02-28-2017 | On March 7, 2016, while performing set-up activities for 2-PT-R084C, "23 EDG 8 Hour Load Test," the normal supply breaker to 480 Volt AC Bus ( ED) 3A tripped on overcurrent. This caused 480 Volt AC Buses 3A and 6A to de-energize since, as part of the test set-up activities, the tie breaker (3AT6A) between Buses 3A and 6A was closed and the normal supply breaker for Bus 6A was opened. This resulted in a loss of both 21 and 22 Residual Heat Removal ( RHR) {BP} pumps. As 'designed, all Emergency Diesel Generators ( EDGs) {EK} received automatic initiation signals to start. All required 480 Volt AC buses automatically re-energized by design, with the exception of Bus 3A, which had an overcurrent lockout. Operators manually started 22 RHR pump to restore RHR cooling.
However, prior to restoring the normal supply power to Bus 3A, 23 EDG tripped on overcurrent which resulted in a second loss of RIM event. The cause for the Bus 3A supply breaker tripping was inadequate procedural guidance resulting in excessive loads being energized on Buses 3A and 6A. The direct cause for 23 EDG tripping was cracked solder joints on the automatic voltage regulator ( AVR). Corrective actions included revising 2-PT-R084C and replacing the voltage regulator. The event had no effect on public health and safety. |
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Category:Letter
MONTHYEARML24011A1982024-01-12012 January 2024 ISFSI, Notice of Organization Change for Site Vice President ML23342A1082024-01-0909 January 2024 Independent Spent Fuel Storage Installation Security Inspection Plan ML23353A1742023-12-19019 December 2023 ISFSI, Emergency Plan, Revision 23-04 L-23-019, Proof of Financial Protection 10 CFR 140.152023-12-18018 December 2023 Proof of Financial Protection 10 CFR 140.15 ML23339A0442023-12-0505 December 2023 Issuance of Amendment No. 68, 301 and 277 Regarding Changes to Implement the Independent Spent Fuel Storage Installation Physical Security Plan ML23326A1322023-12-0505 December 2023 Issuance of Amendment No. 67, 300 & 276 to Implement the Independent Spent Fuel Storage Installation Only Emergency Plan ML23338A2262023-12-0404 December 2023 Signed Amendment No. 27 to Indemnity Agreement No. B-19 ML23356A0212023-12-0101 December 2023 American Nuclear Insurers, Secondary Financial Protection (SFP) Program ML23242A2772023-11-30030 November 2023 NRC Letter Issuance - IP LAR for Units 2 and 3 Renewed Facility Licenses and PDTS to Reflect Permanent Removal of Spent Fuel from SFPs ML23338A0482023-11-30030 November 2023 ISFSI, Report of Changes to Physical Security, Training and Qualification, Safeguards Contingency Plan, and ISFSI Security Program, Revision 28 ML22339A1572023-11-27027 November 2023 Letter - Indian Point - Ea/Fonsi Request for Exemptions from Certain Emergency Planning Requirements for 10 CFR 50.47 and 10 CFR Part 50, Appendix E IR 05000003/20230032023-11-21021 November 2023 NRC Inspection Report Nos. 05000003/2023003, 05000247/2023003, 05000286/2023003, and 07200051/2023003 ML23100A1172023-11-17017 November 2023 NRC Response - Indian Point Energy Center Generating Units 1, 2, and 3 Letter with Enclosures Regarding Changes to Remove the Cyber Security Plan License Condition ML23050A0032023-11-17017 November 2023 Letter - Issuance Indian Point Unit 2 License Amendment Request to Modify Tech Specs for Staffing Requirements Following Spent Fuel Transfer to Dry Storage ML23100A1252023-11-17017 November 2023 Letter and Enclosure 1 - Issuance Indian Point Energy Center Units 1, 2, and 3 Exemption for Offsite Primary and Secondary Liability Insurance Indemnity Agreement ML23100A1432023-11-16016 November 2023 Letter - Issuance Indian Point Energy Center Generating Units 1, 2, and 3 Exemption Concerning Onsite Property Damage Insurance (Docket Nos. 50-003, 50-247, 50-286) ML23064A0002023-11-13013 November 2023 NRC Issuance for Approval-Indian Point EC Units 1, 2 and 3 Emergency Plan and Emergency Action Level Scheme Amendments L-23-012, Master Decommissioning Trust Agreement Changes for Indian Point Nuclear Generating Units 1, 2 and 3, Pilgrim Nuclear Power Station, Palisades Nuclear Plant and the Non-Qualified Trust for Big Rock Point2023-11-13013 November 2023 Master Decommissioning Trust Agreement Changes for Indian Point Nuclear Generating Units 1, 2 and 3, Pilgrim Nuclear Power Station, Palisades Nuclear Plant and the Non-Qualified Trust for Big Rock Point ML23306A0992023-11-0202 November 2023 and Indian Point Energy Center, Notification of Changes in Schedule in Accordance