|Entered date||Site||Region||Reactor type||Event description|
|ENS 55775||9 March 2022 00:47:00||Davis Besse||NRC Region 3||B&W-R-LP|
The following information was provided by the licensee via phone and email: A non-licensed, contract employee supervisor had a confirmed positive for alcohol during a follow-up fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
The following information was received from the licensee via E-mail: This is a retraction of EN55775. The measured Blood Alcohol Level (BAC) of the individual was below the Fitness-For-Duty program limits, so this event did not constitute a violation of the Fitness-For-Duty program. The NRC Resident Inspector has been notified. Notified R3DO (Hills) and the FFD E-mail group.
|ENS 55734||5 February 2022 18:54:00||Davis Besse||NRC Region 3||B&W-R-LP||The following information was provided by the licensee via email: At approximately 1402 EST on 2/5/2022, with the Unit in Mode 1 at approximately 98 percent power, Operations was performing a valve lineup and inadvertently isolated a portion of the Reactor Coolant System (RCS) Letdown System, resulting in the system relief valve lifting and entry into the Makeup and Purification System Malfunction Abnormal Procedure due to loss of letdown. Pressurizer level increased and Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.9 CONDITION A was entered at 1414 EST due to Pressurizer level not below the limit of 228 inches, which has a REQUIRED ACTION to restore Pressurizer level within one hour. A rapid plant down power was initiated at approximately 1430 EST to reduce Pressurizer level. At 1514 EST on 2/5/2022, TS LCO 3.4.9 CONDITION B was entered, which has a REQUIRED ACTION to place the Unit in MODE 3 in 6 hours and in MODE 4 in 12 hours. As the Unit was continuing to down power, this represents initiation of a Technical Specification required shutdown, and this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). At approximately 1542 EST the down power was stopped at 15 percent power. Pressurizer level was restored to less than 228 inches at approximately 1603 EST, and TS LCO 3.4.9 was exited. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.|
|ENS 55346||9 July 2021 00:44:00||Davis Besse||NRC Region 3||B&W-R-LP||At 2154 EDT on 7/8/2021, with the Unit in Mode 1 at 100% power, the reactor automatically tripped due to trip of the main turbine, caused by failure of a non-safety related breaker during functional testing. Following the reactor trip the Steam Feed Rupture Control System automatically initiated on low Steam Generator 1 level, actuating both turbine-driven Auxiliary Feedwater Pumps. The operators subsequently started the high pressure injection pumps manually per procedure in response to overcooling indications. Operations responded and stabilized the plant. Decay heat was initially being removed via the Main Condenser. During post-trip response actions, while attempting to shut down the Auxiliary Feedwater Pumps, a low pressure condition was experienced in Steam Generator 2, resulting in isolation of the Main Condenser and steam being discharged through the Atmospheric Vent Valves for decay heat removal. There is no known primary to secondary leakage. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported in accordance with 10 CFR 50.72(b)(2)(iv)(A) as a four-hour, non-emergency notification of emergency core cooling system (ECCS) discharge into the reactor coolant system, and in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an eight-hour, non-emergency notification of an event that results in a valid actuation of the Auxiliary Feedwater System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.|
|ENS 55321||22 June 2021 16:55:00||Davis Besse||NRC Region 3||B&W-R-LP||At 1208 (EDT) on 6/22/2021, the high-energy line break door separating Auxiliary Feedwater Train Rooms 1 and 2 was not able to be latched following normal usage. The door was able to be closed, protecting Train 1 equipment from a break in Room 2. However, it is assumed a break in Room 1 would push the unlatched door open and allow high-energy fluids to enter Room 2. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. The door was able to be latched at 1215 (EDT) on 6/22/2021 following repairs to the door latch interlocking mechanism. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). No other equipment was inoperable during this event. The NRC Resident Inspector has been notified.|
|ENS 53381||4 May 2018 12:50:00||Davis Besse||NRC Region 3||B&W-R-LP|
On March 8, 2018, an invalid system actuation occurred while preparations were underway to perform Safety Features Actuation System (SFAS) integrated response time surveillance testing during the recent Davis Besse Nuclear Power Station refueling outage. Several minutes after connecting a data recorder to monitor the Emergency Diesel Generator (EDG) 1 start signal, at 1323 hours (EST), the EDG started with no valid actuation signals or test inputs present. The EDG successfully came up to speed and voltage as expected. The associated essential 4160 volt electrical bus remained energized from the normal power supply, therefore, the EDG output breaker did not close to supply power to the bus. Troubleshooting determined the inadvertent actuation was due to a short in the test lead wires at the recorder connection caused by a faulty test lead. The test lead was replaced and the SFAS surveillance testing completed satisfactorily.
This event is being reported as an invalid system actuation per 10 CFR 50.73(a)(2)(iv)(A); this 60-day optional telephone notification is being made per 10 CFR 50.73(a)(i) in lieu of submitting a written Licensee Event Report. The NRC Resident Inspector was notified of the inadvertent EDG start at the time of the event and has been notified of this invalid specified system actuation notification.
|ENS 53191||1 February 2018 13:50:00||Davis Besse||NRC Region 3||B&W-R-LP||A non-licensed (employee) supervisor had a confirmed positive test for alcohol during a random fitness-for-duty (FFD) test. The individual's unescorted access to the plant has been (terminated). The NRC Resident Inspector has been notified.|
|ENS 53051||3 November 2017 14:20:00||Davis Besse||NRC Region 3||B&W-R-LP|
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 52865||20 July 2017 13:30:00||Davis Besse||NRC Region 3||B&W-R-LP||In order to address the concerns outlined in RIS 2015-06 'TORNADO MISSILE PROTECTION,' an evaluation of tornado missile vulnerabilities and their potential impact on Technical Specification (TS) plant equipment was conducted. This evaluation concluded that the following Structures, Systems, and Components (SSCs) are potentially vulnerable to tornado generated missiles: The Davis-Besse Nuclear Power Station (DBNPS) Unit 1 Emergency Diesel Generator (EDG) Fuel Oil Storage Tanks (FOST) (DB-Tl53-l, DB-T-153-2) support the EDG operation for 7 days. The vents on the FOST are necessary to support the transfer of fuel from the FOST to the EDG day tank. These vents are not protected and are vulnerable to a potential tornado-generated missile impact. This postulated strike could impact fuel transfer to the EDG day tank and, therefore does not support operability of both EDGs for Technical Specification 3.8.1. Tornado generated missiles striking the FOST vent piping could potentially affect pump performance and challenge the structural integrity of the tank. This would render both the FOST and corresponding EDG inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(A). The potential vulnerabilities for the FOST vents (as discussed above) are being addressed in accordance with NRC EGM-15-002 Revision 1 and DSS-ISG-2016-01 NRC enforcement discretion and interim guidance documents. Immediate compensatory measures were taken to mitigate the potential consequences of an onsite tornado generated missile impact on the FOST vents. The licensee notified the NRC Resident Inspector.|
|ENS 52701||21 April 2017 07:33:00||Davis Besse||NRC Region 3||B&W-R-LP|
On 4/21/17, high grid voltage conditions were experienced, resulting in voltages higher than those established for operability of the offsite circuits. Grid voltages have been observed at approximately 355.8 kV on the nominal 345 kV system. As a result, both qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System were declared inoperable at 0202 hours (EDT). Main generator voltage control has been lowered to the minimum possible excitation with the unit operating at 100 percent power. The inoperability of both offsite circuits results in a loss of safety function in accordance with NRC reporting guidance. The voltage of the onsite Essential busses remains within acceptable values, and both Emergency Diesel Generators are operable. At 0715 (EDT) on 4/21/17, grid voltage has returned to an acceptable value and the equipment was declared operable. The NRC Resident Inspector has been notified.
