12-20-2017 | On November 3, 2017, at 2022 hours0.0234 days <br />0.562 hours <br />0.00334 weeks <br />7.69371e-4 months <br />, with reactor power at 100 percent, Indian Point Unit 3 experienced an automatic reactor trip on a turbine trip, which was in response to a main generator trip. The main generator trip was initiated by actuation of the Generator Protection System due to a main generator loss of field.
All control rods fully inserted and all required safety systems functioned properly. The plant was stabilized in hot standby with decay heat being removed by the main condenser. The Auxiliary Feedwater System (AFWS) automatically started as expected on steam generator low level to provide feedwater flow to the steam generators. The plant was stabilized in hot standby with decay heat being removed by the main condenser. The direct cause of the loss of main generator field was a failed Thyristor Firing Module drawer which affected proper operation of the redundant Thyristor Firing Module drawer. The root cause was determined to be that the Automatic Voltage Regulator (AVR) Firing Module power supplies have a latent design vulnerability where shared common output nodes are not isolated after a failure. A plant modification is proposed that will eliminate the condition by electrically isolating the AVR Firing Module power supplies upon failure.
This event had no effect on the public health and safety. The event was reported to the Nuclear Regulatory Commission (NRC) on November 3, 2017 under 10 CFR 50.72(b)(2)(iv)(B) and 50.72(b)(3)(iv)(A) as an event that resulted in the automatic actuation of the Reactor Protection System when the reactor is critical and a valid actuation of the AFWS. |
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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] |
comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
05000-286 Indian Point 3 2017 - 00 - 004 Note: The Energy Industry Identification System Codes are identified within the brackets { }.
DESCRIPTION OF EVENT
On November 3, 2017, at 2022 hours0.0234 days <br />0.562 hours <br />0.00334 weeks <br />7.69371e-4 months <br />, with reactor power at 100 percent, the Indian Point Unit 3 (IP3) Control Room operators received a Turbine Trip First Out Annunciator {ALM} and an automatic reactor trip {JC}, initiated by a main generator {TB} Lockout Relay 86BU {RLY, 86} trip. The 86BU relay trip was due to actuation of the main generator Loss of Field Relay 40 {RLY, 40}, and resulted in a direct trip of the 345 kilo-Volt (kV) generator output breakers 1 and 3 {EL, BKR, 52}. All control rods {AB} fully inserted and all required safety systems functioned properly. The plant was stabilized in hot standby with decay heat being removed by the main condenser {SG}. There was no radiation release. The emergency diesel generators {EK, DG} did not start, as offsite power remained available and stable. The Auxiliary Feedwater System (AFWS) {BA} automatically started as expected on steam generator (SG) {AB, SG} low level as a result of void fraction (shrink) effect. Indian Point Unit 2 (IP2) was unaffected and remained at 100 percent power.
The Generator Protection System protects the main generator from internal and external faults by tripping the field excitation breaker {BKR, 41} and the 345 kV generator output breakers 1 and 3. These circuit breakers are tripped by the Primary (86P) and Backup (86BU) Generator Lockout Relays {RLY, 86}, which also initiate a turbine trip {JJ}. The Turbine Protection System energizes solenoid valves 20/AST and 20/ASB {TG, PSV} to dump the autostop oil when a turbine trip is required. This removes the autostop oil pressure, allowing the turbine stop valves {TG, SHV} to close by spring action. A turbine trip can be actuated by a: (1) main generator trip, (2) reactor trip {JC}, (3) safety injection {BQ} actuation, or 4) manual trip. The Primary and Backup Generator Lockout Relays (86P and 86BU) provide the main generator trip signals to energize the 20/AST and 20/ASB solenoid valve for a turbine trip.
