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Category:ABNORMAL OCCURRENCE REPORTS (SEE ALSO LER & RO)
MONTHYEARML20083R6631974-02-19019 February 1974
AO 4-2-8:on 740215,during Reactor Coolant Temp Channel Test, Limiting Safety Sys Setting Read Below Required Tech Spec for Channels II & Iv.Cause Under Investigation
ML20083R6871974-02-19019 February 1974
AO 4-2-7:on 740201,both Doors of Personnel Air Lock to Containment Bldg Opened at Same Time.Caused by Erroneous Operation of Hand Wheel Door Lock by Offsite Employee Which Triggered Failure of Interlock.Interlock Inspected
ML20083R7101974-02-0808 February 1974
AO 4-2-6:on 740125,control Rods Did Not Withdraw from Core to Insertion Limit in Conformance W/Tech Specs.Boron Added to Coolant Sys & Procedures re-emphasized to Personnel
ML20083R6961974-02-0606 February 1974
AO 4-2-7:on 740201,both Doors of Containment Air Lock Opened at Same Period of Time.Caused by Offsite Employee Mistakenly Operating Wrong Hand Wheel Opening Inner Air Lock Door
ML20083R7291974-02-0505 February 1974
AO 4-2-5:on 740123,coolant Sys Pressure Transient Experienced During Reactor Coolant Start.Caused by Insufficient Amount of Nitrogen Added to Pressurizer. Nitrogen Procedure Changed
ML20083R7611974-02-0404 February 1974
AO 4-2-4:on 740121,small Leak Observed at Base of Fillet Weld Attaching Pipe Nozzle to Orifice Flange for Flow Transmitter 946D.Dye Penetrant Insp Indicated Pipe Also Cracked.Faulty Weld Ground Out & Flange Socket Redressed
ML20083R8961974-02-0101 February 1974
AO 4-2-3:insp Determined Snubber on Component Cooling Water Return Line & Snubber on Secondary Blowdown Line Defective. Cause Under Investigation.Snubbers Replaced
ML20083R7181974-01-31031 January 1974
AO 4-2-6:on 740125,reactor Brought Critical W/Control Rods Below Required Insertion Limit.Cause Under Investigation
ML20083R7401974-01-28028 January 1974
AO 4-2-5:on 740123,while Placing Reactor Coolant Pump in Svc,Coolant Sys Pressure Exceeded Tech Spec Limit.Probably Caused by Insufficient Nitrogen Vol
ML20084A0141974-01-25025 January 1974
AO 4-2-2:limiting Safety Sys Settings of Channels I & III Discovered Below Tech Spec Requirement.Cause of Setpoint Drift Problem Under Investigation
ML20083R7681974-01-21021 January 1974
AO 4-2-4:on 740121,small Leak Observed in Upstream Orifice Flange Connection for Flow Transmitter 946D in Return Line to Reactor Coolant Loop.Dye Penetrant Insp of Weld Will Be Made
ML20083R9151974-01-21021 January 1974
AO 4-2-3:insp Completed on 740118 Revealed Accumulator Oil Level Plunger of Two Snubbers Below Criteria Mark.Cause of Failure Will Be Determined During Insp of Disassembled Defective Snubbers.Snubbers Replaced
ML20084A0481974-01-18018 January 1974
AO 4-2-1:low Steam Line Pressure Bistables Found Below Required Limit of 600 Psig.Cause of Setpoint Drift Under Investigation.Bistable Recalibr
ML20084A0261974-01-15015 January 1974
AO 4-2-2:limiting Safety Sys Settings of Channels I & III Found Below Tech Spec Requirement.Cause of Setpoint Drift Under Investigation
ML20084A1411974-01-0808 January 1974
AO 4-2-1:low Steam Line Pressure Bistables Found Below Required Limit of 600 Psig.Cause of Setpoint Drift Under Investigation.Bistable Recalibr
ML20084A1941974-01-0404 January 1974
AO 3-2-18:on 731219,analysis of Monthly Pressurizer Pressure Tests Indicated Low Setting for One Bistable.Bistable Recalibr & Sensitivity of Bistable Setpoints Under Investigation
ML20084A2321973-12-28028 December 1973
AO 3-2-17:on 731217,one Bistable Setting Exceeded Tech Spec Limit of 92%.Cause of Setpoint Drift Under Investigation. Channel Recalibr
