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Category:Drawing
MONTHYEARML22292A1182022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 15, Figures ML22292A0812022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 4, Figures 4.3-1 Through 4.6-9 ML22292A1022022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 6, Appendix 6D, Figures 6.2.1 Through 6C-1 ML22292A1082022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 3, Figures 3.2-1 Thru 3C.3-3 ML22292A1362022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 13, Figure 13.5-1, at the Controls Main Control Room El. 306'-0 ML22292A0772022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 8, Figures ML22292A0662022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 9, Figures ML22292A1272022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 14, Figures ML22292A0932022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 10, Figures ML22292A0892022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 5, Figures ML22292A1372022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 12, Figures ML22292A1102022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 1, Figures 1.2-33 Through End ML22292A1172022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 2, Figures ML22292A1302022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 17, Figures ML22292A1382022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 18, Figures ML22292A1092022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 2, Figures 2K.28A Through End ML22292A1012022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 2, Figures 2H.43 Through 2K-27F ML22292A0632022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 7, Figures ML22292A0992022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 9, Figures 9.4-2e Thru End ML22292A0652022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 1, Figures ML22292A0922022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 11, Figures ML18018B3202018-01-18018 January 2018 Undersized Documents - Transformer 2MTX-XM1B - Oil Cooler Piping, Failed Windings, Tank Bulging, Preparation for Removal, HV Tap Changer, and Lv Leads and Bracing ML18018B1942018-01-18018 January 2018 Undersized Documents, Drawings and Photographs ML17037C7252017-02-0606 February 2017 Drawing; Figure 7.15-1, Rod Worth Minimizer Functional Control Diagram ML17037C5622017-02-0606 February 2017 Plate 4-9 - Geologic Plan View of Fault Cooling Tower Piping Trench ML17037B4552017-02-0606 February 2017 Drawing 12177-EE-1AR-3, 600V One Line Diagram 2EHS*MCC102 Auxiliary Building North - Nine Mile Point Nuclear Station - Unit 2. ML17037A7102017-02-0606 February 2017 Figure 12.7-1a Radiation Zones - Reactor Building. ML17037C6602017-02-0606 February 2017 Figure 5.4-3; Steam Concentration Following a LOCA with CAD Operation ML17037C6592017-02-0606 February 2017 Figure 5.4-2; Hydrogen and Oxygen Concentration in Primary Containment Following a LOCA ML17037C6582017-02-0606 February 2017 Figure 5.3-2; Stand-by Gas Treatment System ML17037C5692017-02-0606 February 2017 Plate 5-10A - Diagram Showing Relationship Between Dip of Bedding and Depth Borings T-4-8 Through T-4-10 ML17037C5672017-02-0606 February 2017 Plate 5-10 - Diagram Showing Relationship Between Dip of Bedding and Depth Borings T-4-1 Through T-4-7 ML17037C5662017-02-0606 February 2017 Plate 5-9A - Diagram Showing Relationship Between Dip of Bedding and Depth Borings T-3-9 and T-3-10 ML17037C5652017-02-0606 February 2017 Plate 5-9 - Diagram Showing Relationship Between Dip of Bedding and Depth Borings T-3-1 Through T-3-8 ML17037C5642017-02-0606 February 2017 Plate 5-1 - Composite Site Stratigraphic Column ML17037C5632017-02-0606 February 2017 Plate 4-10 - Cross Section CT-2 West Wall Cooling Tower Piping Trench ML17037A7122017-02-0606 February 2017 Figure 12.7-2a, Radiation Zones-Turbine Building, Radwaste Building, and Screenwell. ML17037A8182017-02-0606 February 2017 Drawing No. 12177-ESK-4CEC24 ML17037A6892017-02-0606 February 2017 Figure 12.1-1, General Arrangement Reactor Building - Plans ML17037A6902017-02-0606 February 2017 Figure 7.9 - 1A, Feedwater Control System. ML17037A6922017-02-0606 February 2017 Figure 7.6-7, Rod Block Interlocks from Neutron Monitoring System ML17037A8292017-02-0606 February 2017 