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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML20029E2071994-05-0909 May 1994 LER 94-004-00:on 940416,discovered That Surveillance Frequency for Sp 34SV-SUV-008-1S Not Correct for Seven Primary Containment Vent & Purge Valves Due to Personnel Error.Surveillance Brought Up to date.W/940509 Ltr ML20029C8681994-04-25025 April 1994 LER 94-003-00:on 940329,automatic Reactor Shutdown Occurred Due to Trip of Main Turbine.Caused by Loss of Main Generator Field Excitation.Damaged Main Generator Exciter Rotor replaced.W/940425 Ltr ML20029C6981994-04-19019 April 1994 LER 94-002-00:on 940325,personnel Error Resulted in ESF Actuations.Personnel Counseled & Trained & EDG Control Circuit Wire repaired.W/940419 Ltr ML20046D5941993-08-18018 August 1993 LER 93-006-00:on 930721,determined That Valves Could Not Be Closed by Use of Normal Motive Power Due to Inadequate Procedural Controls Resulting in Valve Actuators Being Set Up Improperly.Isolated Affected penetration.W/930818 Ltr ML20045H7441993-07-0909 July 1993 LER 93-012-00:on 930615,automatic Reactor Scram & Isolation of Inboard Group 2 PCIS Valves Occurred.Caused by Loose Packing Nut on Instrument Isolation Valve.Valve Repaired & Similar Installations on Units checked.W/930709 Ltr ML20045B0371993-06-10010 June 1993 LER 93-009-00:on 930514,scram Occurred When Mode Switch Moved to Run Position Due to Blown Fuses in Rps.Procedures 52PM-B21-005-1S,52PM-B21-005-2S & 52GM-MEL-007-0S Revised. W/930610 Ltr ML20045B0401993-06-10010 June 1993 LER 93-010-00:on 930514,unplanned ESF Actuation Occurred Due to Less That Adequate Procedures.Procedures 34SV-B21-001-1S & 34SV-B21-001-2S, MSIV Closure Instrument Functional Test revised.W/930610 Ltr ML20045B0761993-06-10010 June 1993 LER 93-011-00:on 930521,partial Group 1 Primary Containment Isolation Sys Actuation Occurred Due to Component Failure. MSLRM Returned to Svc,Failed electro-pneumatic Control Valve in MSIVs replaced.W/930610 Ltr ML20045B7111993-06-10010 June 1993 LER 93-005-00:on 930521,unplanned Insertion of Manual Scram Initiated Due to Personnel Error.Personnel Involved Temporarily Removed from License Duties & Being Subjected to Formal discipline.W/930610 Ltr ML20045A2121993-06-0303 June 1993 LER 93-008-00:on 930505,determined That B Train of SBGT Sys Had Been Inoperable.Caused by Procedure Error.C/As Included Bringing Missed Surveillances Up to Date,Revising Procedures & Counselling personnel.W/930603 Ltr ML20044F6021993-05-21021 May 1993 LER 93-007-01:on 930504,unplanned ESF Actuations Occurred. Caused by Inappropriate Jumper Placement by Plant Engineer. Personnel Performing LSFTs Made Aware of Event & Instructed Not to Install Jumpers on Relay Contact arms.W/930521 Ltr ML20044F5851993-05-18018 May 1993 LER 93-004-00:on 930419,isolation Valve Unexpectedly Closed While Trip Unit Was Tested.Cause for Valve Closing Undetermined.Logic of Trip Unit Correctly Configured to Prevent Closing.No Corrective Actions taken.W/930518 Ltr ML20044D5931993-05-15015 May 1993 LER 93-005-00:on 930414,fuse 1D11-A-f14B Blew,Resulting in Initiation of Train B of Both Standby Treatment Sys Units & Isolation of Damper B of Both Secondary Containments.Blown Fuse & Several Relays in Logic Replaced ML20044D4881993-05-14014 May 1993 LER 93-004-00:on 930414,unplanned ESF Sys Actuation Occurred When LPCI Valve Automatically Reclosed.Caused by Inadvertent Grounding of Logic Circuit,Resulting in Blown Fuse.Fuse Replaced & Operating Order Issued ML20044D1111993-05-10010 May 1993 LER 93-003-00:on 930412,determined That Monthly Operability Test for DG 1B Not Performed During Required Performance Window on 930328.Caused by Personnel Error.Surveillance Coordinator Aware of causes.W/930510 Ltr ML20044C9891993-05-0303 May 1993 LER 93-003-00:on 930407,scram Time Testing on All Control Rods Not Completed Prior to Exceeding 40% Rated Thermal Power.Caused by Personnel Error.Personnel Counseled & Procedures Will Be revised.W/930503 Ltr ML20024G6961991-04-25025 April 1991 LER 91-007-00:on 910326,unknown Inadequacy in Jumper Connection Results in Scram During Surveillance in Cold Shutdown.Cause Unknown.Functional Test Completed & Surveillance Procedure revised.W/910422 Ltr ML20024G7401991-04-24024 April 1991 LER 91-008-00:on 910327,main Steam Isolation Valve Local Leak Rate Test Failed Due to Normal Equipment Wear Resulting in Degradation of Valve Seating Surfaces.Valves Repaired & retested.W/910424 Ltr ML20029B0691991-03-0505 March 1991 LER 91-002-00:on 910203,partial Outboard Group 2 Primary Containment Isolation Sys Isolation Signal Resulted in Closure of Containment Isolation Valves.Caused by Failed Relay Coil.Coil replaced.W/910301 Ltr ML20029B0621991-02-26026 February 1991 LER 91-001-00:on 910129,determined That Setpoints for Condensate Storage Tank Level Switches Not Set to Initiate Required Transfer When 10,000 Gallons Water Available.Caused by Inadequate Documentation.Setpoints raised.W/910226 Ltr ML20028H8431991-01-27027 January 1991 LER 90-024-00:on 901228,pretreatment Monitoring Station Offgas Samples Not Collected & Analyzed within 4 H of Increased Fission Gas Release.Caused by Misinterpretation of Tech Specs.Personnel instructed.W/910125 Ltr ML20044A6411990-06-22022 June 1990 LER 90-011-00:on 900601,full Reactor Protection Sys Actuation Occurred When Mode Switch Moved to Run Position, Resulting in Scram Signal on MSIVs Less than 90% Open.Caused by Personnel Error.Individual counseled.W/900622 Ltr ML20043G7111990-06-15015 June 1990 LER 90-004-00:on 900521,personnel Error & FSAR Deviation Occurred & Resulted in Tech Spec Violation.Procedure 62CI-OCB-031-OS Incorrectly Directed Personnel to Periodically Open Airlock Doors.Memo issued.W/900615 Ltr ML20043G7141990-06-15015 June 1990 LER 90-009-00:on 900522,determined That Requirements of Tech Spec 3.14.2,Actions 105 & 107 Not Met.Caused by Inadequate Procedure.Normal Range Monitoring Sys Restored to Operable Status & Procedure 64CH-SAM-005-OS revised.W/900615 Ltr ML20043G7581990-06-0808 June 1990 LER 90-001-01:on 900112,component Failure & Inadequate Design Caused Group I Isolation & Scram W/Main Steamline Isolation Valves Less than 90% Open.Root Isolation Valves Replaced W/New Type of valve.W/900608 Ltr ML20043C7281990-05-31031 May 1990 LER 90-008-00:on 900505,determined That Reactor Vessel Head Vent Valves 1B21-F004 & 1B21-F005 Closed Contrary to Tech Spec 3.7.C.2.a(2) Requirements.Caused by Cognitive Personnel Error.Reactor Vessel Head Vent Valves reopened.W/900531 Ltr ML20043A5091990-05-14014 May 1990 LER 90-006-00:on 900418,discovered That Wiring Error Existed in Junction Box Leading to Strip Recorder That Resulted in Inadequate Tech Spec Surveillance.Caused by Personnel Error. Wiring Error Corrected & Personnel