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Pages in category "Report"
The following 200 pages are in this category, out of 12,338 total.
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- 0CAN010503, Special Report
- 0CAN020305, Fitness-for-Duty Performance Data for Period July - December 2002
- 0CAN020403, Transmittal of Fitness-for-Duty Performance Data for Period Ending December 31, 2003
- 0CAN020505, ANO, Units 1 and 2 - Transmittal of Fitness-for-Duty Performance Data for Period July - December 2004
- 0CAN020602, Fitness-for-Duty Performance Data for the Period July - December 2005
- 0CAN021204, Units 1 and 2, Third Five-Year Surveillance of the First Ventilated Storage Cask
- 0CAN021403, Units 1 and 2 - 10 CFR 50.59 Summary Report and Commitment Change Summary Report
- 0CAN021501, Spent Fuel Storage Radioactive Effluent Release Report for 2014
- 0CAN030504, Justification for ANO Exemption Request for Loading of Damaged Fuel
- 0CAN030701, Drug Testing Laboratory Performance Report
- 0CAN031001, Units 1 & 2, Unsatisfactory Laboratory Testing Report
- 0CAN031404, ANO, Units 1 & 2 - Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (Nttf)Review of Insights from the Fukushima
- 0CAN040505, Units 1 and 2, Drug Testing Laboratory Performance Report
- 0CAN050903, Annual 10 CFR 50.46 Report for Calendar Year 2008 Emergency Core Cooling System Evaluation Changes
- 0CAN051704, Engineering Report No. CALC-ANOC-CS-14-00017, Rev 0, 2017 Focused Evaluation for External Flooding at Arkansas Nuclear One.
- 0CAN052102, Annual 10 CFR 50.46 Report for Calendar Year 2020 Emergency Core Cooling System Evaluation Changes
- 0CAN061701, Transmittal of 10 CFR 50.71(e) Report, Revision 17 of the ANO Fire Hazards Analysis
- 0CAN080401, Fitness-for-Duty Performance Data for the Period January - June 2004
- 0CAN080601, Arkansas - Fitness-for-Duty Program Performance Data for the Period January - June 2006
- 0CAN081703, Document 51-9257562-001, Revision 1, Arkansas Nuclear One Hfe - High Frequency Confirmation Report
- 0CAN100901, Units 1 & 2, Unsatisfactory Laboratory Testing Report
- 0CAN101003, 10 CFR 50.59 Summary Report and Commitment Change Summary Report
- 0CAN102102, Units 1 and 210 CFR 50.71(e) Report Revision 20 of the ANO Fire Hazards Analysis
- 0CAN120202, CFR 50.59 Summary Report
- 0CAN121602, Mitigating Strategies Assessment (MSA) Report for the New Seismic Hazard Information Per Nuclear Energy Institute (NEI) 12-06, Appendix H, Revision 2, H.4.3 Path 3
- 0CAN121901, Summary of Lost Specimens Investigation Report
1
- 1CAN010301, Arkansas, Unit 1, Once Through Steam Generator Inservice Inspection Report
- 1CAN031305, Cycle 24 COLR, Revision 6
- 1CAN032001, Supplemental Information Related to License Amendment Request to Revise Loss of Voltage Relay Allowable Values
- 1CAN040302, License Amendment Request to Modify the Fuel Assembly Enrichment, the Spent Fuel Pool (SFP) Boron Concentration TS 3. 7.14, the Loading Restrictions in the SFP in TS 3,7,15m and to Modify the Fuel Storage Design Features in TS 4.3
- 1CAN041105, Request for Use of Non-ASME Code Repair to Service Water Piping in Accordance with Generic Letter 90-05 Relief Request ANO1-R&R-016
- 1CAN051401, Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 Years Arkansas Nuclear One Unit 1
- 1CAN062302, Enclosure 2: ANO-1 SAR Amendment 31 - Redacted Version
- 1CAN062304, Supplement Related to License Amendment Request to Remove Technical Specification Condition Allowing Two Reactor Coolant Pump Operation
- 1CAN091001, 10 CFR 50.46 Report - Significant Change in Peak Cladding Temperature
- 1CAN091301, Updated Seismic Walkdown Report
- 1CAN091601, Submittal of Initial Examination Completion of Post-Examination Analysis
