Emergency Notification System
|Region 1||28||[Region 1 Events]||[RSS]|
|Region 2||19||[Region 2 Events]||[RSS]|
|Region 3||26||[Region 3 Events]||[RSS]|
|Region 4||22||[Region 4 Events]||[RSS]|
by Reactor type
|Entered date||Site||Region||Reactor type||System||Scram||Event description|
|ENS 50130||25 February 2020 10:40:00||Nine Mile Point||NRC Region 1||GE-5||Reactor Building Ventilation||At 0210 (EDT) on May 22, 2014, Nine Mile Point Unit 2, the reactor building vent radiation monitor (Vent WRGMS) was removed from service due to a problem with the check source.
The unplanned isolation of Vent WRGMS is a 8-hour report for 10 CFR 50.72(b)(3)(xiii), any event that results in a major loss of emergency assessment capability. Until the equipment is restored, Chemistry will perform sampling requirements per the ODCM. The NRC Resident Inspector has been notified.The licensee notified the State of New York Public Service Commission.
|ENS 53005||25 February 2020 10:31:00||Millstone||NRC Region 1||Westinghouse PWR 4-Loop||While troubleshooting a hydrogen differential pressure gauge, operators found a hydrogen concentration greater than the lower explosive limit (4% hydrogen concentration). The licensee evacuated the turbine building. The licensee is ventilating the area to allow personnel access to isolate and repair the leak. The plant Fire Brigade is on standby.
This event was declared for GU-2: release of toxic or flammable gas. The main generator pressure and hydrogen concentration are stable. The licensee notified the NRC Resident Inspector. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
The unusual event was terminated at 1617 EDT because the leak was isolated and the hydrogen concentration in the Hydrogen Control Cabinet was zero percent. The licensee notified the NRC Resident Inspector.Notified the R1DO (Jackson), IRD MOC (Stapleton), NRR (King), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
|ENS 53304||25 February 2020 10:00:00||Grand Gulf||NRC Region 4||GE-6||Secondary containment|
Standby Gas Treatment System
|At 0206 (CDT) on March 31, 2018, with the plant in Mode 1 at 100% rated core thermal power, Grand Gulf Nuclear Station experienced a loss of Secondary Containment. During the performance of a Standby Gas Treatment System (SGTS) drawn down test with Auxiliary Building train bay door (1A319A) as the secondary containment boundary, Grand Gulf was unable to maintain secondary containment pressure, as required by SR (surveillance requirement) 184.108.40.206.4, greater than or equal to 0.266 inches of water vacuum for 1 hour. Following initial vacuum draw down, secondary containment pressure degraded to 0.225 inches of water vacuum with operators in the field reporting air leakage from door 1A319A. The test was secured and Secondary Containment was declared inoperable and Technical Specification 220.127.116.11 A.1 was entered. Following completion of the failed surveillance test, Secondary Containment was returned to an operable status at 0315 hours on March 31, 2018, by returning the system to a previously known operable configuration by closing doors 1A310, 1A312 and 1A319. This is being report under 10 CFR 50.72(b)(3)(v)(C). The licensee has notified the NRC Resident Inspector.|
|ENS 53341||25 February 2020 10:00:00||Limerick||NRC Region 1||GE-4||High Pressure Coolant Injection||Unit 1 HPCI (High Pressure Coolant Injection) was declared inoperable due to a Main Pump seal leak that was identified during surveillance testing. Unit 1 HPCI was declared inoperable at 1030 EDT. HPCI was secured and was manually re-aligned to an available status.
At the time of this notification, repairs have been completed and the licensee is making preparations to re-perform the surveillance.The licensee has notified the NRC Resident Inspector.
|ENS 52645||25 February 2020 10:00:00||FitzPatrick||NRC Region 1||GE-4||Service water|
Emergency Diesel Generator
Primary Containment Isolation System
Reactor Core Isolation Cooling
Residual Heat Removal
|The following report is made pursuant to 10 CFR 50.73(a)(2)(iv)(A) due to an unintended initiation signal that occurred on January 31, 2017 with James A. FitzPatrick Nuclear Power Plant (JAF) in Mode 5 at zero (0) percent power.
On January 31, 2017 at 1425 (EST) the control room received multiple annunciations associated with the following Systems / Trains: Primary Containment Isolation System (PCIS) / Trains A and B Residual Heat Removal System (RHR) / Trains A and B Core Spray (CS) / Trains A and B Reactor Core Isolation Cooling (RCIC) All four (4) Emergency Diesel Generators (EDG) auto-started with their associated Emergency Service Water pumps operating. RHR and CS both received initiation signals but were defeated per procedure. The HPCI (High Pressure Coolant Injection) auxiliary oil pump was taken to Pull-to-Lock per procedure, and the RCIC steam isolation valve cycled until the breaker was opened to close the valve. An evaluation concluded that the (Emergency Core Cooling System - ECCS) initiation signals were caused by the opening of a portable job box that was stored near sensitive equipment. Upon opening the job box, the lid bumped a reference leg resulting in the initiation signals. All initiation signals were reset and systems restored to normal shutdown lineups. The licensee notified the NRC Resident Inspector.
To the original report, the licensee added, "This condition recurred at 1624 (EDT on 1/31/17).The licensee notified the NRC Resident Inspector. Notified R1DO (Cook).
|ENS 50542||25 February 2020 09:51:00||Nine Mile Point||NRC Region 1||GE-2||Secondary containment||Nine Mile Point Unit 1 (NMP1) had a momentary loss of Secondary Containment due to both Reactor Building airlock doors being opened at the same time.
At 1042 EDT on 10/16/2014, both Reactor Building airlock doors at NMP1 were opened simultaneously for less than 5 seconds. This results in a momentary loss of Secondary Containment operability (TS 3.4.3). The doors were closed and operability was restored. Secondary Containment being inoperable is an 8 hour notification per 10 CFR 50.72(b)(3)(v)(C), 'any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.' The condition has been entered into the station's corrective action program and the NRC Senior Resident Inspector was notified.The licensee will be notifying the State of New York Public Service Commission.
|ENS 52542||25 February 2020 09:51:00||Dresden||NRC Region 3||GE-3||At 0851 CST on Wednesday, February 8th, 2017, the Dresden Nuclear Power Station (DNPS) Technical Support Center (TSC) Emergency Ventilation System was emergently declared inoperable due to a failure of the outside air damper to reposition. This resulted in the inability for the TSC ventilation to maintain the required air flow to support habitability during emergency conditions. Actions are being taken to repair damper to restore functionality of the TSC ventilation system. In the interim, station procedures provide guidance to relocate the TSC to an alternate facility. This event is being reported under 10 CFR 72(b)(3)(xiii), Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).