counseled.W/900514 Ltr ML20043A5131990-05-14014 May 1990 LER 90-007-00:on 900419,determined That Errors in Calculations for Measuring Feedwater Flow Resulted in Nonconservative Calibr of Flow Transmitters.Caused by Error in Design Calculation.Transmitters recalibr.W/900514 Ltr ML20042E6851990-04-27027 April 1990 LER 90-003-00:on 900328,reactor Scram & Group II Containment Isolation Occurred.Caused by Inadequate Procedure.Procedure Will Be Changed to Require Instruments to Be Pressurized to Process Pressure Before Valved Into svc.W/900423 Ltr ML20042E6841990-04-27027 April 1990 LER 90-005-00:on 900329,safety Relief Valves Experienced Setpoint Drift in Excess of Tolerance.Caused by corrosion- Induced Bonding of Surface Between Pilot Valve Disc & Seat. Valves refurbished.W/900424 Ltr ML20012D8861990-03-19019 March 1990 LER 99-004-00:on 900219,trip Setpoint for Isolation of Liquid Radwaste Effluent Line on Low Dilution Flow Not Set Correctly.Caused by Inadequate Procedure.Procedure Revised temporarily.W/900319 Ltr ML20012C2891990-03-12012 March 1990 LER 90-003-00:on 900212,determined That Surveillance Procedures for Monthly Functional Testing of Drywell High Pressure Instrumentation Logic Channels Less than Adequate. Caused by Personnel Error.Procedures revised.W/900312 Ltr ML20011F4291990-02-26026 February 1990 LER 90-002-00:on 900131,discovered That Functional Test of Turbine Stop Valve Position Limit Switches Not Performed. Caused by Personnel Error When Writing Recent Rev.Rev to Procedure 34SV-C71-001-1S/2S written.W/900226 Ltr ML20006E2891990-02-0707 February 1990 LER 90-001-00:on 900112,reactor Scrammed Because MSIVs Were Less than 90% Open.Caused by Component Failure & Configuration of Condenser Vacuum Sensing Lines & Instruments.Valves replaced.W/900207 Ltr ML20006E0111990-02-0606 February 1990 LER 90-002-00:on 900114,RWCU Experienced High Differential Flow,Indicating Possibility of Leak in Sys.Caused by Component Failure & Less than Adequate Mounting for Relay. Relay Replaced W/Time Delay relay.W/900206 Ltr ML20006A8881990-01-22022 January 1990 LER 90-001-00:on 900104,HPCI Pump Declared Inoperable Due to Rated Flow Not Maintained During Surveillance Testing. Caused by Component Failure.Defective Resistor Replaced & Procedure 34SV-E41-002-1S performed.W/900122 Ltr ML20005E6541990-01-0202 January 1990 LER 89-010-00:on 891204,determined That Plant Was Not Fully Meeting Surveillance Requirements of Tech Spec Table 4.3.6.4-1,item 10.b.Caused by Inadequate Procedure.Recorder Calibr Steps to Be Removed from procedure.W/900102 Ltr ML20005E1851989-12-27027 December 1989 LER 89-009-00:on 891129,reactor Protection Sys Actuation Occurred from Scram Discharge Vol High Level Condition. Caused by Equipment Failure.Backup Temporary Air Compressor Placed Into Svc & Blown Fuse replaced.W/891227 Ltr ML20005E5131989-12-22022 December 1989 LER 89-017-00:on 891128,discovered That Efficiency Factors Used for Old Liquid Radwaste Discharge Radiation Monitors Incorrect.Caused by Personnel Not Incorporating Updated Efficiency Factor Into Sys software.W/891222 Ltr ML19332F8691989-12-14014 December 1989 LER 89-018-00:on 891114,sys High Differential Flow Condition Occurred Causing Actuation of Primary Containment Isolation Sys Valve Group 5 Logic Resulting in Closure of RWCU Valve. Caused by Personnel Error.Personnel counseled.W/891214 Ltr ML19332E6141989-11-30030 November 1989 LER 89-016-00:on 891103,discovered That Procedures 57SV-C51-001-1/2S Did Not Fully Test Rod Block Monitor Function.Caused by Procedural Deficiency.Limiting Condition for Operation Initiated & Procedure revised.W/891130 Ltr ML19332D8791989-11-29029 November 1989 LER 89-008-00:on 891102,RWCU Sys Experienced Partial Primary Containment Isolation Sys Group 5 Isolation Involving Valve 2G31-F004.Caused by Component Failure of Relay 2G31-R616D. Relay replaced.W/891129 Ltr ML19324C3271989-11-0808 November 1989 LER 89-014-00:on 891010,primary Containment Isolation Sys Group 5 Isolation Occurred Due to Opening of Valve 1G31-D002A.Caused by Personnel Error.Personnel Counseled & Memo Issued Re Confirming commands.W/891108 Ltr ML19325F1781989-11-0606 November 1989 LER 89-015-00:on 891009,diesel Generator 1R43-S001B Failed to Start Manually During Monthly Generator Test.Caused by Personnel Error & Incorrect Model Number Assigned to Pump. Pump Replaced & Oil Drained from cylinders.W/891106 Ltr ML19332B6191989-10-31031 October 1989 LER 89-013-00:on 891003,RWCU Sys Valve 1G31-F020 Closed, Rendering Inservice Reactor Coolant Monitor Inoperable. Caused by Cognitive Personnel error.In-line Conductivity Surveillance Initiated & Personnel counseled.W/891031 Ltr ML19325E6911989-10-31031 October 1989 LER 89-012-00:on 891003,plant Operators Received Indication That RWCU Sys Experiencing High Differential Flow Which Resulted in Isolation of Primary Containment Isolation Sys Valves.Caused by Component failure.W/891031 Ltr ML19327B3281989-10-23023 October 1989 LER 89-007-00:on 890926,four Safety Relief Valves Had Exhibited Drift in Mechanical Lift Setpoints in Excess of 3% Tolerance Specified by Inservice Testing Requirements.Root Cause Being Investigated.Valves refurbished.W/891023 Ltr ML19327B2991989-10-23023 October 1989 LER 89-006-00:on 890926,Procedure 34SV-SUV-019-2S, Surveillance Checks Did Not Fully Implement Requirements of Tech Spec Table 4.3.2-1.Caused by Personnel Error. Personnel Counseled & Procedure revised.W/891023 Ltr ML20024F4081983-09-0101 September 1983 LER 83-079/03L-0:on 830809,main Steam Line & Reactor Water Sample Valve Relay 1A71-K7A Determined Operating in Degraded Mode.Caused by Component Failure.Coil & Contacts Replaced. W/830901 Ltr ML20024F3341983-09-0101 September 1983 LER 83-064/03L-0:on 830811,during post-maint Review of DCR 83-76 Determined Torus Vent Valves Instrument Air Piping Returned to Svc W/O Performance of HNP-6907.Caused by Personnel oversight.W/830901 Ltr 1994-05-09
[Table view] Category:RO)