Following the reporting of high switchyard voltage on 4/21/17, the licensee-established voltage limits were re-evaluated. The new high voltage limit has been established at 362.94 kV on the nominal 345 kV system, or 105.2 percent of nominal voltage, as compared to the previous maximum grid voltage of 103.3 percent. This new limit is above the 355.8 kV experienced on 4/21/17. Therefore, the equipment remained operable and no loss of safety function existed for the qualified circuits between the offsite transmission network and the onsite Class 1E AC electrical power distribution system, and the notification made per 10 CFR 50.72(b)(3)(v)(A-D) is being retracted.
An evaluation of the past three years of switchyard voltage data was also performed, and it was concluded the AC power system and its connected safety-related equipment remained capable of performing its required safety functions during the three-year evaluation period. The NRC Resident Inspector has been briefed on the evaluation results and informed of this retraction. Notified the R3DO (Stone).
|ENS 52476||6 January 2017 12:20:00||Bellefonte||NRC Region 2||B&W-R-LP||During a routine inspection on December 6, 2016, Tennessee Valley Authority (TVA) found the Unit 2 Reactor Building Containment Vertical Tendon V281 rock anchor/tendon anchor coupling had failed. The anchor coupling appears to have sheared in the threaded portion allowing the anchor head for the vertical tendon and the anchor head for the rock anchor tendon to separate. TVA had inspected the failed tendon coupling on October 19, 2016, and identified no signs of component specific damage or improper installation creating the potential for an unknown common mode failure. Cause of Deficiency: At this time, the cause for the failure of the V281 rock anchor/tendon anchor coupling is unknown. Safety Significance: As noted previously, the cause for the failure of the V281 rock anchor/tendon anchor coupling is unknown. As a result, the extent of condition can not be determined at this time. If multiple containment tendons are found to be losing the capability to carry tendon design force, and this condition was left uncorrected, it could reduce the capability of the containment structure to perform its design function. TVA had previously completed an analysis of containment structure integrity considering a single tendon coupler failure as a result of a similar failure of a Unit 1 Reactor Building Containment Vertical Tendon V9 in 2009 and determined that the containment structure is maintaining its design capability. Interim Action: Upon discovery on December 6, 2016, the following actions were taken by BLN (Bellefonte) personnel: Access to the Unit 2 tendon gallery was restricted. The area of the V281 tendon failure was subsequently cleaned. Grease samples were obtained and sent to TVA Central Labs for analysis. The couplings from both the rock anchor and tendon anchor locations were removed and sent to TVA Central Labs for metallurgical analysis. Grease samples were also collected from adjacent tendons (V272 through V290) to evaluate if conditions are similar to tendon V281 samples. The failure was entered into the BLN Corrective Action Program (Condition Report 1239343). Update Schedule: TVA plans to provide an update to this report by May 25, 2017 following the completion of the metallurgical and grease analysis. The Licensee has notified the NRC Construction Inspector (Baptist).|
|ENS 52247||16 September 2016 23:35:00||Davis Besse||NRC Region 3||B&W-R-LP|
At 1657 Eastern Daylight Time (EDT) the plant entered Mode 4 (from Mode 5), and subsequently, at 1710 EDT, it was discovered that 480V AC essential busses E1 and F1 were being supplied from the shutdown operations transformers. The essential busses E1 and F1 are required to be aligned to the power operations transformers in Mode 4 for operability in accordance with TS 3.8.9. With both E1 and F1 essential busses aligned to the shutdown operations transformers with the plant in Mode 4, both trains of the essential electrical power distribution system were inoperable, resulting in a loss of safety function. At 1733 EDT both E1 and F1 essential busses were aligned to the power operations transformers as required by TS 3.8.9. This issue is being reported as a loss of safety function of the essential electrical busses. The NRC Resident Inspector has been notified of the event.
Engineering reviewed the actual conditions during the approximate 36 minutes the 480V AC essential busses were being supplied from the shutdown operations transformers. Grid voltages were higher than assumed minimum voltages, and electrical loading during Mode 4 conditions were reduced from expected full power operation loading. As a result, Engineering determined that all equipment remained capable of performing its required functions while connected to the shutdown operations transformers. Because the equipment remained capable of satisfying the requirements for Operability, no condition existed that could have prevented the fulfillment of a safety function. Therefore, no loss of safety function existed for the 480V AC essential buses, and the notification made per 10 CFR 50.72(b)(3)(v)(A-D) by the Davis-Besse Nuclear Power Station on 9/16/2016 (EN# 52247) is being retracted. The NRC Resident Inspector has been briefed on the evaluation results and informed of this retraction. Notified the R3DO (Jeffers).
|ENS 52232||10 September 2016 07:23:00||Davis Besse||NRC Region 3||B&W-R-LP||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 52079||10 July 2016 10:29:00||Davis Besse||NRC Region 3||B&W-R-LP|
At 2342 (EDT), June 30, 2016, the Control Room received panel alarms associated with Safety Features Actuation System (SFAS) Channel 2. Subsequent investigation revealed the alarms were due to a loss of supplied power caused, in part, by a level permissive in the Borated Water Storage Tank (BWST) to be inoperable. With SFAS Channel 1 BWST level transmitter previously declared inoperable for maintenance, the on-shift operating crew did not correctly identify that technical specification (TS) 3.3.5, Condition B applied which is a 6-hour shutdown required action. At 0245, following a duty team call when the condition was re-assessed, the crew entered the proper additional Condition B and correctly identified they were approximately 3 hours into a 6-hour shutdown specification. At 0330 the condition was inappropriately exited on the premise that an operable but degraded situation could be justified. The plant did not initiate a shutdown required by technical specification but, in retrospect, should have initiated and completed a shutdown within 6 hours of 2342.