The November 3, 2017 reactor trip event was reported to the Nuclear Regulatory Commission (NRC) in a 4-hour non-emergency notification under 10 CFR 50.72(b)(2)(iv)(B) for an actuation of the Reactor Protection System (RPS) when the reactor is critical, and included an 8-hour notification for a valid actuation of the RPS and AFWS under 10 CFR 50.72(b)(3)(iv)(A) (Event Log No. 53052). This event notification was updated on November 6, 2017 to revise the report to reflect the actual Reactor Trip First Out Annunciator, which was for Generator Lockout Relay actuation of a turbine trip. The initial notification incorrectly stated that the reactor tripped on 33 SG low level. As previously described, SG low level is an expected post reactor trip transient condition. The event was recorded in the Indian Point Energy Center (IPEC) Corrective Action Program (CAP) as CR-IP3-2017-05133. A post transient evaluation was initiated and completed on November 7, 2017.
Prior to the November 3, 2017 reactor trip event, on August 31, 2017, the IP3 Control Room had received an Exciter Trouble alarm {ALM}. The operators dispatched to investigate discovered that the Drawer Operative light for one of the two Thyristor Firing Module drawers (Module B) was extinguished, indicating a loss of pulse. They also noted an acrid odor in the vicinity of the drawer and burn marks on the underside of the drawer. The Thyristor Firing Modules are part of the Main Generator Exciter System {TL, EXC}. No main generator capability was lost at the time, and IP3 remained online at full power. However, the Exciter Trouble alarm could not be cleared. So Engineering consulted the equipment vendor, Siemens-Allis {S188}, to understand the consequences of the discovered conditions. The vendor stated that the two Thyristor Firing Module drawers are parallel and redundant, and that full generator capability could still be maintained while operating on a single drawer. In addition, the vendor did not recommend opening the failed Thyristor Firing Module drawer because the vendor could not be certain of the material condition of the system, and opening the drawer may introduce a risk of tripping the generator. A Critical Decision Paper (CDP) and an Operator Decision Making Instructions (ODMI) document were prepared using vendor-provided information to assist plant management in determining whether or not to enter a comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
Indian Point 3 2017 - 00 - 004 05000-286
3. LER NUMBER
forced outage to repair the failed Thyristor Firing Module drawer. The decision was ultimately made to wait until the next forced outage or refueling outage to effect the repair. To support this outage repair work, a work order was prepared and the necessary replacement parts were staged.
The Main Generator Exciter System supplies the Direct Current (DC) field excitation current for the generator. It automatically maintains the generator output voltage and controls reactive load in accordance with the setting determined by the operator. It also maintains the generator output within the design capability to keep the generator in synchronism with the electrical transmission grid system. The Exciter System is a rotating, brushless type system that consists of a permanent magnet generator (PMG) {PMG}, an Alternating Current (AC) generator {GEN}, and a rotating rectifier assembly {RECT} mounted on a common shaft. The PMG provides 120 Volt AC, 420 Hertz (Hz), 3-phase power to the exciter field through the field breaker and four Power Amplifiers {AMP}. Two Thyristor Firing Modules control the firing (on) times of the Power Amplifier thyristors (SCRs) {SCR}, and by varying (delaying) the point in the AC sine wave at which the SCRs are pulsed, the magnitude of the current flow to the main generator exciter field can be changed. Each of the two Firing Modules has an internal +/-15 VDC power supply {JX} which is energized by the PMG. The 120 VAC, 420 Hz, 3-phase supply from the PMG is stepped down to 90 VAC inside the Firing Module and is rectified to a DC voltage. The +/-15 VDC voltage is used to power all of the other circuit cards inside the Firing Module. Once the main generator spinning at its rated 1800 revolutions per minute (RPM), the power supply output is always 15 VDC, regardless of generator load.
Two controls are used to transmit and adjust the main generator output terminal voltage demand signals to the Thyristor Firing Modules and Power Amplifiers. These are Base Adjust and Voltage Adjust. The Base Adjuster {EC} is used to generate the signal that determines a base or fixed value for excitation and serves as the means of adjusting voltage and reactive load in the manual mode. The Voltage Adjuster {EC} is used to adjust and, in conjunction with the Automatic Voltage Regulator (AVR) {90}, maintains a predetermined generator output voltage or reactive load. When the AVR is off, only the signal from the Base Adjuster is allowed to be applied to the two Thyristor Firing Modules. With the AVR in service, the AVR output is connected to the Firing Modules. The AVR output signal will add to or subtract from the base signal to maintain the desired main generator output terminal voltage as set by the Voltage Adjuster.