with 10 CFR 50.82(a)(7) ML23063A1432023-11-0101 November 2023 Letter - Issuance Holtec Request for Indian Point Energy Center Generating Units 1, 2, and 3 Exemptions from Certain Emergency Planning Requirements of 10 CFR 50.47 and Part 50 ML23292A0262023-10-19019 October 2023 LTR-23-0211-RI Thomas Congdon, Executive Deputy, Department of Public Service, Chair, Indian Point Decommissioning Oversight Board, Letter Independent Spent Fuel Storage Installation Inspection and Office of the Inspector General Report-RI ML23289A1582023-10-16016 October 2023 Decommissioning International - Registration of Spent Fuel Casks and Notification of Permanent Removal of All Indian Point Unit 3 Spent Fuel Assemblies from the Spent Fuel Pit ML23270A0082023-09-27027 September 2023 Registration of Spent Fuel Casks ML23237A5712023-09-22022 September 2023 09-22-2023 Letter to Dwaine Perry, Chief, Ramapo Munsee Nation, from Chair Hanson, Responds to Letter Regarding Opposition of the Release and Dumping of Radioactive Waste from Indian Point Nuclear Power Plant Into the Hudson River ML23242A2182023-09-12012 September 2023 IPEC NRC Response to the Town of New Windsor, Ny Board Certified Motion Letter Regarding Treated Water Release from IP Site (Dockets 50-003, 50-247, 50-286) ML23250A0812023-09-0707 September 2023 Registration of Spent Fuel Casks ML23255A0142023-08-31031 August 2023 LTR-23-0211 Thomas Congdon, Executive Deputy, Department of Public Service, Chair, Indian Point Decommissioning Oversight Board, Letter Independent Spent Fuel Storage Installation Inspection and Office of the Inspector General Report IR 05000003/20230022023-08-22022 August 2023 NRC Inspection Report 05000003/2023002, 05000247/2023002, 05000286/2023002, and 07200051/2023002 ML23227A1852023-08-15015 August 2023 Request for a Revised Approval Date Regarding the Indian Point Energy Center Permanently Defueled Emergency Plan and Emergency Action Level Scheme ML23222A1442023-08-10010 August 2023 Registration of Spent Fuel Casks ML23208A1642023-07-26026 July 2023 Village of Croton-on-Hudson New York Letter Dated 7-26-23 Re Holtec Wastewater ML23200A0422023-07-19019 July 2023 Registration of Spent Fuel Casks ML23235A0602023-07-17017 July 2023 LTR-23-0194 Dwaine Perry, Chief, Ramapo Munsee Nation, Ltr Opposition of the Release and Dumping of Radioactive Waste from Indian Point Nuclear Power Plant Into the Hudson River ML23194A0442023-07-11011 July 2023 Clarification for Indian Point Energy Center License Amendment Request, Independent Spent Fuel Storage Installation Physical Security Plan ML23192A1002023-07-11011 July 2023 Response to Request for Additional Information Regarding License Amendment Request to Revise the Emergency Plan and Emergency Action Level Scheme ML23171B0432023-06-23023 June 2023 Letter - Indian Point Energy Center - Request for Additional Information for Independent Spent Fuel Storage Installation Facility-Only Emergency Plan License Amendment ML23118A0972023-06-0606 June 2023 06-06-23 Letter to the Honorable Michael V. Lawler, Et Al., from Chair Hanson Regarding Holtec'S Announcement to Expedite Plans to Release Over 500,000 Gallons of Radioactive Wastewater from Indian Point Energy Center Into the Hudson River ML23144A3512023-05-25025 May 2023 Clementina Bartolotta of Pearl River, New York Email Against Treated Water Release from Indian Point Site ML23144A3522023-05-25025 May 2023 Loredana Bidmead of New York E-Mail Against Treated Water Release from Indian Point Site ML23144A3412023-05-25025 May 2023 Dianne Schirripa of Rockland County, New York Email Against Treated Water Release from Indian Point Site ML23144A3472023-05-25025 May 2023 David Mart of Blauvelt, New York Email Against Treated Water Release from Indian Point Site ML23144A3402023-05-25025 May 2023 Melvin Israel of New York Email Against Treated Water Release from Indian Point Site ML23144A3542023-05-25025 May 2023 Terri Thal of New City, New York Email Against Treated Water Release from Indian Point Site ML23144A3532023-05-25025 May 2023 John Shaw of New York Email Against Treated Water Release from Indian Point Site 2024-01-09
[Table view] Category:Licensee Event Report (LER)
