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 Entered dateSiteRegionReactor typeEvent description
ENS 5342927 May 2018 12:42:00FermiNRC Region 3GE-4On May 27, 2018 at 0630 EDT, the Reactor Water Cleanup (RWCU) System Isolation Differential Flow - High function was declared inoperable as a result of indicating downscale. This condition would have prevented the primary containment isolation valves for the RWCU system from automatically isolating on a high differential flow instrumentation signal. At 0753, RWCU was shutdown and the affected penetration flow paths were isolated in accordance with station procedures per Fermi Technical Specifications. The cause of the event is under investigation. There was no radiological release associated with this event. All other RWCU primary containment isolation instrumentation functions remained operable and the associated RWCU system primary containment isolation valves were capable of being remotely closed by the control room operators throughout the event. However, the condition is reportable pursuant to 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector was notified.
ENS 533854 May 2018 16:20:00FermiNRC Region 3GE-4At 1412 EDT, a portable chemical toilet was found tipped over. Approximately 1 gallon of contents spilled to gravel only. A notification to the Michigan Department of Environmental Quality and local health department is required, as well as a press release. This event is being reported pursuant to 10CFR50.72(b)(2)(xi). The licensee will notify the NRC Resident Inspector.
ENS 5335220 April 2018 16:05:00SusquehannaNRC Region 1GE-4A non-licensed supervisory contract worker was found in violation of the Fitness for Duty Program. The individual's access to the plant has been terminated. The licensee notified the NRC Resident Inspector.
ENS 5334117 April 2018 16:29:00LimerickNRC Region 1GE-4Unit 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 5334017 April 2018 12:02:00Browns FerryNRC Region 2GE-4At 0416 CDT on April 17, 2018, the High Pressure Coolant Injection System (HPCI) was isolated due to a water side leak from the gland seal condenser. Unit 1 HPCI remains inoperable pending repairs to the gland seal condenser. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(V)(D), '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 mitigate the consequences of an accident.' This is also reportable as a 60-day written report in accordance with 10 CFR 50.73(a)(2)(V)(D). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified of this event.
ENS 5333614 April 2018 14:34:00FermiNRC Region 3GE-4

At 1040 EDT, Fermi 2 automatically scrammed on RPV (Reactor Pressure Vessel) Level 3 following a loss of the Division 1 Station System Transformer (SST) #64. All control rods fully inserted. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) automatically started as designed on Reactor Water Level (RWL) 2 and restored RWL. The lowest RWL reached was 101.8 inches (above Top of Active Fuel). HPCI injected for approximately 35 seconds. RWL is currently being maintained in the normal level band with RCIC. No Safety Relief Valves (SRVs) actuated. All isolations and actuations for RWL 3 and 2 occurred as expected. Investigation into loss of SST #64 continues. At the time of the scram, all Emergency Core Cooling Systems (ECCS) and Emergency Diesel Generators (EDGs) were operable, and no safety related equipment was out of service. This report is being made in accordance with 10CFR50.72(b)(2)(iv)(A), any event that results in ECCS discharge into the reactor coolant system as a result of a valid signal and 10CFR50.72(b)(2)(iv)(B), any event that results in the actuation of the Reactor Protection System (RPS) when the reactor is critical. Following the loss of power and reactor scram, the Division 2 EECW (Emergency Equipment Cooling Water) Temperature Control Valve (TCV) controller was in Emergency Manual and maintaining max cooling. Operators placed the controller in Auto and the TCV is controlling normally. The NRC Senior Resident has been notified. Decay heat is being removed via Division 2 steam dumps to the condenser. The plant is in a modified shutdown electric lineup with offsite power available and stable. Emergency diesel generators did automatically start and load.

