Subsequent actuation
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NRC Region 1 false
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Subsequent actuation Three Mile Island
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Subsequent actuation Vermont Yankee
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Subsequent actuation Browns Ferry
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Subsequent actuation Watts Bar
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Subsequent actuation Duane Arnold
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Subsequent actuation Fermi
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Subsequent actuation LaSalle
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Subsequent actuation Elk River
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Subsequent actuation La Crosse
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Subsequent actuation Arkansas Nuclear
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Subsequent actuation Comanche Peak
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Subsequent actuation Diablo Canyon
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Subsequent actuation Palo Verde
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Subsequent actuation Quad Cities
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Subsequent actuation South Texas
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Subsequent actuation Victoria
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Subsequent actuation Wolf Creek
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Subsequent actuation Fort Calhoun
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Subsequent actuation Fort Saint Vrain
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Subsequent actuation Hallam
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Subsequent actuation Humboldt Bay
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Subsequent actuation Rancho Seco
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Subsequent actuation San Onofre
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Subsequent actuation Trojan
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NRC Region 4 true
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Subsequent actuation 10
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Subsequent actuation Clinch River
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Subsequent actuation Kemmerer
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NRC Generic Letter 78-09, Multiple-Subsequent Actuations of Safety/Relief Valves Following an Isolation Event
HNP-03-111, Change to Emergency Plan Implementing Procedures for Shearon Harris
M200041, GE-Hitachi ABWR Design Control Document Tier 1 & 2, Rev. 7 - Chapter 05 - Reactor Coolant System and Connected Systems
NRC-16-0034, Fermi, Unit 2, Revision 20 to Updated Final Safety Analysis Report, Chapter 5, Reactor Coolant System and Connected Systems
RS-08-045, LaSalle, Units 1 and 2, Updated Final Safety Analysis Report (Ufsar), Revision 17, Chapter 5.0 - Reactor Coolant System and Connected Systems
RS-13-009, Clinton, Unit 1, Updated Safety Analysis Report, Revision 15, Chapter 5 - Reactor Coolant System and Connected Systems
RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 5 - Reactor Coolant System and Connected Systems
RS-17-126, Quad Cities Nuclear Power Station, Units 1 & 2, Revision 14 to Updated Final Safety Analysis Report, Chapter 5, Reactor Coolant System and Connected Systems
RS-16-196, Clinton - Updated Safety Analysis Report (Usar), Revision 18, Chapter 5 - Reactor Coolant System and Connected Systems
RS-11-159, Quad Cities, Units 1 and 2 - Updated Final Safety Analysis Report (Ufsar), Revision 11, Chapter 05 - Reactor Coolant System and Connected Systems
RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 3 - Design of Structures, Components, Equipment, and Systems