On July 1 at 1351, the BWST level transmitter for SFAS Channel 1 was repaired and declared operable (and exited TS 3.3.5 Condition B), however, the total time exceeded the 6-hour shutdown action. The plant remained stable throughout this event. On July 9, 2016, while internally discussing the event among FENOC senior leadership, it was determined that a 4-hour report would have been made if the shutdown was initiated. Hence, this report is retrospective in that a 10 CFR 50.72(h)(2)(i) required report should have been made upon the initiation of any nuclear plant shutdown required by plant's technical specification.
A Licensee Event Report will be provided pursuant to 10 CFR 50.73(a)(2)(i)(B) as a condition that was prohibited by the plant's technical specification. The NRC Resident Inspector has been notified. (At 1325 EDT on July 1, it was determined that the justification for SFAS channel 2 BWST level permissive to be operable but degraded could not be supported and reentered TS 3.3.5 Condition B.)
|ENS 52010||16 June 2016 14:59:00||Davis Besse||NRC Region 3||B&W-R-LP||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 51837||31 March 2016 00:14:00||Davis Besse||NRC Region 3||B&W-R-LP||On March 30, 2016, at 1715 EDT, with the Unit shutdown and in Mode 6 for refueling, evidence of leakage was identified on a 3/4-inch flexible braided piping connection on Reactor Coolant Pump (RCP) 1-1, and this issue was determined to be reactor coolant system pressure boundary leakage. This flexible piping is for RCP 1-1 first stage seal cavity vent line, and is categorized as ASME Section III Class 2 piping. The leakage was identified due to the discovery of a small amount of boric acid (approximately 1/2 teaspoon) on the welded end connection of the flexible piping. No active leakage was identified at the time of discovery with the Reactor Coolant System depressurized and approximately 110 degrees F. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.13, 'RCS Operational Leakage,' does not apply in the current plant condition (Mode 6). The cause and resolution of the leakage are under evaluation. This event is reportable within 8 hours per 10 CFR 50.72(b)(3)(ii)(A). The NRC Resident Inspector has been notified.|
|ENS 51702||30 January 2016 07:32:00||Davis Besse||NRC Region 3||B&W-R-LP||At 0123 EST, with the unit shutdown in Mode 3 (Hot Standby), during the performance of procedure DB-OP-06910, 'Trip Recovery,' while attempting to restore main feedwater to the Steam Generators, Davis-Besse received a Steam Feedwater Rupture Control System (SFRCS) 'reverse delta pressure' signal to the Auxiliary Feedwater System (AFW). The Auxiliary Feedwater System was operating at the time, feeding the Steam Generators. The SFRCS signal did result in actuation/closure (of) several valves in the Main Steam System, as the SFRCS signal is designed to do. This SFRCS signal/valve actuation was not anticipated. The unit remained in Mode 3 and is stable. This actuation did not have any negative impact to the AFW system and the ability to feed the steam generators. The NRC Resident Inspector has been notified of the event.|
|ENS 51696||29 January 2016 16:43:00||Davis Besse||NRC Region 3||B&W-R-LP||At 1322 EST, with the unit operating at approximately 100% full power, an automatic reactor trip occurred due to actuation of Reactor Protection System (RPS) Channel 4. The cause of the RPS actuation is being investigated at this time. Nuclear Instrumentation calibration for RPS Channel 2 was in progress at the time of the trip, with Channel 2 in bypass and Channel 1 in trip. All control rods fully inserted. Immediately post trip, the Steam Feedwater Rupture Control System actuated due to high Steam Generator 1 level due to unknown causes. The Main Steam Isolation Valves closed and Auxiliary Feedwater started as expected. Secondary side relief valves lifted in response to the trip, with two of the relief valves (one on each header) not properly reseating until operators manually lowered Main Steam Header pressure. The Bayshore 345 kV Offsite Electrical Distribution Circuit automatically isolated at the time of the unit trip. This was unexpected. The remaining offsite circuits remain in service. 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. Both primary Source Range nuclear instruments automatically energized, however, they were previously declared inoperable due to an administrative issue. Both Source Range instruments are functional and indicating properly. Both alternate Source Range instruments are operable, and all required Technical Specification actions have been completed. The NRC Resident Inspector has been notified of the event.|
|ENS 51483||20 October 2015 21:06:00||Davis Besse||NRC Region 3||B&W-R-LP|
Security condition in the Owner Controlled Area outside of the Protected Area. (There is) an unknown vehicle located on the South side of the Intake Canal. The vehicle is locked, the engine is not running, and the parking lights are on. Security is performing an inspection of the vehicle for explosives or other contraband in conjunction with local law enforcement. The Unusual Event was declared based on EAL HU-1. The licensee notified the NRC Resident Inspector. The licensee notified State and local government agencies. Notified (via phone and E-mail): DHS SWO, FEMA Ops Center, and NICC Watch Officer. Notified (via E-mail): FEMA NWC and NuclearSSA.
At 2229 EDT, the Unusual Event for a Security Condition at Davis-Besse Nuclear Power Station was terminated. An inspection of the vehicle in question was performed and it was determined that no threat existed to the site at any time. The licensee notified the NRC Resident Inspector. The licensee notified State and local government agencies. Notified R3DO (Daley), IRD (Stapleton), NRR (Morris) and ILTAB (Tucker). Notified (via phone and E-mail): DHS SWO, FEMA Ops Center, and NICC Watch Officer. Notified (via E-mail): FEMA NWC and NuclearSSA.