  • * * UPDATE ON 4/14/2018 AT 1838 EDT FROM JEFF MYERS TO HOWIE CROUCH * * *

This update provides additional clarification of the applicable reporting criteria for this event associated with Primary Containment Isolation Actuations. All isolations and actuations for RWL (Groups 4, 13, and 15) and RWL 2 (Groups 2, 10, 11, 12, 14, 16, 17, and 18) occurred as expected. This report is also being made in accordance with 10CFR50.72(b)(3)(iv)(A), any event or condition that results in valid actuation of any systems listed in paragraph (b)(3)(iv)(B): RPS, HPCI, and RCIC. RPV pressure is being maintained by the bypass valves to the main condenser. All actuations that occurred were fully completed and restored. The licensee notified the NRC Resident Inspector. Notified R3DO (Stone).

  • * * UPDATE ON 4/15/2018 AT 1950 EDT FROM KELLEY BELENKY TO DAVID AIRD * * *

This update provides additional information regarding the specified system actuations and an additional applicable reporting criteria. The loss of Division 1 Station System Transformer (SST) #64 at 1040 EDT on 4/14/2018 resulted in the automatic initiation of Emergency Diesel Generators (EDG) 11 and 12. The EDGs started as expected and continue to supply their associated busses. This is reportable pursuant to 10CFR50.72(b)(3)(iv)(A), as an event or condition that resulted in a valid actuation of any system listed in paragraph (b)(3)(iv)(B), including EDGs. In addition, the loss of the Division 1 SST #64 resulted in the expected transfer from the normal to alternate power source for the Low Pressure Coolant Injection (LPCI) swing bus, rendering LPCI loop select inoperable. The alternate power source continued to energize the LPCI swing bus throughout the event until the system was realigned to the normal power source at 1239 EDT on 4/14/2018. This condition is reportable pursuant to 10CFR50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Notified R3DO (Stone).

ENS 533197 April 2018 12:10:00BrunswickNRC Region 2GE-4

On April 7, 2018, at 0836 EDT, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped during testing of the stator cooling system. The trip was uncomplicated with all systems responding normally. No safety-related equipment was inoperable at the time of the event. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B).

Operations responded using Emergency Operating Procedures and stabilized the plant in Mode 3. Reactor water level being maintained via normal feedwater system. Decay heat is being removed through the bypass valves.

Reactor water level reached low level 1 (LL1) as a result of the reactor trip. The LL1 signal causes a Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The LL1 isolations occurred as designed; the Group 8 valves were closed at the time of the event. Due to the Primary Containment Isolation System (PCIS) actuation, this event is also being reported as an eight-hour, non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the PCIS. Unit 2 was not affected. There was no impact on the health and safety of the public or plant personnel. The safety significance of this event is minimal. The automatic reactor trip was not complicated and all safety-related systems operated as designed. Investigation of the cause of the Reactor Protection System actuation is in progress. The licensee notified the NRC Resident Inspector.