The NRC Resident Inspector has been notified.
At 0108 CST on February 10, 2017, Dresden TSC ventilation has been restored and is now functional. The NRC Resident Inspector has been notified.Notified R3DO (Kunowski).
|ENS 52523||25 February 2020 09:51:00||Saint Lucie||NRC Region 2||CE||At 1200 (EST) on January 31, 2017, during investigation of potential 1B2 Reactor Coolant Pump seal degradation, a through wall defect was identified on Class 1 piping servicing the Lower Seal Heat Exchanger, which is part of the Reactor Coolant Pressure Boundary (RCPB). The reactor is presently in Mode 3 with decay heat being removed by the atmospheric steam dump valves. The plant is being maneuvered to Mode 5 to affect appropriate repairs. The licensee has notified the NRC Resident Inspector.|
|ENS 50517||25 February 2020 09:51:00||Byron||NRC Region 3||Westinghouse PWR 4-Loop||In accordance with 10 CFR 50.72(b)(3)(ii)(A), this notification reports a degraded condition on Byron Station Unit 2 that occurred on October 7, 2014, when it was determined that the results of planned ultrasonic testing (UT) examinations performed on one CRDM penetration of the reactor vessel head would not meet the applicable acceptance criteria. Byron Station Unit 2 is presently in day eight of a refueling outage. The examinations were
being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1 to ensure the structural integrity of the reactor vessel head pressure boundary. The UT examinations continue for the remaining CRDM head penetrations. The repairs are currently being planned and will be completed prior to unit start-up.The NRC Resident Inspector has been notified of this condition.
|ENS 51862||25 February 2020 09:41:00||Pilgrim||NRC Region 1||GE-3||Emergency Diesel Generator||On April 12, 2016, with the reactor at 100 percent power and the mode switch in RUN, Pilgrim Nuclear Power Station entered an unplanned 24-hour Limiting Condition for Operation (LCO) action statement due to both emergency diesel generators (EDG) being inoperable (Technical Specification 3.5.F.1). At 0050 (EDT) this morning, with EDG B out of service for a planned LCO maintenance window, EDG A was declared inoperable due to a 130 drop per minute leak on a line to a jacket water pressure indicator.
Repairs to EDG A are underway at this time. The following plant equipment has been verified operable: both 345 Kv transmission lines; 23kV transmission line; Station Blackout EDG. This condition is reportable to the NRC Staff as an Event or Condition that Could Have prevented Fulfillment of a Safety Function (Mitigate the consequences of an accident) under 10 CFR 50.72(b)(3)(v)(D), and requires an 8-hour notification. The licensee has notified the NRC Senior Resident Inspector.The licensee will notify the Commonwealth of Massachusetts.
|ENS 51877||25 February 2020 09:41:00||Prairie Island||NRC Region 3||Westinghouse PWR 2-Loop||Missing fire barrier between Fire Area (FA) 59 and 85.
During a walk down of fire barriers for the NFPA 805 project, it was determined that the fire barrier between Fire Area 59 (Unit 1) and 85 (common) is not a rated barrier due to unsealed penetrations in the barrier. Evaluation FPEE 12-006 evaluated the acceptability of the barrier being unrated based on separation of safe shutdown equipment however a review of equipment credited for Appendix R safe shutdown identified that the redundant credited Appendix R equipment is on either side of the fire barrier which is not 3 hour rated. The conclusion of the FPEE is therefore no longer valid. Fire Hazard Analysis Drawings Do Not Match Boundary Description. The plant layout in F5 Appendix F, Rev. 28, Fire Hazard Analysis (FHA), does not agree with the boundary description in the FHA for the Unit 1 and 2 Containment Annulus fire areas, Fire Area (FA) 68 and 72. The layout should but does not show the fire area boundary between the annulus and adjacent fire areas, FA 60 and 75 on 735 (foot) and 61A on 755 (foot), as an Appendix R boundary. The annulus airlock doors are 3-hour fire rated and the airlock is constructed of concrete thick enough to qualify as a 3 hour fire barrier however, there are penetrations in the barrier that are not sealed with fire rated materials or inspected as required by the Fire Protection Program.Therefore, this is an unanalyzed condition reportable under 10 CFR 50.72(b)(3)(ii)(B). This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified.
|ENS 50468||25 February 2020 09:32:00||Wolf Creek||NRC Region 4||Westinghouse PWR 4-Loop||During a review of INPO Event Report 14-33, Direct Current Circuits Challenge Appendix R Fire Analysis, it was determined that portions of the control circuits for the Turbine Generator DC Emergency Lube Oil Pump and the Emergency DC Seal Oil Pump are not properly fused to prevent overload and possible secondary fires. The review found that a fire at the motor starter cabinet in the turbine building could cause specific smart hot shorts that could cause overheating of the control cable and result in secondary fires outside the turbine building. Based on this information, it was determined that this condition is unanalyzed and is potentially reportable per 10 CFR 50.72(b)(3)(ii)(B).
Hourly fire watch compensatory measures are in place in the affected areas of the Turbine Building. The presence of compensatory measures in addition to automatic fire detection and suppression in these fire areas ensures protection of the equipment.The licensee has notified the NRC Resident Inspector.
|ENS 51861||25 February 2020 09:31:00||Palo Verde||NRC Region 4||CE||Reactor Coolant System||The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.