Following the November 3, 2017 reactor trip event, the AVR was removed from service and the Thyristor Firing Module B that had failed on August 31, 2017 was inspected. A visual inspection and subsequent bench testing revealed that internal components of the failed Thyristor Firing Module were damaged. Specifically, one of the phase transformers (T1) used to step down the 120 VAC, 420 Hz, 3-phase supply from the PMG to 90 VAC was damaged and four of the six input diodes (D1, D2, D3, and D5) for the +/-15 VDC power supply were damaged.
The T1 transformer was replaced, along with the other two phase transformers (T2 and T3). The power supply containing the failed diodes was also replaced with a spare power supply. Prior to returning it to service, the repaired Thyristor Firing Module B was tested for functionality. A Failure Modes Analysis (FMA), Fault Tree Analysis, and troubleshooting plan were used to identify and test different segments of the system, as well as the individual subcomponents. Each identified failure mode was systematically eliminated through testing. The AVR system as a whole underwent a full functional test prior to return to service in order to verify its proper operation.
The FMA and Fault Tree Analysis focused on the Main Generator Exciter, PMG, AVR, and their subcomponents for investigation, which collectively work to maintain field excitation in the main generator. The FMA identified a total of 48 potential failure modes, and these were reviewed independently by an Engineering consultant for reasonableness of the basis of each failure mode. Troubleshooting eliminated all but one failure mode, which could neither be eliminated nor confirmed. This potential failure mode describes a condition where the degraded/failed Thyristor Firing Module B caused a failure of the in service/operating parallel redundant Thyristor Firing Module A. This potential failure mode was supported by the investigation and testing of the failed components, which revealed that the power supply internal to Module B did not isolate upon failure and may have comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
05000-286 Indian Point 3 2017 - 00 - 004
3. LER NUMBER
affected the operation of Module A since the two modules share multiple connection points (i.e , the modules are not electrically isolated). A more detailed discussion of the failure mode follows:
During normal full power operations, the delta connected phase transformers T1, T2, and T3 in each Thyristor Firing Module drawer provide 3-phase 90 VAC at the input to the Firing Module power supply. This 90 VAC is rectified to 90 VDC, and Zener diodes regulate voltage to produce 15 VDC between the +15 VDC and common (COM) rails, and between the COM and -15 VDC rails. The COM rails are floating (i.e., not grounded). The Module A and B power supplies are each provided reverse current protection via blocking diodes on their +15 VDC and -15 VDC rails. Additionally, power supply relay K1 is provided to isolate the +15 VDC and -15 VDC output rails when the input voltage is not adequate. The COM rails do not have blocking diodes and are not isolated by the K1 relay, and as such, the COM rails are not electrically isolated on a power supply failure. This power supply circuit design vulnerability, when combined with the failures of the Module B input diodes and T1 transformer, created a low output impedance condition on the COM rail of the failed Module B power supply.
Because the COM rails are shared, and the COM rail of the failed Module B power supply remained unisolated, the low output impedance condition resulted in a backfeed current path to ground via the Module B drawer chassis and damaged T1 transformer core and winding insulation. The backfeed current placed an additional load on the operating Module A power supply, which caused the voltage on the shared COM rail to degrade. The voltage output of the Module A power supply eventually fell below 15 VDC, which caused the pulse generators to fail to provide the required output pulse train to sustain proper main generator field excitation.
Since this final remaining failure mode could not be proven by testing, the Engineering consultant suggested the implementation of a monitoring plan to allow for additional information processing during and following plant start up. This monitoring plan is currently in place via the installation of a high speed recorder, and has been incorporated in an ODMI.
An extent of condition (EOC) review was conducted to determine where the same or similar conditions may exist, but the adverse impact has not yet occurred. For this event, the EOC was limited to the IP2 and IP3 main generator AVRs where a failure could result in a direct plant trip either by a loss of excitation or by other means. The assessment concluded that the extent of condition at IP3 has a low associated risk due to the extensive troubleshooting and testing that occurred on the Exciter System and AVR following the November 3, 2017 reactor trip event, and since the system was returned to service with no known deficiencies. At IP2, although the AVR is of a different design, the recent operating history of the AVR will be reviewed to ensure that any similar conditions are properly evaluated and addressed.