MONTHYEARML20029E2071994-05-0909 May 1994 LER 94-004-00:on 940416,discovered That Surveillance Frequency for Sp 34SV-SUV-008-1S Not Correct for Seven Primary Containment Vent & Purge Valves Due to Personnel Error.Surveillance Brought Up to date.W/940509 Ltr ML20029C8681994-04-25025 April 1994 LER 94-003-00:on 940329,automatic Reactor Shutdown Occurred Due to Trip of Main Turbine.Caused by Loss of Main Generator Field Excitation.Damaged Main Generator Exciter Rotor replaced.W/940425 Ltr ML20029C6981994-04-19019 April 1994 LER 94-002-00:on 940325,personnel Error Resulted in ESF Actuations.Personnel Counseled & Trained & EDG Control Circuit Wire repaired.W/940419 Ltr ML20046D5941993-08-18018 August 1993 LER 93-006-00:on 930721,determined That Valves Could Not Be Closed by Use of Normal Motive Power Due to Inadequate Procedural Controls Resulting in Valve Actuators Being Set Up Improperly.Isolated Affected penetration.W/930818 Ltr ML20045H7441993-07-0909 July 1993 LER 93-012-00:on 930615,automatic Reactor Scram & Isolation of Inboard Group 2 PCIS Valves Occurred.Caused by Loose Packing Nut on Instrument Isolation Valve.Valve Repaired & Similar Installations on Units checked.W/930709 Ltr ML20045B0371993-06-10010 June 1993 LER 93-009-00:on 930514,scram Occurred When Mode Switch Moved to Run Position Due to Blown Fuses in Rps.Procedures 52PM-B21-005-1S,52PM-B21-005-2S & 52GM-MEL-007-0S Revised. W/930610 Ltr ML20045B0401993-06-10010 June 1993 LER 93-010-00:on 930514,unplanned ESF Actuation Occurred Due to Less That Adequate Procedures.Procedures 34SV-B21-001-1S & 34SV-B21-001-2S, MSIV Closure Instrument Functional Test revised.W/930610 Ltr ML20045B0761993-06-10010 June 1993 LER 93-011-00:on 930521,partial Group 1 Primary Containment Isolation Sys Actuation Occurred Due to Component Failure. MSLRM Returned to Svc,Failed electro-pneumatic Control Valve in MSIVs replaced.W/930610 Ltr ML20045B7111993-06-10010 June 1993 LER 93-005-00:on 930521,unplanned Insertion of Manual Scram Initiated Due to Personnel Error.Personnel Involved Temporarily Removed from License Duties & Being Subjected to Formal discipline.W/930610 Ltr ML20045A2121993-06-0303 June 1993 LER 93-008-00:on 930505,determined That B Train of SBGT Sys Had Been Inoperable.Caused by Procedure Error.C/As Included Bringing Missed Surveillances Up to Date,Revising Procedures & Counselling personnel.W/930603 Ltr ML20044F6021993-05-21021 May 1993 LER 93-007-01:on 930504,unplanned ESF Actuations Occurred. Caused by Inappropriate Jumper Placement by Plant Engineer. Personnel Performing LSFTs Made Aware of Event & Instructed Not to Install Jumpers on Relay Contact arms.W/930521 Ltr ML20044F5851993-05-18018 May 1993 LER 93-004-00:on 930419,isolation Valve Unexpectedly Closed While Trip Unit Was Tested.Cause for Valve Closing Undetermined.Logic of Trip Unit Correctly Configured to Prevent Closing.No Corrective Actions taken.W/930518 Ltr ML20044D5931993-05-15015 May 1993 LER 93-005-00:on 930414,fuse 1D11-A-f14B Blew,Resulting in Initiation of Train B of Both Standby Treatment Sys Units & Isolation of Damper B of Both Secondary Containments.Blown Fuse & Several Relays in Logic Replaced ML20044D4881993-05-14014 May 1993 LER 93-004-00:on 930414,unplanned ESF Sys Actuation Occurred When LPCI Valve Automatically Reclosed.Caused by Inadvertent Grounding of Logic Circuit,Resulting in Blown Fuse.Fuse Replaced & Operating Order Issued ML20044D1111993-05-10010 May 1993 LER 93-003-00:on 930412,determined That Monthly Operability Test for DG 1B Not Performed During Required Performance Window on 930328.Caused by Personnel Error.Surveillance Coordinator Aware of causes.W/930510 Ltr ML20044C9891993-05-0303 May 1993 LER 93-003-00:on 930407,scram Time Testing on All Control Rods Not Completed Prior to Exceeding 40% Rated Thermal Power.Caused by Personnel Error.Personnel Counseled & Procedures Will Be revised.W/930503 Ltr ML20024G6961991-04-25025 April 1991 LER 91-007-00:on 910326,unknown Inadequacy in Jumper Connection Results in Scram During Surveillance in Cold Shutdown.Cause Unknown.Functional Test Completed & Surveillance Procedure revised.W/910422 Ltr ML20024G7401991-04-24024 April 1991 LER 91-008-00:on 910327,main Steam Isolation Valve Local Leak Rate Test Failed Due to Normal Equipment Wear Resulting in Degradation of Valve Seating Surfaces.Valves Repaired & retested.W/910424 Ltr ML20029B0691991-03-0505 March 1991 LER 91-002-00:on 910203,partial Outboard Group 2 Primary Containment Isolation Sys Isolation Signal Resulted in Closure of Containment Isolation Valves.Caused by Failed Relay Coil.Coil replaced.W/910301 Ltr ML20029B0621991-02-26026 February 1991 LER 91-001-00:on 910129,determined That Setpoints for Condensate Storage Tank Level Switches Not Set to Initiate Required Transfer When 10,000 Gallons Water Available.Caused by Inadequate Documentation.Setpoints raised.W/910226 Ltr ML20028H8431991-01-27027 January 1991 LER 90-024-00:on 901228,pretreatment Monitoring Station Offgas Samples Not Collected & Analyzed within 4 H of Increased Fission Gas Release.Caused by Misinterpretation of Tech Specs.Personnel instructed.W/910125 Ltr ML20044A6411990-06-22022 June 1990 LER 90-011-00:on 900601,full Reactor Protection Sys Actuation Occurred When Mode Switch Moved to Run Position, Resulting in Scram Signal on MSIVs Less than 90% Open.Caused by Personnel Error.Individual counseled.W/900622 Ltr ML20043G7111990-06-15015 June 1990 LER 90-004-00:on 900521,personnel Error & FSAR Deviation Occurred & Resulted in Tech Spec Violation.Procedure 62CI-OCB-031-OS Incorrectly Directed Personnel to Periodically Open Airlock Doors.Memo issued.W/900615 Ltr ML20043G7141990-06-15015 June 1990 LER 90-009-00:on 900522,determined That Requirements of Tech Spec 3.14.2,Actions 105 & 107 Not Met.Caused by Inadequate Procedure.Normal Range Monitoring Sys Restored to Operable Status & Procedure 64CH-SAM-005-OS revised.W/900615 Ltr ML20043G7581990-06-0808 June 1990 LER 90-001-01:on 900112,component Failure & Inadequate Design Caused Group I Isolation & Scram W/Main Steamline Isolation Valves Less than 90% Open.Root Isolation Valves Replaced W/New Type of valve.W/900608 Ltr ML20043C7281990-05-31031 May 1990 LER 90-008-00:on 900505,determined That Reactor Vessel Head Vent Valves 1B21-F004 & 1B21-F005 Closed Contrary to Tech Spec 3.7.C.2.a(2) Requirements.Caused by Cognitive Personnel Error.Reactor Vessel Head Vent Valves reopened.W/900531 Ltr ML20043A5091990-05-14014 May 1990 LER 90-006-00:on 900418,discovered That Wiring Error Existed in Junction Box Leading to Strip Recorder That Resulted in Inadequate Tech Spec Surveillance.Caused by Personnel Error. Wiring Error Corrected & Personnel counseled.W/900514 Ltr ML20043A5131990-05-14014 May 1990 LER 90-007-00:on 900419,determined That Errors in Calculations for Measuring Feedwater Flow Resulted in Nonconservative Calibr of Flow Transmitters.Caused by Error in Design Calculation.Transmitters recalibr.W/900514 Ltr ML20042E6851990-04-27027 April 1990 LER 90-003-00:on 900328,reactor Scram & Group II Containment Isolation Occurred.Caused by Inadequate Procedure.Procedure Will Be Changed to Require Instruments to Be Pressurized to Process Pressure Before Valved Into svc.W/900423 Ltr ML20042E6841990-04-27027 April 1990 LER 90-005-00:on 900329,safety Relief Valves Experienced Setpoint Drift in Excess of Tolerance.Caused by corrosion- Induced Bonding of Surface Between Pilot Valve Disc & Seat. Valves refurbished.W/900424 Ltr ML20012D8861990-03-19019 March 1990 LER 99-004-00:on 900219,trip Setpoint for Isolation of Liquid Radwaste Effluent Line on Low Dilution Flow Not Set