RS-03-090, Request for Amendment to Technical Specifications Surveillance Requirements for the Main Steam Relief Valves and Relief Request RV-30E
PY-CEI-NRR-1260, Forwards Dcrdr Identifying & Correcting Human Engineering Deficiencies
RS-13-009, Clinton, Unit 1, Updated Safety Analysis Report, Revision 15, Chapter 3 - Design of Structures, Components, Equipment and Systems
RS-11-100, Dresden, Units 2 and 3, Updated Final Safety Analysis Report (Ufsar), Revision 9, Technical Specification Bases
RBG-47603, Revisions to the Technical Requirements Manual and the Technical Specifications Bases
PY-CEI-NRR-2216, Forwards Supplemental Response to NRC Bulletin 96-003, Potential Plugging of ECCS Strainers by Debris in Bwrs. Updated Info to NRC on Several Design Approaches Which Have Been Revised During Final Design Process,Provided
L-MT-12-004, License Amendment Request: Revise Technical Specification Requirements for Testing the Main Steam Safety/Relief Valves
PY-CEI-NRR-1031, Forwards Results of Control Room/Remote Shutdown Room Final Noise Survey & Dcrdr Final Closure of Outstanding Human Engineering Discrepancies
RS-13-009, Clinton, Unit 1, Updated Safety Analysis Report, Revision 15, Chapter 7 - Instrumentation and Control Systems
RS-03-087, Request for Amendment to Technical Specifications Surveillance Requirements for the Main Steam Line Relief Valves and Relief Request RV-30D
GNRO-2011/00021, Response to Request for Additional Information Regarding Extended Power Uprate
W3F1-2017-0067, License Amendment Request to Remove Technical Specification 3/4.3.2 Table 4.3-2 Note 3 Exemption for Testing Relays K114, K305, and K313
PY-CEI-NRR-0010, Forwards Draft FSAR Pages for App 3A & 3B.Change Provides Description of Safety Relief Valve & LOCA-related Hydrodynamic Load Definition for Mark III Containment Re SER Outstanding Issue 9, Pool Dynamic Loads
LR-N17-0088, Hope Creek Generating Station Updated Final Safety Analysis Report, Rev. 22, Chapter 15, Accident Analyses
W3F1-2017-0018, Supplement to Response to RAI on License Amendment Request to Revise Technical Specification 3/4.3.2 to Relocate Surveillance Frequency Requirements for Engineered Safety Features Actuation System Subgroup Relays to Surveillance
NLS2010046, NEDO-33543, Revision 0, Nebraska Public Power District Cooper Nuclear Station Safety Relief Valve Capacity and Setpoint Evaluation, Enclosure 2 to NLS2010046
RS-08-045, LaSalle, Units 1 and 2, Updated Final Safety Analysis Report (Ufsar), Revision 17, Chapter 3.0 - Design of Structures, Components, Equipment, and Systems
NLS9100189, Forwards Revised Page 180 to Tech Specs Re Change to low-low Set Safety/Relief Valve Pressure Switches
NL-03-2266, Follow-Up Written Request for Notice of Enforcement Discretion and Request to Revise Technical Specification Surveillance Requirements 3.3.1.5 and 3.3.3.2, on an Exigent Basis
NRC-19-0054, DTE Energy Company - License Amendment Request to Revise Technical Specifications to Change Surveillance Intervals to Accommodate a 24-Month Fuel Cycle
RBG-46179, Revisions to the Technical Requirements Manual and the Technical Specification Bases
PY-CEI-NRR-0946, Application for Amend to License NPF-58.Amend Adds Ref to Encl Dcrdr - First Refuel Hed Revs Repts, Attachment 1. Encl Supersedes Rept Sheets Encl W/Attachment 2 of Util
NG-18-0024, Response to Request for Additional Information Regarding License Amendment Request (TSCR-170), Revision to Technical Specification 3.5.1, EGGS-Operating
NL-11-1812, Edwin I. Hatch, Unit 2 - 2011 Updated Final Safety Analysis Report, Revision 29, Chapter 6, Engineered Safety Features Through Chapter 7, Instrumentation and Control
U-603781, Transmittal of Revision 10 to the Clinton Power Station Technical Specifications Bases
WO 06-0014, Supplement to Revision to Technical Specifications - Extensions of AC Electrical Power Distribution Completion Times
RS-20-139, Response to Request for Additional Information Regarding License Amendment Requests to Adopt TSTF-505, Revision 2, and 10 CFR 50.69