ENS 533103 April 2018 02:53:00SusquehannaNRC Region 1GE-4On April 3, 2018 at 0019 (EDT), the Susquehanna control room received indication that a loss of Secondary Containment Zone 3 differential pressure had occurred. Control room operators noted the loss following completion of surveillance testing. The cause is under investigation. Zone 3 differential pressure was restored to greater than 0.25 inches WC (water column) at 0145 (EDT). Zone 3 differential pressures being less than 0.25 inches WC constitutes a loss of Secondary Containment based on not meeting requirements of SR (Surveillance Requirement) 3.6.4.1.1. This event is being reported under 10 CFR 50.72(b)(3)(v)(C) and per the guidance of NUREG-1022, Revision 3, Section 3.2.7, as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment system. The NRC Resident Inspector has been notified.
ENS 5330029 March 2018 22:28:00Browns FerryNRC Region 2GE-4At 1344 on March 29, 2018, it was determined (engineering evaluation) that an unanalyzed condition that significantly degraded plant safety previously existed. During a postulated control room abandonment due to a fire, and concurrent with a Loss of Offsite Power (LOOP), the required number of Emergency Equipment Cooling Water (EECW) pumps would not have been available from 10/28/2015 to 3/10/2018. On March 8, 2018, during relay functional testing it was discovered that the C3 Emergency Equipment Cooling Water (EECW) pump closing springs did not recharge with the breaker transfer switch in emergency. On August 23, 2012, a wire modification was performed that contained a drawing error resulting in wire placement on the incorrect connection points for the C3 EECW pump. On March 10, 2018, the C3 EECW pump breaker wiring was corrected and subsequent testing was completed satisfactorily. Prior to 10/28/2015, Brown's Ferry Nuclear Plant (BFN) adhered to Appendix R fire protection requirements which did not credit the C3 EECW pump for fire protection from the backup control location. On 10/28/2015, BFN transitioned to National Fire Protection Association (NFPA) 805 fire protection requirements which takes credit for the C3 EECW pump from the backup control location. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B), 'Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety'. This is also reportable as a 60-day written report in accordance with 10 CFR 50.73(a)(2)(ii)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified of this event.
ENS 5326918 March 2018 16:16:00Browns FerryNRC Region 2GE-4At 1158 CDT on March 18, 2018, the Unit 1 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from High Reactor Steam Dome Pressure in response to Turbine Control Valve Closure. The reactor had been operating at 100 percent power. Investigation is in progress. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Steam Relief Valves (MSRVs) operating on the initial transient as expected. Main Turbine Bypass Valves are currently controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. All safety system operated as expected. At no time was public health and safety at risk. This event is reportable within 4 hours per 10 CFR 50.72(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.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation' and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.
ENS 5326515 March 2018 22:08:00Peach BottomNRC Region 1GE-4At 1524 (EDT) on Thursday, March 15, 2018, Operations was notified of a failure to meet Appendix R requirements for Peach Bottom Atomic Power Station (PBAPS) Unit 2 and Unit 3. Valves associated with the feedwater system for both units were not properly considered as Hi-Lo Pressure interface valves as required by the Appendix R program. This results in the susceptibility to a hot short condition that could open valves, diverting flow from the reactor, damage piping and prevent injection. U3 (Unit 3) Fire Safe Shutdown Credited Reactor Core Isolation Cooling (RCIC) System is affected. U2 (Unit 2) is affected by a potential leak path through the Reactor Water Cleanup system. This event is being reported as an occurrence of an event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety under 10 CFR 50.72(b)(3)(ii). The Station (PBAPS) is performing hourly fire watches for the impacted areas and is also evaluating this condition for corrective action. The licensee notified the NRC Resident Inspector.
ENS 5326115 March 2018 01:39:00CooperNRC Region 4GE-4

At approximately 1711 CDT on 14 MAR 2018, Cooper Nuclear Station was notified by the National Weather Service that the Shubert radio transmission tower was not functioning. This affects the tone alert radios used to notify the public in event of an emergency condition. This condition is reportable under 10CFR50.72(b)(3)(xiii). A backup notification method is available and will be utilized for notifications if needed. The local telephone company is providing troubleshooting and repair services. A return to service time for the Shubert tower is not currently available. The NRC Senior Resident Inspector has been informed. The issue was identified during periodic maintenance. The licensee notified all counties within the 10 mile Emergency Planning Zone.

  • * * UPDATE ON 3/19/2018 AT 1720 EDT FROM STEVE WHEELER TO DONG PARK * * *

This is a follow up notification to update the status of the Shubert radio transmission tower that was reported to be out of service on March 14th per EN53261. The tower was restored to service and determined to be functional at 0759 (CDT) on March 19th, 2018. The licensee notified the NRC Resident Inspector. Notified R4DO (Groom).

