Category:Technical
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Pages in category "Technical"
The following 200 pages are in this category, out of 6,923 total.
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- 0CAN030504, Justification for ANO Exemption Request for Loading of Damaged Fuel
- 0CAN051704, Engineering Report No. CALC-ANOC-CS-14-00017, Rev 0, 2017 Focused Evaluation for External Flooding at Arkansas Nuclear One.
- 0CAN061701, Transmittal of 10 CFR 50.71(e) Report, Revision 17 of the ANO Fire Hazards Analysis
- 0CAN081703, Document 51-9257562-001, Revision 1, Arkansas Nuclear One Hfe - High Frequency Confirmation 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
1
- 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
- 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
- 1CAN091301, Updated Seismic Walkdown Report
- 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
- 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
- 2CAN062103, Request for Alternative ANO2-PT-003 End-of-Interval System Leakage Test for Extended Reactor Coolant Pressure Boundary Piping - Fifth Interval
- 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
- 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
- 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
- 3F0623-02, Maintenance Support Building
- 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.
- 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
- 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
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-2009-70, 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-2012-78, Final Report Kld TR-488, Revision 1, Development of Evacuation Time Estimates, Table K-1. Evacuation Roadway Network Characteristics
- 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-2014-59, 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-2020-23, Request for Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography
- 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-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 05-0079, 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 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 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 10-0118, 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 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 12-0133, 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 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 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
- BVY 04-088, Fitness-for-Duty Program Performance Report for the Period January - June 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 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 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 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 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
- 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 8
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 31, Flow Diagram Instrument Air System. Sheet 3
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 37, Flow Diagram Instrument Air System. Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191161, Rev. 22, Flow Diagram, Maxe-up Water Treatment System.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191163, Rev. 51, Flow Diagram Fire Protection System Inner Loop. Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191163, Rev. 51, Flow Diagram Fire Protection System Inner Loop. Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191164, Rev. 29, Flow Diagram Sampling System. Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191165, Rev. 46, Flow Diagram Sampling System. Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191173, Rev. 12, Flow Diagram Fuel Pool Cooling & Clean Up System. Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191173, Rev. 42, Flow Diagram Fuel Pool Cooling & Cleanup System. Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191177, Rev. 23, Flow Diagram Radwaste Systems. Sheet 3
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191177, Rev. 44, Flow Diagram Radwaste Systems. Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191177G, Rev. 28, Flow Diagram Radwaste Systems. Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191238, Rev. 38, HVAC - Flow Diagram Reactor Building.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191254, Rev. 43, HVAC Heating Flow Diagram & Boiler Room Layout.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191298, Rev. 5, Main One Line Wiring Diagram. Sheet 3
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191299, Rev. 37, 4KV Auxiliary One Line Diagram.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191300, Rev. 27, 480V Auxiliary One Line Wiring Diagram Swgr BUS-8 MCC-8A, 8C. Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191300, Rev. 38, 480V, Auxiliary One Line Diagram MCC-8B, MCC-8E, MCC-89B. Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191301, Rev. 30, 480V Aux. One Line Diagram Swgr BUS-9, MCC-9A, 9C. Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191301, Rev. 39, 480V Aux. One Line Diagram MCC-9B, 9D, 89A. Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191372, 120/240V Vital AC and Instrument AC One Line Diagram, Sheet 4
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191372, 125V DC One Line Diagram, Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191372, 125V DC One Line Diagram, Sheet 3
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191372, 24V DC Neutron Monitoring & 120V AC RPS One Line Diagram, Sheet 5
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191372, Rev. 72, 125VDC One Line Wiring Diagram. Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191451, Circulating Water System Intake Structure-M, Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191452, Circulating Water System Intake Structure-V, Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191453, Circulating Water System Intake Structure-M, Sheet 3
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191461, Circulating Water System Discharge Structure - Sections, Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191463, Circulating Water System Discharge Structure Reinf, Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191483, Reactor Building Foundation Mat Plan - M&R.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191529, Reactor Building Reactor Vessel Pedestal Mat - M&R.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191592, Control Room Building Fl Plan El 248.50 & Fdn Plan-R, Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191595, Control Room Bldg. Ext. Wall Elevs.-M.