CAUSE OF EVENT
The direct cause of the loss of main generator field was a failed Thyristor Firing Module drawer which affected proper operation of the redundant Thyristor Firing Module drawer. The loss of the main generator field actuated Generator Protection System Backup Lockout Relay 86BU, which in turn initiated the turbine trip and reactor trip.
The root cause of this event was that the IP3 AVR Firing Module power supplies have a latent design vulnerability where shared common output nodes are not isolated after a failure. The failure of the components in the Thyristor Firing Module B resulted in a degraded output from both Thyristor Firing Modules that caused the main generator exciter field to collapse. Evidence obtained while troubleshooting the cause of this event led to a plausible sequence of events and failure mode determination.
Significant contributing causes to this event are:
1. Inconsistent evaluation of vendor recommendations. This contributed to this event by not properly comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 00 - 004 2017 validating the vendor recommendation that was used during CDP development. This was also a contributor in that the decision not to upgrade the AVR, contrary to vendor recommendations, allowed the design vulnerability to remain in the system, creating an operating experience knowledge gap by continued use of equipment that has reached obsolescence within the industry.
2. Ineffective implementation of the critical decision process. This contributed to this event by removing the opportunity to challenge assumptions and smaller decisions that factored into the options of the final CDP.
The design documents for the AVR did not contain sufficient detail that would have allowed the organization to identify the latent design vulnerability. If the design vulnerability had been recognized in the recommendations from the CDP, the station would have been given the opportunity to make a different decision.
CORRECTIVE ACTIONS
The following corrective actions have been or will be performed under the Entergy CAP to address the causes of this event.
- Following the IP3 reactor trip event, repairs were made to the failed components and the system was returned to service with no known deficiencies.
- A temporary modification was implemented under Engineering Change (EC)-74798 to install monitoring equipment and record certain key I P3 main generator AVR operating parameters. This monitoring equipment will ensure that important system parameters are logged to assist in any troubleshooting which would follow a future failure of the AVR.
- An ODMI for CR-IP2-2017-04158 (Revision 2) is in effect to provide guidance for operation of the AVR with the Generator Exciter monitoring plan. This ODMI requires monitoring of generator parameters for abnormal operation, including the monitoring equipment installed under EC-74798, and provides direction on actions to be taken in the event that one or more generator parameters deviate outside of their normal operating ranges, or if thyristor power supply voltage is unstable.
- As part of the final system functional test, it was verified that a Thyristor Firing Module drawer could be electrically disconnected without creating a significant system disturbance. This will make future troubleshooting and isolation easier in the event of recurrence of this or a similar failure.
- Revise the critical decision process to require justification of assumptions and evaluate decisions regarding equipment issues for inclusion in the risk register.
- A plant modification is proposed to eliminate the latent design vulnerability condition by electrically isolating the AVR Firing Module power supplies upon a failure, which was the intended original system design.
- Review the recent operating history of the IP2 AVR to ensure that any similar conditions are properly evaluated and addressed.
EVENT ANALYSIS
The event is reportable under 10 CFR 50.73(a)(2)(iv)(A). The licensee shall report any event or condition that resulted in manual or automatic actuation of any of the systems listed under 10 CFR 50.73(a)(2)(iv)(B). Systems to which the requirements of 10 CFR 50.73(a)(2)(iv)(A) apply for this event include the Reactor Protection System (RPS) {JC}, including reactor trip, and AFWS actuation. This event meets the reporting criteria because an automatic reactor trip was initiated on November 3, 2017 at 2022 hours0.0234 days <br />0.562 hours <br />0.00334 weeks <br />7.69371e-4 months <br /> and the AFWS was automatically actuated on a valid low SG water level signal due to shrink effect.