Drawing No. 12177-ESK-4CEC14A ML17037A8382017-02-0606 February 2017 Drawing - (Drawing No. Not Available) ML17037A7142017-02-0606 February 2017 Drawing No. 12241-GSK-31, Over Excavated Area - Reactor Containment - Locations. ML17037A4622017-02-0606 February 2017 Drawing 807E154TY, Revision 3, Leak Detection System Elementary Diagram. Sheet 4 ML17037A8272017-02-0606 February 2017 Drawing No. 12177-ESK-4CEC15 ML17037B4562017-02-0606 February 2017 Drawing 12177-EE-1AQ-3, 600V One Line Diagram 2EHS*MCC101 & 2EHS*MCC301 Screenwell Building - Nine Mile Point Nuclear Station - Unit 2. ML17037A8162017-02-0606 February 2017 Drawing No. 12177-ESK-4CEC25A ML17037A6942017-02-0606 February 2017 Figure 7.4-10, Power Range Neutron Monitoring System Instrument Electrical Diagram ML17037A7052017-02-0606 February 2017 Figure 2.7-16, Screenwell Building Plan and Sections 2022-10-05
[Table view] Category:Graphics incl Charts and Tables
MONTHYEARML22314A2262022-11-10010 November 2022 E-mail Dated 11/10/2022 Relief Request Associated with Pump Periodic Verification Tests of Core Spray System Pumps ML22292A1032022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 2, Appendix 2B, Tables 2B-8 Through 2B-38 ML22292A0982022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 2, Appendix 2B, Table 2B-35, Joint Distribution of Wing Direction and Speed ML22292A0822022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 1, Figures 1.1-1 Thru 1.2-19 Sheet 2 ML22292A0732022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 02, Appendix 2B Table 2B-39 ML22292A0812022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 4, Figures 4.3-1 Through 4.6-9 ML22292A1382022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 18, Figures ML22292A0852022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 2, Appendix 2B Table 2B-40 Thru 2B-55 ML22292A1302022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 17, Figures ML22292A1082022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 3, Figures 3.2-1 Thru 3C.3-3 ML22292A1162022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 2, Appendix 2B, Table 2B-38, Joint Distribution of Wind Direction and Speed Location 100 Ft ML22292A1052022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Chapter 2, Figures 2.5-110 Through 2H-42 NMP1L3339, R. E. Ginna Station - Constellation Energy Group, LLC: Notification of Change in Indirect Ownership2020-04-24024 April 2020 R. E. Ginna Station - Constellation Energy Group, LLC: Notification of Change in Indirect Ownership ML18018A9652018-01-18018 January 2018 Nine Mile Point, Unit 1 - Equipment Qualification Program and Tables I - Equipment Qualification Reports and Table Ii - TMI Action Plan ML18018A9612018-01-18018 January 2018 Preservice Inspection Plan ML17037C6892017-02-0606 February 2017 Table 7.2-1, Process Pipelines Penetrating Primary Containment, 1 of 6 ML17037A6592017-02-0606 February 2017 Stratigraphic Correlation Chart Borings T-4-1 Through T-4-12, Plate 5-3 ML17037C6832017-02-0606 February 2017 Table 7.2-1, Isolation Signal Codes (Cont'D) 6 of 6 ML17037C5692017-02-0606 February 2017 Plate 5-10A - Diagram Showing Relationship Between Dip of Bedding and Depth Borings T-4-8 Through T-4-10 ML17037C5672017-02-0606 February 2017 Plate 5-10 - Diagram Showing Relationship Between Dip of Bedding and Depth Borings T-4-1 Through T-4-7 ML17037C6912017-02-0606 February 2017 Table 7.2-1 (Cont'D) 3 of 6 ML17037C5662017-02-0606 February 2017 Plate 5-9A - Diagram Showing Relationship Between Dip of Bedding and Depth Borings T-3-9 and T-3-10 ML17037C5652017-02-0606 February 2017 Plate 5-9 - Diagram Showing Relationship Between Dip of Bedding and Depth Borings T-3-1 Through T-3-8 ML17037C5642017-02-0606 February 2017 Plate 5-1 - Composite Site Stratigraphic Column ML17037C5632017-02-0606 February 2017 Plate 4-10 - Cross Section CT-2 West Wall Cooling Tower Piping Trench ML17037C5622017-02-0606 February 2017 Plate 4-9 - Geologic Plan View of Fault Cooling Tower Piping Trench ML17037C5612017-02-0606 February 2017 Plate 1-2 - Composite Site Stratographic Column Showing Position of Overcome Borings in Relation to Site Stratigraphy ML17037C7582017-02-0606 February 2017 Table 7.2-1 (Cont'D) 4 of 6 ML17037A6662017-02-0606 February 2017 Relative Pollen Percentages, Spores and