Correctly.Caused by Inadequate Procedure.Procedure Revised temporarily.W/900319 Ltr ML20012C2891990-03-12012 March 1990 LER 90-003-00:on 900212,determined That Surveillance Procedures for Monthly Functional Testing of Drywell High Pressure Instrumentation Logic Channels Less than Adequate. Caused by Personnel Error.Procedures revised.W/900312 Ltr ML20011F4291990-02-26026 February 1990 LER 90-002-00:on 900131,discovered That Functional Test of Turbine Stop Valve Position Limit Switches Not Performed. Caused by Personnel Error When Writing Recent Rev.Rev to Procedure 34SV-C71-001-1S/2S written.W/900226 Ltr ML20006E2891990-02-0707 February 1990 LER 90-001-00:on 900112,reactor Scrammed Because MSIVs Were Less than 90% Open.Caused by Component Failure & Configuration of Condenser Vacuum Sensing Lines & Instruments.Valves replaced.W/900207 Ltr ML20006E0111990-02-0606 February 1990 LER 90-002-00:on 900114,RWCU Experienced High Differential Flow,Indicating Possibility of Leak in Sys.Caused by Component Failure & Less than Adequate Mounting for Relay. Relay Replaced W/Time Delay relay.W/900206 Ltr ML20006A8881990-01-22022 January 1990 LER 90-001-00:on 900104,HPCI Pump Declared Inoperable Due to Rated Flow Not Maintained During Surveillance Testing. Caused by Component Failure.Defective Resistor Replaced & Procedure 34SV-E41-002-1S performed.W/900122 Ltr ML20005E6541990-01-0202 January 1990 LER 89-010-00:on 891204,determined That Plant Was Not Fully Meeting Surveillance Requirements of Tech Spec Table 4.3.6.4-1,item 10.b.Caused by Inadequate Procedure.Recorder Calibr Steps to Be Removed from procedure.W/900102 Ltr ML20005E1851989-12-27027 December 1989 LER 89-009-00:on 891129,reactor Protection Sys Actuation Occurred from Scram Discharge Vol High Level Condition. Caused by Equipment Failure.Backup Temporary Air Compressor Placed Into Svc & Blown Fuse replaced.W/891227 Ltr ML20005E5131989-12-22022 December 1989 LER 89-017-00:on 891128,discovered That Efficiency Factors Used for Old Liquid Radwaste Discharge Radiation Monitors Incorrect.Caused by Personnel Not Incorporating Updated Efficiency Factor Into Sys software.W/891222 Ltr ML19332F8691989-12-14014 December 1989 LER 89-018-00:on 891114,sys High Differential Flow Condition Occurred Causing Actuation of Primary Containment Isolation Sys Valve Group 5 Logic Resulting in Closure of RWCU Valve. Caused by Personnel Error.Personnel counseled.W/891214 Ltr ML19332E6141989-11-30030 November 1989 LER 89-016-00:on 891103,discovered That Procedures 57SV-C51-001-1/2S Did Not Fully Test Rod Block Monitor Function.Caused by Procedural Deficiency.Limiting Condition for Operation Initiated & Procedure revised.W/891130 Ltr ML19332D8791989-11-29029 November 1989 LER 89-008-00:on 891102,RWCU Sys Experienced Partial Primary Containment Isolation Sys Group 5 Isolation Involving Valve 2G31-F004.Caused by Component Failure of Relay 2G31-R616D. Relay replaced.W/891129 Ltr ML19324C3271989-11-0808 November 1989 LER 89-014-00:on 891010,primary Containment Isolation Sys Group 5 Isolation Occurred Due to Opening of Valve 1G31-D002A.Caused by Personnel Error.Personnel Counseled & Memo Issued Re Confirming commands.W/891108 Ltr ML19325F1781989-11-0606 November 1989 LER 89-015-00:on 891009,diesel Generator 1R43-S001B Failed to Start Manually During Monthly Generator Test.Caused by Personnel Error & Incorrect Model Number Assigned to Pump. Pump Replaced & Oil Drained from cylinders.W/891106 Ltr ML19332B6191989-10-31031 October 1989 LER 89-013-00:on 891003,RWCU Sys Valve 1G31-F020 Closed, Rendering Inservice Reactor Coolant Monitor Inoperable. Caused by Cognitive Personnel error.In-line Conductivity Surveillance Initiated & Personnel counseled.W/891031 Ltr ML19325E6911989-10-31031 October 1989 LER 89-012-00:on 891003,plant Operators Received Indication That RWCU Sys Experiencing High Differential Flow Which Resulted in Isolation of Primary Containment Isolation Sys Valves.Caused by Component failure.W/891031 Ltr ML19327B3281989-10-23023 October 1989 LER 89-007-00:on 890926,four Safety Relief Valves Had Exhibited Drift in Mechanical Lift Setpoints in Excess of 3% Tolerance Specified by Inservice Testing Requirements.Root Cause Being Investigated.Valves refurbished.W/891023 Ltr ML19327B2991989-10-23023 October 1989 LER 89-006-00:on 890926,Procedure 34SV-SUV-019-2S, Surveillance Checks Did Not Fully Implement Requirements of Tech Spec Table 4.3.2-1.Caused by Personnel Error. Personnel Counseled & Procedure revised.W/891023 Ltr ML20024F4081983-09-0101 September 1983 LER 83-079/03L-0:on 830809,main Steam Line & Reactor Water Sample Valve Relay 1A71-K7A Determined Operating in Degraded Mode.Caused by Component Failure.Coil & Contacts Replaced. W/830901 Ltr ML20024F3341983-09-0101 September 1983 LER 83-064/03L-0:on 830811,during post-maint Review of DCR 83-76 Determined Torus Vent Valves Instrument Air Piping Returned to Svc W/O Performance of HNP-6907.Caused by Personnel oversight.W/830901 Ltr 1994-05-09
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217D3061999-10-13013 October 1999 SER Accepting Licensee Proposed Changes to Edwin I Hatch Nuclear Plant Emergency Classification Scheme to Add Emergency Action Levels Related to Operation of Independent Spent Fuel Storage Installation HL-5845, Monthly Operating Repts for Sept 1999 for Ei Hatch Nuclear Plant.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Ei Hatch Nuclear Plant.With ML20217A1691999-09-22022 September 1999 Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers ML20212A6641999-09-13013 September 1999 Safety Evaluation Authorizing Relief Request RR-V-16 for Third 10 Yr Interval Inservice Testing Program HL-5836, Monthly Operating Repts for Aug 1999 for Edwin I Hatch Nuclear Plant.With1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Edwin I Hatch Nuclear Plant.With ML20210J9631999-08-0202 August 1999 SER Finding That Licensee Established Acceptable Program to Verify Periodically design-basis Capability of safety-related MOVs at Edwin I Hatch Nuclear Plant,Units 1 & 2 ML20210J9271999-08-0202 August 1999 SER Finds That Licensee Performed Appropriate Evaluations of Operational Configurations of safety-related power-operated Gate Valves to Identify Valves at Plant,Susceptible to Pressure Locking or Thermal Binding HL-5818, Monthly Operating Repts for July 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With HL-5805, Monthly Operating Repts for June 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With ML20207E7631999-06-0303 June 1999 Safety Evaluation Concluding That Licensee Proposed Alternative to Use Code Case N-509 Contained in RR-4 Provides Acceptable Level of Quality & Safety.Considers Rev 2 to RR-4 & RR-6 Acceptable HL-5795, Monthly Operating Repts for May 1999 for Ehnp Units 1 & 