ML19261F295
From kanterella
Revision as of 20:32, 1 February 2020 by StriderTol (talk | contribs) (Created page by program invented by StriderTol)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
| ML19261F295 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 01/06/1978 |
| From: | Ohanlon J METROPOLITAN EDISON CO. |
| To: | |
| References | |
| NUDOCS 7910250749 | |
| Download: ML19261F295 (1) | |
Similar Documents at Three Mile Island | |
|---|---|
Category:REPORTABLE OCCURRENCE REPORT (SEE ALSO AO
MONTHYEARML20024E6601983-08-0404 August 1983
[Table view]RO 83-18:on 830804,during Maint Insps on Hydraulic Snubbers, Two Small Bore Snubbers Identified as Being Installed on RCS Hot Leg Vent Piping Downstream of Valves RC-V15A & B.Caused by Personnel Error ML20050A5431982-03-18018 March 1982 Ro:On 820318,water W/Ph of 3.5 Was Discharged from Secondary Waste Water Neutralizing Tank WT-T-1 Into Plant Discharge Effluent Stream.Caused by Accidental Overflow of Approx 2,500 Gallons of Water from Neutralizing Tank ML20039C6331981-12-23023 December 1981 RO 81-14:nine Pipe Support & Structural Interfacing Areas Identified Requiring Further Evaluation & Potential Corrective Action.Failure of Pipe Supports Could Damage safety-related Components.Followup Rept Expected in 14 Days ML19345A9741980-11-17017 November 1980 RO 80-013/03L-0 Committed to Forwarding Final Valve Testing Evaluation by 801101.Investigation Requires More Time than Anticipated.Requests Extension Until 801219 ML19338F5461980-09-23023 September 1980 RO 80-17:on 800923,determined That There Is Insufficient Documentation to Demonstrate Environ Qualification of Brakes on Motor Operators for Purge Valves AH-V1B & AH-V1C. Cause Not Stated.Corrective Action Will Be Submitted ML19337A8631980-09-23023 September 1980 RO 80-17:determined Insufficient Documentation to Demonstrate That Brakes on Motor Operators for Purge Valves AH-V1B & AH-V1C Environmentally Qualified.Redundant Valves Fully Qualified ML19330C3041980-08-0404 August 1980 Ro:On 800716,during Radiation Emergency Exercise,Use of Emergency Notification Sys Phone Line Became Uncertain, Making Accurate Info Transfers Questionable.Recommends That Microphone Headset Be Made Available to Allow Mobility ML19326E2061980-07-22022 July 1980 RO 80-15/4P:confirming 800722 Telcon,On 800721,river Water Discharge Temp Was More than 3 F Below Inlet River Water Temp for Approx 6-h.Incident Will Be Reviewed to Determine Corrective Action.Review to Be Included in Followup Rept ML19326E0841980-06-27027 June 1980 RO 80-12/IP:during Review of Equipment Required by IE Bulletin 79-01B,found That post-accident Pressure & Temp Would Be Higher than Postulated in NSC 730705 Rept.Short & long-term Action to Be Discussed in Followup Rept ML19323H3421980-06-0606 June 1980 RO-80-10/1P:on 800603,B&W Ltr Informed Util of Potential Reportability of Steam Generator Tube Rupture Safety Analysis.Caused by Nonconservative Error in Fsar.Corrective Action Will Be Addressed in Followup Rept ML19323D5151980-04-23023 April 1980 RO 80-07/TP:on 800423 Results of Radiographic Exams of Piping Welds Associated W/Reactor Bldg Penetrations 415-16 for Leak Rate Test Sys Showed Several Welds Not within Tech Specs.Unit Is Shut Down ML19309A7831980-03-26026 March 1980 RO 80-06/1P:on 800325,during Refueling Frequency Reactor Bldg Local Leak Rate Test of Valve RB-V7,updated Total Leakage Exceeded Tech Specs.Testing Is Continuing W/Approx 1/4 of Test Program Complete ML19322B5081979-11-0808 November 1979 RO 79-17/1P:on 791108 Support Boltings for Rigid Anchor DHH-127A Found to Be Inadequate in Design & Installation. Details & Corrective Action Will Be Addressed in Followup Rept ML19308A5241978-02-0202 February 1978 RO 78-08/1P:on 780202,util Informed by NSSS of Deviation Discovered Between FSAR Steam Line Break Analysis & Actual Operating Parameters.Procedure Mod Will Limit Secondary Site Water Volume ML19308A5081978-01-20020 January 1978 RO 78-05/1P:on 780119,Plant Operations Review Committee Determined That Containment Integrity Had Been Violated by Opening of Valve BS-V47A & B Following Test of BS-P1A & B. Procedure Mod Will Terminate Operation of BS-V-47A & B ML19308A5091978-01-17017 January 1978 Confirms Telcon on 780117 Re Excessive Leak Off from 1A Makeup Pump Discharge Isolation Valve MU-V-74A Resulting in Gaseous Release 500 Times in Excess of 10CFR20 Limits for Xe-133 & Xe-135.Valve Back Seated,Isolating Leak Off ML19308A5111978-01-12012 January 1978 RO 78-2/1P:on 780112,it Was Recognized That Continuous Fire Watch Was Not Established for 63-h After a Emergency Diesel Room Sprinkler Isolation Valve FS-V-156 Was Closed.Caused by Personnel Error.Fire Watch Established on 780112 ML19261F2951978-01-0606 January 1978 RO 78-01/1P:on 780106,util Informed by B&W That During Certain Transients,Difference Between out-of-core Detectors & Heat Balance May Exceed 4% of Full Power.Frequency of Heat Balance Checks Increased ML19261F2961977-12-19019 December 1977 RO 77-29/1P:on 771216,B&W Informed Met Ed That Volume Per Tech Specs Was Insufficient to Bring Reactor to Cold Shutdown condition.TMI-1 Immediately Increased Volume of Source of Concentrated Boric Acid Solution ML19268B9681977-12-19019 December 1977 RO 77-30/1P:on 771216,B&W Informed Met Ed That Software Used to Calculate Incore Imbalance Was in Error,In That Background Correction Was Misapplied as Detector Depletes. Error Will Be Corrected Prior to Cycle 4 Operation ML19308A5131977-12-0505 December 1977 RO 77-27/1P:on 771203,technician Unable to Open Outer Door of Reactor Bldg Personnel Air Lock.Caused by Jammed Linkage. Reactor Shut Down During Repair of Door Closing Mechanism ML19308A5161977-11-15015 November 1977 RO 77-26/1P:on 771114,quadrant Power Tilt Exceeded +2.66% W/Reactor at 15% Full Power During Power Escalation Following Reactor Trip.Within 4-h,quadrant Power Tilt Not Reduced ML19261F3091977-10-0505 October 1977 RO 77-23/1T:on 770920,both Doors of Reactor Bldg Emergency Personnel Access Hatch Open Simultaneously.Caused by Personnel Error.Doors Closed.Forwards LER 77-23/1T ML19261F3121977-09-20020 September 1977 RO 77-23/1P:on 770920,containment Integrity Violated When Both Doors in Reactor Bldg Personnel Access Hatch Were Open for Approx 10 Minutes.Caused by Problem W/Door Interlock. Interlock & Door Seals Tested ML19261F3131977-09-0909 September 1977 RO 77-22/1T:on 770825,B&W-required Operating Restrictions on Decay Heat Removal Pumps Found Contrary to Description of Pump Operating Capability Per Fsar.Caused by B&W Failure to Perform Safety Evaluation.Corrective Action by 770923.W/LER ML19308A5201977-08-26026 August 1977 RO 77-22/1P:confirms 770825 Telcon That as Result of B&W New Decay Heat Pump Operating Limitations,Remedial or Corrective Action Could Be Required to Prevent Unsafe Condition. Procedure Change Initiated to Incorporate New Limitations ML19261F3161977-08-12012 August 1977 RO 77-19/3L:on 770713,leak in Miscellaneous Waste Evaporator Feed Tank Immersion Heater Shorted Out Heater & Feed Pump. Caused by Matl Failure in Heater Temp Element.Heater Replaced & Operating Procedure Modified.Forwards LER ML19261F3191977-08-0909 August 1977 RO 77-21/1T:on 770726,radiation Monitor RM-A2 Out of Svc & Sample of Reactor Bldg Atmosphere Not Taken within 8-h. Caused by Improper Reassembly Following 770725 Surveillance Test.Monitor Repaired & Sample Taken.Forwards LER ML19210A5441977-08-0505 August 1977 Followup Rept to 770707 Ltr Re Ro:On 770613,radwaste Shipment Leak Discovered.Caused by Acidic Attack of Mild Steel Container at Locations of Incomplete Coverage W/Internal Corrosion Resistant Coating ML19261F3221977-07-27027 July 1977 RO 77-18/31:on 770629,Nuclear Svcs River Water Pump NR-P1C Shaft Failed While Operating,Leading to Operation in Degraded Mode.Caused by Mechanical & Procedural Problems. Pump Overhauled,Preventive Maint Procedures to Be Changed ML19308A5211977-07-27027 July 1977 RO 77-21/1P:on 770726,radiation Monitor RM-A2 Out of Svc & Sample of Reactor Bldg Atmosphere Not Taken within 8-h. Caused by Improper Reassembly Following 770725 Surveillance Test.Monitor Repaired & Valid Sample Taken ML19308A5231977-07-20020 July 1977 RO 77-20/4P:on 770719,ETS Violated in That Plant River Water