RBG-20045, Forwards Response to SER Confirmatory Item 5, Effects of Annulus Pressurization. Info Will Be Provided in Amend to FSAR Sections 3.9B for NSSS Scope & 3.9A for balance-of- Plant Scope
L-09-100, Rev 0 to MPL-09-100, Grand Gulf In-Plant Safety Relief Valve Test Fatigue Evaluation Final Rept
W3F1-2022-0054, Revise Technical Specification 3/4.3.2 to Remove Exemption from Testing Certain Relays at Power to Support Elimination of Potential Single Point Vulnerability
HL-4812, Forwards Response to NRC 950310 RAI Re Util 950113 Power Uprate Submittal.Info Relative to Plant Setpoint Methodology & Calculations Also Encl
RS-03-199, Request for License Amendment Related to Revision of Instrument Channel Trip Setpoint Allowable Values
NG-12-0107, Update to License Amendment Request (TSCR-129): Application for Technical Specification Change Regarding Alternative Testing of Safety/Relief Valves Sections Affected: 3.4.3, 3.5.1, and 3.6.1.5 (TAC ME7337 & ME7336)
RAIO-0419-65230, LLC Response to NRC Request for Additional Information No. 512 (Erai No. 9634) on the NuScale Design Certification Application
L-MT-05-097, Additional Information to Support Technical Specification Changes for 24-Month Fuel Cycle License Amendment Request
LR-N17-0088, Hope Creek Generating Station Updated Final Safety Analysis Report, Rev. 22, Chapter 3, Design of Structures, Components, Equipment and Systems, Appendices 3A Through 3I
2CAN069801, Application for Amend to License NPF-6,revising Portion of Plant Protection Sys Design to Ensure That ESFAS Meets Its Applicable Design Requirements.Rev 1 to Engineering Rept 97-R-2015-01 Encl
RBG-24-131, Responds to Mark III Open Items in Suppl 2 to SER (NUREG-0989).Revs to Action Plans 2,5,6 & 8 Encl. Restrictions Against Operation of RHR Sys in Steam Condensing Mode Following LOCA Will Be Added to Procedure
05000220/LER-2012-007, Regarding High Pressure Coolant Injection System Logic Actuation Following an Automatic Turbine Trip Signal Due to High Reactor Water Level
JAFP-24-0010, Response to Request for Additional Information for License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/Rvs) Setpoint Lower Tolerance
PLA-7125, Response to Request for Additional Information on Change to Technical Specification Surveillance Requirement 3.5.1.12
W3F1-2018-0018, Supplemental Information Supporting the License Amendment Request to Remove Technical Specification 3/4.3.2 Table 4.3-2 Note 3 Exemption for Testing Relays K114, K305, and K313
RS-04-005, Request for Amendment to Technical Specifications Surveillance Requirements for the Main Steam Line Valves and Associated Relief Requests
RS-07-118, Application for Technical Specification Change TSTF-423, Risk Informed Modification to Selected Required Action End States for BWR Plants, Using the Consolidated Line Item Improvement Process
NLS8900315, Application for Amend to License DPR-46,consisting of Proposed Change 69,revising Setpoint Tolerance on lo-lo Set Safety/Relief Valve Pressure Switches & Format & Location of lo-lo Set Specs Per STS
Information Notice 2010-25, Inadequate Electrical Connections
05000482/LER-2010-005, Regarding Reactor Trip Due to Low Steam Generator Level from Trip of Main Feedwater Pump
RS-24-032, Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion
RS-05-104, Revised Technical Specification Pages for License Amendment Related to Revision of Instrument Channel Trip Setpoint Allowable Values
SBK-L-05013, License NPF-86 - Supplemental Response to Request for Additional Information Regarding License Amendment Request 04-03. Application for Stretch Power Uprate
BVY-98-124, Responds to NRC Re Violations Noted in Insp Rept 50-271/98-09.Corrective actions:OT-3114,OP-2172 & OP-2143 Were Revised to Allow Only Single Feedwater Pump in Standby Mode at One Time & to Control re-energization of Oil Pump