2. with1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Ehnp Units 1 & 2. with ML20206G1691999-05-0404 May 1999 SER Approving Requirements of Istb 4.6.2(b) Pursuant to 10CFR50.55a(a)(3)(ii) HL-5784, Monthly Operating Repts for Apr 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With HL-5766, Monthly Operating Repts for Mar 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With ML20207M1891999-03-11011 March 1999 SER Accepting Relief Request for Authorization of Alternative Reactor Pressure Vessel Exam for Circumferential Weld HL-5755, Monthly Operating Repts for Feb 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With ML20206P6981999-01-0707 January 1999 Ehnp Intake Structure Licensing Rept HL-5726, Monthly Operating Repts for Dec 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With ML20196J4931998-12-0707 December 1998 Safety Evaluation Accepting Proposed Alternatives in Relief Requests RR-V-12,RR-V-15,RR-P-15,RR-V-7,RR-V-12,RR-V-14 & RR-V-15 HL-5714, Monthly Operating Repts for Nov 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With HL-5706, Monthly Operating Repts for Oct 1998 for Hatch Nuclear Plant Units 1 & 2.With1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Hatch Nuclear Plant Units 1 & 2.With ML20155B6121998-10-28028 October 1998 Safety Evaluation of TR SNCH-9501, BWR Steady State & Transient Analysis Methods Benchmarking Topical Rept. Rept Acceptable HL-5691, Monthly Operating Repts for Sept 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With ML20153G2481998-09-24024 September 1998 SE Concluding That Licensee Implementation Program to Resolve USI A-46 at Plant Adequately Addressed Purpose of 10CFR50.54(f) Request ML20239A2531998-09-0303 September 1998 SER Accepting Licensee Request for Relief Numbers RR-17 & RR-18 for Edwin I Hatch Nuclear Plant,Units 1 & 2.Technical Ltr Rept on Third 10-year Interval ISI Request for Reliefs for Plant,Units 1 & 2 Encl HL-5675, Monthly Operating Repts for Aug 1998 for Ei Hatch Nuclear Plant,Units 1 & 21998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Ei Hatch Nuclear Plant,Units 1 & 2 ML20238F7131998-08-31031 August 1998 9,change 2 to QAP 1.0, Organization HL-5667, Monthly Operating Repts for July 1998 for Ei Hatch Nuclear Plant,Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Ei Hatch Nuclear Plant,Units 1 & 2 HL-5657, Ro:On 980626,noted That Pami Channels Had Been Inoperable for More than Thirty Days.Cause Indeterminate.Licensee Will Replace Automatic Function W/Five Other Qualified Pamis of Like Kind in Drywell & Revised Procedures1998-07-30030 July 1998 Ro:On 980626,noted That Pami Channels Had Been Inoperable for More than Thirty Days.Cause Indeterminate.Licensee Will Replace Automatic Function W/Five Other Qualified Pamis of Like Kind in Drywell & Revised Procedures ML20236W3441998-07-30030 July 1998 Safety Evaluation Accepting Relief Requests for Second 10-yr ISI for Plant,Units 1 & 2 ML20236V5191998-07-28028 July 1998 Safety Evaluation Accepting Proposed License Amend Power Uprate Review ML20236N6751998-07-0909 July 1998 Part 21 & Deficiency Rept Re Notification of Potential Safety Hazard from Breakage of Cast Iron Suction Heads in Apkd Type Pumps.Caused by Migration of Suction Head Journal Sleeve Along Lower End of Pump Shaft.Will Inspect Pumps ML20236L1821998-07-0707 July 1998 Safety Evaluation Accepting 980428 Proposed Alternative to ASME Boiler & Pressure Vessel Code,Section Xi,Repair & Replacement Requirements Under 10CFR50.55a(a)(3) HL-5653, Monthly Operating Repts for June 1998 for Ei Hatch Nuclear Plant,Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Ei Hatch Nuclear Plant,Units 1 & 2 HL-5640, Monthly Operating Repts for May 1998 for Ei Hatch Nuclear Plant,Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Ei Hatch Nuclear Plant,Units 1 & 2 ML20248B8651998-05-15015 May 1998 Quadrennial Simulator Certification Rept HL-5628, Monthly Operating Repts for Apr 1998 for Ei Hatch Nuclear Plant1998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Ei Hatch Nuclear Plant HL-5604, Monthly Operating Repts for Mar 1998 for Edwin I Hatch Nuclear Plant,Units 1 & 21998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Edwin I Hatch Nuclear Plant,Units 1 & 2 ML20216B2711998-02-28028 February 1998 Extended Power Uprate Safety Analysis Rept for Ei Hatch Plant,Units 1 & 2 HL-5585, Monthly Operating Repts for Feb 1998 for Ei Hatch Nuclear Plant,Units 1 & 21998-02-28028 February 1998 Monthly Operating Repts for Feb 1998 for Ei Hatch Nuclear Plant,Units 1 & 2 HL-5571, Monthly Operating Repts for Jan 1998 for Edwin I Hatch Nuclear Plant,Unit 11998-01-31031 January 1998 Monthly Operating Repts for Jan 1998 for Edwin I Hatch Nuclear Plant,Unit 1 HL-5551, Monthly Operating Repts for Dec 1997 for Ei Hatch Nuclear Plant,Units 1 & 21997-12-31031 December 1997 Monthly Operating Repts for Dec 1997 for Ei Hatch Nuclear Plant,Units 1 & 2 ML20199B0561997-12-31031 December 1997 Rev 0 GE-NE-B13-01869-122, Jet Pump Riser Weld Flaw Evaluation Handbook for Hatch Unit 1 HL-5581, Annual Operating Rept for 1997, for Ei Hatch Nuclear Plant Units 1 & 21997-12-31031 December 1997 Annual Operating Rept for 1997, for Ei Hatch Nuclear Plant Units 1 & 2 HL-5533, Monthly Operating Repts for Nov 1997 for Ei Hatch Nuclear Plant,Units 1 & 21997-11-30030 November 1997 Monthly Operating Repts for Nov 1997 for Ei Hatch Nuclear Plant,Units 1 & 2 HL-5514, Monthly Operating Repts for Oct 1997 for Edwin I Hatch Nuclear Plant,Units 1 & 21997-10-31031 October 1997 Monthly Operating Repts for Oct 1997 for Edwin I Hatch Nuclear Plant,Units 1 & 2 ML20212A1981997-10-16016 October 1997 Safety Evaluation Denying Licensee Request for Relief from Implementation of 10CFR50.55a Requirements Re Use of 1992 Edition of ASME Code Section XI for ISI of Containments ML20211M6491997-10-0808 October 1997 Addenda 1 to Part 21 Rept Re Weldments on Opposed Piston & Coltec-Pielstick Emergency stand-by Diesel gen-set lube-oil & Jacket Water Piping Sys.Revised List of Potentially Affected Utils to Include Asterisked Utils,Submitted ML20211H5311997-10-0101 October 1997 Rev 2 to Unit 1,Cycle 17 Colr ML20211H5251997-10-0101 October 1997 Rev 3 to Unit 1,Cycle 17 Colr 1999-09-30
[Table view] |
Text
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APPENDIX 15B i
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DOSE MODELS USED TO EVALUATE !
-
THE ENVIRONMENTAL CONSEQUENCES ,
OF ACCIDENTS ;
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PVNGS FSAR
- .
APPENDIX 15B DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15B.1 INTRODUCTION This section identifies the models used to calculate control room and offsite radiological doses, not calculated in CESSAR, that would result from releases of radioactivity due to various postulated accidents.