Discharge Temp Exceeded Limit of 3 F Below Inlet River Water Temp for 10 Minutes.Caused by Passing Storm Front Which Decreased Ambient Air Temp from 94 to 82 F ML19308A5221977-07-20020 July 1977 RO 77-20/4P:on 770619,ETS Violated in That Plant River Water Discharge Temp Exceeded Limit of 3 F Below Inlet River Water Temp.Caused by Passing Storm Front Which Lowered Ambient Air Temp from 94 to 82 F ML19261F3251977-07-20020 July 1977 RO 77-17/3L:on 770624,inner Door Seal on Personnel Access Hatch Leaked Excesively,Causing Total Reactor Bldg Local Leakage to Exceed Tech Spec 4.4.1.2.3.Caused by Improper Door Installation Procedures.Forwards LER ML19261F3291977-07-0707 July 1977 RO on 770713:Tri-State Motor Transit Co Driver Observed Radwaste Leaking from Container on Truck While at Rest Area. Caused by Pinhole Leak in One Liner.Truck Returned to Station Site ML19261F3301977-06-28028 June 1977 RO 77-15/3L:on 770529,one of Two Units of 230/4.16 Kv Auxiliary Transformer de-energized,leading to Operation in Degraded Mode Permitted by Tech Spec 3.7.2.b.Caused by Wiring Error on Current Transformers.Sys Repaired.Ler Encl ML19261F3331977-06-20020 June 1977 RO 77-14/3L:on 770524,reactor Bldg Isolation Valve RB-V7 Failed to Close on Engineering Safeguards Test Signal, Leading to Operation in Degraded Mode Permitted by Limiting Condition for Operation Per Tech Spec 3.6.6 ML19261F3361977-06-0909 June 1977 RO 77-12/3L:on 770519,engineered Safeguard Pump MU-P-1A Was Removed from Svc Prior to Testing Operability of Redundant Pump.Caused by Personnel Error.Redundant Pump Tested Satisfactorily on 770519.Forwards LER 77-12/03L ML19261F3391977-06-0707 June 1977 RO 77-13/4T:on 770523,analysis of Plant River Water Discharge Showed Concentration of 760 Parts Per Million. Caused Either by Slug of High Concentration of Suspended Solids from Filter Sump or by Poor Sample.Forwards LER ML19322A4381977-05-10010 May 1977 RO 77-07/1P:on 770510,both Reactor Containment Bldg Emergency Personnel Access Doors Open at Same Time.Caused by Closed Outer Door Mechanism While Door Still Was Open ML19322A4391977-04-18018 April 1977 RO 77-04/1P:on 770415,nonconservative Error Found in Safety Analysis to Support Increase in Pressurizer Code Safety Valve Setpoint to 2500 Psig.Caused by Erroneous Assumptions Made for Total Combined Relief Valve Capacity ML19253B9631977-03-22022 March 1977 RO 77-02/04T:on 770304,river Water Discharge Temp Exceeded Inlet River Water Temp Limitation by 14 F Per Ets.Caused by Insufficient Procedures,Personnel Error,Abnormal Ambient Conditions & Low River Water Temp.Ler Encl ML19322A4401977-03-20020 March 1977 RO 77-05/4P:on 770319,river Water Discharge Temp Differential Change More than 2 Degrees in 1-h.Steps Taken to Reduce Decay Heat Removal Load,Reducing River Water Temp Change ML19253B9621977-03-0808 March 1977 RO 77-02/04P:on 770304,river Water Discharge Temp Exceeded Inlet River Water Temp by 14 F,Violating 12 F Limit Per Ets. Caused by Low Inlet Temp & High Ambient Temp.Temp Reduced Below Limit ML19210A2681977-02-0303 February 1977 RO 77-1/1P:on 770202,both a & B Diesel Generators Had Conditions Affecting Starting Capabilities in Response to Loss of Offsite Power.Both Problems Caused by Cranking Timer Timing Out Prior to Oil Low Speed Switch Operation ML19261F1771976-12-29029 December 1976 RO 76-50/4O:on 761221,radionucleide Surface Water Level Ten Times Background Level.Probable Cause Is Chinese Weapons Test,Since Neither Zr-95 Nor Nb-95 Discharged to River from Station During Nov 1976 ML19261F1881976-12-26026 December 1976 RO 76-047/01T:on 761216,sodium Hydroxide Tank Levels Fell Below Required Minimum.Caused by Personnel Misreading Tank Level Data & Technical Analyst Failing to Notice Mistake ML19261F2531976-12-20020 December 1976 RO 76-48/4o:on 761208,isotope Level Samples from TMI Environs Ten Times Background Levels.Chinese Weapons Tests Probably Caused Excesses of I-131,Zr-95,Nb-95,Ce-141 & Ru-103 in Sediment Samples from TMI ML19261F2561976-12-14014 December 1976 Nonroutine 30 Day Rept 75-03,advising That Heated Post Gap Measurements within Acceptable Values.Test Schedule Change Warranted from Weekly to After Each Startup,Once Full Steam Equilibrium Reached & on Monthly Basis Thereafter ML19308A4811976-12-13013 December 1976 RO 76-46/1P:on 761212,boron Concentrations in Borated Water Storage Tank Found Below Required Minimum.Boric Acid Injected to Return Concentration to Necessary Valve.Possibly Caused by Water Leakage from Reactor Coolant Sys 1983-08-04 Category:LER) MONTHYEARML20024E6601983-08-0404 August 1983