MONTHYEARNL-18-039, LER 2018-001-00 for Indian Point, Unit 2 Re Penetration Indications Discovered During Reactor Pressure Vessel Head Inspection2018-05-21021 May 2018 LER 2018-001-00 for Indian Point, Unit 2 Re Penetration Indications Discovered During Reactor Pressure Vessel Head Inspection 05000286/LER-2017-0042017-12-20020 December 2017 Reactor Trip Due to Main Generator Loss of Field, LER 17-004-00 for Indian Point Unit 3, Regarding Reactor Trip Due to Main Generator Loss of Field ML17252A8662017-09-0909 September 2017 Letter Regarding a 04/26/1977 Occurrence Concerning Failure of Number 22 Main Steam Line Isolation Valve to Close to a Manual Signal Initiated by the Control Room Operator - Indian Point Unit No. 2 05000247/LER-2015-0012017-08-29029 August 2017 Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size that Results in Exceeding the Allowed Leakage Rate for Containment, LER 15-001-02 for Indian Point, Unit 2, Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size that Results in Exceeding the Allowed Leakage Rate for Containment 05000286/LER-2017-0032017-08-29029 August 2017 Condensate Storage Tank Declared Inoperable Per Technical Specification, LER 17-003-00 for Indian Point, Unit 3, Regarding Condensate Storage Tank Declared Inoperable Per Technical Specification NL-17-107, LER 15-001-02 for Indian Point, Unit 2, Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate fo2017-08-29029 August 2017 LER 15-001-02 for Indian Point, Unit 2, Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate for 05000247/LER-2017-0032017-08-23023 August 2017 Technical Specification Violation of Section 3.3.1 RPS Instrumentation, LER 17-003-00 for Indian Point Unit 2, Regarding Technical Specification Violation of Section 3.3.1 RPS Instrumentation 05000247/LER-2017-0012017-08-22022 August 2017 Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused By Main Boiler Feedwater Pump Turbine Low Pressure Governor Valves Failed Closed, LER 17-001-00 for Indian Point, Unit 2 Regarding Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused By Main Boiler Feedwater Pump Turbine Low Pressure Governor Valves Failed Closed 05000247/LER-2017-0022017-08-22022 August 2017 Auxiliary Feedwater Flow Indication Inoperable for Longer Than the Allowed Technical Specification Completion Time Due to Failure of Complete Restoration Following Calibration, LER 17-002-00 for Indian Point, Unit 2 Regarding Auxiliary Feedwater Flow Indication Inoperable for Longer Than the Allowed Technical Specification Completion Time Due to Failure of Complete Restoration Following Calibration 05000286/LER-2017-0022017-08-0909 August 2017 Manual Isolation of Chemical and Volume Control System Normal Letdown to Stop a Valve Leak Resulted in an Exceedance of Technical Specification 3.4.9 Condition A Limit for Pressurizer Level, LER 17-002-00 for Indian Point, Unit 3 re Manual Isolation of Chemical and Volume Control System Normal letdown to Stop a Valve Leak Resulted in an Exceedance of Technical Specification 3.4.9 Condition A Limit for Pressurizer Level 05000286/LER-2017-0012017-07-13013 July 2017 Single Flow Barrier Access Point Found Unbolted, LER 17-001-00 for Indian Point, Unit 3 Regarding Single Flow Barrier Access Point Found Unbolted 05000247/LER-2016-0102017-02-28028 February 2017 Safety System Functional Failure Due to an Inoperable Containment Caused by a Through Wall Defect in a Service Water Supply Pipe Elbow to the 24, Fan Cooler Unit, LER 16-010-01 for Indian Point 2 Regarding Safety System Functional Failure Due to an Inoperable Containment Caused by a Through Wall Defect in a Service Water Supply Pipe Elbow to the 24 Fan Cooler Unit 05000247/LER-2016-0022017-02-28028 February 2017 Automatic Actuation of Emergency Diesel Generators (EDGs) Due to 480 VAC Bus Undervoltage Condition and Loss of Residual Heat Removal (RHR) While in Cold Shutdown, LER 16-002-01 for Indian Point, Unit 2 Regarding Automatic Actuation of Emergency Diesel Generators (EDGs) Due to 480 VAC Bus Undervoltage Condition and Loss of Residual Heat Removal (RHR) While in Cold Shutdown NL-16-108, LER 15-001-01 for Indian Point 2 Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate for Conta2016-09-29029 September 2016 LER 15-001-01 for Indian Point 2 Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate for Contai 05000286/LER-2015-0052016-09-14014 September 2016 Automatic Reactor Trip Due to a Turbine-Generator Trip Caused by the Trip of 345kV Main Generator Output Breaker 3 due to a Failure of South Ring Bus 345kV Breaker 5, LER 15-005-01 for Indian Point 3 RE: Automatic Reactor Trip Due to a Turbine-Generator Trip Caused by the Trip of 345kV Main Generator output Breaker 3 due to a Failure of South Ring Bus 345kV Breaker 5 05000286/LER-2015-0042016-09-14014 September 2016 Automatic Reactor Trip Due to a Turbine-Generator Trip Caused by a Failure of the 31 Main Transformer, LER 15-004-01 for