ML20084A2071973-12-21021 December 1973
AO 3-2-18:on 731219,analysis of Monthly Pressurizer Pressure Tests Indicated Low Setting for One Bistable.Channel Recalibr & Detailed Rept Re Instrument Drift Underway
ML20084A2521973-12-18018 December 1973
AO 3-2-17:on 731217,one Bistable Setting Exceeded Tech Spec Limit of 92% Span.Caused by Instrument Drift.Channel Recalibr & Investigation Initiated for Detailed Analysis of Instrument Capabilities
ML20084A2961973-12-0303 December 1973
AO 3-2-16:on 731116,setting for One Bistable Associated W/ Low Pressurizer Pressure Safety Injection Did Not Reach Tech Spec Limit.Cause Not Yet Determined.Channel Recalibr
ML20084A3111973-11-19019 November 1973
AO 3-2-16:on 731116,during Cold Shutdown,Setting for One Bistable Did Not Reach Tech Spec Requirement.Channel Recalibr
ML20084A3351973-11-12012 November 1973
AO 3-2-15:during 731026 Insp,Leaks Observed at Socket Weld Branch Connections for Vents S-47 & S-51.Caused by Fatigue Failure of Pipe to Socket Welds.Metallurgical Exam Initiated,Vent Valve Assemblies Replaced & RCS Inspected
ML20084A3811973-11-0101 November 1973
AO 3-2-14:on 731017,prior to Maint on Pressurizer Safety Valves,Lift Pressure Settings Exceeded Tech Spec Limit. Valves Disassembled,Inspected & Seat Surfaces Repaired
ML20084A3481973-10-31031 October 1973
AO 3-2-15:during 731030 Insp,Leaks Observed at Socket Weld Branch Connections for Vents S-47 & S-51.Cause & Corrective Actions Will Be Provided Upon Completion of Exam
ML20084A4381973-10-26026 October 1973
AO 3-2-13:on 731011,reactor Tripped Off Line Due to Loss of Generator Field Relay Action.Caused by Gaseous & Particulate Radioactivity in Auxiliary Bldg.Investigation Concluded That No Release to Atmosphere Occurred
ML20084A3861973-10-19019 October 1973
AO 3-2-14:on 731017,set Pressure of Safety Valves Exceeded Tech Spec Limit.Valves Will Be Disassembled,Inspected & Seating Surfaces Repaired
ML20084A4111973-10-12012 October 1973
AO 3-2-13:on 731011,reactor Tripped Off Line Due to Loss of Generator Field Relay Action.Caused by Gaseous & Particulate Radioactivity in Auxiliary Bldg.Samples Analyzed & Investigation Initiated
ML20084A3901973-09-27027 September 1973
AO 3-2-12:on 730913,during Maint of Closure Mechanism of 80-ft Elevation Outer Door of Vapor Container,Inner Door Partially Opened for Short Period.Cause Unknown.Procedures Revised & Supplemented to Prevent Recurrence
ML20084A4441973-09-0606 September 1973
AO 3-2-11:on 730822,discovered That Instrument Setting of One of Four Channels of Low Steam Line Pressure Controllers in High Steam Flow Safety Circuit Below Tech Spec Requirements.Caused by Instrument Drift.Channel Recalibr
ML20084A4671973-08-23023 August 1973
AO 3-2-11:on 730822,instrument Setting for One Channel of Low Steam Line Pressure Controllers in High Steam Flow Safety Circuit Found Below Tech Spec Requirements.Caused by Instrument Drift.Channel Recalibr & Satisfactorily Tested
ML20084A4961973-08-16016 August 1973
AO 3-2-10:on 730802,discovered That Limiting Safety Sys Setting of Channel III of overpower-overtemp delta-T Reactor Trip Circuitry Below Tech Spec Requirements.Caused by Improper Adjustment of Channel.Channel Recalibr
ML20084A9691973-07-0303 July 1973
Telecopy AO 3-2-9:on 730702,first Stage Turbine Pressure Setting Discovered Slightly in Excess of Tech Spec Limits. Cause Unknown.Investigation Underway
ML20084B0261973-06-11011 June 1973