[Table view]RO 83-18:on 830804,during Maint Insps on Hydraulic Snubbers, Two Small Bore Snubbers Identified as Being Installed on RCS Hot Leg Vent Piping Downstream of Valves RC-V15A & B.Caused by Personnel Error ML20050A5431982-03-18018 March 1982 Ro:On 820318,water W/Ph of 3.5 Was Discharged from Secondary Waste Water Neutralizing Tank WT-T-1 Into Plant Discharge Effluent Stream.Caused by Accidental Overflow of Approx 2,500 Gallons of Water from Neutralizing Tank ML20039C6331981-12-23023 December 1981 RO 81-14:nine Pipe Support & Structural Interfacing Areas Identified Requiring Further Evaluation & Potential Corrective Action.Failure of Pipe Supports Could Damage safety-related Components.Followup Rept Expected in 14 Days ML19345A9741980-11-17017 November 1980 RO 80-013/03L-0 Committed to Forwarding Final Valve Testing Evaluation by 801101.Investigation Requires More Time than Anticipated.Requests Extension Until 801219 ML19338F5461980-09-23023 September 1980 RO 80-17:on 800923,determined That There Is Insufficient Documentation to Demonstrate Environ Qualification of Brakes on Motor Operators for Purge Valves AH-V1B & AH-V1C. Cause Not Stated.Corrective Action Will Be Submitted ML19337A8631980-09-23023 September 1980 RO 80-17:determined Insufficient Documentation to Demonstrate That Brakes on Motor Operators for Purge Valves AH-V1B & AH-V1C Environmentally Qualified.Redundant Valves Fully Qualified ML19330C3041980-08-0404 August 1980 Ro:On 800716,during Radiation Emergency Exercise,Use of Emergency Notification Sys Phone Line Became Uncertain, Making Accurate Info Transfers Questionable.Recommends That Microphone Headset Be Made Available to Allow Mobility ML19326E2061980-07-22022 July 1980 RO 80-15/4P:confirming 800722 Telcon,On 800721,river Water Discharge Temp Was More than 3 F Below Inlet River Water Temp for Approx 6-h.Incident Will Be Reviewed to Determine Corrective Action.Review to Be Included in Followup Rept ML19326E0841980-06-27027 June 1980 RO 80-12/IP:during Review of Equipment Required by IE Bulletin 79-01B,found That post-accident Pressure & Temp Would Be Higher than Postulated in NSC 730705 Rept.Short & long-term Action to Be Discussed in Followup Rept ML19323H3421980-06-0606 June 1980 RO-80-10/1P:on 800603,B&W Ltr Informed Util of Potential Reportability of Steam Generator Tube Rupture Safety Analysis.Caused by Nonconservative Error in Fsar.Corrective Action Will Be Addressed in Followup Rept ML19323D5151980-04-23023 April 1980 RO 80-07/TP:on 800423 Results of Radiographic Exams of Piping Welds Associated W/Reactor Bldg Penetrations 415-16 for Leak Rate Test Sys Showed Several Welds Not within Tech Specs.Unit Is Shut Down ML19309A7831980-03-26026 March 1980 RO 80-06/1P:on 800325,during Refueling Frequency Reactor Bldg Local Leak Rate Test of Valve RB-V7,updated Total Leakage Exceeded Tech Specs.Testing Is Continuing W/Approx 1/4 of Test Program Complete ML19322B5081979-11-0808 November 1979 RO 79-17/1P:on 791108 Support Boltings for Rigid Anchor DHH-127A Found to Be Inadequate in Design & Installation. Details & Corrective Action Will Be Addressed in Followup Rept ML19308A5241978-02-0202 February 1978 RO 78-08/1P:on 780202,util Informed by NSSS of Deviation Discovered Between FSAR Steam Line Break Analysis & Actual Operating Parameters.Procedure Mod Will Limit Secondary Site Water Volume ML19308A5081978-01-20020 January 1978 RO 78-05/1P:on 780119,Plant Operations Review Committee Determined That Containment Integrity Had Been Violated by Opening of Valve BS-V47A & B Following Test of BS-P1A & B. Procedure Mod Will Terminate Operation of BS-V-47A & B ML19308A5091978-01-17017 January 1978 Confirms