Indian Point Unit No. 3 Regarding Automatic Reactor Trip Due to a Turbine-Generator Trip Caused by a Failure of the 31 Main Transformer 05000286/LER-2015-0072016-09-0606 September 2016 Manual Reactor Trip Due to Decreasing Steam Generator Water Levels Caused by a Miss- Wired Circuit Board in the Main Feedwater Pump Speed Control System, LER 2015-007-01 for Indian Point, Unit 3 Regarding Manual Reactor Trip Due to Decreasing Steam Generator Water Level Caused by a Miss-Wired Circuit Board in the Main Feedwater Pump Speed Control System 05000286/LER-2015-0062016-08-0808 August 2016 Technical SpecificatiOn Prohibited Condition Due to Two Pressurizer Code Safety Valves Discovered Outside their As-Found Lift Setpoint Test Acceptance Criteria, LER 15-006-01 for Indian Point Unit No. 3 Regarding Technical Specification Prohibited Condition Due to Two Pressurizer Code Safety Valves Discovered Outside Their As-Found Lift Setpoint Test Acceptance Criteria 05000286/LER-2014-0042016-08-0101 August 2016 Automatic Reactor Trip as a Result of Meeting the Trip Logic for Over Temperature Delta Temperature during Reactor Protection System Pressurizer Pressure Calibration, LER 14-004-01 for Indian Point Unit 3, Regarding Automatic Reactor Trip as a Result of Meeting the Trip Logic for Over Temperature Delta Temperature During Reactor Protection System Pressurizer Pressure Calibration 05000247/LER-2016-0042016-05-31031 May 2016 Unanalyzed Condition due to Degraded Reactor Baffle-Former Bolts, LER 16-004-00 for Indian Point 2 re Unanalyzed Condition Due to Degraded Reactor Baffle-Former Bolts 05000247/LER-2016-0052016-05-25025 May 2016 Technical Specification (TS) Prohibited Condition Due to a Surveillance Requirement Never Performed for Testing the Trip of the Main Boiler Feedwater Pumps, LER 16-005-00 for Indian Point, Unit 2, Regarding Technical Specification (TS) Prohibited Condition Due to a Surveillance Requirement Never Performed for Testing the Trip of the Main Boiler Feedwater Pumps 05000247/LER-2016-0012016-05-0202 May 2016 Technical Specification Prohibited Condition Caused by One Main Steam Safety Valve Outside Its As-Found Lift Set Point Test Acceptance Criteria, LER 16-001-00 for Indian Point 2 RE: Technical Specification Prohibited Condition Caused by One Main Steam Safety Valve Outside Its As-Found Lift Set Point Test Acceptance Criteria 05000247/LER-2015-0042016-02-18018 February 2016 Safety System Functional Failure Due to an Inoperable Containment Caused by a Flawed Elbow on the 21 Fan Cooler Unit Service Water Motor Cooling Return Pipe, LER 15-004-00 for Indian Point 2 Regarding Safety System Functional Failure Due to an Inoperable Containment Caused by a Flawed Elbow on the 21 Fan Cooler Unit Service Water Motor Cooling Return Pipe 05000286/LER-2015-0082016-02-11011 February 2016 Automatic Reactor Trip Due to a Turbine-Generator Trip as a Result of a Fault on 345 kV Feeder W96 Tower Lines Caused by Pre-Existing Degraded Insulator, LER 15-008-00 for Indian Point, Unit 3, Regarding Automatic Reactor Trip Due to a Turbine-Generator Trip as a Result of a Fault on 345 kV Feeder W96 Tower Lines Caused by Bird Streaming 05000247/LER-2015-0032016-02-0303 February 2016 Manual Reactor Trip Due to Indications of Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to a Power Supply Failure, LER 15-003-00 for Indian Point, Unit 2, Regarding Manual Reactor Trip Due to Indications of Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to a Power Supply Failure NL-15-124, LER 15-001-00 for Indian Point 2, Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate for Cont2015-10-0909 October 2015 LER 15-001-00 for Indian Point 2, Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate for Conta NL-13-166, Report on Inoperable Gross Failed Fuel Detector2013-12-20020 December 2013 Report on Inoperable Gross Failed Fuel Detector NL-13-038, Event Report for Discovery of a Condition That Prevented Immediate Protective Actions to Avoid Exposures to Radiation from Byproduct Material2013-02-19019 February 2013 Event Report for Discovery of a Condition That Prevented Immediate Protective Actions to Avoid Exposures to Radiation from Byproduct Material NL-12-060, Submittal of Report on Inoperable Gross Failed Fuel Detector2012-04-26026 April 2012 Submittal of Report on Inoperable Gross Failed Fuel Detector ML1101906402010-11-0909 November 2010 Event Notification Report; Subject: Power Reactor Indian Point Unit 2 NL-09-108, Submittal of Report on Inoperable Core Exit Thermocouples2009-08-10010 August 2009 