AO 3-2-7:on 730601,one of Two Motor Actuators for MOV 851A Improperly Moved to Closed Position Following Initiation of Automatic Safety Injection as Result of Spurious Signal from High Steam Line Flow Logic.Caused by Logic Malfunction
ML20084A9831973-06-0808 June 1973
Telecopy AO 3-2-8:on 730607,during Performance of Periodic Tests & Calibr Checks Re Reactor Coolant Flow,Flow Trip Settings Discovered Slightly Below Tech Spec Requirements. Cause Unknown.Investigation Underway
ML20084B0391973-06-0101 June 1973
Telecopy AO 3-2-7:on 730601,during Power Level Testing, Spurious Safety Injection Signal Initiated Automatically Starting High Head Pumps.One of Two Valves Tying Discharges Driven to Closed Position.Cause Unknown
ML20084A9921973-05-29029 May 1973
AO 3-2-6:on 730519,within Minutes After Starting High Head Safety Injection Pump 23,accumulator Level Observed Not Changing & Pump Discharge Pressure Decreased to Existing Rcs.Caused by Personnel Error
ML20084B0131973-05-25025 May 1973
AO 3-2-5:on 730518,pressure Transient within RCS Experienced Due to Closure of Certain Air Operated Valves in Reactor Coolant Letdown Sys.Caused by Moisture Freezing in Air Supply Line Refrigerant Dryer of Instrument Air Sys
ML20084B0981973-05-14014 May 1973
Ao:On 730511,during Pressurization,Preparatory to Starting Reactor Coolant Pumps for Performance of Hydrostatic Leak Test of Rcs,Leak Apparently Occured in Valve 204B Located in Loop 21 Charging Line.Caused by Defect in Stud Holes
ML20084B1101973-05-0404 May 1973
Ao:On 730426,MSIV 22 Failed to Close in Response to Manual Signal by Control Room Operator.Caused by Binding of Linkage Connecting Solenoid Armature to Valve.Solenoids Thoroughly Cleaned of All Lubricant
1974-02-08
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML20217H8501999-10-14014 October 1999
Safety Evaluation Supporting Amend 197 to License DPR-64
ML20206U2551999-02-0909 February 1999
Safety Evaluation Supporting Amend 187 to License DPR-64
ML20236Y1571998-08-0303 August 1998
Part 21 Rept Re ASTM A351,GR. CF8 Matl at Indian Point Being Out of Specifications in Molybdenum & Chromium.Cause & Corrective Actions Are Not Stated
ML20236V4281998-07-13013 July 1998
Safety Evaluation of TRs WCAP-14333P & WCAP-14334NP, PRA of RPS & ESFAS Test Times & Completion Times. Repts Acceptable
ML20236T5511998-06-24024 June 1998
Consolidated Edison Co of Ny,Indian Point Unit 2,Drill Scenario Number 1998C
ML20248B2371998-03-31031 March 1998
Revised Monthly Operating Rept for March 1998 for Indian Point Station Unit 2
ML17264A9381997-07-10010 July 1997
Deficiency Rept Re Potential Safety Hazard Associated w/FM-Alco 251 Engin,High Pressure Fuel tube-catalog: 4401031-2 in Which Dual Failure Mode Exists.Caused by Incorrect Forming Process
ML18153A1431997-06-10010 June 1997
Part 21 Rept Re Possible Machining Defect in Certain Stainless Steel Swagelok Tube Fitting Bodies.Facilities Have Been Notified About Possible Problem
ML20210E3591997-03-27027 March 1997
Part 21 Rept Re Sorrento Electronics Inc Has Determined Operation & Maint Manual May Not Adequately Define Requirements for Performing Periodic Surveillance of SR Applications.Caused by Hardware Failures.Revised RM-23A
ML1005008001997-02-28028 February 1997
Conditional Extension of Rod Misalignment TS for Indian Point 3.
ML20115J3981996-07-22022 July 1996
Interim Part 21 Rept Re 3/4 Schedule 80 Pipe Furnished to Consolidated Power Supply.Investigation Revealed Only One Nuclear Customer Involved in Sale of Matl
ML20096E5101995-12-31031 December 1995
Resubmitted Rev 13 to QA Program
ML17059A3611994-07-0606 July 1994
Emergency Action Level Verification & Validation Rept.