Telcon on 780117 Re Excessive Leak Off from 1A Makeup Pump Discharge Isolation Valve MU-V-74A Resulting in Gaseous Release 500 Times in Excess of 10CFR20 Limits for Xe-133 & Xe-135.Valve Back Seated,Isolating Leak Off ML19308A5111978-01-12012 January 1978 RO 78-2/1P:on 780112,it Was Recognized That Continuous Fire Watch Was Not Established for 63-h After a Emergency Diesel Room Sprinkler Isolation Valve FS-V-156 Was Closed.Caused by Personnel Error.Fire Watch Established on 780112 ML19261F2951978-01-0606 January 1978 RO 78-01/1P:on 780106,util Informed by B&W That During Certain Transients,Difference Between out-of-core Detectors & Heat Balance May Exceed 4% of Full Power.Frequency of Heat Balance Checks Increased ML19261F2961977-12-19019 December 1977 RO 77-29/1P:on 771216,B&W Informed Met Ed That Volume Per Tech Specs Was Insufficient to Bring Reactor to Cold Shutdown condition.TMI-1 Immediately Increased Volume of Source of Concentrated Boric Acid Solution ML19268B9681977-12-19019 December 1977 RO 77-30/1P:on 771216,B&W Informed Met Ed That Software Used to Calculate Incore Imbalance Was in Error,In That Background Correction Was Misapplied as Detector Depletes. Error Will Be Corrected Prior to Cycle 4 Operation ML19308A5131977-12-0505 December 1977 RO 77-27/1P:on 771203,technician Unable to Open Outer Door of Reactor Bldg Personnel Air Lock.Caused by Jammed Linkage. Reactor Shut Down During Repair of Door Closing Mechanism ML19308A5161977-11-15015 November 1977 RO 77-26/1P:on 771114,quadrant Power Tilt Exceeded +2.66% W/Reactor at 15% Full Power During Power Escalation Following Reactor Trip.Within 4-h,quadrant Power Tilt Not Reduced ML19261F3091977-10-0505 October 1977 RO 77-23/1T:on 770920,both Doors of Reactor Bldg Emergency Personnel Access Hatch Open Simultaneously.Caused by Personnel Error.Doors Closed.Forwards LER 77-23/1T ML19261F3121977-09-20020 September 1977 RO 77-23/1P:on 770920,containment Integrity Violated When Both Doors in Reactor Bldg Personnel Access Hatch Were Open for Approx 10 Minutes.Caused by Problem W/Door Interlock. Interlock & Door Seals Tested ML19261F3131977-09-0909 September 1977 RO 77-22/1T:on 770825,B&W-required Operating Restrictions on Decay Heat Removal Pumps Found Contrary to Description of Pump Operating Capability Per Fsar.Caused by B&W Failure to Perform Safety Evaluation.Corrective Action by 770923.W/LER ML19308A5201977-08-26026 August 1977 RO 77-22/1P:confirms 770825 Telcon That as Result of B&W New Decay Heat Pump Operating Limitations,Remedial or Corrective Action Could Be Required to Prevent Unsafe Condition. Procedure Change Initiated to Incorporate New Limitations ML19261F3161977-08-12012 August 1977 RO 77-19/3L:on 770713,leak in Miscellaneous Waste Evaporator Feed Tank Immersion Heater Shorted Out Heater & Feed Pump. Caused by Matl Failure in Heater Temp Element.Heater Replaced & Operating Procedure Modified.Forwards LER ML19261F3191977-08-0909 August 1977 RO 77-21/1T:on 770726,radiation Monitor RM-A2 Out of Svc & Sample of Reactor Bldg Atmosphere Not Taken within 8-h. Caused by Improper Reassembly Following 770725 Surveillance Test.Monitor Repaired & Sample Taken.Forwards LER ML19210A5441977-08-0505 August 1977 Followup Rept to 770707 Ltr Re Ro:On 770613,radwaste Shipment Leak Discovered.Caused by Acidic Attack of Mild Steel Container at Locations of Incomplete Coverage W/Internal Corrosion Resistant Coating ML19261F3221977-07-27027 July 1977 RO 77-18/31:on 770629,Nuclear Svcs River Water Pump NR-P1C Shaft Failed While Operating,Leading to Operation in Degraded Mode.Caused by Mechanical & Procedural Problems. Pump Overhauled,Preventive Maint Procedures to Be Changed ML19308A5211977-07-27027 July 1977 RO 77-21/1P:on 770726,radiation Monitor RM-A2 Out of Svc & Sample of Reactor Bldg Atmosphere Not Taken within 8-h. Caused by Improper Reassembly Following 770725 Surveillance Test.Monitor Repaired & Valid Sample Taken ML19308A5231977-07-20020 July 1977 RO 77-20/4P:on 770719,ETS Violated in That Plant River Water Discharge Temp Exceeded Limit of 3 F Below Inlet River Water Temp for 10 Minutes.Caused by Passing Storm