Miscellaneous Microfossil Abundance, Plate 1-3 ML17037A6622017-02-0606 February 2017 Structural Comparison of Fault Zone in Boring Series T-3, Plate 5-11 ML17037A6582017-02-0606 February 2017 Stratigraphic Correlation Chart Borings T-3-1 Through T-3-10, Plate 5-2 ML17037A6632017-02-0606 February 2017 Structural Comparison of Fault Zone in Boring Series T-4, Plate 6-12 ML17037A7152017-02-0606 February 2017 Figure 6-3, Response Test Strip Chart for Unstable Condition. ML17037A7012017-02-0606 February 2017 Table 7.2-1 (Continued) ML17037C6922017-02-0606 February 2017 Table 7.2-1 (Cont'D) 4 of 6 ML17037C7592017-02-0606 February 2017 Table 7.2-1 (Cont'D) 6 of 6 ML17037C7792017-02-0606 February 2017 Table 7.2-1 Isolation Signal Codes(Cont'D) 6 of 6 ML17037C7822017-02-0606 February 2017 Table 7.2-1 Process Pipelines Penetrating Primary Containment 1 of 6 ML17037C7852017-02-0606 February 2017 Table 7.2-1 (Cont'D) 4 of 6 ML16056A1392016-03-11011 March 2016 Correction to the U.S. Nuclear Regulatory Commission Analysis of Licensees' Decommissioning Funding Status Reports ML12138A1352012-05-0707 May 2012 Reactor Pressure Vessel Head Weld Flaw Evaluation - Response to NRC Request for Additional Information ML12045A3892012-01-13013 January 2012 EPIP-EP-001, Attachment 1 EAL Matrix Unit 1, Page 1 of 2 ML12045A3962012-01-13013 January 2012 EPIP-EP-002, Attachment 1 EAL Matrix Unit 2, Page 2 of 2 ML12045A3942012-01-13013 January 2012 EPIP-EP-002, Attachment 1 EAL Matrix Unit 2, Page 1 of 2 ML12045A3932012-01-13013 January 2012 EPIP-EP-001, Attachment 1 EAL Matrix Unit 1, Page 2 of 2 ML12045A3922012-01-13013 January 2012 EPIP-EP-002, Attachment 1 EAL Matrix Unit 2, Page 1 of 2 ML12045A3912012-01-13013 January 2012 EPIP-EP-001, Attachment 1 EAL Matrix Unit 1, Page 2 of 2 ML12045A3902012-01-13013 January 2012 EPIP-EP-001, Attachment 1 EAL Matrix Unit 1, Page 1 of 2 ML12045A3872012-01-13013 January 2012 EPIP-EP-002, Attachment 1 EAL Matrix Unit 2, Page 2 of 2 ML0728203072007-10-11011 October 2007 Electronic Distribtion Initiative Letter, Licensee List, Electronic Distribution Input Information, Division Plant Mailing Lists 2022-11-10
[Table view] Category:Letter
MONTHYEARML24222A6772024-08-0909 August 2024 Response to Request for Additional Information for Application to Revise Technical Specifications to Adopt TSTF-591-A, “Revise Risk Informed Completion Time (RICT) Program” Revision 0 and Revise 10 CFR 50.69 License Condition IR 05000220/20240022024-08-0505 August 2024 Integrated Inspection Report 05000220/2024002 and 05000410/2024002 ML24215A3002024-08-0202 August 2024 Operator Licensing Examination Approval NMP1L3601, Supplemental Information Letter No. 2 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2024-07-31031 July 2024 Supplemental Information Letter No. 2 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation ML24213A1412024-07-31031 July 2024 Requalification Program Inspection NMP2L2883, Fourth Inservice Inspection Interval, Second Inservice Inspection Period 2024 Owner’S Activity Report for RFO-19 Inservice Examinations2024-07-24024 July 2024 Fourth Inservice Inspection Interval, Second Inservice Inspection Period 2024 Owner’S Activity Report for RFO-19 Inservice Examinations ML24198A0852024-07-16016 July 2024 Senior Reactor and Reactor Operator Initial License Examinations RS-24-070, Independent Spent Fuel Storage Installation, Nine Mile Point, Units 1 and 2, Quad Cities, Units 1 and 2, R. E. Ginna - Nuclear Radiological Emergency Plan Document Revisions2024-07-12012 July 2024 Independent Spent Fuel Storage Installation, Nine Mile Point, Units 1 and 2, Quad Cities, Units 1 and 2, R. E. Ginna - Nuclear Radiological Emergency Plan Document Revisions RS-24-061, Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations2024-06-14014 June 2024 Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations NMP1L3584, License Amendment Request to Revise Technical Specifications to Adopt TSTF-230, Revision 1, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling2024-06-13013 June 2024 License Amendment Request to Revise Technical Specifications to Adopt TSTF-230, Revision 1, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling IR 05000220/20244012024-05-30030 May 2024 Security Baseline Inspection Report 05000220/2024401 and 