ENS 5325612 March 2018 14:34:00SusquehannaNRC Region 1GE-4A non-licensed supervisor tested positive for alcohol during pre-access screening. The individual's access to the plant was denied. The NRC Resident Inspector has been notified.
ENS 5325310 March 2018 13:54:00CooperNRC Region 4GE-4HPCI (High Pressure Coolant Injection) was removed from service (unplanned) on 3/10/2018 at time 0709 (CST) by closing HPCI-MO-15, STEAM SUPPLY INBOARD ISOLATION. The inboard steam supply isolation valve is inside Primary Containment. The steam supply valve was closed in an effort to isolate (unidentified) leakage to Primary Containment from a suspected packing leak from HPCI-MO-15. After closing the HPCI-MO-15, Reactor Coolant System Leakage parameters returned to within Technical Specification (TS) LCO (Limiting Condition of Operations) 3.4.4 limits. Entered into Technical Specification LCO 3.5.1 Condition C - HPCI System Inoperable. Required Actions for Condition 'C' are to verify by administrative means RCIC (Reactor Core Isolation Cooling) System is operable within 1 hour and restore HPCI System to operable status within 14 days. RCIC was verified operable by administrative means concurrent with declaration of HPCI inoperable. Normal plant shutdown activities are being planned (for 03/11/2018 at 1200 CDT) to support entry into Primary Containment to initiate any necessary repairs. HPCI is a single train safety system. This report is submitted as a condition that at time of discovery could prevent the fulfillment of the safety function of an SSC (System, Structure, or Component) needed to mitigate the consequences of an accident. This condition has been entered into the CNS (Cooper Nuclear Station) Corrective Action Program. CR-CNS-2018-01346 LCO 3.4.4 was entered during unplanned maintenance of the HPCI system. When the HPCI-MO-15 was cycled from closed to open, unidentified leakage in the containment increased above 2 gallons per minute (gpm) in less than a 24 hour period. Also, total unidentified leakage exceeded 5 gpm. The licensee closed the HPCI-MO-15 valve resulting in a decrease in unidentified leakage below TS shutdown limit. The NRC Resident Inspector was notified.
ENS 532477 March 2018 12:25:00Browns FerryNRC Region 2GE-4

The licensee declared an Unusual Event based on Emergency Action Level (EAL) 6.7.U and entry into the site Security Plan. All required actions or compensatory measures have been completed. The Notice of Unusual Event was terminated at 1142 CST. There was no impact to the operation of any of the units at the Browns Ferry site. The licensee has notified the NRC Senior Resident Inspector. See EN #53248. Notified DHS SWO, FEMA Ops, DHS NICC, FEMA NWC (email) and NuclearSSA (email).

  • * * UPDATE AT 1816 EST ON 03/07/2018 FROM DAVID RENN TO JEFF HERRERA * * *

The licensee provided additional information regarding the event. Notified the R2DO (Musser), IRD MOC (Gott), NRR EO (Miller).