Front Which Decreased Ambient Air Temp from 94 to 82 F ML19308A5221977-07-20020 July 1977 RO 77-20/4P:on 770619,ETS Violated in That Plant River Water Discharge Temp Exceeded Limit of 3 F Below Inlet River Water Temp.Caused by Passing Storm Front Which Lowered Ambient Air Temp from 94 to 82 F ML19261F3251977-07-20020 July 1977 RO 77-17/3L:on 770624,inner Door Seal on Personnel Access Hatch Leaked Excesively,Causing Total Reactor Bldg Local Leakage to Exceed Tech Spec 4.4.1.2.3.Caused by Improper Door Installation Procedures.Forwards LER ML19261F3291977-07-0707 July 1977 RO on 770713:Tri-State Motor Transit Co Driver Observed Radwaste Leaking from Container on Truck While at Rest Area. Caused by Pinhole Leak in One Liner.Truck Returned to Station Site ML19261F3301977-06-28028 June 1977 RO 77-15/3L:on 770529,one of Two Units of 230/4.16 Kv Auxiliary Transformer de-energized,leading to Operation in Degraded Mode Permitted by Tech Spec 3.7.2.b.Caused by Wiring Error on Current Transformers.Sys Repaired.Ler Encl ML19261F3331977-06-20020 June 1977 RO 77-14/3L:on 770524,reactor Bldg Isolation Valve RB-V7 Failed to Close on Engineering Safeguards Test Signal, Leading to Operation in Degraded Mode Permitted by Limiting Condition for Operation Per Tech Spec 3.6.6 ML19261F3361977-06-0909 June 1977 RO 77-12/3L:on 770519,engineered Safeguard Pump MU-P-1A Was Removed from Svc Prior to Testing Operability of Redundant Pump.Caused by Personnel Error.Redundant Pump Tested Satisfactorily on 770519.Forwards LER 77-12/03L ML19261F3391977-06-0707 June 1977 RO 77-13/4T:on 770523,analysis of Plant River Water Discharge Showed Concentration of 760 Parts Per Million. Caused Either by Slug of High Concentration of Suspended Solids from Filter Sump or by Poor Sample.Forwards LER ML19322A4381977-05-10010 May 1977 RO 77-07/1P:on 770510,both Reactor Containment Bldg Emergency Personnel Access Doors Open at Same Time.Caused by Closed Outer Door Mechanism While Door Still Was Open ML19322A4391977-04-18018 April 1977 RO 77-04/1P:on 770415,nonconservative Error Found in Safety Analysis to Support Increase in Pressurizer Code Safety Valve Setpoint to 2500 Psig.Caused by Erroneous Assumptions Made for Total Combined Relief Valve Capacity ML19253B9631977-03-22022 March 1977 RO 77-02/04T:on 770304,river Water Discharge Temp Exceeded Inlet River Water Temp Limitation by 14 F Per Ets.Caused by Insufficient Procedures,Personnel Error,Abnormal Ambient Conditions & Low River Water Temp.Ler Encl ML19322A4401977-03-20020 March 1977 RO 77-05/4P:on 770319,river Water Discharge Temp Differential Change More than 2 Degrees in 1-h.Steps Taken to Reduce Decay Heat Removal Load,Reducing River Water Temp Change ML19253B9621977-03-0808 March 1977 RO 77-02/04P:on 770304,river Water Discharge Temp Exceeded Inlet River Water Temp by 14 F,Violating 12 F Limit Per Ets. Caused by Low Inlet Temp & High Ambient Temp.Temp Reduced Below Limit ML19210A2681977-02-0303 February 1977 RO 77-1/1P:on 770202,both a & B Diesel Generators Had Conditions Affecting Starting Capabilities in Response to Loss of Offsite Power.Both Problems Caused by Cranking Timer Timing Out Prior to Oil Low Speed Switch Operation ML19261F1771976-12-29029 December 1976 RO 76-50/4O:on 761221,radionucleide Surface Water Level Ten Times Background Level.Probable Cause Is Chinese Weapons Test,Since Neither Zr-95 Nor Nb-95 Discharged to River from Station During Nov 1976 ML19261F1881976-12-26026 December 1976 RO 76-047/01T:on 761216,sodium Hydroxide Tank Levels Fell Below Required Minimum.Caused by Personnel Misreading Tank Level Data & Technical Analyst Failing to Notice Mistake ML19261F2531976-12-20020 December 1976 RO 76-48/4o:on 761208,isotope Level Samples from TMI Environs Ten Times Background Levels.Chinese Weapons Tests Probably Caused Excesses of I-131,Zr-95,Nb-95,Ce-141 & Ru-103 in Sediment Samples from TMI ML19261F2561976-12-14014 December 1976 Nonroutine 30 Day Rept 75-03,advising That Heated Post Gap Measurements within Acceptable Values.Test Schedule Change Warranted from Weekly to After Each Startup,Once Full Steam Equilibrium Reached & on Monthly Basis Thereafter ML19308A4811976-12-13013 December 1976 RO 76-46/1P:on 761212,boron Concentrations in Borated Water Storage Tank Found Below Required Minimum.Boric Acid Injected to Return Concentration to Necessary Valve.Possibly Caused by Water Leakage from Reactor Coolant Sys 1983-08-04 Category:TEXT-SAFETY REPORT MONTHYEARML20217K4701999-09-30030 September 1999