05000410/2024401(Cover Letter Only) NMP1L3591, Response to Ny State Pollutant Discharge Elimination System (SPDES) Permit Request for Information & Modification Request2024-05-18018 May 2024 Response to Ny State Pollutant Discharge Elimination System (SPDES) Permit Request for Information & Modification Request NMP1L3589, Special Report: Containment High Range Radiation Monitor Instrumentation Channel 12 Inoperable2024-05-16016 May 2024 Special Report: Containment High Range Radiation Monitor Instrumentation Channel 12 Inoperable NMP1L3582, 2023 Annual Radioactive Environmental Operating Report for Nine Mile Point Units 1 and 22024-05-15015 May 2024 2023 Annual Radioactive Environmental Operating Report for Nine Mile Point Units 1 and 2 NMP1L3582, Annual Radioactive Environmental Operating Report2024-05-15015 May 2024 Annual Radioactive Environmental Operating Report IR 05000220/20240012024-05-10010 May 2024 Integrated Inspection Report 05000220/2024001 and 05000410/2024001 RS-24-049, Updated Notice of Intent to Pursue Subsequent License Renewal Applications2024-05-0909 May 2024 Updated Notice of Intent to Pursue Subsequent License Renewal Applications RS-24-038, Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds2024-05-0202 May 2024 Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds RS-24-041, Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests2024-04-30030 April 2024 Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests NMP1L3581, Independent Spent Fuel Storage Installation (ISFSI) - 2023 Radioactive Effluent Release Report2024-04-30030 April 2024 Independent Spent Fuel Storage Installation (ISFSI) - 2023 Radioactive Effluent Release Report NMP2L2877, 2023 Annual Environmental Operating Report2024-04-19019 April 2024 2023 Annual Environmental Operating Report NMP2L2878, Core Operating Limits Report2024-04-16016 April 2024 Core Operating Limits Report NMP1L3570, Supplemental Information Letter - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2024-02-0101 February 2024 Supplemental Information Letter - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation IR 05000220/20230042024-02-0101 February 2024 Integrated Inspection Report 05000220/2023004 and 05000410/2023004 NMP1L3569, CFR 50.46 Annual Report2024-01-26026 January 2024 CFR 50.46 Annual Report ML24004A2122024-01-0808 January 2024 Senior Reactor and Reactor Operator Initial License Examinations ML23354A0012024-01-0404 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0059 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML23278A1292023-12-14014 December 2023 Units 1 & 2; Limerick, Units 1 & 2; Nine Mile Point, Units 1 & 2; and Peach Bottom, Units 2 & 3 -Revision to Approved Alternatives to Use Boiling Water Reactor Vessel and Internals Project Guidelines IR 05000410/20243012023-12-14014 December 2023 Initial Operator Licensing Examination Report 05000410/2024301 NMP1L3566, Radiological Emergency Plan Document Revision. Includes EP-AA-1013, Revision 10, Radiological Emergency Plan Annex for Nine Mile Point Station2023-12-14014 December 2023 Radiological Emergency Plan Document Revision. Includes EP-AA-1013, Revision 10, Radiological Emergency Plan Annex for Nine Mile Point Station ML23305A1402023-12-13013 December 2023 Units 1 & 2; Nine Mile Point, Unit 2; Peach Bottom, Units 2 & 3; and Quad Cities, Units 1 and 2 - Issuance of Amendments to Adopt Traveler TSTF-580 ML23291A4642023-12-0707 December 2023 Issuance of Amendment No. 251 Regarding the Adoption of Title 10 the Code of Federal Regulations Section 50.69, Risk-Informed Categorization and Treatment of SSC for Nuclear Power Plants NMP1L3564, Supplemental Response to Part 73 Exemption Request - Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements2023-12-0707 December 2023 Supplemental Response to Part 73 Exemption Request - Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements ML23289A0122023-12-0606 December 2023 Issuance of Amendment No. 250 Regarding the Revision to Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b NMP1L3563, Submittal of Relief Request I5R-12, Revision 0, Concerning the Installation of a Full Structural Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208)2023-12-0404 December 2023 Submittal of Relief Request I5R-12, Revision 0, Concerning the