RBG-45946, Entergy Nuclear- South, River Bend Station, Unit-1, License Amendment LAR 2002-11 Modification of the Technical Specification Surveillance Requirements for the Safety/Relief Valves
05000366/LER-2005-003, Unit 2, Manual Reactor Scram Inserted in Response to Reactor Water Chemistry Event Caused by a Condenser Tube Leak
05000530/LER-1991-004-01, :on 910713,spurious Train a Containment Purge Isolation Actuation Signal Initiated on balance-of-plant Esfas,Resulting in Spike of Monitor RU-37.Caused by Failure of Detector Tube.Technicians Briefed
NRC Generic Letter 82-24, Safety Relief Valve Quencher Loads: BWR MARK II and III Containments
TSTF-19-09, Technical Specifications Task Force - Transmittal of TSTF-576, Revision 0, Revise Safety/Relief Valve Requirements
U-600319, Forwards Response to Request for Addl Info Re SER (NUREG-0854) Confirmatory Issue 71 on Humphrey Concerns. Responses Complete & Should Resolve Outstanding Concerns
LR-N24-0030, License Amendment Request to Revise Technical Specification Lift Settings for Reactor Coolant System Safety/Relief Valves
05000391/LER-2021-001, Automatic Reactor Trip on Main Turbine Trip Caused by Main Feed Pump Trip Due to Low Condenser Vacuum
05000373/LER-1998-012-01, :on 980729,reactor Scram During Scram Discharge Volume Surveillance Occurred.Caused by Inadequate Procedures.Revised Subject Surveillance Procedure Accordingly
05000498/LER-1989-008, :on 890212,discovered That Inverter IV-1203 Which Supplies Uninterruptible Power to Distribution Panel DP-1203 Overheating,Causing Actuation of Control Room & Fuel Handling Bldgs HVAC Sys.Transformer Replaced
05000333/LER-1988-004, :on 880523,vendor Informed Util That Three of Five Safety Relief Valves Removed for Test During Jan 1988 Maint Outage Did Not Actuate within Allowed Specs.Caused by Pilot Valve Disc & Seat Bonding or Leakage
05000313/LER-1987-005-01, :on 870825,reactor Trip Occurred Due to Failure of One Phase of Breaker 5118.Caused by Relay Personnel Forgetting to Pull Breaker Failure Module.Maint & Testing Performed & Procedures Under Evaluation
05000413/LER-1987-005, :on 870130,determined That Containment Air Return Fans Could Be Inoperable Due to Collected Spray Flooding.Caused by Design Mfg,Const/Installation Deficiency. Curbs Installed
05000341/LER-1998-002, :on 980220,inadvertent Load Shed of ESF Bus 72E During Performance of Surveillance Test Resulting in ESF Actuations Occurred.Caused by Inadequate Procedure Rev. Power Was Restored to Affected Loads by 1344 Hrs on 980220
05000424/LER-1987-029, :on 870524,reactor Tripped on Steam Generator 3 lo-lo Level While Operators Transferring Feedwater Flow from Auxiliary Feedwater Sys to Main Sys.Caused by Incorrect Closure of Valve Piping.Valve Restored
05000440/LER-1986-078, :on 861106,Div 1 4.16-kV Bus EH11 Deenergized, Causing Deenergization of Annulus Exhaust Gas Treatment Sys Train a & Subsequent Actuation of Standby Train B.Caused by Accidental Bumping of Auxiliary Relay
05000366/LER-1987-003-01, :on 870126,during Main Steam Line Radiation Monitor Functional Test,Primary Containment Isolation Sys Group 1 Isolation Occurred.Caused by Equipment Failure. Temp Sensor Will Be Replaced
05000312/LER-1987-008, :on 870115,discovered Inadequate Sequencing of Diesel Generator High Pressure Pumps During Automatic Loading.Caused by Design Error.Design Control Procedures Implemented
05000346/LER-1986-035, :on 860813,seismic Monitoring Sys Inoperable. Caused by Deficient Surveillance Test Procedures.Strong Motion Accelerometers Removed for Calibr & Test Procedures Revised to Meet Tech Specs
05000366/LER-1998-004-01, :on 981005,noted That full-in Position Indicating Light Did Not Extinguish as Expected When CR 38-15 Was Withdrawn.Caused by Personnel Error.Jumper Was Removed from RPIS Circuit Card for Rod 38-15
05000328/LER-1988-024, :on 880523,reactor Trip Occurred from Low Flow Signal on Rcs.Caused by Procedural Noncompliance.Util Will Review Past Reactor Trips for Similar Occurrence of Common Interaction Between Equipment Causing Trip