- 1CAN100203, Supplemental Response to NRC Bulletin 2002-02 for Arkansas Nuclear One, Unit 1
- 1CAN110203, Response to NRC Request for Additional Information Regarding NRC Bulletin 2002-01 for ANO-1 Incore Instrument Nozzles
- 1CAN111201, Engineering Report CALC-ANO1-CS-12-00002, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Pages 1 Through 331
- 1CAN111201, Engineering Report CALC-ANO1-CS-12-00002, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Pages 332 Through 560
2
- 2CAN010304, Arkansas, Unit 2, License Amendment Request to Change Spent Fuel Pool Loading Restrictions
- 2CAN011003, Submittal of Us Dept. of Commerce, Bureau of Industry and Security. Additional Protocol Report
- 2CAN011202, Additional Protocol Report
- 2CAN011703, Submittal of Additional Protocol Report
- 2CAN022202, Requests for Relief from American Society of Mechanical Engineers Section XI Volumetric Examination Requirements - Fourth 10-Year Interval, Second and Third Periods
- 2CAN040801, Summary of Design and Analyses of Weld Overlays for Hot Leg Nozzle Dissimilar Metal Welds for Alloy 600 Mitigation at ANO-2
- 2CAN040806, 60-Day Report for ANO-2 Reactor Pressure Vessel Head Inspection for Refueling Outage 2R19
- 2CAN060504, Unit 2, Submittal of Report to Provide Information Re Replacement Valve Disc Manufactured by Crane-Aloyco
- 2CAN062103, Request for Alternative ANO2-PT-003 End-of-Interval System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping - Fifth Interval
- 2CAN070804, Cycle 20 Startup Report
- 2CAN091103, CFR 50.59 Summary Report
- 2CAN091302, Updated Seismic Walkdown Report
- 2CAN100802, CFR 50.59 Summary Report for Period Ending October 6, 2008
- 2CAN110502, CFR 50.59 Summary Report
- 2CAN111201, Arkansas, Unit 2, Attachment 1 - Engineering Report CALC-ANO2-CS-12-00001, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, and Attach. 2, List of Regulatory Commitments, Cover
- 2CAN111201, Arkansas, Unit 2, Attachment 1 - Engineering Report CALC-ANO2-CS-12-00001, Rev. 0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Page 364 of 533 Through End
3
- 3F0103-03, Response to Request for Additional Information, Bulletin 2002-01, Reactor Pressure Vessel Degradation & Reactor Coolant Pressure Boundary Integrity
- 3F0112-04, Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR
- 3F0113-08, Attachment D: ANP-3195(NP), Revision 0, Response for Crystal River Unit 3, EPU Licensing Amendment Report NRC Reactor Systems Branch Requests for Additional Information (Non-Proprietary) and Attachment E: Location of Reactor Systems RAI Re
- 3F0119-01, Reference 5 - EPA-600-R-07-020, Performance of Statistical Tests for Site Versus Background Soil Comparisons When Distributional Assumptions Are Not Met.
- 3F0320-01, NRC Commitment Change Report - March 2020
- 3F0508-14, NRC Commitment Change Report - May 2008
- 3F0511-02, Response to Request for Additional Information Required for the Development of the Confirmatory LOCA and Non-LOCA Models
- 3F0512-01, NRC Commitment Change Report - May 2012
- 3F0518-03, Safety Analysis Report and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2018
- 3F0520-01, Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2020
- 3F0522-01, Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2022
- 3F0602-05, Response to Request for Additional Information 263, Revision 0, Relocation of Reactor Coolant System Parameters to Core Operating Limits Report & 20 Percent Steam Generator Tube Plugging
- 3F0616-02, Nrg Commitment Change Report - June 2016
- 3F0623-02, Maintenance Support Building
- 3F0702-11, Reporting Required by Environmental Protection Plan
- 3F0712-03, Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip.
