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| Entered date | Event description | |
|---|---|---|
| ENS 54219 | 14 August 2019 14:47:00 | At 12:01 PM (EDT), on August 14, 2019, all fifty-four (54) of the Davis-Besse Nuclear Power Station Offsite Emergency Notification sirens were inadvertently activated for one minute during a planned silent test. The County Sheriff's Dispatch Office notified FirstEnergy Nuclear Operating Company of the inadvertent actuation. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). All sirens remain functional, and the NRC Resident Inspector has been notified of the issue. The Licensee will be notifying Ottawa and Lucas counties and the state of Ohio. The inadvertent activation was by the county dispatcher. |
| ENS 53051 | 3 November 2017 14:20:00 | On 11/3/17, with the unit operating in Mode 1 at approximately 100 percent power, an issue was identified with the Station Vent Radiation Monitors. The noble gas channels utilize an efficiency factor for isotope Kr-85 instead of the required Xe-133. For the normal range radiation monitors, the efficiency factor is non-conservative, resulting in both monitors being declared inoperable at 1045 hours EDT. As a result, the Normal Control Room Ventilation System was shut down and isolated, and the Control Room Emergency Ventilation System started in accordance with Technical Specification Required Actions at 1122 hours. The inoperability of both Station Vent Normal Range Radiation Monitors represents a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident. The Station Vent Accident Range Monitors also utilize an efficiency factor for Kr-85 instead of Xe-133, but for the Accident Range Monitors the efficiency factor is conservative. Because alternate means exist to determine release rate, which include use of grab samples and field surveys, this degraded capability does not represent a major loss of emergency assessment capability. The NRC Resident Inspector has been notified.
On 11/3/2017, the efficiency factors for the Station Vent Normal and Accident Range Radiation Monitors were revised to the proper setting for the required isotope, the normal range monitors were declared Operable, and the Control Room Normal Ventilation System was returned to service. An evaluation of the issue with the Station Vent Normal Range Radiation Monitors was performed, which determined the Control Room Ventilation isolation setpoint is well below the point at which the dose to the Control Room Operators would exceed General Design Criteria (GDC) limits following a Design Bases Accident. The error introduced from using an incorrect efficiency value did not challenge the margin to the GDC limits; therefore, the Station Vent Normal Range Monitors remained operable, and this issue did not prevent the monitors from fulfilling their safety function to mitigate the consequences of an accident. The NRC Resident Inspector has been notified. Notified R3DO (Stone). |
| ENS 52232 | 10 September 2016 07:23:00 | At 0343 EDT, with the unit operating at approximately 100% full power, an automatic reactor trip occurred due to a Main Generator lock-out. The cause of the generator lock-out is being investigated at this time. All control rods fully inserted. Post trip, the Steam Feedwater Rupture Control System was actuated due to high Steam Generator 1 level. The cause of the high Steam Generator 1 level is being investigated at this time. The unit is currently in Mode 3 (Hot Standby) and stable, at approximately 550 degrees F and 2155 psig. Steam is being discharged through the Atmospheric Vent Valves for decay heat removal. There is no known primary to secondary leakage, and all safety systems functioned as expected. The NRC Resident Inspector has been notified of the event. The licensee notified the State of Ohio, Ottawa and Lucas County. |
| ENS 52010 | 16 June 2016 14:59:00 | Upon review of recent industry operating experience, an issue was identified for the potential impact of the low barometric pressure associated with a tornado on the Emergency Diesel Generators (EDGs). The Davis-Besse Nuclear Power Station EDGs are equipped with a crankcase positive pressure trip with a set point of approximately 1 inch of water. This crankcase pressure trip is bypassed during an emergency start signal of the EDG from the Safety Features Actuation System or from an essential bus under voltage condition. Engineering has determined that a design basis tornado could create sufficient low pressure to potentially actuate the crankcase positive pressure trip due to different vent paths between the EDG Room and the EDG crankcase. If the crankcase pressure trip occurs before the EDG starts on an emergency signal due to the tornado, the crankcase pressure trip would cause an EDG lockout condition. The EDG lockout condition would then prevent either normal or emergency start of the EDG until operators could manually reset the lockout condition locally at the EDG. This condition could potentially affect both EDGs simultaneously. No severe weather warnings or watches are forecast in the local areas that could challenge the crankcase pressure trip. Compensatory measures are being established that upon notification of a Tornado Watch or Tornado Warning that would be implemented to defeat the crankcase pressure trip function and allow the EDGs to perform their required safety function during a potential tornado. The NRC Resident Inspector has been notified. |
| ENS 49369 | 20 September 2013 13:28:00 | A press release is being made today by the FirstEnergy Nuclear Operating Company regarding routine inspections of the Davis-Besse Nuclear Power Station's concrete shield building. These routine inspections of the Davis-Besse Nuclear Power Station's concrete shield building conducted to date have confirmed that the building continues to maintain its structural integrity and ability to safely perform its functions. The NRC Resident Inspector has been informed.
