P.4-28o-29
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creation date: 2025-12-08T03:36:19-05:00
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P.4(o)
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05000482/LER-2010-002, Regarding Turbine Trip Function of Reactor Trip P-4 Interlock Defeated During Entry Into and in Mode 3
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CNS-16-015, Technical Specification Bases Changes
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PLA-2756, Forwards Postulated Pipe Rupture Calculations M-MSS-010 & M-MSS-006,per NRC 861112 Telcon Request
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ML18037A299
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CNS-15-098, Technical Specification Bases Changes
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CP-201201129, License Amendment Request (LAR) 12-005, Revision to Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation and 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation
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NL-11-0280, License Amendment Request for Steam Generator Water Level High-High Setpoint Change
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NL-17-0534, Joseph M. Farley Nuclear Plant, Technical Specifications Bases
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NL-16-2280, Vogtle Electric Generating Plant, Units 1 & 2, Updated Final Safety Analysis Report, Bases for Improved Technical Specifications
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NL-15-156, Sequoyah Nuclear Plants, Units 1 and 2 - Technical Specifications Conversion to NUREG-1431, Rev. 4.0 (SQN-TS-11-10) - Supplement 3. Part 2 of 9
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RS-12-221, Byron, Units 1 and 2, Updated Final Safety Analysis Report, Revision 14, Chapter 2.0 - Site Characteristics, Table of Contents Through Table 2.5-35
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NRC 2009-0123, License Amendment Request 261, Extended Power Uprate, Response to Request for Additional Information for Calculation of the Auxiliary Feedwater Pump Low Suction Pressure Setpoint
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PLA-6561, Amendment Request No. 296 to Unit 1 and Amendment Request No. 266 to Unit 2 Re Emergency Core Cooling System Instrumentation - Technical Specification (TS) Table 3.3.5.1-1 and Editorial Change to TS 3.10.8.F, Response to RAI
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ML091280276
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RS-08-161, Braidwood, Units 1 and 2 - Technical Specification Bases
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GNRO-2008/00063, UFSAR Update for Grand Gulf, Unit 1, Nuclear Station for September 2008
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CP-201300130, Comanche Peak, Units 1 & 2, Technical Specification Bases, Revision 67
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CP-201101065, Comanche Peak Units 1 & 2 Final Safety Analysis Report (Fsar), Amendment 104, Technical Specifications Bases
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CP-201000212, Technical Specifications Bases for Comanche Peak Steam Electric Station, Units 1 and 2
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CP-200801015, Technical Specification Bases for Comanche Peak Steam Electric Station Units 1 & 2
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ET 10-0028, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation
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ET 10-0014, Application to Revise Technical Specification 3.3.2, Engineered Safety Feature Actuation System Instrumentation, Table 3.3.2-1
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L-04-040, License Amendment Request Nos. 321 and 193, Modifying Technical Specification Requirements for Mode Change Limitations in Specifications 3.0.4 and 4.0.4 and the Associated Technical Specification Bases
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CNL-20-058, Tennessee Valley Authority, Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications for Function 6.e of Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation (WBN-TS-19-27)
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RA-20-0207, Duke Energy - Application to Revise the Technical Specification for Engineered Safety Feature Actuation System (ESFAS) Instrumentation
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RBG-27292, Forwards Initial Response to NRC Compliance Bulletin 87-002, Fastener Testing to Determine Conformance W/Applicable Tech Specs. Final Response to Bulletin Will Be Submitted by 880129
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85-03.Corrective Actions: Notification Procedure Revised & Main Control Room Modified to Provide Addl Space & Orientation for Communicator
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B11449, Forwards Response to 840802 SER Open Items Re Seismic & Dynamic Qualification Program
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B13137, Requests Exemption from Certain Requirements of 10CFR50, App J,Sections III.C.1 & III.C.2 Re Type C Testing Requirements.Diagrams for Each Penetration Included in Attachment I
