L-2017-058, License Amendment Request to Revise the National Fire Protection Association Standard 805 Modifications
2CAN101601, License Amendment Request to Revise the National Fire Protection Association (NFPA) Standard 805 Modifications
GNRO-2014/00068, License Amendment Request to Revise Technical Specification 2.1.1.2
W3F1-2014-0025, Updated Responses to Request for Additional Information Regarding Adoption of National Fire Protection Association Standard NFPA 805 License Amendment Request (LAR)
BSEP 13-0083, Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805
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 1 of 9
NL-14-121, Indian Point, Unit 2, Updated Final Safety Analysis Report (Ufsar), Revision 25, Technical Specification Bases List of Effective Sections
LR-N12-0124, Salem Generating Station, Units 1 and 2 - Attachment 1 to LR-N12-0124, Generic Letter 2004-02, Updated Supplemental Response for Salem Regarding Generic Letter 2004-02
NL-10-107, Indian Point 2 (IP2) Nuclear Power Plant Amendment Update to the Final Safety Analysis Report (Fsar), Revision 22, Improved Technical Specifications Bases
NL-09-1510, Vogtle Electric Generating Plant, Units 1 and 2 - Inservice Inspection Plan Third Inspection Interval Volume 1, Version 3.0
NRC-2009-0046, Annual Monitoring Report, Cover Letter Through Part D, Groundwater Monitoring
L-2009-084, Response to Nrc'S Request for Additional Information
0CAN020803, Notification of Submission of Final Supplemental Response within 90 Days of Completion of Generic Letter 2004-02 Actions
RC-18-0091, License Amendment Request - LAR-16-01490 National Fire Protection Association Standard 805 Program Revisions
L-2019-012, Response to Follow-on NRC RAI No. B.2.3.9-1 a
L-03-113, License No.DPR-60, Cycle 16 Steam Generator Tube Inspection 90 Day Report
BVY 20-025, Response to Request for Information 10 CFR 20.2002 Request for Alternate Waste Disposal at Us Ecology, Idaho (EPID L-2020-D20-0000)
NLS8600148, Nebraska Public Power District,1985 Annual Rept
3F1086-09, Forwards BAW-1976, Small Break LOCA for B&W 177FA Lowered Loop Plant in Response to NUREG-0737,Item II.K.3.31
SNRC-1210, Responds to NRC Re Violations Noted in Insp Rept 50-322/85-31.Corrective Actions:Procedure Changes Initiated, Friction Testing Repeated on Rod 18-27 W/Cooling Water Valve Fully Closed & Test Data Evaluated
2CAN039204, Application for Amend to License NPF-6,revising TS 4.4.5 to Allow Installation of Tube Sleeves as Alternative to Plugging Defective SG Tubes.Encl Rept 51-1212539-00 Withheld
RA-22-0313, Emergency Action Level (EAL) Technical Basis Document, Revision 4, EAL Wallchart, Revision 3 and Alternate TSC and OSC Renovations
LD-84-069, Forwards marked-up CESSAR FSAR Pages Reflecting Review of Tech Specs Re First Sys 80 Plant.Rev Involves Interface Requirements for MSIV & Main Feedwater Isolation Valve Closure Time
W3P84-3343, Forwards Addl Info Re Design Adequacy & as-built Conditions for Masonry Block Wall,Per 841128 Request.Clarification of Stresses Originated for 6-ft Pipe Through Wall SI Encl
W3P82-1868, Confirms 820707 Telcon Agreement Re Schedule for Equipment Audit.Audit Will Start on 820920
1CAN069803, Forwards Relief Requests 1-98-001 & 1-98-002 Re Insps Performed During Second ten-yr ISI Interval at ANO-1,which Ending 970601
05000213/LER-1987-014, :on 870811,fire Barrier Breaches Occurred During Outage Const Activities in Switchgear Room.Caused by Inadequate Contractor Training.Openings Permanently Resealed.Employee/Contractor Program Enhanced
05000328/LER-1986-006-04, :on 861106,during Review of Surveillance Instruction 234.1, Tech Spec Fire Detectors, Eight Fire Detection Instruments Listed as Operable Instead of Nine. Caused by Personnel Error.Instrument Tested
05000281/LER-1986-014-02, :on 861010,results of Type C Containment Leakage Testing Indicated That Inside & Outside Containment Sump Trip Valves Had Leakage Greater than 300 Scfh.Caused by Erosion of Plug & Seat.Valves Will Be Replaced
RC-99-0173, Requests That Info Listed in Rvid,Version 2,be Amended to Reflect Date for VC Summer Nuclear Station,As Marked in Encl to Ltr