05000366/LER-1998-004-04, :on 981005,condition Prohibited by TS Was Noted.Caused by Personnel Error.Coached Involved Personnel & Briefed Other Technicians on Event.With
05000483/LER-1988-010, :on 880902,unplanned Reactor Trip Occurred, Followed by Automatic Actuation of ESF & Auxiliary Feedwater & Feedwater Isolation.Caused by Personnel Error & Leaking Valve.Switchgear Doors Will Be Locked
05000368/LER-1986-001-01, :on 860211,MSIV 2CV-1010 a Spuriously Closed, Resulting in Reactor Trip & Insertion of Control Element Assemblies.Caused by Malfunction in motor-driven Relay (Mdr).Mdr Replaced & Tested
05000328/LER-1998-001-05, :on 980827,turbine Trip Occurred Followed by Reactor Trip.Caused by Failure of Sudden Pressure Relay on B Phase Main Transformer.Control Room Operators Responded & Removed,Inspected & Replaced Failed Relays
05000298/LER-1988-019, :on 880715,reactor Manually Scrammed Subsequent to Discovery of Electrical Arcing at Bolted Connection Between Removable Covers of Isolated Phase Bus Duct.Caused by Broken Ground Strap.Sys Restored to Svc
05000416/LER-1997-002-04, :on 970619,SRV Test Switch for Div II relief/low-low Set Logic Placed in Test Position.Caused by Div Switch Already in Test. Div II SRV Group Test Switch Immediately Returned to Normal
05000298/LER-1988-002, :on 880128,reactor Scram Occurred Due to Average Power Range Monitor High Flux.Caused by Locking Screw Used to Secure Disc Loosening & Vibrating Off.Group Isolations Reset & Diesel Generators Secured
05000370/LER-1986-002-02, :on 860123,reactor Trip Occurred as Result of intermediate-range (IR) high-flux Reactor Trip Signal.Caused by Defective Procedure.Ir Detectors Recalibr & Trip Reset Point Changed to 15% Power.Procedures Revised
05000368/LER-1985-014, :on 850718,reactor Trip Occurred.Caused by Inadvertent Actuation of Engineered Safeguards During Plant Protection Sys Surveillance Testing.Trip Matrix Test Pushbuttons Installed in All Four Channels
05000440/LER-1992-024, :on 921121,two SRVs Unexpectedly Opened During Maint/Surveillance Activity,Introducing Pressure Transient Which Caused Reactor Thermal Power to Increase.Cause Not Determined.Two Slave Trip Units Replaced
05000362/LER-1985-005-01, :on 850131 & 0201,Train B Containment Purge Isolation Sys Spuriously Actuated as Result of Instrument Intermittent Failure on Containment Area Radiation Monitor. Caused by Failure of Keylock Function Switch
05000259/LER-1984-027, :on 840627,main Steam Relief Valve Lifted & Leaked.Unit Manually Scrammed & Valve Replaced.Testing of Valve Revealed No Electrical/Mechanical Anomalies.No Further Action Necessary
05000361/LER-1984-053, :on 840919,fuel Handling Isolation Sys Train B Spuriously Actuated When Power Supply to Control Module for Fuel Handling & Vent Airborne Monitor Gas Detector 2RE-7823-2 Deenergized for Sys Mod Work
05000333/LER-1995-006-01, :on 950315,reactor SRV Setpoint Drift Occurred. Caused by Corrosion Bonding Between SRV Pilot Disc & Pilot Seat.Performed New analysis,GE-NE A00-05785-1,Rev 0 & Replaced Pilot Assemblies W/Refurbished Assemblies
05000366/LER-1991-020-01, :on 911105,protective Breaker on Output of motor-generator Set B in Reactor Protection Sys Tripped, Causing Loss of Power to Radiation Monitors.Caused by Spurious Actuation of Breaker.Voltage Adjusted
05000285/LER-1991-026, :on 911105,discovered That Senior Reactor Operator Did Not Have Current NRC Medical Exam.Caused by Lack of Single Point Control of NRC Medical Exam Program. Personnel Training Info Sys Upgraded
05000414/LER-1991-011-03, :on 910913,RCS Pressure Dropped Below 600 Psig, Causing Cold Leg Accumulators to Begin Injecting Borated Water Into Rcs.Caused by Design Deficiency.Station Problem Rept for Spray Valves Initiated
05000321/LER-1991-014-01, :on 910810,component Failure Occurred Resulting in Unplanned ESF Actuations.Failed Components Replaced, Operating Order Issued & Design & Field Review of SGTS Actuation Logic Completed
