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Category:Topical Report
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Pages in category "Topical Report"
The following 200 pages are in this category, out of 723 total.
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- BSEP 02-0169, GE Nuclear Energy Report NEDO-33063, Safety Analysis Report for Brunswick Steam Electric Plant, Units 1 and 2, Maximum Extended Load Line Limit Analysis Plus.
- BSEP 07-0053, Areva Np, Inc., Topical Report, ANP-2637, Revision 1, Boiling Water Reactor Licensing Methodology Compendium.
- BSEP 10-0071, ANP-2920(NP), Revision 0, Brunswick Unit 2 Cycle 20 Fuel Cycle Design.
- BSEP 10-0093, ANP-2936(NP), Revision 0, Brunswick, Unit 2, Thermal-Hydraulic Design Report for Atrium Tm 10XM Fuel Assemblies.
- BSEP 10-0112, ANP-2943(NP), Revision 0, Brunswick Units 1 and 2, LOCA-ECCS Analysis MAPLHGR Limit for Atrium 10XM Fuel
- BSEP 10-0126, ANP-2956(NP), Rev. 0, Brunswick Unit 2 Cycle 20 Reload Safety Analysis, Enclosure 6 to BSEP 10-0126
- BSEP 18-0027, ANP-3655NP, Revision 0, Brunswick Mellla+ CRDA Assessment with Draft Criteria.
- BVY 04-009, Vermont Yankee - Technical Specification Proposed Change No. 263, Supplement No. 4 Regarding Extended Power Uprate, NRC Acceptance Review
- BYRON 2016-0078, WCAP-18056-NP, Rev. 0, Analysis of Capsule Y from the Exelon Generation Byron Unit 2 Reactor Vessel Radiation Surveillance Program.
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- CNL-14-226, Organization Topical Report, TVA-NPOD89-A
- CNL-15-060, Technical Specifications Change No. WBN2-TS-15-16 - Revise Technical Specifications for Use of Steam Generator Alternate Repair Criterion F
- CNL-15-247, Brown Ferry Unit 1, Submittal of Response to NRC Request for Additional Information Re Application to Modify Technical Specification 2.1.1.2, Reactor Core Minimum Critical Power Ratio Safety Limits (TS-506). Enclosure 2 ANP-3458NP Enclose
- CNL-15-249, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 1, Spent Fuel Pool Criticality Safety Analysis Information
- CNL-15-250, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 2, MICROBURN-B2 Information, Including Enclosures 2, 4, 6, and 7
- CNL-16-056, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 9, Responses to Requests for Additional Information
- CNL-17-012, Organization Topical Report, TVA-NPOD89-A
- CNL-18-069, Tennessee Valley Authority - Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revisions 35 and 36
- CNL-19-001, Sequoyah Nuclear Plant and Watts Bar Nuclear Plant - Organization Topical Report, TVA-NPOD89-A
- CNL-19-067, Application to Revise Watts Bar Nuclear Plant (WBN) Unit 2 - Technical Specifications for Steam Generator Tube Repair Sleeve (WBN-TS-391-19-13)
- CNL-21-007, Organization Topical Report, TVA-NPOD89-A, Revision 24
- CNL-23-003, Browns Nuclear Plant, Units 1, 2 & 3; Sequoyah Nuclear Plant, Units 1 & 2; and Watts Bar Nuclear Plant, Units 1 & 2 - Organizational Topic Report TVA-NPOD89-A
- CNL-23-044, Transmittal of Revision 3 to WCAP-18774-P and WCAP-18774-NP, Addendum to the Rotterdam Dockyard Company Final Stress Report for 173 P.W.R. Vessels TVA III & IV (Report No. 30749-B-030, Rev. 3) - Evaluation of One Closure Stud Out
- CNRO-2002-00055, Rev to Tsar in Support of Appendix K Measurement Uncertainty Recovery Power Uprate Request
- CNS-15-024, WCAP-17993-NP, Revision 0-B, Justification for the Use of RAPTOR-M3G for the Catawba, Unit 1, Measurement Uncertainty Recapture (Mur) Power Uprate Fluence Evaluations.