ML17311A0181994-05-13013 May 1994
New York State EAL Upgrade Project Verification & Validation Rept.
ML20029C7801994-03-31031 March 1994
Monthly Operating Rept for Mar 1994 for Indian Point Unit 1. W/940415 Ltr
ML20056E5171993-08-31031 August 1993
Technical Review Rept, Tardy Licensee Actions
ML20062J2281993-07-23023 July 1993
Consolidated Edison Co of Ny Indian Point Unit 2,Drill Scenario 1993
ML20118A2681992-12-31031 December 1992
Consolidated Edison Co of Ny Indian Point,Unit 2 Exercise Scenario,1992
ML20125C7161992-12-0707 December 1992
Part 21 Rept Re Possibility for Malfunction of Declutching Mechanisms in SMB/SB-000 & SMB/SB/SBD-00 Actuators. Malfunction Only Occurs During Seismic Event.Balanced Levers May Be Purchased from Vendor.List of Affected Utils Encl
ML20127P5861992-11-23023 November 1992
Followup to 921005 Part 21 Rept Re Potential Defect in SB/SBD-1 Housing Cover Screws.Procedure Re Replacement of SBD-1 Spring Cover Bolts Encl.All Fasteners Should Be Loosened & Removed.List of Affected Utils Encl
ML20096H2301992-05-21021 May 1992
Special Rept:On 920504,south Side Lower Electrical Tunnel Detection Sys 8 Taken Out of Svc for Mod to Reposition Detection Sys Run of Conduit.Detection Sys Declared Operable on 920520 After Mod Completed & Sys Retested
ML20079F9181991-05-31031 May 1991
Structural Evaluation of Indian Point,Units 2 & 3 Pressurizer Surge Lines,Considering Effects of Thermal Stratification
ML20059E9461990-08-31031 August 1990
Nonproprietary Rev 2 to Indian Point 2 Tube Fatigue Reevaluation
ML20059G2011990-07-31031 July 1990
Final Rept on Steam Generator Insp, Repair & Restoration Efforts During 1990 Midcycle Insp
ML20058K4151990-06-30030 June 1990
Steam Generator Insp,Repair & Restoration Program Presentation to Nrc
ML20058K4121990-06-30030 June 1990
Status Rept,Indian Point Unit 2 Mid-Cycle Steam Generator Insp Presentation to Nrc
ML20043A4891990-05-30030 May 1990
Nonproprietary Indian Point Unit 2 Steam Generator Insp, Repair & Restoration Program.
ML20042F1441989-12-31031 December 1989
New York Power Authority Annual Rept for 1989. W/900430 Ltr
ML19332B9371989-11-30030 November 1989
Nonproprietary Info Presented to NRC Re Indian Point Unit 2 Steam Generator Secondary Side Loose Objects.
ML19332D6661989-10-31031 October 1989
Nonproprietary Rev 2 to Indian Point Unit 2 Steam Generator Girth Weld/Feedwater Nozzles Rept Spring,1989 Outage.