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, DSAR Changes
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, DSAR Drawing List
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, G-191237, Rev. 12, HVAC - Flow Diagram Turbine, Service & Control Room Buildings. Sheet 2
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, G-191237, Rev. 47, HVAC - Flow Diagram Turbine, Service & Control Room Buildings. Sheet 1
- BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, UFSAR to DSAR Conversion Matrix
- BVY 20-026, Defueled Safety Analysis Report, Revision 2
- BW210047, ER-BR-330-1008, Revision 0, Snubber Program Plan for the Fourth 10-Year Interval
- BW210065, Pressure and Temperature Limits Report, Revision 8
- BW220062, Pressure and Temperature Limits Report (Ptlr), Revision 9
- BYRON 2002-0032, Cycle 12 Core Operating Limits Report, Revision 1 for Byron Station Unit 1
- BYRON 2003-0064, Revision to Reactor Coolant System Pressure & Temperature Limits Report
- BYRON 2004-0035, Regulatory Commitment Change Summary Report
- BYRON 2005-0071, Steam Generator Inservice Inspection Summary Report
- BYRON 2005-0077, Day Response to First Revised NRC Order EA-03-009, Issuance of First Revised NRC Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors.
- BYRON 2007-0051, Reactor Coolant System (RCS) Pressure and Temperature Limits Reports (PTLR)
- BYRON 2010-0001, Form NIS-2 Owner'S Report for Repair/Replacement Actitivity as Required by the Provisions of the ASME Code Section XI
- BYRON 2014-0003, Pressure and Temperature Limits Report (PTLR) for Measurement Uncertainty Recapture (Mur) Power Uprate
- BYRON 2014-0040, Pressure and Temperature Limits Report (PTLR) Revised for Negative Pressure Application During Reactor Coolant System Vacuum Fill
- BYRON 2020-0084, 10 CFR 72.48 Evaluation Summary Report
- BYRON 2020-0085, 10 CFR 50.59 Summary Report
- BYRON 2022-0071, Materials Reliability Program: Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement (MRP-335, Revision 3-A) Final Causal Report and Description of Corrective Action Assignment
C
- 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 Acc
- CNL-14-089, (Bfn), Units 1, 2, and 3 - Application to Modify Technical Specification 2.1.1, Reactor Core Safety Limits (BFN-TS-492)
- 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 Pl
- CNL-15-039, Severe Accident Management Alternatives for Reactor Coolant Pump Seals
- CNL-15-056, Application to Modify Technical Specification 2.1.1.2, Reactor Core Minimum Critical Power Ratio Safety Limits (TS-506)
- CNL-15-085, Response to NRC Request for Additional Information Regarding Proposed Technical Specification Change to Modify Technical Proposed Technical Specification Change to Modify Technical
- CNL-15-097, Flood Hazard Reevaluation Report for Watts Bar Plant, Response to NRC Request for Information Per 10CFR50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident
- CNL-15-143, the Tennessee Valley Authority (TVA) Nuclear Power Group Commercial Grade Dedication Recovery Project - Closure Report
- CNL-15-204, Response to NRC Question Regarding the Revised Containment Analysis, Including Enclosure 3, WCAP-17834-NP, Revision 2, Wcobra/Trac Long Term LOCA M&E and Containment Integrity Analysis and Enclosure 4, Affidavit
- 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 Revision
- CNL-17-050, Submission of Technical Reports to Support a Public Meeting Regarding the Transition to the Reactor Oversight Process for the Mitigating Systems Performance Index
- CNL-17-134, Transmittal of WCAP-18191-NP, Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations
- CNL-18-059, Submittal of Revised Site Safety Analysis Report Subsection 2.4.2.2, Flood Design Considerations, and Subsection 2.4.3.5, Probable Maximum Flood Flow, and Figure 2.4.3-3 in Support of Early Site Permit Application for Clinch River Nuclear
- CNL-18-112, Extended Power Uprate - Flow Induced Vibration Summary Report
- CNL-18-134, Extended Power Uprate - Replacement Steam Dryer Revised Analysis and Limit Curves 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 Conditions
- 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 Information
- CNL-19-041, Extended Power Uprate - Unit 1 Flow Induced Vibration Summary Report
- CNL-19-074, Extended Power Uprate - Flow Induced Vibration Summary Report
- CNL-19-082, License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)
- CNL-20-074, Submittal of Additional Supplement to License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)
- CNL-20-102, 10 CFR 71.95 Report for 3-60B Casks User
- CNL-21-011, Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002)
- CNL-21-018, Application to Adopt TSTF-205-A, Revision of Channel Calibration, Channel Functional Test, and Related Definitions, and TSTF-563-A, Revise Instrument Testing Definitions to Incorporate the Surveillance .
