SBK-L-17156, Response to Request for Additional Information Regarding License Amendment Request 16-03 Related to Alkali-Silica Reaction
TMI-14-051, 2013 Radiological Environmental Monitoring Program Report
TMI-02-066, Submittal of 2011 Radiological Environmental Monitoring Program Report
TMI-10-030, Annual Radiological Environmental Operating Report
TMI-09-053, Annual Radiological Environmental Operating Report for 1 January Through 31 December 2008
NOC-AE-09002400, Technical Specification Bases Control Program
LIC-08-0037, Calculation FC05694, Calculation of Minimum Reactor Coolant Time Using Shutdown Cooling System, Attachment 2, Pages 51 Through 90
L-06-075, Discharge Monitoring Report Permit No. PA0025615
L-05-202, Discharge Monitoring Report (NPDES) Permit No. PA0025615
NRC 2004-0064, Transmittal of Point Beach, Units 1 and 2 - Emergency Plan Implementing Procedure, Revision to Emergency Action Levels
L-PI-03-118, Request for Relief No. 14, Revision 1 for the Third 10-Year Interval Inservice Inspection Program
ELV-02272, Requests Authorization to Conduct After Hours Unannounced Drill During CY91 in Lieu of After Hours Unannounced Emergency Prepared Exercise.Items That Drill Will Provide Listed
E901-03-018, Radiological Environmental Monitoring Report
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
NRC-03-0018, Submittal of Changes to Emergency Response Data System for Integrated Plant Computer System Replacement
E910-02-019, Radiological Environmental Monitoring Report
IA-88-555, List of Notices of Violation for Insps W/Transmittal Dates of Jan 1978 to Present
RA-21-0097, Notification of Permit Revision Request Regarding Copper and Zinc Limits
FVY-86-117, Forwards Response to NRC 861024 Request for Addl Info Re Containment Safety Study.Engineering Scoping Study Examining Feasibility of Improving Reliability of Valves in Reactor Bldg Complete.Review of Proposed Mods Underway
NPL-97-0794, Provides Revised Response to GL 96-06, Assurance of Equipment Operability & Containmnt Integrity During Design Basis Accident Conditions, as Suggested in 971104 & 13 Ltrs
PY-CEI-NRR-0015, Forwards Public Version of Rev 1 to Emergency Plan OM-15A, Incorporating Encl Responses to Questions in NRC 811221 & s & Addressing SER Outstanding Issue 19
B12610, Responds to Questions Posed During 870701 & 21 Telcon Re Fluence Calculations & Util Intentions Re Evaluation of PTS Upon Conversion from Stainless Steel to Zircaloy Clad Fuel. NRC Should Approve RT PTS Evaluation Per 10CFR50.61
FVY-87-109, Forwards Info on Regional Insp Re Mark I Containment Mods (USI A-7) Conducted on 871005-09
B12293, Forwards Reviews for Isap Topics 1.06, Wind & Tornado Loadings/Tornado Missiles & 1.31, Pipe Break Outside Containment, Per 850517 Commitment
05000315/LER-1986-002, :on 860217,discovered That Three Reactor Trip Sys Interlock Setpoints Not Set Per Tech Spec Requirements. Caused by Misinterpretation of Tech Specs.Procedures Revised to Agree W/Tech Specs
05000305/LER-1985-016, :on 850725,control Operator Observed That Position Indication for Containment Isolation Valves & Steam Exclusion Dampers Changed Position.Caused by Power Interruption & Loss of Sys Redundancy
HBL-85-035, Forwards Complete Response to Item 75 of NRC 850123 & 0214 Requests for Addl Info Re Safstor Decommissioning of Facility.Util Response to NRC Questions on Safstor Decommissioning Complete
05000397/LER-1985-007, :on 850131,one Generator Primary Protection Lockout Relay,Spuriously Tripped Causing Abnormal Electrical Operations.Caused by 500 Kv Breakers Opening & Causing Overspeed Condition
05000331/LER-1984-023, :on 840619,w/plant at 75% Power,Surveillance Test Revealed Three of Four Reactor High Pressure Switches Out of Tolerance in Nonconservative Direction.Switches Recalibr,Returned to Svc & Retested
05000369/LER-1991-009-01, :During Wk of 910708,hardware Limits for Certain Process Control Equipment in Reactor Protection Sys Exceeded.Caused by Design Deficiency Due to Deficient Documentation.Resistors Changed
05000331/LER-1991-007, :on 910806,inadvertent Inboard High Pressure Coolant Injection Steamline Isolation Occurred.Caused by Personnel Error.Rcic Speed Sensors Replaced & RCIC Turbine Insulation Pad Stenciled