15B.2 ASSUMPTIONS The following assumptions are basic to the model for the whole body dose due to immersion in a cloud of radioactivity and to the model for the thyroid dose due to inhalation of radio-activity:
~ A. All radioactive releases are treated as ground level
{
[ releases regardless of the point of discharge.
B. The dose receptor is a standard man, as defined by the International Commission on Radiological Protec-tion (ICRP) ,. (reference 1) .
C. No credit is taken for cloud d,epletion by ground deposition and radioactive decay during transport to the exclusion area boundary (EAB) or the outer boundary qf'the low-population zone (LPZ).
D. Radionuclide data, including decay constants and decay energies presented in table 15B-1, are taken from references 2 through 6.
15B.3 WHOLE BODY GAMMA AND BETA SKIN DOSE
-
The whole body gamma dose delivered to an offsite dose receptor is ca}cy% lated by assuming the receptor to be immersed T^ in a hemispherical radioactive cloud that is infinite in all 1
0302 -
elW4P 15B-1
PVNGS FSAR APPENDIX 15B Table 15B-1 RADIONUCLIDE PARAMETERS Average MeV/ Disintegration MeV/ Disintegration Nuclide Half-Life (gamma) (beta)
I-131 8.06 D 0.381 0.194 I-132 2.28 H 2.333 0.519 I-133 21 H 0.608 0.403 I-134 52 M 2.529 0.558 I-135 6.7 H 1.635 0.475 Kr-83m 1.86 H 0.002 0.037 Kr-85m 4.48 H 0.159 0.253 Kr-85 10.73 Y 0.002 0.251 Kr-87 76.31 M 0.793 1.324 Kr-88 2.80 H 1.950 0.375 Kr-89 3.16 M 1.712 1.001 Xe-131m 11.9 D 0.02 0.143 Xe-133m 2.25 D 0.0416 0.190 Xe-133 5.29 D 0.0454 .
0.135 Xe-135m 15.65 M 0.432 0.095 Xe-135 9.15 H 0.247 ,
0.316
'
Xe-137 3.83 M 0.194 l.642 Xe-138 14.17 M 1.183 '
O.606 l T.
directions above the ground plane; i . e . , 'a. emi-infinite cloud. The concentration of radioactive mat'drial within this 1 cloud is uniform and equal to the maximum centerline ground level concentration that would exist in the cloud at the appropriate distance from the point of release.
0303 ,
. g'
' ]
15B-2 ;
,
PVNGS FSAR APPENDIX 15B The gamma dose to an offsite receptor due to gamma radiation for a given time period is:
DCF
.
where D
wb
= whole body dose to an offsite receptor from gamma radiation, (rem) x/O = site atmospheric dispersion factor effective during the time period at the point of exposure, 3
(s/m )
DCF whole body dose conversion factor for the semi-bi =
infinite cloud model for nuclide i, (rem-m 3 /Ci-s).
(See table 15B-2)
Q. = total activity of nuclide i released during the j time period, (Ci)
V The gamma dose to the control room personnel is calculated assuming a finite hemispherical cloud model. The gamma dose due to gamma radiation in the control room for a given time period is:
D _ (CRV'O$) 0. 338 DCF wb1
.
( i} (
wb ~ 1173 i (CRVOL) ( 0. 02832 )
where D
wb
= whole body gamma dose to control room personnel from gamma radiation, (rem)
CRO = the control room occupancy factor $1 3600 = conversion factor, s/h
.02832 = conversion factor, ft /m CRVOL = control room volume, ft
,.:. 4 :n '~0304 i g 15B-3
-. . _ . - _ . _ . - - - . - - .. -
. , . . ..
PVNGS FSAR APPENDIX 15B Table 15B-2 WHOLE BODY GAMMA AND BETA SKIN DOSE CONVERSION FACTORS Beta Skin DCF Whole Body Gamma DCF Radionuclide (rem - m3/Ci - h) (rem - m3/Ci - s)
'I-131' l.14E2 8.72E-2 I-132 4.75E2 5.13E-1 I-133 2.65E2 1.55E-1 I-134 3.32E2 5.32E-1 I-135 4.64E2 4.21E-1 Kr-83m 0 5.02E-6 Kr-85 1.53E2 5.25E-4 Kr-85m 1.67E2 3.72E-2 Kr-87 1,llE3 1.87E-1 Kr-88 2.70E2 4.64E-1 Kr-89 1.15E3 5.25E-1 Xe-131m 5.43El 2.92E-3 Xe-133m 1.13E2 8.00E-3 lh Xe-133 3.49El 9.33E-3 Xe-135m 8.llEl 9.92E-2 Xe-135 2.12E2 5.72E-2 Xe-137 1.39E3 4.53E-2 Xe-138 4.71E2 2.81E-1 1
'
10 1 = total integrated activity for nuclide i in control room for the time period, (Ci-hr)
DCF the semi-infinite cloud whole body gamma dose wbi =
conversion factor for nuclide i, (rem-m 3/Ci-s).
(See table 15B-2)
- 0 The expression (CR O is a geometrical correction factor to ratio a finite cloud to infinite cloud (reference 7).
O
.
,
-
oaos 9 15B-4
. _ . . - -
f PVNGS FSAR APPENDIX 15B 0(./ The beta skin dose to control room personnel is calculated .
assuming a tissue depth of 7 mg/cm2 . The beta skin dose to control room personnel for a given time period is:
D ED -
Of (3)
Ss (CRVOL 02832) 1 Ssi where D =
the beta skin dose conversion factor for nuclide i, 6si 3
(rem-m /Ci-h). (See table 15B-2 for factor) and all other parameters are as previously defined. ;
15B.4 THYROID INHALATION DOSE The thyroid dose to an offsite receptor for a given time period is obtained from the following expression:
D = X/Q B I ( -
DCFf) (4) x
-
where:
D = thyroid -inhalation dose, (rem)
X/0 =
site. atmospheric dispersion factor during the time period, (s/m 3)
B =
breathing rate during the time period, (m /s)
, (See table 15B-3) 01 =
total activity of nuclide i released during time period, (Ci)
...
DCF =
thyroid dose conversion factor for nuclide i, i
(rem /Ci inhaled). (See table 15B-4)
The radionuclide data are given in table 15B-1. The atmospheric dispersion factors used in the analysis of the environmental consequences of accidents are given in section 2.3.
Y&
- o 6
15B-5
_
PVNGS FSAR APPENDIX 15B Breathing rates and dose conversion factors for radioactive iodines required for computing thyro'id inhalation doses are tabulated in tables 15B-3 and 15B-4, respectively.
Table 15B-3 BREATHING RATES ("}
Time After Accident m /s 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.47(-04) 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.75(-04) 1 to 30 days 2.32(-04)
- a. From Regulatory Guide 1.4
.
Table 15B-4 IODINE DOSE CONVERSION FACTORS ("}
'
Iodine Isotope (rem-thyroid / Curie Inhaled) ,
I-131 1. 4 8 (+0 6)
I-132 5. 3 5 (+0 4 ),
I-133 4.00(+05)
I-134 2. 5 0 (+ 0,4 )
I-135 1. 2 4 (+05 )~
- a. See reference 8
.
ISB.5 CONTROL ROOM DOSE *
,
During the course of an accident, control room personnel may receive doses from the following sources: "
.
A. Direct whole body gamma dose from the radioa,ctivity present in the containment building ,
B.. Direct whole body gamma dose from the radioactive
'~, cloud surrounding the control room 0307 M.
15B-6 !
PVNGS FSAR APPENDIX 15B
(
C. Whole body gamma, thyroid inhalation, and beta skin
( )
doses from the airborne radioactivity present in the control room.
