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| Site | Start date | Title | Description | |
|---|---|---|---|---|
| 05000286/LER-2017-004 | Indian Point | 3 November 2017 20 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 | On November 3, 2017, at 2022 hours, 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. |
| 05000352/LER-2014-004 | Limerick | 4 March 2014 | Valid Manual Actuation of the Reactor Protection System With the Reactor Critical Due to Closure of Turbine Valves | A valid manual actuation of the reactor protection system (RPS) was initiated due to an unexpected closure of all six main turbine intercept valves (TVs). The cause of the main turbine IVs closure was due to a degraded Electro-hydraulic Control (EHC) 30 VDC house power supply. The Unit 1 EHC system was replaced with a Digital EHC (DEHC) system during the subsequent refueling outage 1R15. The Unit 2 EHC system is scheduled to be replaced with a DEHC system during the next refueling outage 2R13 in April 2015. The Unit 2 EHC house and permanent magnet generator (PMG) power supplies as-found voltages were verified to be within the calibration procedure limits and the as-left voltages were adjusted to the middle of the procedure acceptable band during the planned maintenance outage 2M49. |
| 05000339/LER-2013-001 | North Anna Power Station Unit 2 | 10 May 2013 | 1 OF 4 | On May 10, 2013, at 0612 hours with Unit 2 in Mode 1, 60 percent power following a refueling outage, a manual reactor trip was initiated as a result of increased vibrations on the number 9 main turbine/generator bearing and a report of a luminous discharge in the main generator exciter enclosure. All systems responded as expected following the manual trip. The auxiliary feedwater (AFW) pumps received an automatic start signal due to the resulting low-low level in "C" Steam Generator (SG). The AFW System operated as designed with no abnormalities noted and was subsequently returned to automatic operation. The SG levels were restored to normal operating level. When resetting AMSAC, a second AFW system automatic start occurred due to a procedure sequence error and was determined to be an invalid signal. At 0820 hours, a 4 hour report was made in accordance with 10CFR50.72(b)(2)(iv)(B) for Reactor Protection System (RPS) actuation and 8 hour report in accordance with 10CFR50.72(b)(3)(iv)(A) for the first AFW pump automatic start. This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) for a condition that resulted in automatic actuation of the RPS and AFW System. The health and safety of the public were not affected by the event since the equipment responded as designed. |
| 05000458/LER-2002-001 | River Bend Station | 18 September 2002 | Automatic Reactor Scram Due to Main Turbine Electro-hydraulic Control Malfunction | On September 18, 2002, at 8:25 p.m., an automatic reactor scram occurred while the plant was operating at 100 percent power. This event is being reported in accordance with 10CFR50.73(a)(2)(iv)(A) as an event that resulted in the automatic actuation of the reactor protection system and a manual actuation of the reactor core isolation cooling system. The sequence of events leading to the scram began with a voltage transient in the main turbine electro-hydraulic control system. This resulted in a "close" signal to the main turbine control valves, which caused reactor steam pressure to rise. The scram was initiated by a high neutron flux signal in the average power range monitors resulting from the rise in reactor steam pressure. Following the scram, a valve in the main condensate pump discharge header closed unexpectedly, shutting off flow to the reactor feedwater pumps. All three reactor feedwater pumps tripped in sequence due to low suction pressure, as designed. Operators initiated the reactor core isolation cooling system to provide makeup water to the reactor. This event is bounded by the analyses contained in the River Bend Updated Safety Analysis Report, and thus was of minimal significance with respect to the health and safety of the public. |
| 05000249/LER-2002-002 | Dresden Nuclear Power Station Unit 3 | 21 July 2002 | Reactor Scram due to Main Shaft Oil Pump Failure ' | On July 21, 2002, at 2033 hours Dresden Nuclear Power Station received an automatic reactor scram as the result of the main turbine tripping. The turbine trip was due to low discharge pressure from the turbine shaft bearing oil pump. Previous to the scram, on July 15, 2002, the turbine was taken off-line following identification of damage to the Permanent Magnet Generator (PMG). Following inspection of the PMG, the unit was returned to service on July 16, 2002. On July 21, 2002, low discharge pressure from the turbine shaft bearing oil pump caused the Unit 3 turbine to trip and consequently, a reactor scram occurred. The cause was determined to be the degradation of the auxiliary control rotor gear coupling insulation, resulting in current flow through the gear shafts which caused electrolysis and accelerated wear of the bearings. The corrective actions include revising turbine front standard procedures to include dismantling and inspection of the extension tube, inspecting and meggering the insulated coupling splines, and inspecting the front standard insulation kit per General Electric specifications. At no time did this condition compromise the health and safety of the public. The failure of the Main Shaft Oil Pump (MSOP) resulted in low turbine oil pressure, which initiated a turbine trip and reactor scram as designed. All safety systems functioned as designed following the reactor trip. '- NRC FORM 366A U.S. NUCLEAR REGULATORY APPROVED BY OMB N0.3150-0104 EXPIRES 07/31/2004 collection request 50 hrs Reported lessons learned are incorporated into the licensing process and fed back to industry Forward comments regarding burden estimate to the Information and Records Management Branch (t-6 f33), US Nuclear Regulatory Commission, Washington, DC 20555.0001, and to the Paperwork Reduction Project (3150-0104), Office Of Management And Budget, Washington, DC 20503. If an information collection does not display a currency valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection |
| 05000339/LER-2001-005 | Document Number | 22 December 2001 | On December 22, 2001, at 1542 hours, with Unit 2 in Mode 1 operating at 100% power, an automatic reactor trip occurred due a to failure in the turbine control electro hydraulic control (EHC) power supply system. A momentary fault on the control system power supply created a situation where the EHC system controller changed operating modes and reset its demand to zero. Consequently, the turbine valves drifted closed. The loss of load transient caused a reactor trip on low-low steam generator level. The reactor trip signal initiated a turbine trip. A non-emergency 4-hour notification was made to the NRC, at 1920 hours, on December 22, 2001, in accordance with 10CFR50.72(b)(2)(iv)(B). During the event, the auxiliary feedwater (AFW) system and accident mitigation system actuation circuitry (AMSAC) actuated. A non-emergency 8-hour notification was also made to the NRC, at 1920 hours, on December 22, 2001, in accordance with 10CFR50.72(b)(3)(iv)(A). The cause of the event was a failure of components in both the normal and backup turbine control EHC power supplies. Repairs to the components in both the normal and backup turbine control EHC power supply were performed prior to unit startup. No significant safety consequences resulted from this event because the reactor protection system and ESF systems functioned as designed following the reactor trip. The health and safety of the public were not affected at any time during this event. | |
| 05000339/LER-1993-002 | North Anna | 12 August 1993 | LER 93-002-01:on 930416,automatic Trip Occurred Due to Malfunction in Main Generator Voltage Regulator Circuitry. Performed Insp of turbine/generator.W/930812 Ltr |