At approximately 2335 (MST) on April 10, 2016, while performing planned routine visual examinations of Unit 1 components in the Containment Building, engineering personnel identified white residue on the piping instrument nozzle for the reactor coolant system (RCS) 2B cold leg resistance temperature detector, 1JRCET121Y. The white residue was dry and no active leakage was noted on the instrument nozzle. Preliminary chemical analysis for radionuclide and boron content of the white residue determined the existence of boron and the isotopic content was consistent with RCS fluid. At 0535 (MST) on April 11, 2016, it was determined the residue resulted from RCS pressure boundary leakage, based on results of the chemical analysis and additional examination of the piping and instrument nozzle by qualified engineering personnel. Technical Specifications Limiting Condition of Operation (LCO) 3.4.14 permits no RCS pressure boundary leakage and therefore, the discovery of leakage from the instrument nozzle represents a degradation of a principal safety barrier. This notification is being made for a degraded condition pursuant to the requirements of 10CFR 50.72(b)(3)(ii)(A). The unit has been shut down for its 19th refueling outage since 4/9/16 at 0000.The NRC resident inspectors have been informed of this condition.
|ENS 51353||25 February 2020 09:31:00||Duane Arnold||NRC Region 3||GE-4||Secondary containment||At 0752 CDT on 8/27/2015, both doors of a Secondary Containment Airlock were reported to be open simultaneously for a period of approximately five seconds. The brief time that the doors were simultaneously open constituted an inoperable condition of Secondary Containment. The airlock interlock was verified to operate correctly, and Secondary Containment has been restored to an operable status. The licensee has notified the NRC Resident Inspector.|
|ENS 48618||25 February 2020 09:05:00||Monticello||NRC Region 3||GE-3||Emergency Response Data System||At approximately 0730 CST, Xcel Energy Monticello Nuclear Generating Plant was informed of a loss of internet and data services due to equipment problems with the Wide Area Network (WAN). The loss of data services resulted in the loss of ERDS communication capabilities with the NRC. Compensatory actions were established to communicate plant parameters verbally via the Emergency Notification System, which was verified functional at 0836 CST.
The WAN and ERDS were restored at 0921 CST and ERDS was verified functional at 0948 CST. This issue is being reported as a loss of communications capability non-emergency event notification under 10CFR50.72(b)(3)(xiii) as defined in NUREG 1022, Rev. 2.The licensee has notified the NRC Resident Inspector.
|ENS 50795||25 February 2020 09:04:00||Grand Gulf||NRC Region 4||GE-6||Feedwater|
Emergency Diesel Generator
Main Steam Isolation Valve
|A reactor SCRAM occurred at 1856 CST on 2/7/15 from 100 percent core thermal power. The cause of the SCRAM appears to be a Generator/Turbine trip, but it is still under investigation.
Appropriate off-normal event procedures were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF power occurred. No ECCS initiation signals were reached, and no ECCS or Emergency Diesel Generator initiations occurred. Main Steam Isolation Valves remained open and Safety Relief Valves lifted and reseated as designed. Currently, reactor water level is being maintained by the Condensate and Feedwater system in normal band, and reactor pressure is being controlled via turbine bypass valves to the main condenser. Following the reactor SCRAM, all rods fully inserted and all systems functioned as expected. The plant is in a normal electrical lineup.The licensee has notified the NRC Resident Inspector.
|ENS 50716||25 February 2020 09:04:00||South Texas||NRC Region 4||Westinghouse PWR 4-Loop||At approximately 1545 CST on 1/5/15, an employee was found unconscious by fellow teammates. His teammates, who are certified in cardiopulmonary resuscitation (CPR), commenced resuscitation until Matagorda County EMS arrived on the scene and transported the individual to Matagorda County Regional Hospital for medical treatment. The individual was evaluated by the Emergency Room Physician and was later pronounced deceased due to Cardiopulmonary Arrest.
STP is investigating the incident. STP has reported this to the Occupational Safety and Health Administration (OSHA) and as such is reporting this in accordance with 10 CFR 50.72 (b)(2)(xi). A press release is not planned at this time.The NRC Resident Inspector was notified.
|ENS 51080||25 February 2020 09:04:00||Fermi||NRC Region 3||GE-4||At approximately 1130 EDT on May 19, 2015, Fermi-2 determined that a sanitary sewer system leak had occurred from underground lines beneath the parking lot near Warehouse E. Some of the sewage entered the storm sewer. The leak has been secured. A local sanitary contractor is currently responding to the site to clean the affected areas. The duration and quantity of the spill is unknown, however, current estimates of the sewage entering the storm sewer are less than 200 gallons. Reports to the Michigan Department of Environmental Quality (MDEQ), the local health department and the local news media are required in accordance with Michigan Act 451, Public Acts of 1994, as amended, Part 31, Section 342.3112a. Since reports are in the process of being made, this is considered a News Release or Notification of Other Government Agency, and therefore is reportable under 10 CFR 50.72(b )(2)(xi).
The NRC Resident Inspector has been notified.The release was onsite only and did not reach navigable waters.
|ENS 50917||25 February 2020 09:04:00||Oconee||NRC Region 2||B&W-L-LP||At approximately 0605 EDT on March 23, 2015, the Oconee Nuclear Station main control room and Security received an emergency call for an employee experiencing a non-work related medical issue. Site first responders were dispatched in conjunction with a request for off-site medical assistance. The individual was transported by ambulance to the Oconee Medical Center and was pronounced dead at 0717 EDT. The individual was outside of the protected area (within the owner controlled area) and no radioactive material or contamination was involved. The cause of death has not been determined.
This notification is being made in accordance with 10 CFR 50.72(b )(2)(xi) for situations related to the health of on-site personnel for which a notification to other government agencies has been made. The South Carolina Occupational Safety and Health Administration (SCOSHA) was notified at 0920 EDT. The NRC Resident Inspector has been notified.The licensee notified Pickens County, South Carolina and Oconee County Emergency Managements.
|ENS 50209||25 February 2020 09:04:00||Millstone||NRC Region 1||CE||At 0832 EDT on 6/18/2014, Millstone Station Unit 2 removed the Main Steam Line Radiation Monitors RM-4299B and RM-4299C from service for pre-planned maintenance. Portable handheld radiation detectors will be used to assess radiation releases. Both radiation monitors will be restored no later than 2200 EDT on 6/19/14. The licensee informed both State and local agencies and the NRC Resident Inspector.|
|ENS 51697||25 February 2020 09:04:00||Duane Arnold||NRC Region 3||GE-4||Emergency Diesel Generator||At 0920 CST on 1/29/16, while performing main generator voltage adjustments at the direction of the transmission operator, Duane Arnold Energy Center (DAEC) switchyard voltage lowered to the calculated point where offsite sources were considered inoperable. During this event, both onsite emergency diesel generators were fully operable and capable of performing their intended safety function.