05000327/LER-1988-032-01, :on 880903,during Calibr of Auxiliary Bldg Vent Radiation Monitor,Auxiliary Bldg Isolation Occurred.Caused by Procedure Inadequacy & Design of Control Circuitry. Handswitch Placed in Correct Position
05000328/LER-1988-036, :on 880830,determined Failure to Comply W/Tech Spec Action Statement for Reactor Protection Sys Instrumentation Channel.Caused by Personnel Error.Training Will Be Provided & Event Reviewed
05000285/LER-1992-028, :on 920822,RPS Automatically Tripped Reactor on Thermal Margin/Low Pressure,Causing Increase in RCS Pressure Caused by Failure of Ac to Dc Power Converter in EHC Panel. Power Source Modified & Pressures Adjusted
05000482/LER-1991-002-01, :on 911112,Tech Spec 4.8.1.1.2 Could Not Be Satisfied for Emergency Diesel Generator B.Caused by Only Load Description & Not Numercial Value of Load Put Into Surveillance Procedures.Change to TS Submitted
05000482/LER-1992-003, :on 920213,failure to Maintain Proper CR Evs Resulted When Train a Operated in Lieu of Train B.Caused by Inattention to Detail.Standing Order for CR Operators & Sys Operating Procedure to Be Revised
PY-CEI-NRR-1716, Forwards Response to NRC 930825 RAI Re Plant Level 2 PRA Submitted 920715.Development of HPCS Fault Tree & Support Sys to Frontline Sys Dependency Matrix Also Encl
05000259/LER-1981-085-03, /03L-0:on 811211,HPCI Suppression Chamber High Level Switches Discovered to Be Set 8 Inches Above Normal Level.Caused by Change to Tech Specs Not Being Reflected in Table 3.2.B.Switches Physically Lowered
A16000, Nondestructive Exam of 51 Fuel & Reflector Elements from Fort St Vrain Core Segment 1
05000328/LER-1993-004-01, :on 930323,discovered That Surveillance Interval Specified in TS 4.6.5.1 Exceeded.Caused by Ineffective Communication Between Chemistry Personnel. Procedure Revised
05000328/LER-1993-002-03, :on 930318,during Integrity Surveillance,Five 1/2-inch Drain Valves Discovered to Be Misfigured.Caused by Valves Unlocked.Valves Placed in Correct Position & Secured
05000259/LER-1993-001-02, :on 930110,RB Crane SI Could Not Be Verified to Have Been Performed,Due to Ambiguous or Missing Instruction. Operations Personnel Will Be Instructed to Verify Latest SI Performance Through Work Control
05000311/LER-1981-045-99, /99X-0:on 810423,inadvertent Safety Injection Occurred During Unit Cooldown.Caused by Personnel Error in Decreasing Automatic Setpoint.Personnel Counseled on Requirements to Follow Procedures
L-82-520, Interim Deficiency Rept Re Failure of Reactor Shutdown Cooling Bypass Valves to Respond to Command Signals. Initially Reported on 820630.Review Continuing.Next Rept Will Be Submitted by 830115
NLS8900401, Submits Response to Generic Ltr 88-20 & Suppl 1, Individual Plant Exam for Severe Accident Vulnerabilities. Licensee Will Perform Individual Plant Exam by Utilizing Level I Pra. Containment Performance Analysis Will Be Done Per App 1
PNO-III-80-018, on 800124,Cessna 421 Crashed on take-off South of Port of Columbus,Oh W/Cargo of 2 Mci Nonhuman Use Cr-51 & 6 Mci Human Use Ga-67.State of Oh Disaster Svcs Agency Will Perform Radiation Surveys of Crash Site
L-85-406, Forwards Addl Info Re 840522 Application for Amend to License NPF-16,increasing Esfas/Auxiliary Feedwater Actuating Sys Actuation Logic Surveillance Interval from 6 Months to 18 Months
BW080110, Thirteenth Refueling Outage Steam Generator Tube Inspection Report
CY-05-059, License Termination Plan Supplemental Information - Results of the Unconfined Aquifer Pumping Test Conducted in the Industrial Area
NRC 2005-0026, Clarification to Information Regarding the Point Beach Nuclear Plant License Renewal Application
E910-03-031, FSS Report for Cv Interior, 774 Elev and Below, Revision 1
L-PI-04-005, Request for Relief No. 16, Revision 0 for the Third 10-Year Interval Inservice Inspection Program
NL-16-109, Indian Point 2 Improved Technical Specifications Bases
NSD-NRC-97-5348, Forwards Assessment on Surge Line Flooding Effect on Low Power & Shutdown Risk in Response to 970905 Telcon Request
05000261/LER-1976-008, :on 760404,rod L-9 Dropped Due to Loose Electrical Connector.Radial Tilt Induced Resulting in Operation in Degraded Mode.Caused by Electrical Connector at Rv Head Failing Open