Submittal of Report on Inoperable Core Exit Thermocouples ML0509600412004-12-17017 December 2004 Final Precursor Analysis - IP-2 Grid Loop ML0509600512004-12-17017 December 2004 Final Precursor Analysis - IP-3 Grid Loop NL-03-136, LER 03-03-00 for Indian Point Unit 3 Regarding Automatic Turbine Trip/Reactor Trip Due to Fault in 345kV Generator Output Breaker 32003-08-21021 August 2003 LER 03-03-00 for Indian Point Unit 3 Regarding Automatic Turbine Trip/Reactor Trip Due to Fault in 345kV Generator Output Breaker 3 ML0209104352002-03-19019 March 2002 LER 98-001-01 for Indian Point Unit 3 Re Potential Failure or Inadvertent Operation of Fire Protection Systems, Caused by Personnel Error in Design ML17252A8951979-05-25025 May 1979 Letter Reporting a 05/18/1973 Occurrence of a Pressure Transient within the Reactor Coolant System Due to the Closure of Certain Air Operated Valves in the Reactor Coolant Letdown System - Indian Point Unit 2 ML17252A8461974-02-19019 February 1974 Letter Regarding Performance of a Surveillance Test PT-M2 Reactor Coolant Temperature Analog Channel Functional Test - Delta T Overtemperature and T Overpower - Indian Point Unit No. 2 ML17252A8481974-02-19019 February 1974 Letter Regarding a February 1, 1974 Occurrence Where Both Door of the 80 Foot Elevation Personnel Air Lock to the Containment Building Were Inadvertently Open at the Same Time for a Period of About Thirty Seconds - Indian Point Unit. 2 ML17252A8471974-02-0808 February 1974 Letter Regarding an Occurrence on 1/25/1974 at the Indian Point Unit No. 2 Reactor Was Brought Critical in Preparation for Placing the Plant Back in Service Following Completion of Repairs Associated with No. 22 Steam Generator Feedwater Li ML17252A8491974-02-0606 February 1974 Letter Regarding an Occurrence Where Both Doors of the 80 Foot Elevation Personnel Air Lock to the Containment Building Were Inadvertently Opened at the Same Time for About Thirty Seconds - Indian Point Unit 2 ML17252A8501974-02-0505 February 1974 Letter Regarding an Occurrence Where a Slight Reactor Coolant System Pressure Transient Was Experienced in the Course of Placing a Reactor Coolant Pump in Service - Indian Point Unit No. 2 ML17252A8511974-02-0101 February 1974 Letter Regarding an Inspection of All Bergen-Paterson Hydraulic Shock and Sway Arrestors (Snubbers) Located in the Vapor Containment Was Performed and Two Did Not Meet the Established Criterion for Operability - Indian Point Unit No. 2 ML17252A8521974-01-31031 January 1974 Letter Regarding an Occurrence Where the Reactor Was Brought Critical Preparatory to Placing the Plant Back in Service Following Completion of Repairs Associated with the 11/13/1973 Feedwater Line Break Incident - Indian Point Unit No. 2 ML17252A8591974-01-28028 January 1974 Letter Regarding an Occurrence 01/23/1974 Where a Slight Reactor Coolant System Pressure Transient Above the Technical Specifications Limit Was Experienced in the Course of Placing a Reactor Coolant Pump in Service - Indian Point Unit No. 2 ML17252A8721974-01-18018 January 1974 Letter Regarding Analysis of Results of Monthly Periodic Surveillance Test PT-M11 (Steam Line Pressure Analog Channel Function Test) Indicated That One of the Low Steam Line Pressure Bistables Associated with High Steam - Indian Point Unit ML17252A8761973-12-28028 December 1973 Letter Regarding 12/17/1973 Analysis of the Results of Periodic Tests and Calibration Checks Relating to Pressurizer Level Indicated a Setpoint Drift - Indian Point Unit 2 ML17252A8771973-12-18018 December 1973 Letter Regarding a 12/17/1973 Analysis of Results of Periodic Tests and Calibration Checks Relating to Pressurizer Level Indicated a Setting for One of the Bistables Was Above the Technical Spec. Limit - Indian Point Unit 2 ML17252A8791973-12-0303 December 1973 Letter Regarding a 11/18/1973 Occurrence Relating to the Discovery of the Erroneous Setting for 1 of the Bistables Associated with Low Pressurizer Safety Injection Required by the Technical Specifications - Indian Point Unit No. 2 ML17252A8781973-11-30030 November 1973 Letter Providing Supplemental Information Concerning the 11/13/1973 Incident at Indian Point Unit No. 2 ML17252A8821973-11-19019 November 1973 Letter Concerning a 11/16/1973 Occurrence Regarding Periodic Tests and Calibration Checks Indicating the Setting for 1 of the Bistable Device Was Below the Technical Specification Requirements - Indian Point Unit 2 2018-05-21
[Table view] |
Note: The Energy Industry Identification System Codes are identified within the brackets {}.