While no safety concerns arose, Technical Specification (TS) Limiting Condition for Operability (LCO) 3.8.1.a Condition C was entered due to two offsite electric power circuits being inoperable. Immediate actions were taken to adjust main generator voltage to restore switchyard voltage. In less than 10 minutes, both offsite circuits were declared operable and LCO 3.8.1.a Condition C was exited. This resulted in a reportable event pursuant to 10CFR50.72(b)(3)(v)(D). During the time of reduced grid voltage, no bus low-voltage alarms nor any equipment issues occurred. The cause of this event is under investigation, and there are not any current operability concerns with the offsite power circuits. The licensee has notified the NRC Resident Inspector.
The purpose of this notification is to retract a previous report made on 1/29/16 at 1709 (EST). NRC notification was initially made as a result of offsite power sources being declared inoperable due to the identification of a low voltage condition by the grid operator. Subsequent to the initial report, NextEra Energy Duane Arnold (NextEra) has determined that the offsite power grid conditions were above the required limit and were therefore operable. It was determined that a trip of the DAEC turbine/generator would not have led to fluctuations in voltages in the DAEC switchyard below the trip setpoints for Loss of Power. Based on this, the safety-related buses and loads were fully capable of performing their intended design basis function. The degraded voltage relays setpoint were not reached; therefore, the essential electrical buses remained connected to offsite power. Therefore, the low voltage condition does not meet the reporting requirements of 10 CFR 50.72 or 10 CFR 50.73 The Resident Inspector has been notified.Notified R3DO (Kozak).
|ENS 51064||25 February 2020 09:04:00||Beaver Valley||NRC Region 1||Westinghouse PWR 3-Loop||At 1629 EDT on May 11, 2015, Beaver Valley Power Station (BVPS) Unit 1 (Mode 6 Refueling) opened circuit breaker PCB-92 to de-energize system service transformer 1A due to an emergent issue with a transformer bushing. This resulted in a loss of power to the ERF (Emergency Response Facility) substation which powers various plant support equipment at both units. The BVPS Unit 2 Digital Radiation Monitoring System (DRMS) was non-functional for approximately 21 minutes. This was a loss of radiation monitor capability to alarm and indicate in the control room. Following power restoration to the ERF substation, the radiation monitor system was restored at 1650 EDT. Both Unit 1 and Unit 2 remain stable.
Since the BVPS Unit 2 DRMS was non-functional, this event resulted in a loss of emergency assessment capability and is reportable per 10 CFR 50.72(b)(3)(xiii).The NRC Resident Inspector has been notified.
|ENS 50753||25 February 2020 08:52:00||Brunswick||NRC Region 2||GE-4||At approximately 0531 EST on January 23, 2015, the Brunswick Nuclear Plant main control room received an emergency call for a contract employee experiencing a non-work related medical issue. Site first responders were dispatched in conjunction with a request for off-site medical assistance. At approximately 0613 EST, the responding off-site paramedics determined that the efforts to revive the patient were unsuccessful. The individual was outside of the protected area (within the owner controlled area), and no radioactive material or contamination was involved. The cause of death has not been determined.
This notification is being made in accordance with 10 CFR 50.72(b)(2)(xi) for situation related to the health of on-site personnel for which a notification to other government agencies is planned. The Occupational Safety and Health Administration (OSHA) will be notified. A press release is not planned at this time.The NRC Resident Inspector has been notified.
|ENS 50235||25 February 2020 08:52:00||Diablo Canyon||NRC Region 4||Westinghouse PWR 4-Loop||At approximately 1535 Pacific Daylight Time (PDT) the Diablo Canyon Power Plant (DCPP) Emergency Planning Supervisor received a notification from an offsite DCPP employee that one of the emergency plan sirens had inadvertently actuated. The DCPP Shift Manager was notified of the situation by approximately 1545 PDT.
The County of San Luis Obispo was notified of the inadvertent actuation of the single siren. At approximately 1550 PDT the County of San Luis Obispo sent out a county wide alert stating, 'Civil Emergency in this area until (1610) PDT prepare for action'. At approximately 1600 PDT the County of San Luis Obispo sent out another county wide alert stating, 'An early warning system siren was sounded in error. There is no emergency.' Field technicians have been sent out to the siren location and will shut the siren off. The cause of the inadvertent actuation of the siren is not known at this time.The licensee notified the NRC Resident Inspector and the County of San Luis Obispo.
|ENS 50787||25 February 2020 08:52:00||Zion||NRC Region 3||Westinghouse PWR 4-Loop||On 2/3/2015 at 1901 (CST), equipment important to safety failed to function as designed: 33 NAC MAGNASTOR systems containing Spent Nuclear Fuel, and 1 NAC MAGNASTOR system containing GTCC (Greater than Class C) waste, had air inlet vent partial blockage, totaling 50 percent or greater, due to environmental conditions, i.e., snow. (Please Note, no VCC systems reached full blockage.)
NO personnel radiation exposure. Actions Taken lAW the NAC FSAR, Rev. 5, which included actions to immediately return the systems to less than 50 percent blockage and ongoing maintenance activities to return all systems to full operability. Additional information: 1. On 2/4/2015 at 0445 hrs. All 34 failed systems were restored to less than 50 percent blockage. 2. lAW the NAC FSAR we have 30 days from time of discovery to return each system to full operability. 3. A report will be followed up within 60 days per 10 CFR 72.75.g The licensee will notify the NRC Inspector and the State of Illinois.