- CNS-15-078, WCAP-18060-NP, Revision 1, Response to RAIs Concerning the Use of RAPTOR-M3G for the Catawba Unit 1 Measurement Uncertainty Recapture (Mur) Power Uprate Fluence Evaluations
- CP-200900748, WCAP-17072-NP, Rev. 0, H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model D5)
- CPSES-200402317, Transmittal of Unit 2 Reactor Vessel Surveillance Capsule X Analysis Report, WCAP-16277-NP
- CPSES-200500255, Submittal of Supplement to the CPSES Loss of Coolant Accident (LOCA) Analysis Methodologies - Topical Report ERX-04-004, Revision 0
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- DCL-02-097, License Amendment Request 02-05, Revision to Technical Specification Table 3.3.1-1, Reactor Trip System Instrumentation, & Revised Reactor Coolant System Flow Measurement
- DCL-03-052, Rev 0 to WCAP-15958, Analysis of Capsule V from Pacific Gas & Electric Co Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program, Appendix a Through E
- DCL-03-132, to WCAP-15919-NP, Steam Generator Tube Repair for Westinghouse Designed Plants with 7/8 Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves.
- DCL-10-030, Diablo Canyon Power Plant - Topical Report, Process Protection System Replacement Diversity & Defense-in-Depth Assessment, Revision 0
- DCL-10-114, Diablo Canyon - Topical Report, Process Protection System Replacement Diversity & Defense-in-Depth Assessment, Revision 1
- DCL-15-150, WCAP-17462-NP, Revision 1, Program Plan for Aging Management of Reactor Vessel at Diablo Canyon, Unit 1, and WCAP-17463-NP, Revision 1, Program Plan for Aging Management of Reactor Vessel Internals at Diablo Canyon, Unit 2
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- GNRO-2010/00056, NEDO-33601, Revison 0, Engineering Report Grand Gulf Replacement Steam Dryer Fatigue Stress Analysis Using Pble Methodology. Appendix E Through Appendix G
- GO2-15-056, Enclosure 2: GE-Hitachi Nuclear Energy 001N6043.4-NP, Rev. 1, Energy Northwest Columbia Generating Station Jet Pump 17/18 Riser Evaluation at 106% Rated Core Flow (115 Mlbs/Hr).
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- JAFP-19-0016, Summary of Changes to Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-10
- JAFP-20-0006, Summary of Changes to Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-10
- JAFP-20-0030, Response to Request for Additional Information in Support of License Amendment Request - Proposed Changes to the Technical Specifications to Primary Containment Hydrodynamic Loads
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- L-02-006, Part 1 of 2, Diablo Canyon Independent Spent Fuel Storage Installation - Submittal of Diablo Independent Spent Fuel Storage Installation Safety Analysis Report Reference Documents
- L-02-012, Analysis of Loaded HI-STORM 100 System Under Drop & Tipover Scenarios
- L-03-064, Supplemental Information Supporting Proposed Alternative Repair Methods for Reactor Vessel Head Penetrations
- L-04-089, Generic W* Tube Plugging Criteria for 51 Series Steam Generator Tubesheet Region Wextex Expansions.
- L-04-125, Extended Power Uprate Licensing Report, Supplemental Information
- L-17-292, Modified Rt PTS Values and Reactor Vessel Surveillance Capsule Withdrawal Schedule
- L-18-030, WCAP-18102-NP, Rev 1, Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation.
- L-20-208, WCAP-18558-NP, Analysis of Capsule Y from the Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program. Part 4 of 4
- L-2002-082, Part 2 of 2, St. Lucie, Unit 1 - Reactor Vessel Surveillance Capsule, Report of Test Results - Revision 1, Appendix B, Charpy V-Notch Plots for Each Capsule Using Hyperbolic Tangent Curve-Fitting Method
- L-2003-002, to Topical Report WCAP-15975-NP, NDE Inspection Strategy for the Tubesheet Region in St. Lucie Unit 2.