Installation of a Full Structural Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208) IR 05000220/20234022023-11-28028 November 2023 Security Baseline Inspection Report 05000220/2023402 and 05000410/2023402 NMP1L3557, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-22022 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums ML23305A0052023-11-0101 November 2023 Operator Licensing Examination Approval IR 05000220/20234202023-11-0101 November 2023 Security Baseline Inspection Report 05000220/2023420 and 05000410/2023420 IR 05000220/20230032023-10-25025 October 2023 Integrated Inspection Report 05000220/2023003 and 05000410/2023003 IR 05000220/20235012023-10-17017 October 2023 Emergency Preparedness Biennial Exercise Inspection Report 05000220/2023501 and 05000410/2023501 IR 05000220/20230112023-10-16016 October 2023 Comprehensive Engineering Team Inspection Report 05000220/2023011 and 05000410/2023011 RS-23-097, Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans2023-10-12012 October 2023 Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans NMP1L3554, Submittal of Revision 28 to the Final Safety Analysis Report (Updated), Fire Protection Design Criteria Document, 10CFR50.59 Evaluation Summary Report, 10CFR54.37(b) Aging Management Review, and Technical Specifications with Revised Bases2023-10-0606 October 2023 Submittal of Revision 28 to the Final Safety Analysis Report (Updated), Fire Protection Design Criteria Document, 10CFR50.59 Evaluation Summary Report, 10CFR54.37(b) Aging Management Review, and Technical Specifications with Revised Bases C IR 05000220/20233032023-09-20020 September 2023 Retake Operator Licensing Examination Report 05000220/2023303 ML23250A0822023-09-19019 September 2023 Regulatory Audit Summary Regarding LARs to Adopt TSTF-505, Rev. 2, and 10 CFR 50.69 ML23257A1732023-09-14014 September 2023 Requalification Program Inspection IR 05000220/20230052023-08-31031 August 2023 Updated Inspection Plan for Nine Mile Point Nuclear Station, Units 1 and 2 (Report 05000220/2023005 and 05000410/2023005) RS-23-080, Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs2023-08-30030 August 2023 Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs 2024-08-09
[Table view] Category:Safety Evaluation
MONTHYEARML24145A1862024-05-30030 May 2024 Authorization and Safety Evaluation of Alternative Relief Request I5R-12 Concerning the Installation of a Full Structural Weld Overlay on RPV Recirculation Inlet Nozzle N2E ML23305A1402023-12-13013 December 2023 Units 1 & 2; Nine Mile Point, Unit 2; Peach Bottom, Units 2 & 3; and Quad Cities, Units 1 and 2 - Issuance of Amendments to Adopt Traveler TSTF-580 ML23291A4642023-12-0707 December 2023 Issuance of Amendment No. 251 Regarding the Adoption of Title 10 the Code of Federal Regulations Section 50.69, Risk-Informed Categorization and Treatment of SSC for Nuclear Power Plants ML23289A0122023-12-0606 December 2023 Issuance of Amendment No. 250 Regarding the Revision to Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b ML23151A3472023-08-21021 August 2023 Issuance of Amendments to Adopt TSTF-295-A, Modify Note 2 to Actions of PAM Table to Allow Separate Condition Entry for Each Penetration ML23131A4242023-06-23023 June 2023 Issuance of Amendment No. 249 Regarding the Revision to Technical Specification 3.3.1 to Adopt Technical Specifications Task Force Traveler TSTF-568 ML23156A6832023-06-22022 June 2023 Authorization and Safety Evaluation of Alternative Relief Request I5R-14 ML23156A6822023-06-22022 June 2023 Authorization and Safety Evaluation of Alternative Relief Request I5R-11 ML23114A2522023-04-28028 April 2023 Request to Use a Provision of a Later Edition of the ASME Boiler & Pressure Vessel Code, Section XI ML23081A0382023-04-25025 April 2023 County, 1 & 2; Nine Mile Point, 2; and Quad Cities, 1 & 2 - Issuance of Amendments to Adopt TSTF-306, Rev. 2, Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration ML23103A1912023-04-20020 April 2023 Request for Threshold Determination Under 10 CFR 50.80 ML23082A3502023-04-17017 April 2023 Issuance of Amendment No. 192, Revision of Surveillance Requirements Associated with Emergency Diesel Generator Testing ML23094A1422023-04-0505 April 2023 Approval of Alternative Request I5R-13 to Utilize Specific