[Table view]Monthly Operating Rept for Sept 1999 for TMI-1.With ML20211H5111999-08-31031 August 1999 Non-proprietary Rev 1 to MPR-1820(NP), TMI Nuclear Generating Station OTSG Kinetic Expansion Insp Criteria Analysis ML20211Q3551999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Tmi,Unit 1.With ML20210R4791999-08-13013 August 1999 Update 3 to Post-Defueling Monitored Storage SAR, for TMI-2 ML20210U4791999-07-31031 July 1999 Monthly Operating Rept for July 1999 for TMI-1.With ML20209G0011999-07-0909 July 1999 Staff Evaluation of Individual Plant Exam of External Events Submittal on Plant,Unit 1 ML20210K7651999-07-0909 July 1999 Rev 2 to 86-5002073-02, Summary Rept for Bwog 20% Tp Loca ML20209H8251999-07-0101 July 1999 Provides Commission with Evaluation of & Recommendations for Improvement in Processes Used in Staff Review & Approval of Applications for Transfer of Operating Licenses of TMI-1 & Pilgrim Station ML20209H1421999-06-30030 June 1999 Monthly Operating Rept for June 1999 for TMI-1.With ML20195H0751999-06-0808 June 1999 Drill 9904, 1999 Biennial Exercise for Three Mile Island ML20195H9261999-05-31031 May 1999 Monthly Operating Rept for May 1999 for TMI-1.With ML20209G0351999-05-31031 May 1999 TER on Review of TMI-1 IPEEE Submittal on High Winds,Floods & Other External Events (Hfo) ML20207B6621999-05-27027 May 1999 SER Finding That Licensee Established Acceptable Program to Periodically Verify design-basis Capability of safety-related MOVs at TMI-1 & That Util Adequately Addressed Actions Required in GL 96-05 ML20206R0571999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Tmi,Unit 1.With ML20206D4201999-04-20020 April 1999 Safety Evaluation Granting Exemption from Technical Requirements of 10CFR50,App R,Section III.G.2.c for Fire Areas/Zones AB-FZ-4,CB-FA-1,FH-FZ-1,FH-FZ-6,FH-FZ-6, IPSH-FZ-1,IPSH-FZ-2,AB-FZ-3,AB-FZ-5,AB-FZ-7 & FH-FZ-2 ML20209G0071999-03-31031 March 1999 Submittal-Only Screening Review of Three Mile Island,Unit 1 Individual Plant Exam for External Events (Seismic Portion) ML20205K6851999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Tmi,Unit 1.With ML20210C0161999-03-0101 March 1999 Forwards Corrected Pp 3 of SECY-98-252.Correction Makes Changes to Footnote 3 as Directed by SRM on SECY-98-246 ML20207M8461999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for TMI-1.With ML20196K3561999-01-22022 January 1999 Safety Evaluation Concluding That Although Original Licensee Thermal Model Was Unacceptable for Ampacity Derating Assessments Revised Model Identified in 970624 Submittal Acceptable for Installed Electrical Raceway Ampacity Limits ML20207A9291998-12-31031 December 1998 1998 Annual Rept for TMI-1 & TMI-2 ML20196G4661998-12-31031 December 1998 British Energy Annual Rept & Accounts 1997/98. Prospectus of British Energy Share Offer Encl ML20196F6861998-12-0202 December 1998 Safety Evaluation Accepting Licensee Second 10-yr Interval ISI Program Plan Request for Alternative to ASME B&PV Code Section XI Requirements Re Actions to Be Taken Upon Detecting Leakage at Bolted Connection ML20198B8641998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for TMI-1.With ML20195C6921998-11-12012 November 1998 Safety Evaluation Supporting Amend 52 to License DPR-73 ML20195J8591998-11-12012 November 1998 Rev 11 to 1000-PLN-7200.01, Gpu Nuclear Operational QA Plan ML20196B7191998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for TMI-1.With ML20203G1211998-10-30030 October 1998 Informs Commission About Staff Preliminary Views Concerning Whether Proposed Purchase of TMI-1,by Amergen,Inc,Would Cause Commission to Know or Have Reason to Believe That License for TMI-1 Would Be Controlled by Foreign Govt ML20155E7511998-10-15015 October 1998 Rev 1 to Form NIS-1 Owners Data