|ENS 51185||27 June 2015 07:16:00||Davis Besse||NRC Region 3||B&W-R-LP||On June 26, 2015 at 2335 (EDT), with Auxiliary Feedwater (AFW) train 1 declared inoperable for scheduled surveillance testing, AFW train 2 was declared inoperable as a result of the supply breaker for SW1395, Service Water Loop 2 secondaries isolation valve, being found open, i.e. out of its required position. Limiting Condition for Operation (LCO) 3.7.5 Condition D was entered for two Emergency Feedwater Trains inoperable. AFW Train 1 and the non-safety related motor driven AFW pump were available to provide emergency feedwater if required. The breaker was verified to be functioning as required and then closed, restoring the safety function. All associated LCOs were exited by 0133 (EDT) on June 27, 2015. The licensee notified the NRC Resident Inspector.|
|ENS 51061||9 May 2015 19:45:00||Davis Besse||NRC Region 3||B&W-R-LP|
At 1855 EDT, a steam leak from the #1 moisture separator reheater in the turbine building was reported to the control room. Operators performed a rapid down power to approximately 30% at which time the reactor was manually tripped. At 1910 EDT an Unusual Event was declared. The steam feed rupture control system was manually initiated (this includes actuation of both turbine-driven Auxiliary Feedwater Pumps) and the steam leak was isolated. Station air compressor #2 (non-safety related) tripped. Station air compressor #1 automatically started. The unit is currently in mode 3 (Hot Standby) and stable. Steam is being discharged through the atmospheric dumps as a means of decay heat removal. There is no known primary to secondary leakage. All systems functioned as expected. There were no reported injuries and personnel accountability is in progress.
The licensee notified state and local agencies and informed the NRC Resident Inspector. Notified DHS SWO, FEMA Ops Center, NICC Watch Officer and FEMA NWC and NuclearSSA via email.
The licensee exited the Unusual Event at 2121 EDT based on the following: At 2121 hours EDT, the Unusual Event at the Davis-Besse Nuclear Power Station was terminated. The steam leak has been isolated and plant conditions are stable. Cooling continues to be maintained via the auxiliary feedwater system. The initiation of auxiliary feedwater at the start of the event is reportable as a Specified System Actuation per 10CFR50.72(b)(3)(iv)(A). The licensee notified state and local agencies and informed the NRC Resident Inspector. Notified R3DO (Skokowski), NRR EO (Morris) and IRD (Grant). Notified DHS SWO, FEMA Ops Center, NICC Watch Officer and FEMA NWC and NuclearSSA via email.
|ENS 50639||25 November 2014 16:32:00||Davis Besse||NRC Region 3||B&W-R-LP||On Tuesday, November 25, 2014, at 1200 EST, the FirstEnergy Nuclear Operating Company (FENOC) reviewed AREVA 10CFR50.46 Notification Letter FAB14-00625 for the Davis-Besse Nuclear Power Station (DBNPS). This letter indicates that certain non-conservatisms were discovered in the methodology application and inputs used by AREVA for nuclear fuel core configurations with Mark-B-HTP fuel when operated under certain conditions. When corrected, this increases the Peak Cladding Temperature (PCT) in excess of the value prescribed in 10CFR50.46(b)(1) under Loss of Coolant Accident (LOCA) conditions. The DBNPS reactor core contains Mark-B-HTP fuel. 10CFR50.46 paragraph (b) defines the acceptance criteria for the LOCA analysis process. The DBNPS licensing basis PCT is evaluated for compliance with and must not exceed the criterion prescribed in 10CFR50.46(b)(1). AREVA had provided compensatory measures in terms of plant axial imbalance limits and Fq linear heat rate limits associated with reductions in LOCA linear heat rates so that the DBNPS operates within 50.46 limits. FENOC implemented the compensatory measures at the DBNPS on October 23, 2014, per AREVA recommendations, and as a result the errors reported have no impact on current plant operation or public health and safety. Preliminary analysis of past operating conditions indicate that the DBNPS did not exceed the 50.46(b)(1) criteria for PCT. This 8-hour notification is being reported in accordance with 10CFR50.72(b)(3)(ii)(B). Based on 50.46(a)(3)(ii) criteria, FENOC will submit a report within 30 days for the DBNPS. FENOC has notified the DBNPS NRC Senior Resident Inspector.|
|ENS 50381||19 August 2014 14:29:00||Davis Besse||NRC Region 3||B&W-R-LP|
At 1925 EDT on 08/18/2014 an equipment failure prevented a boundary door to the Shield Building Negative Pressure Area to latch closed upon egress, thereby preventing fulfillment of the Station Emergency Ventilation System safety function. Necessary door repairs per normal station practices were completed at 1935 EDT to establish full safety system function. This event was previously considered not reportable. Subsequent review determined the event reportable. The NRC Resident has been notified of the event. The failure to meet the 8-hour reporting requirement has been entered into the Corrective Action Program. The licensee will notify the State, Ottawa, and Lucas counties
At approximately 0413 (EDT) on 8/20/14, the boundary door to the Shield Building Negative Pressure Area again failed to latch closed upon egress. The door was able to be closed and latched at 0419, restoring the Station Emergency Ventilation System safety function. Door use will be limited to essential activities until final repairs to the door closure and latching mechanism are complete. The NRC Resident has been notified of the event. The licensee will notify the State of Ohio and local authorities. Notified R3DO (Passehl).
|ENS 50263||8 July 2014 10:06:00||Davis Besse||NRC Region 3||B&W-R-LP|
Unusual Event declared at 0935 EDT on July 8, 2014 due to a single fire alarm in containment. ELA HU4. No abnormal condition or plant impact at this time. The plant remains stable at 100% power. A single smoke detector inside containment alarmed at 0925 EDT. There are no other indications of smoke or fire at this time. Adjacent smoke detectors are not in alarm. A containment entry team is being assembled to verify that there is no fire inside containment. The smoke detector alarm cleared at 1006 EDT. The licensee notified the NRC Resident Inspector. The licensee will make appropriate notifications to state and local government agencies. Notified DHS, FEMA, and NICC via email and phone. Notified FEMA NWC and Nuclear SSA via email.
The Unusual Event was terminated at 1328 EDT on 7/8/14. A containment entry was performed and no indications of a fire existed. The fire alarm was spurious and the detector has been disabled. There was no notification of other government agencies and there is no media /press release planned. The plant maintained stable operations at 100% power. The licensee notified the NRC Resident Inspector, the state of Ohio and local officials. Notified R3DO (Hills) NRR EO (McGinty), IRC MOC (Grant). Notified DHS, FEMA, and NICC via email and phone. Notified FEMA NWC and Nuclear SSA via email.