Provisions of Code Case N-716-3 ML23025A4122023-02-28028 February 2023 Issuance of Amendment No. 248 to Revise Alternative Source Term Calculation for Main Steam Isolation Valve Leakage and Non-MSIV Leakage ML22339A0582022-12-21021 December 2022 Relief Request Associated with Pump Periodic Verification Tests of Core Spray System Pumps ML22293B8052022-11-30030 November 2022 Constellation Energy Generation, Llc_Fleet - Request to Authorize Use Honeywell Mururoa V4F1 R Supplied Air Suits ML22090A0862022-04-29029 April 2022 Amendments to Adopt TSTF-541,Rev.2,Add Exceptions to Surveillance Requirements for Valves,Dampers Locked in Actuated Position ML22094A0012022-04-15015 April 2022 Constellation Energy Generation, LLC - Proposed Alternative for Repair of Water Level Instrumentation Partial Penetration Nozzles (Epids L-2021-LLR-0057 and L-2021-LLR-0058) ML22061A0402022-03-11011 March 2022 Relief Request Associated with Excess Flow Check Valves ML22033A3102022-03-0404 March 2022 Issuance of Amendment No. 190, Changes to Reactor Pressure Vessel Water Inventory Control Technical Specification Requirements ML21347A0382022-01-13013 January 2022 Issuance of Amendments to Revise Reactor Coolant Leakage Requirements ML21277A2482021-11-16016 November 2021 Letter with Enclosure 4, Safety Evaluation for Transfer of Licenses and Draft Conforming License Amendments (Public Version) ML21295A7342021-11-15015 November 2021 Issuance of Amendment No. 187 Adoption of Technical Specification Task Force (TSTF) Traveler TSTF-501, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control ML21256A1792021-11-0202 November 2021 Issuance of Amendment No. 246 Changes to Reactor Pressure Vessel Water Inventory Control Technical Specification Requirements ML21280A0782021-10-27027 October 2021 Proposed Alternative to Use of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-893 (Epids L-2020-LLR-0147 and L-2020-LLR-0148) ML21230A2062021-09-0303 September 2021 Proposed Alternative to Use ASME OM Code Case OMN-28 ML21216A2202021-08-0505 August 2021 Proposed Alternative to Eliminate Certain Documentation Requirements for Pressure Retaining Bolting ML21166A1682021-06-25025 June 2021 ML21140A1532021-06-0303 June 2021 Correction to April 22, 2021 Safety Evaluation for Requests for Alternative to the Inservice Testing Requirements of the ASME OM Code for the Fifth 10-Year Program Interval ML21082A2212021-04-29029 April 2021 Issuance of Amendment No. 186, Adoption of TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - 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CONTROL NO. 8857 FILE F ROM. Niagara Mohawk Power DATE OF DOC DATE R EC'D LTR TWX RPT OTHER Corp. Syracuse, N.Y. 13 02 8-8-75 8-20-75 TO: ORIG CC OTHER SENTNRC PDR Mr. George Lear 1 signed SENT LOCAL PDR CLASS UNCLASS PROPINFO INPUT NO CYS REC'D DOCKET NO:
50<<220 DESCRIPTION: L'tr re our 7-8-75 submittal... ~ ENCLOSURES: Safety Evaluation of Additional trans the following: Spent Fuel Storage Racks k for Nine Mile Pt.
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@Qg P4 August 8, 1975 Mr. George E. Lear, Chief Operating Reactors Branch I/3 Division of Reactor Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Re: Nine Mile Point Unit 1 Docket No. 50-220
Dear Mr. Lear:
The enclosed submittal is in response to your letter of July 8, 1975. That letter requested a conclusion on whether the planned modification to the spent fuel pool constitutes an unreviewed safety question and the basis therefore. We have concluded that the planned addition of spent fuel racks to the Nine Mile Point Unit 1 spent fuel pool, as described in our submittal of June 10, 1975, does not constitute an unreviewed safety question pursuant to 10 CFR Part 50, Paragraph 50.59.
The Site Operations Review Committee and the Safety Review and Audit Board have reviewed this safety evaluation and concur with its conclusions.
Additional information in support of this conclusion is attached.
No licensing action is requested. This material is supplied for your information.
Very truly yours, NIAGARA MOHAWK POWER CORPORATION Geral IC. ode Vice President-En ering NLR/sz Enclosure
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regulatory Docket SAFETY EVALUATION OF ADDITIONAL SPENT FUEL STORAGE RACKS NINE NILE POINT UNIT 1 I.