Rept for Isi,Rept on 1997 Outage 12R EC Exams of TMI-1 OTSG Tubing ML20154L5541998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for TMI Unit 1.With ML20153A9941998-09-16016 September 1998 Safety Evaluation Denying Request to Remove Missile Shields from Plant Design ML20151U8821998-09-0808 September 1998 SER on Revised Emergency Action Levels for Gpu Nuclear,Inc, Three Mile Island Nuclear Plant Units 1 & 2 ML20151V2811998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Tmi,Unit 1.With ML20237A8331998-08-12012 August 1998 Safety Evaluation Accepting USI A-46 Program Implementation at Plant,Unit 1 ML20237C6411998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Tmi,Unit 1 ML20236R2201998-06-30030 June 1998 Monthly Operating Rept for June 1998 for TMI-1 ML20236W9961998-06-0909 June 1998 1998 Quadrennial Simulator Certification Rept ML20248F7441998-05-31031 May 1998 Reactor Vessel Working Group,Response to RAI Regarding Reactor Pressure Vessel Integrity ML20249A1061998-05-31031 May 1998 Monthly Operating Rept for May 1998 for TMI-1 ML20247G0761998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Three Mile Island Nuclear Station,Unit 1 ML20212A2191998-04-22022 April 1998 Rev 3 to Gpu Nuclear Post-Defueling Monitored Storage QAP for Three Mile Island Unit 2 ML20248H6991998-04-0808 April 1998 Requests,By Negative Consent,Commission Approval of Intent to Inform Doe,Idaho Operations Ofc of Finding That Adequate Safety Basis Support Granting Exemption to 10CFR72 Seismic Design Requirement for ISFSI to Store TMI-2 Fuel Debris ML20216K1061998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Three Mile Island Nuclear Station,Unit 1 ML20217E0811998-03-24024 March 1998 Rev 0 to TR-121, TMI-1 Control Room Habitability for Max Hypothetical Accident ML20212E2291998-03-0404 March 1998 Rev 11 to 1000-PLN-7200,01, Gpu Nuclear Operational QAP, Consisting of Revised Pages & Pages for Which Pagination Affected ML20216F0981998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Three Mile Island Nuclear Station,Unit 1 ML20202F8121998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for TMI Nuclear Station, Unit 1 ML20199G8371998-01-22022 January 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Three Mile Island Nuclear Station,Unit 1 ML20198N2901998-01-12012 January 1998 Form NIS-1 Owners' Data Rept for Isi ML20199J3251997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Three Mile Island Nuclear Station,Unit 1 1999-09-30 |
Text
'
C,65uYMAG ftLEJ Yz, / MIL
/
METROPOLITAN EDISON COMPANY sussrounvorcentur.rveucunima conronnos CST OFFICE BOX 480 MfDOLETOWN. PENNSYLVANIA 17057 TELEPHONE 717.N4 4041
~~
January 6. 1978 -
Mr. Boyce H. Grier _
Directorate of Regulatory Operations Region I Nuclear Regulatory Commission 631 Park Avenue King of Prussia. Pennsylvania 19406 Operating License DPR-50 Docket 50-289
Subject:
Reportable Occurrence No. 78-01/IP
Dear Mr. Grier:
This telegram is to confirm the jelecon between Mr. P. J. Kellogg, Region I NRC, and Mr. J. P..O'Hanlon. Unit I Superintendent, at 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br /> on 6 January 1978.
This item is considered potentially reportable in accordance with Technical Spec i fica tion -4.9.2. A.9.
Metropolitan Edisoh has been nformed by the THI-l NSSS vendor, Babcock and Wilcox, that during certain plant transients. the difference between the out-of-core detectors and the heat balance may exceed four percent full power. The maximum difference assumed in the Safejy Analysis is four percent. -
THI-l has increased the frequency of heat balance checks to insure that thethan is less diffarence between two percent fullthe out-of-core detectors and the heat balance power. Technical Specification Table 4.1-1 requires recalibration of the out-of-core detectors if indicated power differs from heat balance power by more than two percent full power.
Further corrective action will be discussed in the followup report.
Very truly yours,
\AO "1g6 pg
. P. O'Hanlon \,
Unit,I, Superintendent -
K JP0/ job 0
?9102s0 7 V P J
Retrieved from "https://kanterella.com/w/index.php?title=ML19261F295&oldid=2382001"