- 3F0902-09, Submittal of Core Operating Limits Report, Cycle 13, Revision 1, for Crystal River Unit 3
- 3F0910-01, CFR 50.46 Notification of Change in Peak Cladding Temperature for Small Break Loss of Coolant Accident Analysis
- 3F0912-01, ANP-3156 Np, Crystal River 3 EPU Boric Acid Precipitation RAI Responses, Attachment C
- 3F1011-08, ANP-3052, Rev. 0, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip
- 3F1102-08, Sea Turtle Mortality Report Submitted to the U. S. National Marine Fisheries Service
- 3F1107-06, CFR 50.46 Loss-of-Coolant Accident Evaluation Model Change and Peak Cladding Temperature Change Report
- 3F1112-01, Alion Technical Report ALION-PLN-ENER-8706-02, Rev. 0, Crystal River 3: Bypass Fiber Quantity Test Plan
- 3F1112-02, 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 174 of 250 Through Page 250 of 250
- 3F1205-03, Special Report 05-01: Once-Through Steam Generator (OTSG) Notifications Required Prior to Mode 4
- 3F1205-04, 10 CFR 50.46 Loss-Of-Coolant Accident Evaluation Model Change and Peak Cladding Temperature Change Report
- 3F1207-03, Engineering Report ER-608NP, Revision 2, LEFM + Meter Factor Calculation and Accuracy Assessment for Crystal River Unit 3 Nuclear Power Station.
- 3F1209-11, 10 CFR 50.46 Loss-of-Coolant Accident Evaluation Model Change and Peak Cladding Temperature Change Report
- 3F1211-14, Summary of Changes to Evaluation Models and Peak Cladding Temperature for Large Break Loss of Coolant Analysis and Small Break Loss of Coolant Analysis
A
- A000412, National Pollutant Discharge Elimination System (NPDES) Permit Renewal Application
- A25396, Msre Tru Waste Determination LA-UR-10-07278
- AEP-NRC-2008-11, Completion of Commitment Regarding Small Break Loss-of-Coolant Accident Analysis 8.75-Inch Case
- AEP-NRC-2009-25, Small Break Loss-of-Coolant Accident Evaluation Model Reanalysis
- AEP-NRC-2009-70, Revised Technical Justification for Deviation from EPRI MRP-139 Inspection Requirements for Reactor Vessel Alloy 600/82/182 Welds at DC Cook Nuclear Plant
- AEP-NRC-2012-38, Response to Request for Information, 10 CFR 50.46 Report for Emergency Core Cooling System Model Change or Error Associated with Thermal Conductivity Degradation
- AEP-NRC-2012-78, Final Report Kld TR-488, Revision 1, Development of Evacuation Time Estimates, Table K-1. Evacuation Roadway Network Characteristics
- AEP-NRC-2012-83, Communications Assessment Requested by Nuclear Regulatory Commission Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulation 50.54(f) Regarding Recommendation 2.1, 2.3, and 9.3.