DESCRIPTION OF EVENT
On March 7, 2016 at approximately 10:18 hours, with Indian Point Unit 2 in Cold Shutdown, Mode 5, Operations test personnel were performing set-up activities for surveillance procedure 2-PT-R084C, "23 EDG 8 Hour Load Test," when the normal supply breaker to 480 Volt AC Bus {ED} 3A tripped on overcurrent. This caused both 480 Volt AC Buses 3A and 6A to de-energize since, as part of the load test set-up activities, the tie breaker (3AT6A) between Buses 3A and 6A was required to be closed and the normal supply breaker for Bus 6A was required to be opened. The 8-hour load test was designed such that 23 EDG would power the loads on 480 Volt AC Buses 3A and 6A simultaneously. Approximately 14 minutes after cross-tying Bus 3A to Bus 6A and opening the Bus 6A normal supply breaker, the normal supply breaker to Bus 3A tripped on overcurrent. This resulted in a loss of assigned loads for both Bus 3A and 6A including 21 and 22 Residual Heat Removal (RHR) {BP} pumps, and 21 Spent Fuel Pool (SFP) {DA} pump. Technical Specification LCO 3.4.7 requires one RHR loop to be operable and in operation, and either the non-operating RHR to be operable and capable of being powered, or the secondary side water level in at least two steam generators to be greater than or equal to 0-percent narrow range. Technical Specification 3.4.7 Condition C was entered and operations personnel immediately initiated actions to restore one RHR loop to operation. There were no SSCs that were inoperable at the beginning of the event which contributed to the event. As designed, 21, 22, and 23 Emergency Diesel Generators (EDGs) {EK} received automatic engineered safety feature (ESF) start signals because of the loss of voltage on 480 Volt AC Bus 6A. As part of the load test set-up activities 23 EDG had already been running, although not tied to Bus 6A yet. At the time that both RHR pumps were de-energized, 24 Reactor Coolant Pump (RCP) {AB} was in operation and providing forced circulation in the reactor coolant system. The Main Steam System {SB} was available and the steam generators were coupled (i.e. pressurizer level and steam generator levels were adequate) thus decay heat removal was never lost. All 480 Volt AC buses re-energized automatically by design, with the exception of Bus 3A. Bus 3A had an overcurrent lockout that prevented 22 EDG from automatically loading onto the bus. At approximately 10:19, 22 RHR pump was started to restore RHR cooling. This event was recorded in the Indian Point Energy Center corrective action program as CR-1P2-2016-01256.
On March 7, 2016 at approximately 11:32 hours, before operators were able to complete the restoration of the normal supply power to Bus 3A, 23 EDG un-expectantly tripped on overcurrent. This resulted in a second automatic EDG engineered safety feature start signal which de-energiZed 480 Volt AC Buses 5A, 2A, 3A and 6A, and generated a start signal to the EDGs. This event was recorded in the Indian Point Energy Center corrective action'program as CR-1P2-2016-01260. Both loops of RHR cooling were lost because of 480 Volt AC Buses 3A and 6A being de-energized. At approximately 11:35 hours, Operations personnel manually started 21 RHR to restore residual heat removal capability as required by Technical Specification 3.4.7 Condition C.
On March 9, 2016 at approximately 20:41 hours, Operations test personnel had Commenced surveillance test 2-PT-R014, "Automatic Safety Injection System Electrical Load and Blackout Test". During the test, the voltage on 480 Volt AC Bus 6A dropped to approximately 200 volts when 23 Auxiliary Feedwater (AFW) Pump was sequenced to the bus. This event was recorded in Indian Point Energy Center corrective action program as CR-IP2-2016-01430.
On March 11, 2016 further investigation of the condition concluded that the automatic voltage regulation (AVR) system for 23 EDG was not functioning properly.
The AVR was replaced and 23 EDG was tested successfully and declared operable. The specific failure mechanism in the AVR was under investigation by a vendor. This condition was recorded in Indian Point Energy Center corrective action program as CR-IP2-2016-01260.
