that total core bypass flow (excluding water rod flow) fraction at rated conditions changes from [ ] of rated core flow during the transition from a full GE14 core to a full ATRIUM-10 core (bottom-peaked power shape). [
] In summary, adequate bypass flow will be available with the introduction of the ATRIUM-10 fuel design and applicable design criteria are met.
3.6 Stability Each new fuel design is analyzed to demonstrate that the stability performance is equivalent to or better than an existing (NRC-approved) AREVA fuel design. The stability performance is a function of the core power, core flow, core power distribution, and to a lesser extent, the fuel design. [
] A comparative stability analysis was performed with the NRC-approved STAIF code (Reference 9). The study AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page 3-6 shows that the ATRIUM-10 fuel design has decay ratios equivalent to or better than other approved AREVA fuel designs.
As stated above, the stability performance of a core is strongly dependent on the core power, core flow, and power distribution in the core. Therefore, core stability is evaluated on a cycle-specific basis and addressed in the reload licensing report.
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Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page 3-7 Table 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM-10 Fuel Assembly Report Section Description Criteria Results or Disposition Thermal and Hydraulic Criteria 3.1 / 3.2 Hydraulic Hydraulic flow resistance Verified on a plant-specific basis.
compatibility shall be sufficiently similar to existing fuel ATRIUM-10 demonstrated to be such that there is no compatible with GE14.
significant impact on total core flow or flow [
distribution among assemblies.
3.3 Thermal margin Fuel design shall be SPCB is applied to both the performance within the limits of ATRIUM-10 and GE14 fuel.
applicability of an approved CHF correlation.
< 0.1% of rods in boiling Verified on cycle-specific basis for transition. Chapter 14 analyses.
Fuel centerline No centerline melting. Refer to the mechanical design temperature report.
3.4 Rod bow Rod bow must be The lateral displacement of the fuel accounted for in rods due to fuel rod bowing is not of establishing thermal sufficient magnitude to impact margins, thermal margins.
3.5 Bypass flow Bypass flow Verified on a plant-specific basis.
characteristics shall be similar among Analysis results demonstrate that assemblies to provide adequate bypass flow is provided.
adequate bypass flow.
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Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM -10 Fuel Assemblies Page 3-8 Table 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM-10 Fuel Assembly (Continued)
Report Section Description Criteria Results or Disposition Thermal and Hydraulic Criteria (Continued) 3.6 Stability New fuel designs are ATRIUM-10 channel and core stable in the approved decay ratios have been power and flow operating demonstrated to be equivalent to or region, and stability better than other approved AREVA performance will be fuel designs.
equivalent to (or better than) existing (approved) Core stability behavior is evaluated AREVA fuel designs. on a cycle-specific basis.
LOCA analysis LOCA analyzed in Approved Appendix K LOCA accordance with model.
Appendix K modeling requirements. Criteria Plant- and fuel-specific analysis defined in 10 CFR 50.46. with cycle-specific verifications.
CRDA analysis < 280 cal/gm for Cycle-specific analysis is coolability. performed.
ASME over- ASME pressure vessel Cycle-specific analysis is pressurization core requirements shall performed.
analysis be satisfied.
Seismic/LOCA Assembly remains Refer to the mechanical design liftoff engaged in fuel support. report.
AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUMTM-1 0 Fuel Assemblies Page 3-9 Table 3.2 Comparative Description of Browns Ferry Unit 1 ATRIUM-10 and GE14 Fuel Fuel Parameter ATRIUM-10 GE14 Number of fuel rods Full-length fuel rods 83 78 PLFRs 8 14 Fuel clad OD, in 0.3957 0.404 Number of spacers 8 8 Active fuel length, ft Full-length fuel rods 12.454 12.500 PLFRs 7.5 7.0 Hydraulic resistance characteristics Table 3.3 Table 3.3 Number of water rods 1 2 Water rod OD, in 1.378* 0.980
- Square water channel outer width.
AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-1 0 Fuel Assemblies Page 3-10 Table 3.3 Hydraulic Characterization Comparison Between Browns Ferry Unit I ATRIUM-10 and GE14 Fuel Assemblies I
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Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUMTM-1 0 Fuel Assemblies Page 3-11 Table 3.4 Browns Ferry Unit I Thermal-Hydraulic Design Conditions Reactor conditions 100%P / 100%F 54.3%P / 37.3%F Core power level, MWt 3952 2146 Core exit pressure, psia 1060 987 Core inlet enthalpy, Btu/lbm 523.2 492.2 Total core coolant flow, Mlbm/hr 102.5 38.2 Axial power shape Bottom-peaked Bottom-peaked (Figure 3.1) (Figure 3.1)
AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUMTM-10 Fuel Assemblies Page 3-12 Table 3.5 Browns Ferry Unit 1 Transition Core Thermal-Hydraulic Results at Rated Conditions (100%P / 100%F)
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Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUMTM-10 Fuel Assemblies Page 3-13 Table 3.6 Browns Ferry Unit 1 Transition Core Thermal-Hydraulic Results at Off-Rated Conditions (54.3%P / 37.3%F)
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AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-1 0 Fuel Assemblies Page 3-14 Table 3.7 Browns Ferry Unit I Thermal-Hydraulic Results at Rated Conditions (100%P / 100%F) for Transition to ATRIUM-10 Fuel I
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Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-1 0 Fuel Assemblies Page 3-15 Table 3.8 Browns Ferry Unit 1 Thermal-Hydraulic Results at Off-Rated Conditions (54.3%P / 37.3%F) for Transition to ATRIUM-10 Fuel I
AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page 3-16 I
Figure 3.1 Axial Power Shapes AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-1 0 Fuel Assemblies Page 3-17 I
Figure 3.2 Transition Core:
Hydraulic Demand Curves 100%P/100%F AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-1 0 Fuel Assemblies Page 3-18 I
Figure 3.3 Transition Core:
Hydraulic Demand Curves 54.3%P/37.3%F AREVA NP Inc.
Browns Ferry Unit 1 ANP-2807(NP)
Thermal-Hydraulic Design Report Revision 0 for ATRIUM TM-10 Fuel Assemblies Page 4-1 4.0 References
- 1. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic MechanicalDesign Criteriafor BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
- 2. XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.
- 3. XN-NF-79-59(P)(A), Methodology for Calculationof PressureDrop in BWR Fuel Assemblies, Exxon Nuclear Company, November 1983.
- 4. XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon NuclearMethodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.
- 5. Letter, R.A. Copeland (AREVA) to R.C. Jones (USNRC), "Explicit Modeling of BWR Water Rod in XCOBRA," RAC:002:90, January 9,1990.
- 6. Letter, R.C. Jones (USNRC) to R.A. Copeland (AREVA), no subject (regarding XCOBRA water rod model), February 1, 1990.
- 7. EMF-2209(P)(A) Revision 2, SPCB CriticalPower Correlation,Framatome ANP, September 2003.
- 8. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation'sCriticalPower Correlationsto Co-Resident Fuel, Siemens Power Corporation, August 2000.
- 9. EMF-CC-074(P)(A) Volume 1, STAIF - A ComputerProgram for BWR StabilityAnalysis in the FrequencyDomain; and Volume 2, STAIF - A Computer Programfor BWR Stability Analysis in the Frequency Domain - Code QualificationReport, Siemens Power Corporation, July 1994.
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