|ENS 50252||3 July 2014 10:34:00||Davis Besse||NRC Region 3||B&W-R-LP||At 0336 (CDT) on 5/21/14, during activities to isolate the preferred power supply for wiring modification, all control room annunciators cleared and lost power due to a disconnect switch failure associated with the alternate power supply. The normal power supply was restored at 0349 (CDT) to restore the annunciators to functional status. During the brief period of time the annunciators were unavailable, redundant assets as described in existing station documentation and/or the control room alarm typer remained functional along with the station computer alarms and Safety Parameter Display System to provide backup assessment capability. This event was previously determined to not be reportable, however, following additional review of the significance this is reportable. The NRC Resident has been notified of the event. The late reporting of this event has been entered into the Corrective Action Program. The licensee notified the State of Ohio as well as Lucas and Ottawa counties.|
|ENS 50143||26 May 2014 10:39:00||Davis Besse||NRC Region 3||B&W-R-LP||At time 0252 EDT, the control room overhead annunciators malfunctioned resulting in the inability to receive more than one station alarm input and could not be acknowledged using the normal station alarm acknowledge pushbutton. This condition existed from time 0252 EDT until time 0516 EDT on 5/26/2014 at which time the system was restored. During the entire period, backup assessment capability was functional, dependent upon redundant assets as described in existing station documentation and/or functionality of the control room alarm typer. Additionally, the station computer alarms and Safety Parameter Display System remained functional during the entire time period. The licensee will continue to investigate the annunciator system. The licensee notified the NRC Resident Inspector. The licensee will also notify State and local government agencies.|
|ENS 50097||8 May 2014 17:46:00||Davis Besse||NRC Region 3||B&W-R-LP||On 5/4/14 while the plant was in Mode 3 and the reactor not critical, unexpected position indications were observed on a Control Rod while withdrawing an Axial Power Shaping Rod (APSR). Due to the uncertainty of rod positions, the APSR was inserted into the core. The reactor trip breakers were then opened from the Control Room using the manual trip pushbuttons. All Control and Safety Rods were unlatched and fully inserted into the reactor core before the reactor trip breakers were opened. This manual initiation of the Reactor Protection System with the reactor not critical is being reported per 10 CFR 50.72(b)(3)(iv)(A). The reportability of this event was determined based on an extent of condition review for Event Number 50086 that occurred 5/5/14. The failure to meet the 8-hour reporting requirement has been entered into the Corrective Action Program. The licensee notified the NRC Resident Inspector, the State of Ohio, and Ottawa and Lucas Counties.|
|ENS 50086||5 May 2014 21:17:00||Davis Besse||NRC Region 3||B&W-R-LP||During planned testing of the Control Rod Drive (CRD) system, the reactor trip breakers were opened via the manual reactor trip push buttons to de-energize a CRD motor in response to a high temperature. The partially withdrawn control rods fully inserted and all other rods remained in their initial positions. This manual Reactor Protection System (RPS) actuation while the reactor was not critical is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A). The licensee notified the NRC Resident Inspector, the State of Ohio, and Ottawa and Lucas Counties.|
|ENS 49833||16 February 2014 21:37:00||Davis Besse||NRC Region 3||B&W-R-LP||At time 1725 (EST) on 02/16/2014 a contract worker fell ten to fifteen (10-15) feet off of a ladder while descending the ladder above the 565 foot level in the Containment Building. The individual received lacerations to the forehead and potential other injuries. For his work activities the individual was dressed in anti-contamination clothing. The individual was transported via off-site ambulance in his anti-contamination clothing to H.B. Magruder hospital in Port Clinton, Ohio. A First Energy (FENOC) Radiation Protection (RP) Technician accompanied the injured person in the ambulance during the transfer to the hospital. A FENOC RP Supervisor along with the RP Technician assisted at the hospital while the injured individual received medical treatment. The protective anti-contamination clothing was removed from the individual. A full body frisk was performed and showed no contamination on the individual. Smear readings on the anti-contamination booties and gloves showed (approximately) 1500 dpm / 100 square cm. The clothing was bagged and subsequent readings and surveys on the outside of the bag showed no readings above background. The bagged anti-contamination clothing will be transported back to Davis-Besse NPS by FENOC RP Supervision. The H.B. Magruder Hospital Treatment Room and the off-site ambulance also were frisked and showed no residual contamination. The licensee notified the State of Ohio and Lucas and Ottawa Counties. The licensee notified the NRC Resident Inspector.|
|ENS 49828||14 February 2014 11:14:00||Davis Besse||NRC Region 3||B&W-R-LP|
On 02/14/2014, an unfilled area was discovered in the concrete along the top of the shield building construction opening on the annulus side. The condition was discovered during the current steam generator replacement outage, and is likely due to not completely repouring the shield building wall opening in 2011. Analysis shows this condition is bounded by previous calculations that demonstrate the containment function is maintained such that the protection of the health and safety of the public was not in question. Further analysis is planned to reconfirm previous calculations. The NRC Resident Inspector has been notified.
The FirstEnergy Nuclear Operating Company is retracting the 8-hour non-emergency notification made on 02/14/2014 (EN# 49828). Engineering evaluation of the unfilled concrete area along the top of the shield building construction opening determined the condition did not prevent the shield building from performing all design functions as described in the Updated Safety Analysis Report. Therefore, this issue did not represent a condition that significantly degraded plant safety, and the notification made per 10CFR 50.72(b)(3)(ii)(B) is being retracted. The NRC Resident Inspector has been briefed on the evaluation results and informed of this retraction. Notified the R3DO (Kunowski).