05000366/LER-1995-001-02, :on 950411,personnel Error Resulted in Automatic Reactor Shutdown.Operator Was Disciplined Under Util Positive Discipline Program
05000321/LER-1994-010, :on 940808,work in Control Room Panel W/Limited Access Leads to Actuation of Esf.Isolation of RWCU Believed to Be Spurious.Util Investigating Feasibility of Relocating Relay to Another Location
05000483/LER-1990-015, :on 901119,ESF Auxiliary Feedwater Actuation Occurred.No Cause for V55 Flashover Relay Actuation Could Be Determined
05000333/LER-1990-019, :on 900530,automatic Depressurization Sys Safety Relief Valve Accumulator Pressure Discovered to Be Lower than Test Required Psig.Caused by Large Pressure Drop Across Check Valves.Check Valves Replaced
05000366/LER-1993-010-02, :on 931108,spurious Radiation Monitor Trip Resulted in Automatic Actuation of Esf.Caused by Electronic Noise Induced by Operator Tapping on Radiation Monitor. Radiation Monitor Was Recalibrated
05000325/LER-1993-003-03, :on 930608,RPS MG a Tripped,Resulting in Successful ESF Actuations & Isolations Due to Defective Procedure.Voltage Adjust Potentiometer for MG a Voltage Regulator Replaced on 930617
AECM-90-0089, Forwards Draft Tech Specs for Power Distribution Limits,Rcs, ECCS & Plant Sys as Part of Util Involvement W/Bwr Owners Group as BWR-6 Lead Plant
AECM-81-341, Forwards Proposed FSAR Changes & Response to NRC Questions Re in-plant Safety/Relief Valve Test Program,Reactor Level Ref Leg boil-off & LOCA Recirculation Valve Failure
AECM-81-290, Forwards Response to Unresolved Safety Issues A-1, waterhammer;A-9,ATWS;A-11,reactor Vessel Matls Toughness; A-17,sys Interaction in Nuclear Power plants;A-39,safety Relief Valve Pool Dynamic Loads & A-40,seismic Design
NPL-01-008, In-plant Safety Relief Valve Test Plan
RS-22-085, Application to Revise Design Basis to Allow Use of Plastic Section Properties in Lower Downcomer Braces Analysis
05000261/LER-2018-001, Valid Actuation of Emergency Diesel Generator
RAIO-0518-60191, LLC Response to NRC Request for Additional Information No. 418 (Erai No. 9511) on the NuScale Design Certification Application
05000260/LER-1992-003-03, :on 920423,during Performance of HPCI Test,An Unexpected Group 4 PCIS Received,Resulting in Closure of Inboard & Outboard HPCI Valves.Caused by Incorrect Plug Installation.Test Plug Replaced
05000260/LER-1992-003-02, :on 920423,unexpected Group 4 Primary Isolation Signal Received,Resulting in Closure of HPCI Steam Supply Isolation Valves.Caused by Inadvertent Installation of Test Plug.Test Plug Marked
05000400/LER-1997-012, :on 970505,determined That Previous Auxiliary Control Panel Had Not Verified Operability of Interposing Relays.Caused by Misinterpretation of Tss.Reviewed Other Remote Shutdown Panel Transfer Circuitry
HNP-96-200, Responds to NRC Re Violations Noted in Insp Rept 50-400/96-09.Corrective Actions:Co Responsible for Inadvertent Filling & Resultant Dilution of Rwst,Was Counseled on Use of Error Prevention Techniques
05000250/LER-1994-005, :on 941103,both Units Outside Design Basis Due to Design Defect in Safeguards Bus Sequencer Test Logic. Resumed Monthly Manual Testing of Sequencer
L-92-275, Special Rept:On 920918,overpressure Mitigating Sys Actuation Occurred Due to Increase in RCS Pressure Experienced as motor-operated Valves Approached Fully Closed Position. Night Order Issued & Inservice Testing Procedure Revised
05000250/LER-1988-028, :on 881117,control Room Ventilation Sys Shifted to Recirculation Mode While Conducting Test of Channel B Air Intake Radiation Monitor RAI 6642.Caused by Maladjustment of Monitor During Calibr.Instruction Revised
05000397/LER-1998-011, :on 980617,ECCS Pump Room Flooding Was Noted Due to FP Sys Pipe Break.Caused by Inadequate Design of FP Sys.Detailed Review of FP Sys Design Was Conducted