- AEP-NRC-2012-86, Flooding Walkdown Report in Response to the 50.54(F) Information Request Regarding Near-Term Task Force Recommendation 2.3: Flooding for the D.C. Cook Nuclear Power Plant
- AEP-NRC-2012-87, Seismic Walkdown Report SD-121023-001, Rev. 0 in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Appendix C, Page C-437 Through C-486
- AEP-NRC-2013-07, SD-121023-001, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Page D-405 Through End
- AEP-NRC-2013-74, SD-121023-001, Rev. 1, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Page C-99 Through Page C-198
- AEP-NRC-2014-08, SD-121023-001, Rev. 2, in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the D. C. Cook Unit 1 and 2, Page D-322 Through Page D-404
- AEP-NRC-2014-15, 30 Day Report of Changes to or Errors in an Evaluation Model
- AEP-NRC-2014-59, Enclosure 3: Babcock & Wilcox Report, S-1473-002, Revision 0, Examination of Clevis Bolts Removed from D. C. Cook Nuclear Plant. Part 3 of 3
- AEP-NRC-2015-83, Revision 1 of Final Integrated Plan Regarding March 12, 2012, NRC Order Regarding Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)
- AEP-NRC-2018-21, 30-Day Report of Changes to or Errors in an Evaluation Model
- AEP-NRC-2018-36, Notification of Initial Renewable Operating Permit
- AEP-NRC-2020-23, Request for Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography
- AEP-NRC-2020-28, CFR 72.48(d)(2) Summary Report of Completed Changes, Tests, and Experiments 1O CFR 72.48 Evaluations
- AEP-NRC-2021-07, Supplement to Report Per Technical Specification 5.6.6, Lnoperability of Unit 1, Post Accident Monitoring, Containment Water Level
- AEP-NRC-2021-44, Form OAR-1, Owner'S Activity Report
- AEP-NRC-2022-46, Notification of Deviation from Electric Power Research Institute (EPRI) Materials Reliability Program 2019-008, Interim Guidance for NEI 03-08 Needed Requirements for Us PWR Plants for Management of Thermal Fatigue in
- AEP-NRC-2022-58, U1C31 Steam Generator Tube Inspection Report
- AEP-NRC-2022-66, Report Per Technical Specification 5.6.6 for Inoperability of Unit 2 Post Accident Monitoring Neutron Flux Monitoring
B
- BSEP 02-0078, Occupational Radiation Exposure Report for 2001 from Carolina Power & Light Co
- BSEP 02-0151, Extended Power Uprate Implementation Test Report - Phase 1
- BSEP 02-0186, Response to Request for Additional Information, Proposed License Amendment to Revise Pressure - Temperature Curve Limits
- BSEP 05-0018, Occupational Radiation Exposure Report for 2004
- BSEP 05-0079, Inservice Inspection Program for the Third Year Interval-Refueling Outage 16 Owner'S Activity Report
- BSEP 05-0081, Extended Power Uprate, Phase 2 Implementation Test Report
- BSEP 05-0103, Special Report - Technical Requirements Manual Section 3.4, Accident Monitoring Instrumentation
- BSEP 05-0110, Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors
- BSEP 05-0137, Core Operating Limits Report, Revision 2, Cycle 15
- BSEP 06-0075, Groundwater Questionnaire
- BSEP 06-0095, Report of 10 CFR 50.59 Evaluation and Commitment Changes
- BSEP 07-0067, BSEP 07-0067 Enclosure 9; Areva Report ANP-2624(NP), Revision 0, Brunswick, Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM-1O Fuel, Dated June 2007
- BSEP 07-0075, Areva Report ANP-2638NP, Revision 0, Applicability of Areva Np BWR Methods to Extended Power Uprate Conditions, Enclosure 6
- BSEP 07-0102, ANP-2661 (Np), Revision 0, Brunswick Nuclear Plant New Fuel Storage Vault Criticality Safety Analysis for ATRIUM-10 Fuel.
- BSEP 08-0121, Submittal of Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors
- BSEP 09-0022, Special Report - Technical Requirements Manual Section 3.4, Accident Monitoring Instrumentation
- BSEP 09-0034, Enclosure 10 to BSEP 09-0034 - ANP-2771(NP), Rev. 0, Brunswick, Unit 2 Cycle 19 Reload Safety Analysis, Dated January 2009
- BSEP 10-0097, Report of 10 CFR 50.59 Evaluations and Commitment Changes
- BSEP 10-0118, ANP-2950NP, Revision 0, Atrium 10XM Fuel Rod Thermal and Mechanical Evaluation for Brunswick Unit 2 Cycle 20 Reload BRK2-20.