|ENS 49546||17 November 2013 19:27:00||Davis Besse||NRC Region 3||B&W-R-LP||A station annunciator system malfunction caused all Control Room annunciator indications to be in alarm status. This condition resulted in a loss of normal audible and visual plant condition assessment capabilities and was assessed as being a significant loss of assessment capabilities. Backup assessment capability existed and was dependent upon redundant assets as described in existing station documentation and/or functionality of the Control Room Alarm Typer. At 1659 EST full annunciator capability was restored following cleaning of a disconnect switch. The NRC Resident Inspector has been notified." State and local officials will be notified.|
|ENS 49369||20 September 2013 13:28:00||Davis Besse||NRC Region 3||B&W-R-LP|
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 49163||1 July 2013 15:27:00||Davis Besse||NRC Region 3||B&W-R-LP||On 07/01/2013 at 1103 (EDT), inspection personnel identified leakage from a 3/4 inch small-bore pipe socket weld for RCP 1-2 first stage seal cavity vent line. At current Reactor Coolant System conditions (Normal Operating Pressure and Temperature), the leak rate is approximately 8 to 9 drops per minute. The plant entered Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.13, 'RCS Operational Leakage', Condition B. Repairs to the line are being evaluated. This pressure boundary leakage is reportable per 10 CFR 50.72(b)(3)(ii)(A). The NRC Resident Inspector has been notified. The LCO requires the licensee to be in Mode 5 by 2303 EDT on 7/2/13.|
|ENS 49159||29 June 2013 22:48:00||Davis Besse||NRC Region 3||B&W-R-LP||Automatic trip of Reactor Coolant Pump 1-2 due to an electrical differential current fault resulted in an RPS actuation on Flux/Delta Flux/Flow. Startup Feedwater Valve 1 did not respond as expected post-trip and has been placed in manual control. All secondary side steam reliefs initially re-seated following reactor trip. Subsequent Main Steam Line #1 Safety Valve leakage mitigated during post-trip recovery actions. All other systems have functioned as expected. The plant is stable in Mode 3 - Hot Standby. All rods inserted into the core during the trip. Decay heat is being removed via turbine bypass valves to the main condenser with normal feedwater to the steam generators. The plant is in its normal shutdown electrical lineup. The licensee characterized the trip as uncomplicated. The licensee will be notifying Lucas and Ottawa counties, the State of Ohio and will be issuing a press release. They have notified the NRC Resident Inspector.|
|ENS 48000||7 June 2012 02:39:00||Davis Besse||NRC Region 3||B&W-R-LP||On June 6, 2012, at 1956 EDT, with the Unit shutdown for refueling, leakage was identified from a 3/4-inch weld during Reactor Coolant System (RCS) walkdown inspections. The leakage amount was approximately 0.1 gpm pinhole spray. During the performance of MODE 3 engineering walkdown inspections in accordance with procedure DB-PF-03010 (ASME Section III, Class 1 and 2), with the RCS at Normal Operating Temperature and Pressure, a pressure boundary leak was identified on the Reactor Coolant Pump (RCP) 1-2 1st seal cavity vent line upstream weld of 3/4 inch small bore pipe socketweld at a 90 degree elbow between the RCP pump and valve RC-407 (1st Seal Cavity Vent Isolation). The plant was in MODE 3 at Normal Operating Pressure and Normal Operating Temperature (NOP/NOT) for the inspections. The plant entered Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.13, 'RCS Operational Leakage,' Condition B and procedure DB-OP-02522. 'Small RCS Leaks,' abnormal operating procedure. Plant cooldown to comply with LCO 3.4.13, Condition B, Required Action B.2 is in progress. The cause and resolution are under evaluation. This event is reportable within 8 hours under 10CFR50.72(b)(3)(ii)(A). The NRC Resident Inspector has been notified. This condition has been documented in the Davis-Besse Corrective Action program as Condition Report 2012-09381. The plant is required to be in MODE 5 within 36 hours.|
|ENS 47670||16 February 2012 15:03:00||Davis Besse||NRC Region 3||B&W-R-LP|
At approximately 0542 hours (EST), the Safety Parameter Display System (SPDS) became non-functional, and therefore was not available in the Control Room or in the emergency response facilities. Efforts are underway to restore the system, but as of 1342 hours (EST) these efforts have not been successful. The loss of SPDS for more than 8 hours is being reported in accordance with 10 CFR 50.72(b)(3)(xiii). Troubleshooting efforts continue to restore the SPDS. The Emergency Response Data System (ERDS) remains functional, and the station remains capable of performing dose assessment using manual inputs per site procedures. The Control Room also continues to have the capability to retrieve plant data inputs to assess plant conditions and perform core damage assessments. The NRC Senior Resident Inspector has been notified, and the State of Ohio Emergency Management Agency will be notified.
As of 2100 EST on 02/16/12 the SPDS has been restored to service and is functioning properly. The licensee will inform the NRC Resident Inspector. Notified R3DO (Passehl).
|ENS 47443||16 November 2011 03:04:00||Davis Besse||NRC Region 3||B&W-R-LP|
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||Davis Besse||NRC Region 3||B&W-R-LP||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||Davis Besse||NRC Region 3||B&W-R-LP||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 46576||31 January 2011 10:14:00||Davis Besse||NRC Region 3||B&W-R-LP||At 0849 (EST) on 1/31/2011 the Emergency Notification System Sirens inadvertently actuated for 3 minutes. We (Davis Besse) have been informed by Ottawa county that the actuation signal appears to have originated from the Ottawa County Sherriff's dispatcher console. They also informed Davis Besse that Vendor Support is being requested to assist in the determination of the cause. The Ottawa County Emergency Management Agency is planning a news release for the inadvertent actuation. The Emergency Notification System sirens were secured and are considered to be operable. The licensee notified the NRC Resident Inspector.|
|ENS 46551||19 January 2011 03:10:00||Davis Besse||NRC Region 3||B&W-R-LP|
An electrical fire and explosions were reported near the Containment Access Facility construction area. An Unusual Event was declared based on HU4. Temporary electrical power at service disconnect DSLM3-3 was isolated. The fire was out at 0243 EST. The fire was extinguished using dry chemical. The fire was reported at 0232 EST on 1/19/11. The cause of the fire has not been determined at this time. The fire and explosions were initially reported by site security personnel. The licensee declared the NOUE at 0243 EST based on criteria HU4. The licensee initially called for offsite assistance in putting out the fire, however, the fire was extinguished by plant personnel and the offsite assistance was turned back. The licensee posted a reflash watch. The fire reportedly involved temporary cables and possibly a transformer supplying power to the construction area which is located inside the protected area outside the auxiliary building. The licensee notified the NRC Resident Inspector.
The licensee terminated the Notification of Unusual Event at 0358 EST. No additional information is available at this time. Notified R3DO (Bloomer), NRR EO (Skeen), IRD (Gott), DHS (Stringfield), and FEMA (Casto).