In calculating the exposure to control room personnel, occupancy factors were obtained from reference 7 as follows:
0 to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: occupancy factor = 1 1 to 4 days: occupancy factor = 0.6 4 to 30 days: occupancy factor = 0.4 The dose model for each of the radiation sources are discussed below:
A. Direct whole body gamma dose from the radioactivity present in the containment building (direct containment dose).
Time integrated (0 to 30 days) radionuclide concentra-
,y
- tions in the containment are calculated. For
(_) conservati< redit is taken for reduction of the
.containmen- ..etivity by means other than radioactive decay. The containment is modeled by an equivalent volume cylindrical source having a diameter of 146 feet and a height of 155 feet. The radioactivity present in the containment is assumed to be uniformly distributed in the cylindrical source. Shielding is provided by the 4-foot concrete containment walls, 120 feet of air separating the containment building
,from the control building, and 2-foot thick control room walls.
.
No credit is taken for any shielding that would be provided by the auxiliary building.
B .- Direct whole body gamma dose from the radioactive cloud surrounding the control room (outside cloud dose) .
-,
,,
~
15B-7
PVNGS FSAR APPENDIX 15B Leakage from the containment building, or any building h
~
will result in the formation of a radioactive plume.
For conservatism it is assumed that this plume forms a cloud surrounding the control room. Gamma radiation from this cloud, although attenuated, can penetrate the control room roof and walls resulting in a whole body gamma dose to control room personnel. The radius of the cloud is computed using a mass balance of the radioactivity released due to leakage and the volume of the cloudt therefore, the radioactive cloud is time variant and expands for the duration of the accident.
Radioactivity concentration (Ci/m ) in the radio-active cloud surrounding the control room is the product of the building leak rato (Ci/s) and the control room atmospheric dispersion factor, X/0 (s/m 3),
Exclusion area boundary and low population zone X/Q's are presented in section 2.3. A tabulation of control room X/Q's is presented in table 15B-5.
The calculational model for the r;.untrol room is an equivalent volume hemisphere of radius 42 feet. Credit is taken for concrete shielding provided by the control room walls and ceiling.
Table 15B-5 ATMOSPHERIC DISPERSION FACTORS
.
Time Period Control Room X/Q (s/m )
0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.97(-3) 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.16(-3) 1 to 4 days 4.53(-4) 4 to 30 days 1. 3 0 (~4 )
O
>
- ,
15B-8
PVNGS FSAR
- APPENDIX 15B
\ '
/ C. Dose from the airborne radioactivity present in the control room (occupancy dose).
Airborne radioactivity will be drawn into the control room due to the intake of outside air required to maintain a positive pressure in the control room.
This contributes to the whole body gamma, thyroid inhalation, and beta skin doses. The major parameters of the control room ventilation system are presented in table 15B-6.
The whole body gamma dose is computed using a finite cloud model. The calculational model is an equivalent volume hemisphere of 42-foot radius.
A thyroid inhalation dose results from the radioactive iodine present in the control room. The control room habitability system, designed to renave iodine from the air, is described in table 15E-6.
b('N 15B.6 . ACTIVITY RELEASE MODELS 15B.6.1 GENERAL EQUATION The activity released from a postulated accident is calculated by using the following matrix equation for each isotope and each specie of iodine:
hf+cX=S; Initial Condition K(t )g = K o (5)
Q=L - AI where:
K(t) =
(a1(t))
ag = the activity in the ith node, (Ci)
C = (C matrix
-* 1 >:
i
.'g)
( )
</
2\9 15B-9
PVNGS FSAR APPENDIX 15B Table 15B-6 ,
CONTROL ROOM ESSENTIAL VENTILATION SYSTEM PARAMETERS ("
Parameter Assumption Number of emergency ventilation systems 1 operating Filtered Intake rate, standard ft / min 1,000 Unfiltered intake rate, standard ft / min 0 Intake cleanup filter efficiency Iodine, elemental, % 95 Iodine, organic, % 95 Iodine, particulate, % 99 Recirculation rate, standard ft / min 27,410 Recirculation cleanup filter efficiency Iodine, elemental, % 95 Iodine, organic, % 95 Iodine, particulate % 99 3
Leak rate, standard ft / min (out leakage) 1,000 Control room volume, standard ft 161,000
- a. There are two completely redundant emergency control room ventilation systems.
For a more detailed description of this system, refer to section 9.4.2. The dose model employed in this analysis is consistent with the thyroid inhalation model discussed in section 15B.4.
The beta skin dose model is consistent with the ,
,
" infinite hemispherical cloud" model described in ;
section 15B.3.
l l
0311
'
g 15B-10 1
PVNGS FSAR APPENDIX 15B
) C g = the transfer rate from the ith noJe to the jth node, (s-1)
(S1) vector S =
Sg = the production rate in the ith node (Ci/sec)
Q = the activity released to the environment over the time period t g to t g, (Ci) 5 =
(ty) matrix 1
1
= the leak rate from the ith node to environment
( /sec)
E =
A(t) dt (Ci-sea) o Each node represents a volume where activity can be accumulated.
The environment and the control room are each represented by a node. To ensure that the system of differential equations has constant coefficients, the time scale is broken up into time intervals over which all parameters are constant. Thus, all coefficients and sources are assumed to be representable by step functions.
The matrix equation is solved using matrix techniques. The particular solution is obtained by Gaussian elimination. The homogenous solution is obtcined by solving for the eigenvectors and the eigenvalues of the coefficient matrix C. They are determined by using QR transformation techniques.
The following sections describe how the coefficient matrix and the source vector are calculated for the different accident calculations.
'
0312
15B-ll l
PVNGS FSAR
,
APPENDIX 15B 15B.6.2 THE MODEL FOR CONTAINMENT LEAKAGE The model for LOCA containment leakage is shown in figure 15B-1.
The system of differential equations for estimating the released activity is as follows:
dA y+A A (6a) dt d1 -b21^2 -b31^3 =0 dA dt + ( d+ s +b21 + L23)^2 -b32^3 =0 (6M dA dt
-b 2 3' 2 + I d+L31 + b32)^3 = 0 (6c) dA dt
-
Q Ib u * (1 L f b
21 ^2 -h u+ IL b f 31 ^3 (6d)
,
+ (Lg+L +fRpc+ d} ^4 =0 t
O= (L21 ^2
Ay(t) = activity in the environment, (Ci)
A2(t)
= ctivity in the sprayed region of the containment, (Ci)
= activity in the unsprayed region of the contain-A3 ( t) ment, (Ci)
A4(t)
= activity in the control room, (Ci)
A d
= radioactive decay constant, (s-l)
T 21 b
21
(100)(24)(3600), (s-1)
T = leak rate from the sprayed volume to the environ-21 ment (%/ day)
T b
31
"
31 (100)(24)(3600), s -l)
T = leak rate from the unsprayed volume to the environ-31 ment (%/ day) ,
.
5B-12 0313 Y L
.. -.. - .
PVNGS FSAR APPENDIX 15B
-[ A s
= th spray removal constant, (s-1)
L 23 "
(v ) (60) , (s )
2 T = transfer rate from the sprayed region to the 23 3
unsprayed region, (ft / min) 3 V = volume of the sprayed region, (ft )
- 2 T
32 -1 32
"
(v 3) (60) ' U I T = transfer rate from the unsprayed region to the 32 sprayed region, (ft / min) 3 V = v lume f the unsprayed region, (ft )
3 T -
(.3048)3 L
u
"
60 , (m /s)
T = unfiltered inleakage into the control room, u
(ft /3 min)
60 , ( /sec)
T = filtered air intake rate into the control room, f
(ft /3 min) fg = filter efficiency of the filters on the intake units a
X/O = atmospheric dispersion factor for the control room, (s/m 3)
T r
R c
"
(V ) (60) ,
s -1) c T = filtered recirculatior, rate in the control room, R
(ft /3 min)
V = control room free. volume, (ft 3) c f
3
= filter efficiency of the filter on the recircula-tion unit Q = activity released to the environment, (Ci) a
,
...,,
u,.,,
a s '0314 15B-13
. - . .-
-. . .