- BSEP 10-0141, ANP-2978NP, Rev. 0, Areva Responses to RAIs on the Atrium 10XM Compliance Audit and Brunswick Lars.
- BSEP 11-0031, Areva Report ANP-2992NP, Revision 0, Areva Response to Additional RAI on the Brunswick RODEX4 LAR
- BSEP 11-0038, Renewed Facility Operating License and Cycle 20 Core Operating Limits Report
- BSEP 12-0031, Areva Document ANP-3086(NP), Brunswick Units 1 and 2 SLMCPR Operability Assessment Critical Power Correlation for Atrium 10XM Fuel - Improved K-factor Model, Enclosure 20 to BSEP 12-0031
- BSEP 12-0040, ANP-3061(NP), Revision 0, Brunswick, Unit 1, Cycle 19 Reload Safety Analysis.
- BSEP 12-0127, Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident
- BSEP 12-0128, Special Report - Technical Requirements Manual Section 3.4, Accident Monitoring Instrumentation
- BSEP 12-0133, Document No. 51-9195058-000, Areva Input to Process Energy: Follow-Up to SAFLIM3D LAR Request for Additional Information 3, Enclosure 3 to BSEP 12-0133
- BSEP 13-0002, Technical Data Record, 12-9197120-000, Brunswick SAFLIM3D LAR Support - Calculation Error Impact
- BSEP 13-0030, Overall Integrated Plan in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events
- BSEP 14-0007, 1707-01-F03, Rev. 3, Operability Assessment (CR2013-8241)
- BSEP 14-0028, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi A
- BSEP 14-0093, Report of 10 CFR 50.59 Evaluations and Commitment Changes
- BSEP 14-0101, Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors
- BSEP 14-0131, Expedited Seismic Evaluation Process Report in Response to 10 CFR 50.54(f) Request for Information Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident
- BSEP 15-0004, Enclosure 3 - Areva Operability Assessment. (CR 2014-7395)
- BSEP 16-0056, ANP-3108(NP), Revision 1, Applicability of Areva BWR Methods to Brunswick Extended Power Flow Operating Domain.
- BSEP 17-0069, Seismic Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Appendix H, Revision 2, H.4.4 Path 4:GMRS ≪ 2xSSE
- BSEP 18-0044, Application of Dissimilar Metal Weld Full Structural Overlay - Reactor Pressure Vessel Nozzles N4A and N4D
- BVY 03-069, Fitness-for-Duty Program Performance Report for Period January - June 2003
- BVY 04-088, Fitness-for-Duty Program Performance Report for the Period January - June 2004
- BVY 05-018, Fitness-for-Duty Program Performance Report for the Period July - December 2004
- BVY 05-030, Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263 - Supplement No. 25, Extended Power Uprate - Station Blackout and Appendix R Analyses
- BVY 05-066, Comments Regarding License Amendment No. 223
- BVY 05-078, Fitness-for-Duty Program Performance Report for the Period January - June 2005
- BVY 05-089, Entergy Nuclear Northeast - Proof of Financial Protection
- BVY 06-016, Entergy Nuclear Northeast - Fitness-for-Duty Program Performance Report for the Period July 2005 - December 2005
- BVY 06-019, Vermont Yankee, Extended Power Uprate - Regulatory Commitment Information Regarding Steam Dryer Monitoring and Fiv Effects
- BVY 06-025, Regulatory Commitment Change Schedule for Installation of Capacitor Banks
- BVY 06-031, Revision 1 to Steam Dryer Monitoring Plan
- BVY 06-033, Cycle 24 10 CFR 50.59 Report
- BVY 06-068, Cycle 25 Startup Test Report Following Power Ascension and Testing
- BVY 07-004, Emergency Response Data System (ERDS) Data Point Library Update
- BVY 07-010, BVY-07-010 EPRI FAC Program Information; 05000271/2007-006
- BVY 07-023, NEDC-33291, Revision 0, GNF2 Lead Use Assembly (Lua) for Vermont Yankee Plant.