NRC-99-0050, Forwards Rev 7 to ITS Submittal of 980403,to Provide Remaining Responses to NRC RAIs Re ITS Sections 3.5 & 3.8. Revision Instructions Included
05000280/LER-1982-089-03, /03L-0:on 820826,B Core Cooling Monitor Failed, Causing Spurious Alarms.Caused by Relay Failure on Printed Circuit Cards A-1 & A-2.Boards Repaired & Core Cooling Monitor Returned to Svc
05000366/LER-1982-137-03, /03L-0:on 821206,w/unit in Hot Shutdown Following Reactor Trip,Specific Iodine Activity of Reactor Coolant Exceeded Tech Specs.Caused by Leaking Fuel Rod.Sampling & Analyses Performed Every 4 H.Software Corrected
05000373/LER-1982-075-03, /03L-0:on 820726,air Hose to Automatic Dispatch Sys Valve 1B21-FO13V Ruptured So That in Compressors Could Not Maintain Sys Pressure.Probably Caused by Hose Age, Exceeding Min Bend Radius & Exposure to High Temp
05000325/LER-1981-042-01, /01T-0:on 810327,during Investigation of Ctip Machine Isolation Ball Valve Operability,Power to Retract TIP Probe Was Lost.Caused by Design Deficiency Which Allows Removal of Power to Probe Drive Motor While de-energized
PNO-II-82-042, on 820417,fire Discovered in Steam Generator Repair Bldg Located Near Unit 1.Both Reactors Not Affected by Fire.Fire Extinguished by Surry County Fire Dept & Plant Fire Brigade.Cause Not Known
05000327/LER-1980-103-03, /03L-0:on 800627,containment Relief Switches 1-PS-30-46 A&B, 47 A&B,48 A&B Did Not Continue to Operate Throughout Loca.Caused by Switches Not Being Tested to Potential LOCA Radiation Levels.Switches Will Be Evaluated
05000366/LER-1980-070-03, /03L-0:on 800422,while Performing Surveillance Procedure HNP-2-3405,reactor Core Isolation Cooling Trip & Throttle Valve Closed When Turbine Speed Reached 2,000 Rpm. Cause Not Identified.Trip & Throttle Valve Reset
05000321/LER-1980-001-01, /01T-0:on 800103,while Performing Maint on RHR LPCI Injection Valve,Clevis Pin Found Missing from Strut on Drywell Penetration Restraint.Caused by Improper Installation.New Clevis Pin Installed
05000321/LER-1979-065-01, /01T-0:on 790823,during Hanger Surveillance,Piping Section in Plant Instrument Air Sys Was Found Not to Be Seismically Supported Per Original Design.Caused by Failure to Install Proper Hanger.Proper Hanger Installed
05000245/LER-1979-028-03, /03L-0:on 790828,after Venting Pressure Suppression Chamber,Vent Bypass Valve Failed to Close.Caused by Rust Scale Contamination in Pivot Mechanism of Main Valve Preventing Valve Plug from Rotating Completely Closed
05000338/LER-1979-079-03, /03L-0 on 790608:quadrant Power Tilt Ratio of 1.0292 in Upper Core & 1.0285 in Lower Core Obtained from ex-core Detector Current Calculations.Figures Exceed Tech Spec Limits.Cause Unknown.Investigation Initiated
NL-14-0143, Refueling Outage 1R25 Inservice Inspection Report
L-10-014, Technical Specification 5.6.6.2.1 - Steam Generator Inspection Report
BYRON 2009-0003, Steam Generator Inservice Inspection Summary Report for Refueling Outage 14
L-05-177, Responses to a Request for Additional Information (RAI Dated November 1, 2005) in Support of License Amendment Request Nos. 302 and 173
L-2005-171, PTN-ENG-SEIJ-04-073, Rev a, Florida Power & Light Co. Proposed License Amendment Support Evaluation of Design Inputs Utilized for Westinghouse Licensing Report
BW050064, Eleventh Refueling Outage Steam Generator Inservice Inspection Summary Report
ET 05-0013, Response to Request for Additional Information Regarding the Steam Generator Tube Inspection Summary Reports for the Fall 2003 Outage
BW040017, Tenth Refueling Outage Steam Generator Inservice Inspection Summary Report
HNP-04-076, One Year Special Report, Steam Generator Tube Inspection Results
NUREG/CR-5176, Updated Board Notification 89-01 Re Lll NUREG/CR-5176, Seismic Failure & Cask Drop Analyses of Spent Fuel Pools at Two Representative Nuclear Power Plants
B12724, Revised Inservice Insp Testing Program. W/12 Oversize Drawings
BECO-87-158, Responds to 870714 Request for Addl Info Re Second 10-yr Interval Inservice Insp Program Plan.Ref List of Calibr Blocks,Including Identifications,Matl Specs & Sizes Encl