The press release originally provided to the NRC was revised prior to release to the public to update the inspections completed to date. The NRC Resident Inspector has been informed. Notified R3DO (Riemer). |
| ENS 47443 | 16 November 2011 03:04:00 | At 0222 EST on November 16, 2011, an ALERT was declared due to an electrical fire in the auxiliary building which houses safety related equipment. The apparent cause of the fire was due to an unknown source of water leaking on a breaker, thus causing an arc. The electrical fire is out. The plant was at 0% power and will remain shutdown in Mode 5. There was no impact on core cooling, or emergency power supplies. The licensee has notified the NRC Resident Inspector and state and local agencies.
At 0443 EST on November 16, 2011, Davis Besse, Unit 1 exited their ALERT. The electrical short affected the Control Room Emergency Ventilation Fan #1 Damper. The licensee has notified the NRC Resident Inspector. Notified R3DO (Hills) and Canada Nuclear Safety Commission (Jim Sandlef). |
| ENS 47096 | 26 July 2011 16:47:00 | Information was received in regards to an old design issue identified in a Component Design Basis Inspection Unresolved Item. Two issues were identified with the Safety-Related Direct Current (DC) System: 1. The plant's licensing basis states that non-safety-related electrical equipment, whose failure under postulated environmental conditions could prevent satisfactory accomplishment of the specified safety-related electrical equipment required safety functions, is qualified as required. However, the Reactor Coolant Pump (RCP) backup lift oil pump motors and the Containment Emergency Lighting Panel L49E1 are located inside containment and are not environmentally qualified. This could challenge the adequacy of electrical separation between the potentially grounded non-safety related equipment and the safety related batteries. 2. Automatic transfer switches are installed to automatically transfer non-safety related loads such as non-nuclear instrumentation, station annunciators, plant computer, and integrated control system between two non-safety related inverters, which receive power from the safety-related DC power system. If a ground fault existed on one of these switches, the fault could be transferred from one power source to the redundant source, potentially impacting the ability of both safety-related DC power sources to perform their required functions. This type of transfer is not permitted by the plant's licensing basis. The breakers for the 4 RCP backup lift oil pump motors and for the Containment Emergency Lighting were opened. One train of instrumentation power was placed on its alternate power source from the Alternating Current (AC) system, eliminating the potential to impact both trains of the DC power system. This condition is being reported per 10 CFR 50.72(b)(3)(ii)(B) as a condition that results in the plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v)(A-D) as an event or condition that could have prevented fulfillment of a safety function. The licensee has notified state and local authorities and the NRC Resident Inspector. |
| ENS 46653 | 3 March 2011 20:11:00 | While testing fire detection systems, a radio was keyed in the vicinity of the Auxiliary Shutdown Panel. Control Room alarms that occurred at the same time led to a review of plant data. This review revealed two momentary events (approximately 8 and 19 seconds) over an approximate two minute period that caused momentary reductions in the control signals to the Auxiliary Feedwater Pump and Motor-Driven Feedwater Pump discharge control valves. These momentary signal reductions resulted in all trains of Emergency Feedwater being inoperable for approximately two minutes, pending further evaluation. With all trains of Emergency Feedwater inoperable, this event is being reported in accordance with 10 CFR 50.72(b)(3)(v) as a momentary loss of safety function for equipment needed to (A) shut down the reactor and maintain it in a safe shutdown condition and to (B) remove residual heat. Fire detection testing has been completed, and a sign placed on the Auxiliary Shutdown Panel Room door stating that no radio usage is permitted inside the room. All trains of Emergency Feedwater are now operable. The licensee has notified the NRC Resident Inspector. |
| ENS 41698 | 13 May 2005 16:04:00 | On March 23, 2005, with the plant at 100% power, an issue was discovered questioning the design of an approximately 3-inch diameter module of containment electrical penetration (PBP5D) used to feed a non-safety related lighting panel (L49E1) inside the containment building. Specifically, it was identified that this penetration module was protected from overcurrent by a single 30-amp breaker (BF503). Action was taken at that time to open the supply breakers de-energizing the lighting panel to prevent any potential fault on the subject penetration module while the issue was being evaluated. Evaluations were initiated to determine the potential for the containment electrical penetration module to be thermally or mechanically damaged by a postulated fault. During this evaluation, no weak link upstream of the penetration module could be found that could be credited for preventing a fault in the penetration module in the event of a random failure of the breaker. The next breaker upstream of this circuit is too large to provide adequate overcurrent protection if the 30-amp breaker fails to open. Furthermore, it could not be shown that the electrical penetration module could withstand the maximum possible fault current while maintaining its containment integrity function. Therefore, it is now assumed that this postulated fault current could have damaged the penetration module to the extent that it would no longer have performed its containment integrity function. This condition is being reported within 8 hours of discovery in accordance with 10CFR50.72(b)(3)(ii)(B) as the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. Specifically, it was discovered that a system required to meet the single failure criterion does not do so. The licensee notified the NRC Resident Inspector. |