- BVY 07-051, Cycle 25 10CFR50.59 Report, Performed Between November 12, 2005 and June 6, 2007
- BVY 07-079, Update of Aging Management Program Audit Q&A Database
- BVY 08-010, Submittal of Report, Decommissioning Cost Analysis, Pursuant to 10CFR50.75(f)(3)
- BVY 11-021, NEDO-33618, Revision 0, Vermont Yankee Core Plate Bolt Stress Analysis Report, Attachment 2 to Bvy 11-021
- BVY 12-080, Engineering Report VY-RPT-12-00019, Vermont Yankee Seismic Walkdown Submitted Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Page 205 of 505 Through Page 464 of 505
- BVY 12-080, Engineering Report VY-RPT-12-00019, Vermont Yankee Seismic Walkdown Submitted Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Page 465 of 505 Through End
- BVY 12-081, Flooding Walkdown Report - Entergy'S Response to NRC Request for Information Pursuant to 10CFR50.54(f) Flooding Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident
- BVY 13-061, VY-RPT-12-00019, Revision 1, Vermont Yankee Seismic Walkdown Submittal Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Cover Through Page 159 of 505 of Attachment C
- BVY 13-061, VY-RPT-12-00019, Revision 1, Vermont Yankee Seismic Walkdown Submittal Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Page 160 of 505 of Attachment C Through Page 369 of 505 of Attachment C
- BVY 13-061, VY-RPT-12-00019, Revision 1, Vermont Yankee Seismic Walkdown Submittal Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Page 370 of 505 of Attachment C Through 82 of 198 of Attachment D
- BVY 13-061, VY-RPT-12-00019, Revision 1, Vermont Yankee Seismic Walkdown Submittal Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Page 83 of 198 of Attachment D Through End
- BVY 13-090, Submittal of 10 CFR 71.95 Report Involving 8-120B Cask
- BVY 14-011, Revision to Flooding Walkdown Report - Entergy'S Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from Fukushima ...
- BVY 14-011, Vermont Yankee, Revision to Flooding Walkdown Report - Entergy'S Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from
- BVY 14-085, Update to Irradiated Fuel Management Program Pursuant to 10 CFR 50.54(bb)
- BVY 15-018, Transmittal of Biennial 10 CFR 50.59 Report
- BVY 15-046, Rev. 0 to Defueled Safety Analysis Report, Drawing G-200347, Circulating Water System Aerating Structure-MAS & Reinf.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report - List of Effective Pages
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing 5920-00526, Process Radiation Monitoring System
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing 5920-06245, Plan Showing Property Lines & Plan Site.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191142, Site Plot Plan.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191144, General Arrangement Turbine Building Ground Floor Plan.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191145, General Arrangement Turbine Building Operating Floor Plan.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191148, Rev. 23, General Arrangement Reactor Building Plans Sheet 1.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191149, Rev. 27, General Arrangement Reactor Building Plans Sheet 2.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191150, Rev. 20, General Arrangement Reactor Building Section.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191151, Rev 19, General Arrangement Radwaste Building Plans.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191152, Rev. 14, General Arrangement Radwaste Building Section.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191159, Rev. 100, Flow Diagram Service Water System. Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191159, Rev. 92, Flow Diagram Service Water System. Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 24, Flow Diagram Instrument Air System. Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 25, Flow Diagram Instrument Air System. Sheet 4
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 25, Flow Diagram Instrument Air System. Sheet 5
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 25, Flow Diagram Instrument Air System. Sheet 6
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 25, Flow Diagram Instrument Air System. Sheet 7