- CNL-21-040, Supplement to Expedited Application for Approval to Use a Growth Rate Temperature Adjustment When Implementing the Generic Letter 95-05 Analysis for the (WBN TS-391-21-002)
- CNL-21-055, Revision to Refueling Outage 3 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Final Report
- CNL-21-072, TVA Nuclear Calculation Coversheet/ Ecm Metadata Update
- 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 Use
- 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 Information
- 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)
- CNL-23-002, Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Sche
- CNL-24-016, Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14)
- CNRO-2002-00052, Proposed Alternative to ASME Examination Requirements for Repairs Performed on Reactor Vessel Head Penetrations
- CNRO-2003-00020, Arkansas, Unit 2 and Waterford, Unit 3, Relaxation Requests to NRC Order EA-03-009
- CNRO-2003-00030, Arkansas, Unit 2 and Waterford, Unit 3, Letter CNRO-2003-00027 to NRC, Relaxation Requests to NRC Order EA-03-009, Dated July 1, 2003
- CNRO-2003-00033, Rev. 1 to Engineering Report M-EP-2003-002, Fracture Mechanics Analysis for the Assessment of the Potential for Primary Water Stress Corrosion Crack Growth Un-Inspected Regions of the Control Element..., Appendix C, Attachment 47 Thr
- CNRO-2003-00038, Rev. 0 to M-EP-2003-004, Fracture Mechanics Analysis for the Assessment of the Potential for Primary Water Stress Corrosion Crack Growth Un-Inspected Regions of the Control Element Drive Mechanism At..., Appendix D, Attachment 5 Thro
- CNRO-2003-00049, Letter Transmitting Mark-Up of Engineering Report M-EP-2003-004, Rev. 0 Fracture Mechanics Analysis for the Assessment Potential for Primary Water Stress Corrosion Crack (PWSCC) Growth, Un-Inspected Regions..., Pages 43 Through 57
- CNRO-2004-00008, Rev. 0 to M-EP-2004-001, Fracture Mechanics Analysis for the Assessment of the Potential for Primary Water Stress Corrosion Crack Growth in the Un-Inspected Regions of the Control Rod Drive Mechanism Nozzles at ANO, Unit 1, App. C Th
- CNRO-2005-00048, Relaxation Request 5 to NRC First Revised Order EA-03-009 for the Control Element Drive Mechanism Nozzles
- CNRO-2006-00039, Results of the Waterford 3 Pressurizer Flaw Evaluation
- CNS-13-005, License Amendment Request for Methodology Report DPC-NE-3001-P, Revision 1, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology
- CNS-14-038, DUKCORP042-PR-001, Seismic Hazard & Screening Report in Response to the 50.54(f) Information Request Re Fukishima Near-Term Task Force Recommendation 2.1 Seismic
- CNS-14-129, Transmittal of Industry High Density Polyethylene (Hdpe) Test Results
- CNS-16-017, DPC-NE-2005-A, Rev 4a, Thermal-Hydraulic Statistical Core Design Methodology.
- CNS-17-036, Mid-Cycle Power Uprate Ascension Startup Report
- CNS-17-042, 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