- L-2003-101, WCAP-16038-NP, Revision 0, Structural Integrity Evaluation or Reactor Vessel Upper Head Penetrations to Support Continued Operation: St. Lucie Unit 2, Table of Contents Through Figure 6.24
- L-2004-233, Westinghouse WCAP-15918-NP, Revision 2, Steam Generator Tube Repair for Combustion Engineering and Westinghouse Designed Plants with 3/4 Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves.
- L-2006-067, Request for Approval of FPL Quality Assurance Topical Report
- L-2006-164, Topical Quality Assurance Report (Fpltqar 1-76A)
- L-2006-246, Response to RAI Follow-Up Questions Regarding Fpl'S Common Quality Assurance Topical Report Request
- L-2007-110, Topical Quality Assurance Report (Fpltqar 1-76A)
- L-2008-150, Duane Arnold, and Point, Beach, Units 1 and 2, Quality Assurance Topical Report (QATR-FPL-1) Revisions 1 & 2
- L-2009-134, Submittal of Quality Assurance Topical Report (QATR FPL-1) Revision 4
- L-2010-078, ANP-2903(NP), Rev. 0, St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding.
- L-2010-078, WCAP-17197-NP, Rev. 0, St. Lucie Unit 1 RCS Pressure & Temperature Limits & Low-Temperature Overpressure Protection Report for 54 Effective Full Power Years.
- L-2010-113, Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 10; WCAP-17094-NP, Revision 2, New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis
- L-2010-138, Submittal of Quality Assurance Topical Report (QATR FPL-1) Revision 7
- L-2010-259, Extended Power Uprate Licensing Report, Attachment 5; Appendix G, WCAP-17197-NP, Revision 0, RCS Pressure and Temperature Limits and Low-Temperature Overpressure Protection Report for 54 Effective Full Power Years
- L-2010-299, License Amendment Request for Extended Power Uprate, Attachment 12; WCAP-17070-NP, Revision 0, Westinghouse Setpoint Methodology for Protection Systems
- L-2011-418, ANP-2903Q1(NP), Revision 0, St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding.
- L-2013-208, Submittal of Quality Assurance Topical Report (QATR FPL-1) Revision 13
- L-2015-053, WCAP-17973-NP, Revision 1, Turkey Point, Units 3 and 4, Pressurizer Heater Sleeve Flaw Evaluation to Support Half-Nozzle Repairs, Enclosure 1 (Non-Proprietary)
- L-2015-160, WCAP-17939-NP, Revision 0, Analysis of Capsule 97 Degrees from the Florida Power & Light Company St. Lucie, Unit 2, Reactor Vessel Radiation Surveillance Program, Part 1 of 3
- L-2015-160, WCAP-17939-NP, Revision 0, Analysis of Capsule 97 Degrees from the Florida Power & Light Company St. Lucie, Unit 2, Reactor Vessel Radiation Surveillance Program, Part 2 of 3
- L-2015-160, WCAP-17939-NP, Revision 0, Analysis of Capsule 97 Degrees from the Florida Power & Light Company St. Lucie, Unit 2, Reactor Vessel Radiation Surveillance Program, Part 3 of 3
- L-2016-129, NextEra Energy Quality Assurance Topical Report (QATR FPL-1) Revision 19 Annual Submittal
- L-2018-166, WCAP-15355-NP, Revision 0, a Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casings.