Indian Point 3 05000-286 comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
Indian Point 3 05000-286 - 00 PAST SIMILAR. EVENTS A review was performed of the past five years for IP2 and IP3 Licensee Event Reports (LERs) that reported a reactor trip resulting from a failure of the Main Generator Excitation System, including the exciter and voltage regulator. The review identified IP2 LER 2012-006 that reported an automatic reactor trip which occurred on June 6, 2012. The reactor trip resulted from a turbine trip due to a trip of Backup Generator Lockout Relay 86BU on loss of main generator field excitation. Investigation determined the 86BU relay actuation was triggered by relay 62BU1/AUX {RLY, 62}, which serves as a time delay for the KLF-40 loss of field relay {RLY, 40}. The actuation of the loss of field relay was in response to a loss of generator field excitation from the Generrex voltage regulator system {TL, EXC}. The Generrex is the IP2 main generator exciter regulating system manufactured by the General Electric Company {G080}.
The direct cause of the reactor trip was loss of generator field excitation due to failure of the Generrex C-Phase Trigger Generation Card. The root cause was indeterminate, but was most likely due to premature failure of the U5 operational amplifier (op-amp) on the C-Phase Trigger Generation Card causing the U3 and U6 op-amps to also degrade. Corrective actions included replacement of the C-Phase Trigger Generator and AC/DC Gate cards with new cards which were then calibrated and monitored for proper operation. The failed C-Phase Trigger Generator card was shipped to a vendor for an equipment failure analysis, and it was later confirmed that the card failed due to failure of the op-amps.
SAFETY SIGNIFICANCE
This event had no effect on the health and safety of the public. There were no actual safety consequences for the event because it was an uncomplicated automatic reactor trip with no other transients or accidents, and the required primary safety systems performed as designed. The AFWS actuated and provided required feedwater flow to the SGs. The AFWS actuation was an expected reaction to the low SG water level caused by SG void fraction (shrink).
This occurs after a reactor trip due to main steam {SB} back pressure that results from the rapid reduction of steam flow following turbine control valve {TA, FCV} closure. A reactor trip with the reduction in SG level and AFWS actuation are conditions for which the plant is analyzed. This event was bounded by the analyzed event described in IP3 Updated Final Safety Analysis Report (UFSAR) Section 14.1.9, Loss of Normal Feedwater. The AFWS has adequate redundancy to provide the minimum required flow assuming a single failure. The UFSAR analysis demonstrates that the AFWS is capable of removing the stored and residual heat plus reactor coolant pump {P} waste heat following a loss of normal feedwater event, thereby preventing over-pressurization of the Reactor Coolant System (RCS) {AB} and preserving reactor coolant inventory.
The analysis in UFSAR Section 14.1.8, Loss of External Electrical Load, concludes that an immediate reactor trip on a turbine trip is not required for reactor protection. A reactor trip on a turbine trip is provided to anticipate probable plant transients and to avoid the resulting thermal transient. If the reactor {AC} is not tripped by a turbine trip, the over temperature delta temperature (OTDT) or over pressure delta temperature (OPDT) trip would prevent safety limits from being exceeded. This event was bounded by the analyzed event described in UFSAR Section 14.1.8.
The response of the plant is evaluated for a complete loss of steam load or a turbine trip from full power without a direct reactor trip. The analysis shows that the plant design is such that there would be no challenge to the integrity of the RCS or main steam system {SB} and no core safety limit would be violated.
For this event, all control rods inserted as required upon initiation of the reactor trip. The RCS pressure remained below the setpoint for pressurizer power operated relief valve (PORV) {AB, RV} and code safety valve {AB, RV} operation, and above the setpoint for automatic safety injection actuation. Following the reactor trip, the plant was stabilized in hot standby with decay heat being removed by the main condenser.
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05000286/LER-2017-001 | Single Flow Barrier Access Point Found Unbolted LER 17-001-00 for Indian Point, Unit 3 Regarding Single Flow Barrier Access Point Found Unbolted | 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-2017-001 | 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 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000286/LER-2017-002 | 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 | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 | 05000247/LER-2017-002 | 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 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000286/LER-2017-003 | 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 | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(V)(B) | 05000247/LER-2017-003 | 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 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000286/LER-2017-004 | 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 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
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