SUMMARY
- n. ~Ei i O*i The spent fuel storage pool is a reinforced concrete structure lined with stainless steel plate. The pool is 33 feet 2 inches wide, 37 feet 5 1/2 inches long, and 38 feet 10 inches in depth. It was designed specifically to maintain the mean temperature of the pool below 125 F. In so doing, it was originally considered that the
'pool should accommodate full core discharge capability and the associated maximum heat load. The spacing of the fuel bundles is maintained such that the Effective Multiplication Factor (k ff) is always less than 0.90.
B. Descri tion of llodification Up to seventeen new spent fuel storage racks will be installed in the spent fuel pool. Our June 10, 1975 submittal analyzed an addition of twelve spent fuel racks to the spent fuel pool. How-ever, the original design allowed for spacing of up to seventeen racks. Therefore, we have re-analyzed to accommodate the addition of seventeen spent fuel racks. The seventeen racks will provide capacity for an additional 340 spent fuel bundles. To accommodate the new racks, thirteen control rod racks will be relocated to an aisle area, where they will be bolted to a support base, and braced laterally for seismic restraint.
These additional racks will be placed in locations provided in the original design and construction. Swing bolt assemblies to accommodate these additional racks were installed in the pool during original construction.
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C. Pu'r ose of Modification The spent fuel storage pool, with the additional seventeen storage racks, will provide sufficient capability to store a full core discharge should the need arise following the Fall, 1975 refueling.
Additional racks are required due to the recent removal of six racks for installation of the Cask Drop Protection System, and also because of the unavailability of off-site storage and fuel reprocessing services.
4 D. Results of Safet Evaluation The Safety Evaluation demonstrates that the planned addition of spent fuel storage racks does not constitute an unreviewed safety question pursuant to 10 CFR Part 50, paragraph 50.59:
'. Neither the probability of occurrence nor the consequences of an accident or malfunction of equipment important to safety will be changed.
- 2. No possibility for an accident or malfunction of a different type than previously evaluated has been created.
- 3. No margin of safety defined in the basis for any technical specification has been reduced.
II . DESIGN CONS I DERATIONS A. ~0 Table 1 outlines the past, present and future condition of the equipment in the pool based upon this submittal. The seventeen racks identified as "New" in Figure 1 will be identical to those existing and will be installed on existing swing bolt mountings on the pool floor. The thirteen control rod racks will be placed in the aisle area as shown in Figure 1. There, they will be bolted to a support base braced lateral'ly for seismic restraint.
B. Seismic and Structural Considerations
- 1) Static -'The original design accounted for the weight loadings associated with the additional racks'n that the entire floor was assumed to be supporting fuel including the weight of the cask in the northwest corner.
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- 2) Dynamic - As in the static design, the original dynamic design of a horizontal earthquake of 0.25g included the. pool filled with spent fuel plus the cask. The control rod racks, relocated to the aisle of the pool as proposed will not over-stress the supports of the adjacent spent fuel racks.
'. Eft'ective Multi lication Factor (k ff)
The original calculations which limit kerf to less than 0.9 are based on the spacing of the spent fuel storage racks and not the
'number. The racks being added are designed and will be placed in the pool such that spacing is identical to that of the original racks. Therefore, keff will remain unchanged.
D. ~Shi el din Presently the reactor floor area is a controlled "Radiation Area".
As defined in the Station's radiation protection procedures, a "Radiation Area" is one where the radiation level is from 5. mrem/hr to 100 mrem/hr. Presently the dose level around the spent fuel pool area is about 5 mrem/hr, largely a result of the radioactivity in the spent fuel pool water. The increase in the number of spent fuel pool bundles may increase this level slightly but will not cause the control of this area to be changed.
E. U-235 and B -Product Inven'tor Operating License DPR-63 allows for the possession of U-235 and the associated by-products from reactor operation. Calculations show that even after the Spring, 1977 refueling including the increased number of spent fuel bundles, less than 3800 kg of U-235 will be located at the Nine Mile Point Unit 1 facility.
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F. S ent Fuel Pool Coolin The maximum pool heat input under normal conditions will be less than 9 million BTU per hour. Only one of the two pool filtering and cooling systems will be required to remove this heat and maintain bulk pool temperature at or below 125 F.