This phone notification is being made to retract an earlier '24 Hour Report' made on February 4, 2015 at 1815 CST, NRC event #50787. On February 4, 2015 at 1810 CST, Zion Station notified NRC that on February 3, 2015 at 1901 CST, Equipment Important To Safety failed to function as designed that being 33 NAC MAGNASTOR systems containing Spent Nuclear Fuel and 1 NAC MAGNASTOR containing GTCC waste had their Inlet vents partially blocked, totaling 50% or greater, due to environmental conditions, i.e. Snow. This appeared to meet criteria of Non-Emergency notification under 10 CFR 72.75d(1)(i). NAC International has provided documentation stating we have 58 hours to perform the immediate action to unblock the VCC vents to greater than 50% and 30 days to completely unblock all the vents. This documentation is attached to this notification. Upon further review by Exelon Management, Zion Solutions Management and NAC, using criteria established in Sections 18.104.22.168 and 12.2.2 of the SER for CoC 72-1031, Amendment 0, no limits were exceeded. Actions were immediately initiated to restore the 34 failed systems to less than 50% blockage and were completed in less than 10 Hours. This is well within the 58 hour SAR basis. This is the basis for retracting the notification. Based on the above, (the licensee) management concludes this event does not meet the 24-hour reporting criteria established in 10 CFR 72.75d(1)(i).The R3DO(Pelke) was notified.
|ENS 51033||25 February 2020 08:52:00||Palisades||NRC Region 3||CE||At 1241 EDT, Operations staff at Palisades declared an Unusual Event under EAL HU1.1 due to seismic activity felt on site. No seismic alarms were initiated. No plant equipment was affected. The epicenter of the 4.2 magnitude earthquake was located south of Galesburg, MI. Palisades continues to operate at 100% power.
The licensee has notified the NRC Resident Inspector.
The licensee terminated the Unusual Event at 1541 EDT on 5/2/15. The licensee has notified the NRC Resident Inspector and the state and local government. Notified R3DO (Orlikowski), IRD MOC (Stapleton), NRR EO (Morris), NRR ET (Dean), and R3RA (Pederson).Notified other Federal Agencies (DHS SWO, FEMA Ops, FEMA NWC, NICC Watch Officer and NuclearSSA).
|ENS 50446||25 February 2020 08:52:00||Prairie Island||NRC Region 3||Westinghouse PWR 2-Loop||Shield Building||At 1009 CDT on September 11, 2014, 1R-22 Shield Building Vent Gas Radiation Monitor was removed from service for planned maintenance. This monitor has no compensatory measure that will allow timely classification of two Emergency Action Levels (EALs) - NUE (Notification of Unusual Event) and Alert classifications when out of service. It is also used for offsite dose projection calculations. This results in a Loss of Emergency Assessment Capability while 1R-22 is out of service. This is a reportable condition in accordance with 10 CFR 50.72(b)(3)(xiii).
Unit 1 Shield Building Ventilation Stack is also monitored by high range monitor, 1R-50, which is used for the same purpose in Site Area or General Emergency classifications. 1R-50 is being monitored and is indicating normal values. There are no radioactive leaks that will impact the Shield Building as evidenced by normal readings on 1R-22 prior to its removal from service. The duration of this maintenance is scheduled for 8 hours and will continue until the monitor is returned to service. Maintenance will not result in the unplanned release of radioactivity to the environment and will not affect the safe operation of the plant or health and safety of the public.The licensee has notified the NRC Resident Inspector.
|ENS 51695||25 February 2020 08:52:00||Three Mile Island||NRC Region 1||B&W-L-LP||Steam Generator|
|On December 1, 2015, at 2119 EST, with Unit 1 in power operation mode, during a planned maintenance activity, an invalid Heat Sink Protection System (HSPS) actuation occurred. At the time of the event, electrical maintenance technicians were verifying a HSPS relay contact state using an electrical test meter. The contact was being verified open by recording both voltage and resistance readings across the contact. The technicians first measured voltage. No voltage was found, indicating the relay contact was open, as expected. The technicians then measured for resistance across the open contact. Test meters have lower circuit impedance when measuring resistance as opposed to voltage, which can result in electrically bridging across open contacts. When the meter was installed across the open contact to obtain the resistance reading, the HSPS actuation circuit logic was completed and the inadvertent HSPS actuation occurred. The HSPS actuation resulted in the steam driven Emergency Feedwater (EFW) pump automatically starting and control valves receiving actuation set point signals. There was no emergency feed water injection into the steam generators. At the time of the inadvertent HSPS actuation, steam generator operating levels were above the HSPS actuation setpoint.
The specific train and system that actuated was the Heat Sink Protection System, Emergency Feedwater System Actuation on Loss of All Reactor Coolant Pumps (RCP) Train A. The HSPS Loss of all RCP Train A actuation was complete. The EFW valves and EFW steam driven pump started and functioned successfully. This is reported under 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of HSPS Loss of all RCP Train A and in accordance with 10 CFR 50.73(a)(1), this notification of the invalid actuation is provided in lieu of a written LER.The Licensee notified the NRC Resident Inspector.
|ENS 50771||25 February 2020 08:52:00||Pilgrim||NRC Region 1||GE-3||High Pressure Coolant Injection|
Emergency Diesel Generator
Reactor Core Isolation Cooling
|Automatic Scram||On Tuesday, January 27, 2015, at 0948 EST, with the Reactor Mode Select Switch (RMSS) in the Shutdown position and Pilgrim Nuclear Power Station (PNPS) at 0% core thermal power, the High Pressure Coolant Injection (HPCI) system was isolated by the main control room operating crew and declared INOPERABLE. HPCI had been in service for reactor pressure control following the automatic reactor scram experienced during winter storm 'Juno' reported in EN# 50769. It appears there was a malfunction of the HPCI turbine gland seal condenser blower or associated condensate pump. Reactor pressure control was transitioned to the safety relief valves and the reactor cooldown was continued. The plant is stable. The Emergency Diesel Generators are powering the safety related 4KV buses and reactor water level is being maintained by the Reactor Core Isolation Cooling (RCIC) system. HPCI is required to be OPERABLE in accordance with Technical Specification 3.5.C.1. Since HPCI is a single train system, the INOPERABILITY is reportable in accordance with 10CFR50.72(b)(3)(v)(D). The cause of the HPCI malfunction is not known at this time and troubleshooting continues.
This event had no impact on the health and/or safety of the public. The USNRC Senior Resident Inspector has been notified.Shutdown cooling is in service.
|ENS 51340||25 February 2020 08:52:00||Arkansas Nuclear||NRC Region 4||CE|
|This notification is being made in accordance with 10 CFR 50 72(b)(3)(xiii) as an event that will result in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g. a significant portion of control room indication, Emergency Notification System or offsite notification system.)