|ENS 46063||1 July 2010 13:58:00||Davis Besse||NRC Region 3||B&W-R-LP||(This is a) historical condition previously reported in LER 2009-001 (that) should also have been reported in accordance with 10 CFR 50.72(b)(3)(v)(D). Due to misapplication of Potter and Brumfield Rotary Relays, both operating Containment Air Cooler Fans were declared inoperable on October 13, 2009, and the fans switched from their normal fast speed alignment to the slow speed alignment used for accidents, which eliminated the relay issue and allowed them to be declared operable. This issue was reported in LER 2009-001 as an operation prohibited by the Technical Specifications on December 14, 2009. The fan control circuitry was modified to correct the condition. Upon further review, this condition should have been reported per the requirements of 10 CFR 50.72(b)(3)(v)(D) due to both required Containment Air Cooler Fan trains being declared inoperable for the condition. A revision to LER 2009-001 will be submitted per 10 CFR 50.73(a)(2)(v)(D). The NRC Senior Resident Inspector has been notified.|
|ENS 45764||13 March 2010 04:45:00||Davis Besse||NRC Region 3||B&W-R-LP|
On March 12, 2010, during the Davis-Besse Nuclear Power Station Unit No. 1 (DBNPS) refueling outage, the documented results of planned ultrasonic (UT) examinations performed on the Control Rod Drive Mechanism (CRDM) nozzles penetrating the reactor vessel closure head (RVCH) identified that two of the nozzles inspected to date did not meet the applicable acceptance criteria. Each of these two nozzles have similar indications that appear to penetrate into the nozzle walls from a lack of fusion point at the outer diameter of the nozzle and the J-Groove weld. Leak path detection was performed on both nozzles with the results showing no leak path. Bare metal visual examinations are scheduled to be performed on all nozzles to determine if there is any pressure boundary leakage. Both indications will require repair prior to returning the vessel head to service. There are sixty-nine nozzles and all will be subject to these UT inspections. It is important to note that this notification is being provided prior to the completion of all of the required UT and VT (Visual Tests) examinations for these two nozzles and the remaining nozzles. The indications were detected with a blade probe used to detect axially-oriented indications. The remaining probe that is used to complete the UT examination in each of these two penetrations is a blade probe that is used to identify circumferentially-oriented indications. The examinations are being performed to meet the requirements of 10CFR50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1, to identify potential flaws/indications well before they grow to a size that could potentially affect the structural integrity of the reactor vessel head pressure boundary. The licensee notified the NRC Resident Inspector and will notify the State of Ohio and both Lucas and Ottawa Counties.
On March 13, 2010, additional Control Rod Drive Mechanism nozzles (4 and 59) were identified that did not meet the applicable acceptance criteria. The indications on these nozzles are similar in nature and location to the indications previously reported for nozzles 28 and 33. These indications will also require repair prior to returning the vessel head to service. Like the two previously reported nozzles, this notification is being provided prior to the completion of all of the required examinations for the nozzle and the remaining nozzles. The indications were detected with a blade probe used to detect axially-oriented indications. The remaining probe that will be used to complete the examination in this penetration is a probe sensitive to circumferentially-oriented indications. Additionally, during the bare metal visual examination of the outer surface of the reactor vessel closure head, boric acid deposits were found at CRDM nozzles 4 and 33 that are indicative of primary water leakage. The visual examination of the RVCH is continuing. The licensee notified the NRC Resident Inspector, State of Ohio, and local government. Notified the R3DO (Phillips).
Update to Davis-Besse event #45764 initially reported 3/13/2010 at 0445 and adding an additional 10 CFR 50.72 reporting criteria. On March 15, 2010, the FirstEnergy Nuclear Operating Company is issuing a press release regarding the Davis-Besse Nuclear Power Station Unit No. 1 reactor vessel head (RVCH) Control Rod Drive Mechanism nozzles that have indications that do not meet the applicable acceptance regulatory criteria. The results of the ongoing planned ultrasonic (UT) examinations performed on these nozzles penetrating the RVCH have identified indications that will require repair prior to returning the vessel head to service. Currently, there are twelve nozzles which will be repaired, two of which have evidence of primary water leakage found during the bare metal visual examination of the outer surface of the reactor vessel closure head. There are sixty-nine nozzles and all are subject to these UT inspections. It is important to note that this media release is being provided prior to the completion of all of the required examinations for these and the remaining nozzles. The licensee notified the NRC Resident Inspector and will be notifying the State of Ohio and both Lucas and Ottawa Counties. Notified R3DO (Monte Phillips)
|ENS 45700||15 February 2010 16:42:00||Davis Besse||NRC Region 3||B&W-R-LP||A non-licensed, contract supervisor had a confirmed positive for a controlled substance during a random fitness-for-duty test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details.|
|ENS 45559||10 December 2009 16:23:00||Bellefonte||NRC Region 2||B&W-R-LP||Inspection of failed Unit 1 Reactor Building Containment Vertical Tendon V9 coupling indicates a potential for an unknown common mode failure mechanism for BLN Containment vertical tendon rock anchor couplings. Unit 1 Reactor Building Containment Vertical Tendon V9 experienced a failure of the rock anchor/tendon anchor coupling on August 17, 2009 at approximately 1400 CDT. The time of failure was identified based on a loud noise bang reported by several individuals. Initial investigation failed to reveal the source of the noise. The failed tendon was discovered on August 24, 2009 during a tour of U1 Tendon Gallery, elevation 607. Unsafe conditions previously precluded an inspection of the failed coupling for proper installation or component specific damage. The failed tendon coupling was inspected on 11/23/2009 and showed no signs of component specific damage or improper installation creating the potential for an unknown common mode failure. Safety significance: Until the mechanism of failure is identified the extent of condition will not be known. If multiple containment tendons are found to be losing the capability to carry tendon design force and this condition was left uncorrected, this could jeopardize the ability of the containment structure to perform its design function. Causes of deficiency: The cause of this deficiency is unknown at this time. Further analysis is in progress and when completed, an update to this report will be provided. Interim progress: Grease from the lower anchor head can has been analyzed for moisture content. Results were within vendor specifications. Additional samples have been sent for further analysis as described in Regulatory Guide 1.25 'In-service Inspection of Ungrouted Tendon in Prestressed Concrete Containments.' After successful safe securing of the tendon load, the failed coupling was visually inspected. The visual inspection of the failed coupling did not indicate a component-specific failure mechanism or indication of visually apparent common mode failure mechanism. Based on this inspection visual inspection of additional tendon coupling tendon couplers is not warranted at this time. The coupling has been removed from both the tendon anchorhead and the rock anchor tendon anchorhead and sent to the TVA Central Lab for metallurgical analysis. Records are being reviewed to identify previous non-conformance reports and certificates of compliances for the coupler. An extent of condition and extent of cause investigation will apply to vertical tendons are similar in design, these tendons do not utilize an anchorhead coupler in the design. However, these tendons will be considered in the analysis. Future updates: TVA plans to provide an update to this report by March 31, 2010 following the completion of the metallurgical analysis.|