.-
PVNGS FSAR APPENDIX 15B The coefficient matrix is:
C=
- __
+A -b -b 0 d 21 31 0 0
+(Ad+ s+b21+b23 - 32 0 -L 0 23 *I d+b31+b32) 0
-h(L+II-fILIb u L f 21 -hIb+II-E) u L bib 31 f
+( f+bu+fR c+ d}
-. _
Af ter solving for A(t) , the integrated activity in each node can then be calculated.
From the integrated activity, the offsite doses and the doses to the operators in the control room can be calculated using the dose models given in sections 15B.3 and 15B.4.
15B.6.3 THE MODEL FOR RECIRCULATION LOOP LEAKAGE The model for LOCA leakage in recirculation loops outside containment is shown in figure 15.B-2. The activity released due to the operational leakage of the engineered safety feature (ESF) components during the recirculation mode of the l postulated LOCA is calculated from the following equations:
dA dt
+
dlA - Il-II b 21 A 2
-0 (8a) dA
+ *8 dt (+ d+L21) A 2 2 t
1 Q= (1-f) L 21 A 2 dt (9) tg where:
A1 = the activity in the environment, (Ci)*
A2 = the activity in the ESF component rooms, (Ci) 0315 15B-14
. . - _. .. _. - .- .. . __- . . _ . . . _ . . _ .
,
PVNGS:FSAR APPENDIX 15B
/'
Ad = decay constant, (s~)
L21 = filtered leak rate to the environment, (ESF room vol/s) ,
f = filter efficiency of the filters on the ESF room purge i
- units.
i A T U 8
S2=P -
y s
Ag = activity in the recirculation water, (Ci)
,
P = iodine partition factor 3
T = twice the maximum operational leak rate, (cm /s) s
! 3 V g = total volume of recirculation water, (cm )
Q = activity released to the environment, (Ci)
! The coefficient matrix is:
"A -(1-f)L 21 d
C=
0 (A d+L21)
_
, +
I The source vector is
_ _
S=
8
- L 2_
j 15B.6.4 THE MODEL FOR THE FUEL HANDLING ACCIJENT IN THE FUEL BUILDING WITH ESF SAFEGUARDS ACTUAT'.ON
'
The model for the release of activity from **.te fuel building during a postulated fuel handling accident is shown in I' figure 15B-3. The activity released to the environment is l estimated-from the following equations:
.dA y+AA
.
!
dt- dl- Il~f) L 21A 2 = 0 , (10a) i
,
. (J c1Gu
(
9
-15B ,. - - , . _ . - . . .. : - .---~. ._ .... - . _ - - - .
PVMGS FSAR APPENDIX 15B dA dt 2+ I d+b21)^2 = 0 (10b) ty ;
O= L A dt 21 2 (11) o where:
Ay = activity in the environment, (Ci)
A = a tivity in the fuel building atmosphere, (Ci) 2 A
d
= decay constant, (s-1)
L 21 = purg rate to the environment, (s-1) f = filter efficiency of the filters on the ventilation unit Q = activity released to the environment, (Ci)
The resultant coefficient matrix is:
"A.d - (1-f) L 21 O
_
(Ad+ 21}
15B.6.5 OTHER ACCIDENT MODELS Other accidents can be conservatively modeled as simulated instantaneous releases to the environment. This is simulated as a large transfer rate to the environment. The model is shown in figure 15B-3. The system of differential equations is:
dA y+A u dt dl -b21^2 =0 (12a) dA dt 2* IAd+b21) ^2 =0 (12b) fty L (13)
Q=J 21 ^2 dt o
.
0317 O
'
15B-16
,.
PVNGS FSAR APPENDIX 15B I :
1/ where:
Ay = activity in the environment, (Ci)
A = activity to be released to the envitoninent, (Ci) 2 A
d
= decay constant, (s-1)
L21 = very large transfer rate to the environment, (s-1)
Q = activity released to the environment, (Ci)
The resultant coefficient matrix is:
"A -b d 21 C=
0
_
(Ad+ 21'_
15B.7 REFERENCES
- 1. " Report of ICRP Committee II, Permissible Dose for Internal
(}
'^'
Radiation (1959)," Health Physics, 3, p 30, 146-153, 1960.
,
- 2. Martin, M. J. and Blichert-Toft, P. H., Radioactive Atoms, Auger-Electron, L, B, y, and X-Ray Data, Nuclear Data Tables A8, 1, 1970.
- 3. Martin, M. J., Radioactive Atoms - Supplement 1, ORNL-4923, August 1973.
- 4. Bowman, W. W. and MacMurdo, K. W., " Radioactive Decay Gammas, Ordered by Energy and Nuclide," Atomic Data and Nuclear Data Tables 13, 89, 1974.
- 5. Meek, M. E. and Gilbert, R. S. , " Summary of Gamma and Beta Energy and Intensity Data /' NEDO-12037, January 1970.
- 6. Lederer, C. M., Hollander, J. M., and Perlman, I., Table of the Isotopes, 6th edition, March 1968.
- 7. Murphy, K. G. and Campe, K. M., " Nuclear Power Plant Control Room Ventilation System Design for Meeting General
.f ^' Criterion 19," Thirteenth AEC Air Cleaning Conference.
..
D318 -
M 15B-17
x PVNGS FSAR APPENDIX 15B
- 8. Di Nunno, J. J., et. al., " Calculation of Distance Factors for Power and Test Reacter Sites," TID 14844, March, 1962.
O
.
I I
!
.. i e g a ; . '
$
0 3,ai a 15B-18
_ _ _ _ - _ .
(3 CONTAINMENT UPSPR AYED REGION L21 V
32 J L '23 L CONTROL ATMOSPHER E V ROOM
'31 V
SPRAYED REGION O
Q DIRECT UNFILTERED LE AKAGE FR ACTION FROM v UNSPR AYED R EGION Q DIRECT UNFILTERED LEAKAGE FRACTION FROM v SPRAYED REGION
.c TR ANSFER R ATE FROM SPRAYED REGION TO UNSPR AYED REGION TR ANSFER R ATE FROM UNSPR AYED REGION TO SPRAYED REGION Palo Verde Nuclear Generating Station FSAR I ,
, ( s' CONTAINMENT LEA E I
DOSE MODEL 04'O Figure 15B-1 r, g
..-. - _ . . = . . . .. _ _ . _ . _ - . . . - . - - . _ - . . -. __ . . . . . - ~ .. - . _ . - . - . _ . . - . .. .
i
! l
!
. f i
!
1
!
i
{
'
1
!
- :
i i
i RO MS ATMOSPHERE
,
I '
r r FILTER -
r SOURCE i
!
1 i
i i
!
i l
!
- Q RECIRCULATION OF SUMP WATER TO ESF !
4 .
v OOMPON ENTS '
h DIRECT FILTERED LEAKAGE FRACTION FROM ESF ROOM i
i f
I-
!
!
,
!
.
Palo Verde Nuclear Generating Station FSAR
'
l ESF ROOM LEAKAGE DOSE MODELS
- 0321 Figure 15B-2
^
...
__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . . _ . _ . . _ . - - _ . . . . . - . _ . _ . _ _ _ _ _ . . . . . _ _ . . - - _ . . _ . . . _ . _ . , _ _ _ _ . . _ _ . _ _ . _ _,.. ____.- _ ,
__
h s
e si m v
PR IM ARY HOLDUP ATMOSPHERE SYSTEM y FILTER y (G
,l v
Q DIRECT UNFILTERED LEAKAGE FRACTION FROM V PP' MARY HOLD-UP SYSTEM Q DIRECT FILTERED LEAKAGE FR ACTION FROM v PRIMARY HOLDUP SYSTEM l
,
Palo Verde Nuclear Generatistg Station l
C g
0322 OTHER ACCIDENT DOSE MODEL Figure 15B-3 1
Jaur, I