The emergency preparedness plan requires seismic monitoring instruments to diagnose an earthquake for emergency actions levels (EAL) HU6 (Natural or destructive phenomena affecting protected area) and HA6 (Natural or destructive phenomena affecting vital areas). At 1020 CDT on August 24, 2015 the Semi-Annual Seismic System Functional Test commenced. While this test is in progress, seismic alarm capability is not available for EAL declaration purposes. ANO procedures provide compensatory measures of using offsite sources to obtain seismic data. It should be noted that seismic data will still remain capable of being recorded, only alarm capability is lost.The licensee notified the NRC Resident Inspector.
|ENS 49599||25 February 2020 08:52:00||Byron||NRC Region 3||Westinghouse PWR 4-Loop||On December 2, 2013, activities are scheduled that will render the Technical Support Center (TSC) non-functional by removing the normal and emergency ventilation system from service. These activities are being performed in support of planned preventive maintenance. In preparation for these normal and emergency ventilation system outages, the TSC emergency responders were notified that if an emergency occurred during this outage the Emergency Coordinator and the TSC staff involved with classification, notification and PARS should report to the Work Execution Center. All other TSC personnel should report to the Operational Support Center. The duration of this TSC outage is expected to be less than 36 hours. The NRC Operations Center will be provided an update to this notification when the TSC normal and emergency ventilation is restored. This 8 hour notification in accordance with 10 CFR 50.72(b)(3)(xiii).
The licensee notified the NRC Resident Inspector.
The TSC is fully functional. The licensee will notify the NRC Resident Inspector.Notified R3DO (Riemer).
|ENS 50045||25 February 2020 08:52:00||Millstone||NRC Region 1||CE||Main steam line radiation monitors RM-4299A, RM4299B, RM4299C will be removed from service for a period of approximately 4 days during planned maintenance on the power supplies. This constitutes a loss of assessment capability. The radiation monitors will be returned to service following maintenance. The licensee has notified the NRC Resident Inspector, the State of Connecticut, and the local government.|
|ENS 50252||25 February 2020 08:46:00||Davis Besse||NRC Region 3||B&W-R-LP||Safety Parameter Display System||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 50165||25 February 2020 08:46:00||Millstone||NRC Region 1||CE||At 0933 EDT, Millstone Station Unit 2 removed the Stack High Range Radiation Monitor RM-8168 from service for planned maintenance. Maintenance and testing were completed and the Stack Radiation Monitor returned to service at 1252 EDT. The licensee informed both State and local agencies and the NRC Resident Inspector.|
|ENS 49369||25 February 2020 08:26:00||Davis Besse||NRC Region 3||B&W-R-LP||Shield Building||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 49422||25 February 2020 08:26:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits in the control room has determined that the condition described below to be applicable to Callaway Nuclear Plant resulting in an unanalyzed condition with respect to 10 CFR 50 Appendix R analysis requirements. The original plant wiring design and associated analysis for the Class 1E Train B batteries and chargers (including the B Swing charger) control room ampere indications do not include overcurrent protection features to limit the fault current.
In the postulated event, a fire in the control room could cause one of the ammeter wires to hot short to the ground plane; simultaneously, the fire causes another DC wire from the opposite polarity on the same battery to also hot short to the ground plane. This would cause a ground loop through the unprotected ammeter wiring. This event could result in excessive current flow (heating) in the ammeter wiring to the point of causing a secondary fire in the raceway system. The secondary fire could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10CFR50 Appendix R. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The NRC Resident Inspector has been notified.Similar Events: EN #49411 and EN #49419
|ENS 49178||25 February 2020 08:26:00||Saint Lucie||NRC Region 2||CE||On July 8, 2013, at 0830 EDT, the National Response Center and the State Watch office were notified of a leak greater than the limit of 109 gal of 11% Sodium Hypochlorite onto the ground. Total amount spilled is approximately 200 gallons. While the leak is currently active, the leak has being routed to an approved injection point.
This notification is being made in accordance with 10CFR 50.72 (b)(2)(xi) due to notification of off-site agencies.The NRC Resident Inspector has been notified.
|ENS 49495||25 February 2020 08:25:00||Prairie Island||NRC Region 3||Westinghouse PWR 2-Loop||Main Steam Isolation Valve||During Unit 2 refueling outage (currently defueled) preventative maintenance on CV-31117, Loop B Main Steam Isolation Valve (MSIV), an opening in the valve was discovered without direct administrative controls to ensure the opening could be closed within 6 minutes following a Loss of Coolant Accident on Unit 1. Addition of this opening to other openings created a total of greater than 10 square feet of non-closable openings and required declaring the Auxiliary Building Special Ventilation System (ABSVS) boundary inoperable. The inoperable ABSVS boundary caused both trains of ABSVS to be declared inoperable and required entry into Technical Specification (TS) 3.7.12, Condition B. This could have prevented the ability to control the release of radioactive material and is considered a potential loss of safety function per 10CFR50.72(b)(3)(v)(C).
Administrative controls in the ABSVS Boundary were re-established by installing a closure device in the opening on CV-31117 at 1246 CDT. The ABSVS boundary and both trains of ABSVS were declared operable at that time. During the time that the ventilation system was out of service, no evolutions were in progress that could have resulted in an unmonitored release.
After further evaluation, the non-closable boundary openings for the Auxiliary Building Special Ventilation Zone (ABSVZ) including CV-31117, Loop B Main Steam Isolation Valve (MSIV), at the time of discovery is calculated to be 9.48 square feet. This is below the 10 square feet required in TS 3.7.12, thus both trains of ABSVS were operable and no loss of safety function existed. The licensee has notified the NRC Resident Inspector.Notified the R3DO (Cameron).
|ENS 49414||25 February 2020 08:25:00||Oyster Creek||NRC Region 1||GE-2||Reactor Protection System|
|Manual Scram||Today at approximately 1040 (EDT), during a planned reactor power ascension with reactor power at approximately 20% of rated thermal power, main condenser vacuum began to lower. In accordance with the abnormal operating procedure for degrading vacuum, Operators inserted a manual scram of the reactor at 1130 (EDT). The cause of the degraded vacuum is currently under investigation.