|ENS 45523||27 November 2009 08:40:00||Davis Besse||NRC Region 3||B&W-R-LP||On November 27, 2009, at approximately 0734 hours, an accidental discharge of a security officer's sidearm occurred at the Davis-Besse Nuclear Power Station (DBNPS) resulting in a leg injury to the officer. The accidental discharge occurred in the Primary Access Facility. The onsite DBNPS first aid team and the Caroll Township Emergency Medical Services responded to the Primary Access Facility. The injured officer was in stable condition and transported off site (at 0800) to St. Vincent Hospital (in Toledo, OH). Both the Ottawa County Sheriffs Office and the Caroll Township Police Department responded to the DBNPS site to obtain information. The Ottawa County Emergency Management Agency (EMA), the Lucas County EMA, and the Ohio EMA were notified of this incident. The NRC Resident Inspector has been notified. Plant operation was not affected by this incident. (The plant has received calls from the media however,) no press release is planned at this time.|
|ENS 45162||25 June 2009 11:44:00||Davis Besse||NRC Region 3||B&W-R-LP||A transitory ALERT (condition was determined to have existed) based on Emergency Action Level 7.D.2 - 'Onsite Explosion Affecting Plant Operation'. At time 0049 on 06/25/09 a catastrophic failure-explosion of the Constant Current Potential Device (CCPD) on 'J' Bus near Air Circuit Breaker (ACB) 34563 resulted in a loss of switchyard 345 KV Bus 'J'. This event de-energized Startup Transformer 01 which is a tie from offsite sources to the Unit 13.8 KV Busses. The unit entered (Technical Specification) LCO 3.8.1, Condition A (due to the loss of one offsite power source). The Unit remains stable and in operation at 100% RTP (reactor thermal power). A problem solving decision making team is working on (the) troubleshooting/repair/restoration activities. This event did not impact any plant safety systems or result in any release of radioactive material. Failure of the CCPD caused automatic opening of the breakers on both sides of the 'J' bus which was configured as part of a switchyard ring bus at the time of the event. This resulted in the loss of one of the offsite power ties. However, startup Transformer 02 is still energized from offsite power and remains available for plant operations. Other than the de-energized startup transformer, onsite electrical configurations are normal including availability of emergency diesel generators. The licensee is in a 72 hour LCO per Tech Spec 3.8.1, Condition A, to restore the lost offsite power source. The licensee is inspecting the switchyard for collateral damage to other equipment from the failure of the CCPD. The licensee believes the CCPD failure is likely a result of equipment failure and not the result of any equipment tampering. The licensee stated that initially, the severity of the CCPD failure was not recognized because of the night time conditions and minimal lighting in the area. After daylight examination of the location of the event, it was determined that the failure of the CCPD should have been classified as an explosion affecting plant operation under EAL 7.D.2. Consequently, the licensee made the after-the-fact declaration. Licensee has notified the NRC Resident Inspector, and will be notifying State and local authorities.|
|ENS 45066||14 May 2009 11:45:00||Bellefonte||NRC Region 2||B&W-R-LP||DESCRIPTION OF DEFICIENCY Configuration control was not maintained and physical equipment issues were not documented under a Quality Assurance Plan for the period of time from in which Construction Permits CPPR-122 and CPPR-123 were withdrawn until they were reinstated. SAFETY IMPLICATIONS There are no safety implications because controls are in place under TVA's corrective action program that will prohibit current plant structures, systems or components, including those affected in the course of resource recovery activities, from being placed into service without being further evaluated and having been fully restored or replaced. This deficiency has been entered into TVA's Corrective Action Program as Problem Evaluation Report 170988.|
|ENS 45032||30 April 2009 12:50:00||Davis Besse||NRC Region 3||B&W-R-LP||At 0855 six sirens had been activated by the Ottawa County dispatch console for 3 minutes. At 0905 Ottawa County Sheriff Dispatch Center notified (the licensee) that sirens in Ottawa County had been inadvertently activated. It appears this was caused by the county radio service vendor resetting the dispatch center consoles during trouble shooting of the sheriff's radio system. Immediate actions taken: RA-EP-00420, Response to Prompt Notification System Malfunction, was implemented. The siren system was polled and the data from the Emergency Operations Facility (EOF) siren computer was reviewed. The computer data indicated at 08:55:55 the six sirens located in Bay Township had been activated by the Ottawa County Dispatch Console for 3-minutes. Fleet siren maintenance was contacted and requested to come to Ottawa County to meet with the radio service vendor to determine the cause of the inadvertent activation. In addition, NOP-LP-5001, Communicating Events of Public Interest, was implemented and associated notifications were made. These notifications included the State of Ohio, Ottawa County, and Lucas County. The NRC Resident Inspector was notified by the licensee.|
|ENS 44596||23 October 2008 11:59:00||Davis Besse||NRC Region 3||B&W-R-LP||This is a report of a situation, related to the protection of the environment, for which a notification to other government agencies is being made, as described in 10 CFR 50.72(b)(2)(xi). On October 22, 2008, excavation within the Protected Area was ongoing to identify a leak in the Fire Protection System. At approximately 1600 hours, the Turbine and Water Treatment Building sump discharge line was identified as leaking. This three-inch carbon steel pipe routes the sump discharge to the settling basins, where it is eventually discharged via a monitored outfall to the environment. The cause of the leakage is unknown, but due to the condition of the piping, it is believed to have existed prior to the excavation activities. The amount of leakage is therefore conservatively assumed to be more than 100 gallons, but this cannot be quantified at this time. Analysis of a water sample from the sump discharge line leak determined that the water leaking from the pipe contains approximately 37,500 picocuries per liter (pCi/l) tritium. These tritium levels are consistent with tritium levels in the Condensate/Feedwater Systems in the Turbine Building. Actions are underway to remove the piping from service and re-route the sump pump discharge. Analysis of routine groundwater well samples taken earlier this month is being expedited, and additional well samples are being planned. The State of Ohio, Lucas County and Ottawa County government agencies were contacted regarding the above information at 0900 on October 23, 2008. The Resident Inspector has also been briefed on the issue.|
|ENS 43880||4 January 2008 08:05:00||Davis Besse||NRC Region 3||B&W-R-LP||While in Mode 6 with the Reactor Head in place and fuel in the Reactor Vessel, weld overlay was in progress on the common Decay Heat Suction line from the Reactor Coolant System. The Reactor Coolant System is in a drained condition with RCS level approximately 24 inches above the Hot Leg Centerline. During the first weld pass on the suction line weld located in Containment, the welding operator noticed water seeping from the weld. Visual inspection revealed a small leak. The leakage is too small to quantify the leakrate. Surface examinations prior to welding revealed no abnormal conditions or leakage. Welding operations on this weld have been suspended. Planned Refueling activities continue to defuel the reactor vessel. One train of Decay Heat Removal remains in service providing core cooling. The second train is aligned in standby for Decay Heat Removal. Both Trains of Decay Heat Removal have been declared inoperable due to this leak, however both trains of Decay Heat Removal remain functional. The licensee notified the NRC Resident Inspector.|