Normal pool heat input is based on storage of 500 spent fuel bundles (94 percent of core) with decay as follows:
300 bundles (56 percent of core) with one year or more decay 200 bundles (38 percent of core) with twelve days decay The maximum pool heat input with 1140 fuel bundles in the spent fuel pool will be 27.3 million BTU per hour. Both pool filtering and cooling systems will be required to remove thi s heat and maintain bulk pool temperature at or below 125 F. If one pool cooling loop becomes inoperable, pool bulk temperature will not exceed 150 F. Because the fuel must be irradiated to generate decay heat, the worst case heat load is assumed to occur if the reactor is shutdown seven days after startup. Twelve days was assumed as the time necessary to unload the core into the spent fuel pool.
III. ACCIDENT ANALYSIS A. Seismic Event Because all of the new fuel racks will be installed onto existing mountings, the seismic adequacy of. the mountings will be unaffected.
The additional racks would 'not affect the iritegrity of the pool in the design event. As discussed in IIB above, both the static and dynamic loadings due to this modification wi'jl not overstress any structural components in the spent fuel pool.
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B. S ent Fuel Cask Dro Over Pool The cask, drop protection system precludes the drop of a spent fuel cask onto spent fuel in the pool. Further, this system prevents damage to the pool and its contents by guiding and decelerating the cask, should it be dropped. Operation of the cask drop protection system is independent of the number of fuel racks or the amount of fuel in the pool. Therefore, the probability of such an accident and the consequences thereof would not be affected by the addition of spent fuel racks. (Reference letters - Hay 31, 1973, P. D.
Raymond to D. L. Ziemann and September 29, 1972, T. J. Brosnan to J. F. O'eary.)
'. Refuelin Accident As discussed in Appendix E to the FSAR, the Second Supplement to the FSAR, the Technical Supplement to Petition to Increase Power Level, and
. Amendment No. 1 to Application to Convert Provisional Operating License to Full Term Operating License, the most severe refueling accident remains the drop of a fuel bundle over the reactor core. The .
additional spent fuel storage racks will not alter this conclusion.
The offsite dose resulting from this accident, or from dropping a fuel bundle into the spent fuel pool will not be changed. The probabilities of these accidents will not be affected.
D. S ent Fuel Pool Coo'lin Loss of coolant from the pool is precluded because all penetrations are located at least one foot above the top of the fuel. The pool cooling water pumps, and the makeup water valve would be powered from the emergency diesel generators in the event of loss of offsite power.
As discussed above, the normal pool heat load requires operation of only one of the two cooling systems. Should one cooling system become unavailable, bulk pool temperature would not ex'ceed 150 F.
As discussed in the Fourth Supplement to the FSAR, the large volume of pool water and the availability of makeup ensure a slow rate of temperature rise. In the unlikely event that both cooling systems become unavailable, continuous availability of makeup water to the pool is assured. Sufficient time would be available to repair the postulated malfunction.
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The additional spent fuel racks will not affect the probability of a cooling system malfunction. In such an event, adequate measures would be available to provide or restore cooling such that safe spent fuel storage is maintained.
IV. TECHNICAL SPECIFICATIONS The applicable Technical Specification is Specification 5.5, which requires a keff less than 0.9. As discussed in Section IIC above, the modification will not affect keff.
V. CONCLUSIONS The Safety Evaluation demonstrates that the planned addition of spent fuel storage racks does not constitute an unreviewed safety question pursuant to 10 CFR Part 50, paragraph 50.59 because:
- l. As discussed above, all systems, associated with the spent fuel pool will perform their design function without exceeding their originally assigned margins of conservatism. Therefore, the probability o' occurrence and the consequence of an accident or malfunction of equipment important to safety has not been changed.
- 2. The addition of spent fuel racks does not affect the probability of occurrence or consequences of any accidents discussed in the FSAR or supplements thereto. This is discussed in Section III above.
Also, no accident or malfunction. of a different type than previously evaluated in the FSAR has been created,
- 3. The conservative margins built into the design of equipment important to maintain safe operation of the spent fuel pool will not be exceeded. Both Keff and the maximum pool temperature have not been altered, Therefore, the margin of safety defined in the basis for all technical specifi-cations has not changed.
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TABLE 1 Pool Conditions Mith Pro osed Addition Future Deleted For Post 1975 Capaci Ori inal Desi n Cask S stem Present Condition ~Refuel in Tota i Spent Fuel Racks 44 40 57 Spent Fuel Bundles 880 120- 300 500 1140 Control Rod Racks 14 13 13 13 Control Rods 140 130 Channel Racks Fuel Channels 120 293 20
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