All rods inserted into the core and all systems functioned as expected during the scram. No electromatic (EMRVs) or safety relief valves lifted during the transient. The plant is currently shutdown and parameters are stable. The plant is in its normal shutdown electrical lineup and decay heat is being removed via steam bypass valves to the main condenser. This event is reportable within 4 hours per 10CFR50.72(b)(2)(iv)(B) - any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.The licensee has notified the NRC Resident Inspector and will be notifying state authorities.
|ENS 49465||25 February 2020 08:25:00||Susquehanna||NRC Region 1||GE-4||Secondary containment||On October 23, 2013 at 0620, Susquehanna Steam Electric Station operators observed secondary containment differential pressure was at negative 0.17 inches water gauge for Zone II (Unit 2 Reactor Building). Tech Spec Secondary Containment Operability requires a negative pressure of at least 0.25 inches water gauge. Zone I (Unit 1 Reactor Building) and III (Common Refuel Floor Area) ventilation remained in service and stable.
Zone II differential pressure was restored to within the required band by manual damper adjustment in about 15 minutes and was verified to be stable. LCO 22.214.171.124 was entered for both units at 0620 and exited at 0635. This event is being reported under 10 CFR 50.72(b)(3)(v) and per the guidance of NUREG 1022 Rev 3 section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment System.The licensee has notified the NRC Resident Inspector.
|ENS 49509||25 February 2020 08:25:00||Millstone||NRC Region 1||CE||The Millstone Station stack radiation monitor, RM-8169, was removed from service for preplanned maintenance at 1028 (EST) and returned to service at 1513 (EST).
The Unit 2 high range stack radiation monitor, RM-8168, was removed from service for preplanned maintenance at 1543 (EST) and will be removed from service for approximately 3 days.The licensee has notified the NRC Resident Inspector and applicable state and local authorities.
|ENS 49492||25 February 2020 08:25:00||Susquehanna||NRC Region 1||GE-4||Secondary containment|
Reactor Protection System
|On November 1, 2013 at 0309 EDT, Secondary Containment Zone I (Unit 1 Reactor Building) differential pressure was lost following a routine transfer of Reactor Protection System Power supplies. Upon restoration from the power supply transfer, one of the Reactor Building Exhaust Fans tripped. There were no obvious malfunctions associated with the equipment and fan was able to be restarted. Zone II (Unit 2 Reactor Building) and III (Common Refuel Floor Area) ventilation remained in service and stable.
Zone I differential pressure recovered within a few minutes and was verified to be stable. LCO 126.96.36.199 was entered for both units at 0309 EDT and exited at 0315 EDT. Tech Spec Secondary Containment Operability requires a negative pressure of at least 0.25 inches water gauge. There have been no further perturbations in differential pressure and secondary containment remains operable. This event is being reported under 10 CFR 50.72(b)(3)(v) and per the guidance of NUREG 1022 Rev 3 section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment System.The licensee notified the NRC Resident Inspector. See similar event number #49489.
|ENS 49545||25 February 2020 08:25:00||Columbia||NRC Region 4||GE-5||Primary containment||This notification is being made due to a loss of emergency assessment capability. At 0955 (PST) on 11/17/2013,
it was discovered that the CMS-SR-13 sample valves would not align to allow monitoring the drywell. The redundant sample rack, CMS-SR-14, is already out of service due to an issue with the sample pump, which is currently awaiting replacement parts. The capability to monitor the drywell for H2 and O2 remotely is non-existent if an emergency situation were to arise because of these two issues. The only method currently available to sample the drywall for H2 and O2 is manually via chemistry procedures. This method would be unavailable in accident conditions because the sample point isolation valves isolate on an accident signal (F or A). Even if the sample points were not isolated, the Chemist would be unable to draw a sample safely during an accident due to radiation levels in the Reactor Building and/or H2 and O2 levels in containment. PPM 13.1.1 Table 6 specifies DW/WW (Drywell/Wetwell) hydrogen and oxygen measurements as a method for determining potential loss of Primary Containment. PPM 13.14.11 does not specify an alternative method for the loss of CMS-SR-13 & 14. Based on this information and discussions with Licensing and Emergency Preparedness, the station has determined this issue is a loss of emergency assessment capability per 10 CFR 50.72(b)(3)(xiii).
The NRC Resident Inspector has been notified.
At 1351 PST on 11/18/13, repairs on CMS-SR-13 were completed, returning the monitor to operable status of being able to sample both the drywell and wetwell for hydrogen and oxygen as required for emergency assessment capability. The NRC Resident Inspector has been notified.Notified R4DO (PROULX).
|ENS 49029||25 February 2020 08:24:00||Nine Mile Point||NRC Region 1||GE-2||Emergency Cooling Loop 11 High Steam Flow Isolation capability not functional due to inadequate water level in differential pressure transmitter reference legs.
Emergency Cooling Loop 11 High Steam Flow transmitter for channel 11 alarmed due to gross failure at 0840 hours (EDT) on May 14, 2013. At 1154 hours channel 12 Emergency Cooling Loop 11 High Steam Flow transmitter experienced a gross failure. With both transmitters failed, the isolation capability of Emergency Condenser Loop 11 on high steam flow is not available. At 1215 hours, a manual isolation of Emergency Cooling Loop 11 was initiated fulfilling the safety function. Subsequent troubleshooting has revealed that the reference legs of the differential pressure transmitters for high steam flow were not filled properly. This is an 8 hour notification per 10 CFR 50.72(b)(3)(v) for a condition could have prevented the mitigation of the consequences of an accident.The licensee notified the NRC Resident Inspector and will notify the State of New York.
|ENS 49351||25 February 2020 08:24:00||Susquehanna||NRC Region 1||GE-4||At 1455 EDT, (on 91/16/13), (Susquehanna) Engineering determined the leakage from the 2B RHR Pump Suction Relief valve caused the Engineered Safety Feature (ESF) Leakage to exceed the 2.5 gpm which was provided to the NRC during the implementation of the Alternate Source Term (AST) submittal. The calculated leakage rate was 7.5 gpm.