The onsite AC power distribution system includes 480 Volt AC Buses 5A, 6A, 2A and 3A which are divided into three safeguards power trains. The three safeguards power trains are Train 5A (Bus 5A and 21 EDG), Train 6A (Bus 6A and 23 EDG), and Train 2A/3A (Bus 2A and 3A and 22 EDG). The 480 Volt AC buses receive power from 6.9 kV bus sections through their respective Station Service Transformer {FK} (SST) or from associated onsite EDGs. The 480 Volt AC buses are designed with protection against undervoltage (UV) and degraded grid voltage (DGV) using relays that sense UV or DGV conditions.
In Mode 5 with the reactor coolant system (RCS) loops filled, the reactor coolant is circulated by means of two RHR loops connected to the RCS, each loop containing an RHR heat exchanger, an RHR pump, and appropriate flow and temperature instrumentation for control, protection, and indication. One RHR pump circulates the water through the RCS at a sufficient rate to prevent boric acid stratification.
The number of loops in operation can vary to suit the operational needs. The intent of LCO 3.4.7 is to provide forced flow from at least one RHR loop for decay heat removal and transport. The flow provided by one RHR loop is adequate for decay heat removal. The other intent of LCO 3.4.7 is to require that a second path be available to provide redundancy for heat removal.
Cause of Event
The direct cause of the event that occurred on March 7, 2016 at approximately 10:18 hours was a loss of power to 480 Volt Bus 3A when the normal supply breaker from Station Service Transformer (SST-3) tripped on overcurrent due to excessive loads energized on Bus 3A and Bus 6A. As designed, this caused the 480 Volt AC buses to strip loads, resulting in the loss of both 21 and 22 RHR Pumps, 21 SFP Pump, lighting and various other loads.
The apparent cause was inadequate guidance in 2-PT-R084C resulting in excessive loads energized on Buses 3A and 6A. Procedure 2-PT-R084C contained a precaution and limitation to limit current on the 3AT6A cross tie breaker but contained no guidelines on limiting current on the Bus 3A normal supply breaker (i.e. breaker overcurrent trip point was listed but this was for information only and there are no meters where this current can be read). At the time, SST-3 was carrying approximately 260 amps.
Procedure 2-PT-R084C was subsequently revised to include a precaution and limitation to maintain SST loads less than 200 amps.
The direct cause for the second transient that occurred on March 7, 2016 at approximately 11:32 hours was determined to be cracked solder connections on the 23 EDG AVR circuit card. Following a microscopic inspection of the solder joints for the magnetic amplifier by the vendor, it was determined that all nine terminals for the Ll magnetic amplifier exhibited solder cracking. Based upon the cyclic heating and cooling of the Ll magnetic amplifier, it was concluded that thermal growth of the Ll magnetic amplifier relative to the card could have caused cyclic stress on the solder joints. TherefOre, thermally-induced stress is the most likely cause of the solder joint cracks. The apparent cause was the failure to establish a PM strategy in accordance with a Part 21 notification and vendor's maintenance bulletin for the AVR card defect. Based on the intermittent nature of this failure in addition to elimination of all remaining potential causes, it is with high confidence that the cause for the overcurrent trip event which occurred on March 7, 2016 was the result of the intermittent symptoms associated with the degraded connections on the AVR assembly.
This condition was recorded in the Indian Point Energy Center corrective action program (CAP) as CR-1P2-2016-02944.
Corrective Actions
The following corrective actions have been or will be performed under the Corrective Action Program (CAP) to address the causes of this event.
- Revised 2-PT-R084A, B and C to include maintaining SST loads less than 200 amps.
This revision also added critical steps, one of which is verifying that SST load will be less than 200 amps after bus cross-tie is performed.
- The entire scope of Operations test activities was reviewed. Activities that had an impact on a shutdown key safety function were identified and reviewed to ensure that procedure quality was adequate.
- Past operability of 23 EDG was assessed. The AVR assembly was removed from 23 EDG and was tested successfully in accordance with the manufacturer's factory acceptance testing procedures the results from the vendor's surveillance tests, there was tripping on March 7, 2016.
Event Analysis
in spite of the cracked solder joints. Based upon failure analysis and previous satisfactory no indication that 23 EDG was inoperable before The event is reportable under 10 CFR 50.73(a)(2)(iv)(A). The licensee shall report any event or condition that resulted in the manual or automatic actuation of any system listed in 10 CFR 50.73(a)(2)(iv)(B). The systems to which the requirements of 10 CFR 50.73(a)(2)(iv)(A) apply include (#8) Emergency AC electrical power systems including emergency diesel generators. The actuation and start of the EDGs on the two occasions on March 7, 2016 at 10:18 hours and 11:32 hours meet the reporting criteria.