ML20247J8161989-07-31031 July 1989
Safety Evaluation for UHS Temp Increase to 95 F at Indian Point Unit 3
ML20248B3171989-06-30030 June 1989
Rev 1 to Nonproprietary WCAP-12294, Indian Point Unit 2 Steam Generator Girth Weld/Feedwater Nozzles Rept,Spring 1989 Outage
ML20248D3631989-06-30030 June 1989
Rev 1,to Indian Point Unit 3 Reactor Vessel Fluence & Ref Temp PTS Evaluations
ML20247N5331989-05-31031 May 1989
Nonproprietary Indian Point Unit 2 Steam Generator Girth Weld/Feedwater Nozzles Rept Spring,1989 Outage
ML20247J8071989-05-31031 May 1989
Containment Margin Improvement Analysis for Indian Point Unit 3
ML20247G5171989-04-30030 April 1989
Monthly Maint Category I Rept Pages from Monthly Operating Rept for Apr 1989 for Indian Point
ML20244C3311989-04-10010 April 1989
Safety Evaluation Supporting Amend 137 to License DPR-26
ML20248F4211989-03-31031 March 1989
NSSS Stretch Rating-3,083.4 Mwt Licensing Rept
ML20235V5931989-03-0202 March 1989
Special Rept:During Cycle 6/7 Refueling Outage Scheduled from Feb-May 1989,openings Will Be Made in Plant Penetration Fire Barriers in Order to Install Various Mods. Fire Watches Posted & Fire Detection Tests Completed
ML20248F3001988-12-31031 December 1988
10CFR50.59(b) Rept of Changes,Tests & Experiments Completed in 1988
ML20246E2711988-12-31031 December 1988
Con Edison 1988 Annual Rept
ML20196D3011988-10-31031 October 1988
Reactor Vessel Matl Surveillance Program for Indian Point Unit 2 Analysis of Capsule V
ML20155H2541988-09-30030 September 1988
Rev 2 to Indian Point Unit 2 (NRC Bulletin 88-008 Thermal Stresses in Piping Connected to RCS) Indentification of Unisolable Piping & Determination of Insp Locations
ML20154M5661988-08-31031 August 1988
Monthly Operating Rept for Aug 1988 for Indian Point Station Unit 2
ML20154M5691988-07-31031 July 1988
Revised Page to Monthly Operating Rept for Jul 1988 for Indian Point Station Unit 2
ML20151N7881988-07-31031 July 1988
Rev 1 to Charpy Toughness & Brittle Transition Temp Characterization of HAZ High Hardness Zone of Indian Point Unit 2 Steam Generator Girth Weld by Gleeble Weld Thermal Cycle Simulation
ML20006B4741988-05-31031 May 1988
Nonproprietary Indian Point Unit 2 Evaluation for Tube Vibration Induced Fatigue.
ML20151N7821988-05-31031 May 1988
Fracture Sensitivity Study of Girth Weld 6 Repaired Configuration,Indian Point Unit 2
ML20153F9161988-04-28028 April 1988
Changes,Tests & Experiments - 1986
ML20151T4681988-01-31031 January 1988
Experimental & Finite Element Evaluation of Spent Fuel Rack Damping & Stiffness
1999-02-09
[Table view] |
Text
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February 4, 1974 Re: Indian Point Unit No. 2 Facility Operating License No. DPR-26, AO 4-2-4 Mr. John F. O' Leary, Director 50-247 Directorate of Licensing -
Office of Regulation U.S. Atomic Energy Commission Washington, D.C. 20545 Dra Mr. O' Leary:
The following report is provided pursuant to the requirements of Section 6.12.2 (a) of the Technical Specifications to Facility Operating License DPR-26.
During the course of a routine plant inspection on January 21, 1974, a cmall leak was observed at the upstream orifice flange connection for flow transmitter 946D in the six-inch Residual Heat Removal System return' line (Line No. 335) to No. 21 reactor coolant loop. The leak was at the b:se of a fillet weld attaching a 3/4" pipe nozzle to the orifice flange and was on the pipe side of the weld. At the time of this observation, the reactor was in the cold shutdown condition with the Residual Heat R:moval System in service and pressurized to about 400 psig. After depressurizing and draining the section of piping at the orifice flange, a dye penetrant inspection of the weld area was performed to determine the extent of the crack. The results of this inspection indicated a crack at the base of the fillet weld approximately 1200 around the circumference. The faulty weld was then ground out. A dye penetrant check of,the end of the pipe indicated that it was also cracked.
Approximately a one-inch length of this 3/4" pipe, including the section that had a crack in it, was then cut off and sent to our metallurgical laboratory for further investigation and study. The socket in the flange was redressed and the new end of the pipe was reinserted and rewelded to the flange.
This leak did not adversely' affect the safe operation of Indian Point 1 Unit No. 2'and was well within the makeup capabilities of the Chemical and Volumo Control System. Had the plant been operating, the Resid al Heat 8304040525 740204 9. % &
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Re: Indian Point Unit No. 2 Facility Operating License No. DPR-26, AO 4-2-4 (cont'd.)
R:moval System would have been isolated from the Reactor Coolant System cnd, therefore, would not have experienced primary system pressure. For these reasons, the safety implications of the leakage are considered negligible.
f,g very truly yours, t bWe dLkLsl ,
Warren R. Cobean, Jr.1 J
jm cc: James P. O'Reilly i
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