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U-600975, Forwards Dcrdr Final Rept Addressing Completion of Items Listed in Sser 5,Section 18.3.Attachment 2 Summarizes Human Engineering Observations & Deficiencies Identified & Assessed as Result of Review
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LER-1997-001, :on 970117,missed Surveillance on Auxiliary Contacts of Reactor Trip Breakers Discovered.Caused by Adequate PMT for Reactor Trip Breaker Was Not Performed. Multi-disciplinary Team Formed to Perform Investigation
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AECM-88-0064, Application for Amend to License NPF-29,revising Tech Specs Re Mode Changes,Inconsistent Application of Exceptions & Unnecessary Shutdowns When Surveillance Intervals Inadvertently Exceeded,Per Generic Ltr 87-09.Fee Paid
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03L-0:on 791213,during Surveillance Test,Steam Line Differential Pressure Safety Injection Signal Comparator Pb 486B Tripped Below Tech Spec Limit.Cause Not Stated.New Card Installed & Calibr.Channel Operable
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05000263/LER-1983-023-03, /03L-0:on 831026,flow Through Train B Standby Gas Treatment Less than Tech Spec Requirement.Caused by Stack Exhaust Flow Incorrectly Adjusted.Stack Flow Corrected
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NUREG-0840, Comments Re Peer Group Review of Draft NRC Fuel Testing Task Force rept.In-vessel Core Damage Progression Must Be Experimentally Addressed
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LER-1991-006, :on 910202,inboard Isolation Valves Closed on RWCU Delta Flow High Signal During Removal of RWCU Filter from Svc.Caused by Personnel Error.Sys Operating Instruction Being Revised
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JAFP-83-0888, Effluent & Waste Disposal Semiannual Rept,Jan-June 1983
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05000348/LER-1990-002, :on 900222,two Cases Found Where Surveillance Testing Procedures Did Not Perform All Tests Required by Tech Specs.Caused by Inadequate Procedural Guidance. Procedures Revised
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05000316/LER-1981-067-03, Containment Isolation Valve QCR-300 Blocked Open to Allow Isolation of Associated Air Header for Repair.Caused by Personnel Error.Shift Supervisor Reinstructed
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L-97-204, Forwards Info Copy of State of Fl Domestic Wastewater Facility Permit for South Hitchinson Island Wastewater Treatment Facility Id FL0139475
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Appropriate ASME Code.Minor Mod to V-2526 Being Processed,To Satisfy NRC 830215 Concern.Completion Expected Prior to Core Load
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IR 05000387/1981011
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05000410/LER-1993-011, :on 931119,reactor Scram & ESF Actuation Occurred.Caused by Poor Work Organization & Planning. Work Planning Guide Revised & Procedure Changes Will Be Evaluated
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IR 05000410/1981009
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LR-N12-0220, Loss of Coolant Accident Peak Cladding Temperature Margin Tracking - Annual Report
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05000482/LER-2012-002, Regarding One Train of Automatic Safety Injection Blocked During Entry Into Mode 3 Due to Procedural Weakness
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05000482/LER-2009-009-01, For Wolf Creek, Regarding Defeating Feedwater Isolation on Low Tavg Coincident with P-4 Function Results in Missed Mode Change
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NRC 2010-0113, License Amendment Request 261 Extended Power Uprate, Transmittal of Sample Reactor Protection System/Engineered Safety Features Actuation System Instrumentation Setpoint Calculations
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05000482/LER-2009-009, Regarding Defeating Feedwater Isolation on Low Tavg Coincident with P-4 Function Results in Missed Mode Change
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B18629, Response to Request for Additional Information Technical Specifications Change Request 3-2-00 Emergency Diesel Generator Allowed Outage Time
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Mechanical Engineering Branch
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label: RBG-21487, Forwards Final Response to NRC 850430 Request for Addl Info Re Feedwater Line Break Outside Containment.Confirmatory Analysis Encl to Suppl Util 850520 Response,Per Discussion W/Mechanical Engineering Branch
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ML20215B304
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ML20115A763
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ML20064G884
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Cross cutting aspect
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