- L-2021-115, NextEra Energy Quality Assurance Topical Report (FPL-1) Revision 26 Annual Submittal and Florida Power and Light Company Quality Assurance Program Description for 10 CFR Part 52 Licenses (FPL-2) Revision 10
- L-2021-234, NextEra Energy Quality Assurance Topical Report (FPL-1) Revision 27 Update
- L-2022-068, NextEra Energy Duane Arnold Quality Assurance Topical Report (FPL-3) Revision 2
- L-2022-082, NextEra Energy Seabrook, LLC, and NextEra Energy Point Beach, LLC, Quality Assurance Topical Report FPL-1, Revision 28
- L-2023-071, NextEra Energy Quality Assurance Topical Report (FPL-1) Revision 29 and Florida Power and Light Company Quality Assurance Program Description for 10 CFR Part 52 Licenses (FPL-2) Revision 11, Annual Submittal
- L-2023-173, Quality Assurance Topical Report (FPL-1) Revision 30 Update
- L-2024-114, Quality Assurance Topical Report (FPL-1 Revision 31 Annual Submittal
- L-21-093, Enclosure B - Affidavit of Philip A. Opsal and Enclosure C: Rev. 1 to ANP-3438NP, MRP-227-A - Applicant/Licensee Action Item 7 Analysis for Davis-Besse Nuclear Power Station Unit 1 (Non-Proprietary)
- L-HU-04-039, Nuclear Management Company, Llc., Response to Request for Additional Information Request for Approval of Nuclear Management Company Quality Assurance Topical Report, Dated August 18, 2004
- L-MT-14-044, ANP-3135NP, Applicability of Areva BWR Methods to Extended Flow Window for Monticello.
- L-MT-15-074, Enclosure 7, WCAP-18604-NP, Revision 0, Monticello EPU Main Steam Line Strain Data Evaluation Report.
- L-MT-23-030, Subsequent License Renewal Application Supplement 3
- L-PI-03-040, Praire Island, Unit 1 - Transmittal of Westinghouse Lower Row Tube Dent Root Cause Analysis
- L-PI-04-017, Development and Qualification of a Gothic Containment Evaluation Model for the Prairie Island Nuclear Generating Plants.
- L-PI-05-110, Supplement to License Amendment Request (LAR) to Revise the Spent Fuel Pool Criticality Analyses and Technical Specifications (TS) 3.7.17, Spent Fuel Pool Storage and 4.3, Fuel Storage
- L-PI-21-006, License Amendment Request to Revise Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, to Incorporate Installed Bypass Test Capability for the Power Range RTS Functions
- L-PI-22-005, WCAP-18660-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, Appendix C, Part 5 Through Appendix D
- L-PI-23-007, Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
- LIC-10-0065, LTR-PAFM-10-123-NP, Rev. 0, Technical Justification to Support Alternative Visual Examination Intervals for Fort Calhoun Reactor Vessel Outlet Nozzle to Safe End Dissimilar Metal Welds, Enclosure 2
- LR-N03-0406, GNF-A Proprietary Report, NEDC-33107P, GEXL80 Correlation for SVEA96+ Fuel. (Non-Proprietary Version)
- LR-N04-0199, Response to Request for Additional Information Regarding Relief Request S1-RR-13-B21 NRC Order EA-03-009 on Reactor Pressure Vessel Head Inspections
- LR-N04-0378, NEDO-33153, SAFER/GESTR-LOCA Loss of Coolant Accident Analysis for Hope Creek Generating Station.
- LR-N09-0290, NEDO-33529, Rev. 0, Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (Itas) in Hope Creek Generating Station
- LR-N12-0156, Submittal of Updated Final Safety Analysis Report, Revision 26, Salem Units 1 & 2 Technical Specification Bases Changes, 10 CFR 54.37(b) Review Results and PSEG Nuclear LLC Quality Assurance Topical Report, NO-AA-10, Revision 82
- LR-N15-0020, License Amendment Request to Revise Technical Specification 3/4.3.1, Reactor Trip System Instrumentation
- LR-N20-0072, License Amendment Request to Amend Tech Specs to Revise and Relocate the Reactor Coolant System Pressure & Temperature Limits & Pressurizer Overpressure Protection System Limits to a Pressure & Temperature Limits Report
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