This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii). Unit 2 is currently in Mode 4 (cold shutdown) for a maintenance outage. This leaking RHR pump suction relief valve, previously identified in EN #49344, is being evaluated and repaired.The licensee has notified the NRC Resident Inspector and the state.
|ENS 49048||25 February 2020 08:12:00||Monticello||NRC Region 3||GE-3||This is a non-emergency 8-hour notification for a planned loss of emergency assessment capability. This event is reportable in accordance with 10CFR50.72(b)(3)(xiii) because the work activities affects the functionality of the Plant Process Computer System. Monticello Nuclear Generating Plant will remove the Plant Process Computer System (PPCS) from service on 5/17/13 at 1527 (CDT) to perform system upgrades and planned maintenance. The PPCS system is planned to be non-functional for less than 4 hours. While the system is out of service, the Emergency Plan can still be implemented as assessment capabilities are available under alternate means and communication of the assessment results using communication equipment. ERDS will be out of service during this period. Compensatory measures for the loss will be implemented. The NRC Resident Inspector has been notified.|
|ENS 48910||25 February 2020 08:12:00||Summer||NRC Region 2||W-AP1000||50.55(e) initial notification for failure to comply with requirements of 10 CFR 50 Appendix B, Criterion VII for procurement of safety-related components associated with AP1000 Nuclear Power Plant construction by CB&I (formerly Shaw Nuclear).
This 50.55(e) initial notification addresses a failure to comply by CB&I, an agent of South Carolina Electric & Gas (SCE&G) for Virgil C. Summer 2 & 3, to meet the requirements of Appendix B, Criterion VII. It is concluded that the QA programmatic issues, as identified by the root cause analysis associated with NRC violation 05200025/2012-004-02, could have produced a defect and this condition is reportable in accordance with 10 CFR 50.55(e)(3)(iii)(C). The root cause of the programmatic procurement problems was that the existing Shaw Nuclear procurement and quality oversight and inspection program did not include a sufficiently strategic, integrated, and graded approach to assure the required quality of material, equipment, and services. This notification closes the interim report submitted on February 4, 2013 by SCE&G. This 50.55(e) initial notification is being submitted pursuant to the requirements of 10 CFR 50.55(e)(3)(iii)(C).The licensee has notified the NRC Resident Inspector.
|ENS 49084||25 February 2020 08:12:00||Nine Mile Point||NRC Region 1||GE-5||Reactor Core Isolation Cooling|
High Pressure Core Spray
Residual Heat Removal
|This 60-day telephone notification is being made per the reporting requirements specified in 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to describe an invalid actuation signal affecting containment isolation valves in more than one system.
On April 2, 2013, Nine Mile Point 2 (NMP2) received a Division 2 reactor building area high ambient temperature isolation signal when lifting a lead for trip unit E31-N638B while performing surveillance N2-IPS-LDS-Q010, Reactor Building General Area Temperature Instrumentation Channel Functional Test. The isolation signal provided a closure signal to two Reactor Core Isolation Cooling System (RCIC) valves, and three Residual Heat Removal (RHR) system containment isolation valves. As a result of the isolation signal one of the RCIC containment isolation valves, 2ICS*MOV128 closed. The other four valves were already in their normal closed position. The RHR system valves are associated with the RHR Shutdown Cooling System and second RCIC isolation valve is used to warmup and place the RCIC system in standby following an isolated condition. All affected isolation valves responded as designed. As a result of 2ICS*MOV128 closing the RCIC system was declared inoperable. Technical Specification 3.5.3, RCIC System, Condition A was entered. Action A.1 required verifying the High Pressure Core Spray System (HPCS) was operable immediately. Action A.2 requires restoring RCIC to operable within 14 days. After the instrumentation system was restored to normal, the RCIC system was subsequently restored to available later that day at 1205 (EDT) and operable at 1500 (EDT). The actuation signal was not valid because it resulted from maintenance activities when leads were lifted, and the trip unit had not been bypassed as required by the procedure. There were no isolation logic signals in response to actual plant conditions or parameters. This event was entered into the corrective action system as Condition Report (CR) 2013-002461. There were no actual safety consequences or impact on the health and safety of the public as a result of this event.The licensee notified the NRC Resident Inspector and the State.
|ENS 49055||25 February 2020 08:12:00||Hatch||NRC Region 2||GE-4||On March 24, 2013, at 1009 EDT, while personnel were entering the torus compartment to perform planned maintenance activities via permanently installed plant ladder, the sensing line to transmitter 2E41-N062D was inadvertently bumped. (This) was confirmed to be the cause for an invalid torus high water level alarm and a HPCI (High Pressure Core Injection) pump suction swap. This resulted in the HPCI suction swapping from its normal lineup, condensate storage tank (CST), to the torus as designed. Once actuated the suction swap occurred as designed. The cause was attributed to the close proximity between an individual descending/ascending the fixed ladder and the affected sensing line in conjunction with a loose tubing restraint which made the line more sensitive to being bumped. After confirming that the actuation on high torus level was invalid, HPCI suction was realigned to the CST.
The HPCI pump suction was subsequently realigned to the CST and the loose tubing restraints were tightened.The licensee notified the NRC Resident Inspector.
|ENS 49143||25 February 2020 08:12:00||Diablo Canyon||NRC Region 4||Westinghouse PWR 4-Loop||Emergency Diesel Generator||At 2120 PDT, Diablo Canyon Power Plant experienced a loss of the offsite 230 kV startup power source due to an offsite transmission system relay actuation, resulting in valid anticipatory starts of Units 1 and 2 three emergency diesel generators on each unit. All diesels successfully started but were not loaded. All systems operated as designed with no problems observed.
The 230 kV startup power source is the only offsite power system designed to be immediately available following an accident. However, the safety related onsite emergency diesel system would have provided power to mitigate the consequences of an accident while the 230 kV was unavailable. Restoration of the 230 kV offsite power system is in progress." The licensee will notify the NRC Resident Inspector.
230 kV was restored to Operable on 06/24/13 at 0200 PDT. Normal 500 kV Offsite Power remained operable and was unaffected by this event. Following further review, Diablo Canyon Power Plant (DCPP) has determined that the loss of the 230 kV system was not a condition that could have prevented fulfillment of the safety function of a system credited to mitigate the consequences of an accident. This power source is not considered to be a safety-related system that is credited to mitigate any accident as described in the DCPP UFSAR, Chapters 6 and 15 accident analyses. PG&E concludes that the emergency diesel generators are the only power source needed to fulfill the accident mitigation function, and they did not become inoperable as a result of this event.The licensee will be notifying the NRC Resident Inspector. Notified R4DO (Werner).
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