Pursuant to 10 CFR 50.73(a)(2)(v)(B), the loss of both 21 and 22 RHR loops was not considered to be a loss of safety function needed for residual heat removal, since at least one RHR pump was always capable of being powered by either the onsite or offsite power sources.
In accordance with 10 CFR 50.72(b)(3)(iv)(A), on March 7, 2016, at 17:12 hours, an eight hour non-emergency notification (#51775) was made for an event or condition which resulted in a valid actuation of the EDGs. The event was recorded in the Indian Point Energy Center corrective action program (CAP) as CR-IP2-2016-01256.
Past Similar Events
A review of the past five years of Licensee Event Reports (LERs) for events reporting valid Emergency Diesel Gerierator automatic actuations occurring during surveillance testing did not find any similar events.
Safety Significance
This event had no safety consequences on the health and safety of the public. In general, when the unit is shut down, the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents.
However, assuming a single failure and concurrent loss of all offsite or all onsite power is not. required. The rationale for Xhis is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, 3, and 4 have no specific analyses in MODES 5 and 6. Worst case bounding events are deemed not credible in MODES 5 and 6 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and the corresponding stresses result in the probabilities of occurrence being significantly reduced or eliminated, and have minimal consequences.
There were no actual implications from the event since there were no DBAs or radiological releases. During this event there were always two EDGs capable of supplying two safeguards power trains of the onsite AC electrical power distribution subsystems. Power from the offsite sources was available, and the ESF actuation circuitry and EDGs performed in accordance with design. The minimum safeguards power was available to power safety loads. At the time that both RHR pumps were de- energized, 24 Reactor Coolant Pump (RCP) was in operation and providing forced circulation in the reactor coolant system. The Main Steam System {S13} was available and the steam generators were coupled (i.e. pressurizer level and steam generator levels were adequate) thus decay heat removal was never lost.
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05000247/LER-2016-010 | Safety System Functional Failure Due to an Inoperable Containment Caused by a Through Wall Defect in a Service Water Supply Pipe Elbow to the 24, Fan Cooler Unit LER 16-010-01 for Indian Point 2 Regarding Safety System Functional Failure Due to an Inoperable Containment Caused by a Through Wall Defect in a Service Water Supply Pipe Elbow to the 24 Fan Cooler Unit | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000286/LER-2016-001 | Safety System Functional Failure Due to an Inoperable Containment Caused by a Flaw on the 31 .Fan Cooler Unit Service Water Return Coil Line Affecting Containment Integrity | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000247/LER-2016-001 | Technical Specification Prohibited Condition Caused by One Main Steam Safety Valve Outside Its As-Found Lift Set Point Test Acceptance Criteria LER 16-001-00 for Indian Point 2 RE: Technical Specification Prohibited Condition Caused by One Main Steam Safety Valve Outside Its As-Found Lift Set Point Test Acceptance Criteria | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000247/LER-2016-002 | Automatic Actuation of Emergency Diesel Generators (EDGs) Due to 480 VAC Bus Undervoltage Condition and Loss of Residual Heat Removal (RHR) While in Cold Shutdown LER 16-002-01 for Indian Point, Unit 2 Regarding Automatic Actuation of Emergency Diesel Generators (EDGs) Due to 480 VAC Bus Undervoltage Condition and Loss of Residual Heat Removal (RHR) While in Cold Shutdown | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat | 05000247/LER-2016-003 | Technical Specification (TS) Prohibited Condition Due to an Inoperable 21 Main Boiler Feedwater Pump Discharge Valve for Greater Than the TS Allowed Outage Time | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000247/LER-2016-004 | Unanalyzed Condition due to Degraded Reactor Baffle-Former Bolts LER 16-004-00 for Indian Point 2 re Unanalyzed Condition Due to Degraded Reactor Baffle-Former Bolts | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000247/LER-2016-005 | Technical Specification (TS) Prohibited Condition Due to a Surveillance Requirement Never Performed for Testing the Trip of the Main Boiler Feedwater Pumps LER 16-005-00 for Indian Point, Unit 2, Regarding Technical Specification (TS) Prohibited Condition Due to a Surveillance Requirement Never Performed for Testing the Trip of the Main Boiler Feedwater Pumps | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000247/LER-2016-007 | Safety System Functional Failure and Common Cause Inoperability of the Emergency Core Cooling System Due to Violation of Containment Sump Debris Barrier Integrity | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | 05000247/LER-2016-008 | Technical Specification Required Shutdown Due to Not Completing Repairs 'to,a,Refect in a Service Water Pipe to the 21 Component Cooling Water Heat Exchanger Within .tlie'TS AOT | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000247/LER-2016-009 | Automatic Reactor Trip Due to Actuation of the Trip Logic of the Reactor Protection System During Preparation for Testing | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
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