GO2-82-120, Forwards marked-up Response to Containment Sys Branch Issue 47a-f, Acceptable SRV Design Load Spec from 810914-17 Meeting.Revised Page 5 to Response to Issue 48,submitted 820113,encl
GO2-82-804, Forwards Final Response to 820708 Concerns Re Adequacy of Design Margins of Mark II Containment Sys.Issues Raised by GE Do Not Represent Significant Safety Concerns for Facility
GO2-81-196, Forwards Addl Info Re Safety Relief Valve Methodology,In Response to NRC 810528 Request.Info Will Be Incorporated Into FSAR within 4 Months
GO2-81-183, Forwards Responses to Round 2 Questions from Containment Sys Branch.Responses Will Be Incorporated Into FSAR Amend within 4 Months.Ge Proprietary Figures Sent Under Separate Cover by
05000335/LER-1996-008, :on 960703,inadvertent Actuation of Safety Injection Actuation Signal & Containment Isolation Signal Occurred Due to Loss of 15 Vdc Regulated Power Supply During Maint.Design Review Performed
05000335/LER-1996-008-01, :on 960608,inadvertent Actuation of Safety Injection Actuation Signal & Containment Isolation Actuation Signal Occurred.Caused by Loss of Power Supply.Reviewed Power Supply Design & Conducted Visual Insp
HNP-17-048, Transmittal of 10 CFR 50.54(q) Evaluation and Revision 28 to PEP-230, Control Room Operations
05000387/LER-2016-009, Regarding Valid Primary Containment Isolation Actuation During Local Leak Rate Testing Due to Human Performance Error
05000259/LER-2014-006, Regarding Main Steam Relief Valves Lift Settings Outside Technical Specifications Required Setpoint
05000296/LER-2014-002, Regarding Automatic Reactor Scram Due to an Anticipated Transient Without Scram/Alternate Rod Insertion Signal Generated During Functional Test
05000206/LER-1988-011-01, :on 880724,containment Isolation Sys Train a Actuation Occurred Due to Loss of Power to Vital Bus (VB) 1. Caused by Improper Installation of Breaker/Bus Connection. VB 1 Supply Breaker Replaced
05000482/LER-2010-006, Regarding Manual Reactor Trip Due to Trip of Main Feedwater Pump
05000482/LER-2009-001, For Wolf Creek, Regarding Reactor Protection System Actuation and Reactor Trip Due to Main Feedwater Regulating Valve Failing Closed
05000482/LER-2008-003, Re Manual Reactor Trip Due to Loss of Steam Generator Level
05000255/LER-2005-005, Regarding Reactor Protection System and Auxiliary Feedwater System Actuation
05000255/LER-2005-001, Regarding Reactor Protection System and Auxiliary Feedwater System Actuation
05000255/LER-2004-001, Regarding Reactor Protection System and Auxiliary Feedwater System Actuation
05000255/LER-2002-002, Automatic Reactor Trip and Safety System Actuation
05000390/LER-2002-003, Automatic Turbine/Reactor Trip Due to Main Transformer Protection Circuit Ground Due to Inadequate Cable Splice
05000390/LER-2001-004, Turbine/Reactor Initiated by an Invalid AMSAC Signal
RAIO-0518-60110, LLC - Response to NRC Request for Additional Information No. 394 (Erai No. 9407) on the NuScale Design Certification Application
RS-09-020, Request for a License Amendment to Technical Specification 3.5.1, Emergency Core Cooling Systems (ECCS) Operating
TSTF-21-03, TSTF Response to NRC Questions on TSTF-576, Revision 0, Revise Safety/Relief Valve Requirements and Submittal of Revision 1
05000318/LER-1996-001-01, :on 960227,breakers 552-41,552-21 & 552-61 Tripped Open in Plant Switchyard.Caused by Failure of Auxiliary Relay Card in Breaker 552-41.Failed Relay Card Replaced
05000321/LER-1991-017-01, :on 910911,reactor Feedwater Pumps & Main Turbine Tripped on False High Reactor Water Level Signal, Resulting in Reactor Scram.Caused by Personnel Error.Health Physics Technician Counseled
NLS2025035, Application to Revise Technical Specifications to Adopt TSTF-576, Revise Safety/Relief Valve Requirements
TSTF-23-06, TSTF Response to NRC Request for Additional Information on TSTF-576, Revision 2, Revise Safety/Relief Valve Requirements
TSTF-23-04, Technical Specifications Taskforce Transmittal of TSTF-576, Revision 2, Revise Safety/Relief Valve Requirements