ENS 532361 March 2018 13:53:00FermiNRC Region 3GE-4A non-licensed, supervisory employee was determined to be under the influence of alcohol during a random test. The employee's unescorted access was terminated. The NRC Resident Inspector has been notified.
ENS 5321415 February 2018 17:36:00FermiNRC Region 3GE-4A can of alcohol (8.4 ounces) was discovered unopened in a refrigerator inside the protected area. Site security took possession of the can of alcohol. The owner of the can of alcohol is unknown. This report is being made under 10 CFR 26.719(b)(1) as a 24 hour telephone notification. The can had an expiration date of April 2017. The licensee notified the NRC Resident Inspector and the Regional Inspector.
ENS 5320211 February 2018 23:36:00SusquehannaNRC Region 1GE-4On February 11, 2018 at 2203 (EST), the Susquehanna Control Room received indication that a loss of Secondary Containment Zone 2 differential pressure (DP) had occurred. Control Room operators noted a differential pressure of <.25" WC (inches Water Column) for several seconds. System DP was restored to normal in 1 minute. The cause of the pressure swings is under investigation. Zone 2 differential pressures being less than 0.25" WC constitutes a loss of Secondary Containment based on not meeting requirements of SR 3.6.4.1.1. This event is being reported under 10 CFR 50.72(b)(3)(v)(c) 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.
ENS 5317822 January 2018 15:09:00FermiNRC Region 3GE-4At 1115 EST, on 1/22/18, Fermi 2 determined that the site was in violation of its National Pollutant Discharge Elimination System (NPDES) permit due to an oil sheen being observed in a overflow canal that had breached the installed oil booms and entered navigable waterways. Approximately 5-10 gallons of oil has reached navigable water, which resulted in exceeding State limits. The oil is currently contained with no additional leakage to navigable waters and cleanup is in progress. The cause of the oil entering the overflow canal is under investigation. Reports will be made to the Michigan Department of Environmental Quality (MDEQ) and other local agencies. Since these reports are in the process of being made, this is considered a News Release or Notification to Other Government Agencies, therefore this event is reportable under 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. The Licensee has notified the National Response Center.
ENS 5316716 January 2018 10:06:00Duane ArnoldNRC Region 3GE-4This 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of a containment isolation signal affecting more than one system. At 2230 CST on November 30, 2017, with the Duane Arnold Energy Center (DAEC) operating at 100 percent power, an invalid Group 3 isolation on the 'B' side of the Primary Containment Isolation System (PCIS) occurred. Group 3 isolation signals were generated for Primary Containment Isolation Valves for Drywell and Torus Ventilation and Purge, Containment Nitrogen Compressor Suction and Discharge, Recirculation Pump Seals, and Post Accident Sample System. This event was caused by a fault on the 1D25 Instrument AC Inverter. The fault was caused by an insufficient design clearance to ground and was corrected by increasing the clearance. All equipment responded in accordance with the plant design. Specifically, all actuations were complete and successful. There were no safety consequences or impacts on the health and safety of the public. The event was entered into DAEC's corrective action program for resolution. The NRC Resident Inspector has been notified.
ENS 5316511 January 2018 16:06:00FermiNRC Region 3GE-4On January 11, 2018, at 1041 EST, a planned train swap of the Reactor Building Heating Ventilation and Air Conditioning (RBHVAC) system resulted in the Technical Specification (TS) for secondary containment pressure boundary not being met for less than one minute. The maximum secondary containment pressure observed during that time was approximately 0.117 inches of vacuum water gauge. Secondary containment pressure was returned to within the TS operability limit of greater than or equal to 0.125 inches of vacuum water gauge per TS Surveillance Requirement (SR) 3.6.4.1.1 by starting Division 1 of the Standby Gas Treatment System (SGTS) in addition to the RBHVAC system already in operation. Secondary containment pressure is currently stable. Secondary containment was declared Operable at 1045 EST. There were no radiological releases associated with this event. Declaring secondary containment inoperable as a result of not meeting TS SR 3.6.4.1.1 is reportable under 10CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The licensee has notified the NRC Resident Inspector.
ENS 5316210 January 2018 13:53:00Browns FerryNRC Region 2GE-4At 0928 CST on January 10, 2018, the Unit 3 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from Turbine Control Valve Emergency Trip System pressure low. The reactor had been operating near 73 percent power for an emergent issue for Turbine Control Valve (TCV) No. 3. With TCV No. 3 out of service and closed, the unit was operating with RPS in a half scram condition. A subsequent failure of the TCV No. 2 sensing line resulted in RPS coincidence logic being met for TCV fast closure SCRAM. The investigation of the TCV No. 2 sensing line failure continues. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Turbine Bypass Valves controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. Neither High Pressure Coolant Injection nor Reactor Core Isolation Cooling initiation signals were received. This event is reportable within 4 hours per 10 CFR 50.72(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 preplanned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident inspector has been notified.
ENS 5313320 December 2017 21:22:00CooperNRC Region 4GE-4During review of the documentation for the 11/16/17 outage of the NOAA/NWS (National Oceanographic and Atmospheric Administration/National Weather Service) tower, it was identified that there was also record of a trouble ticket being issued on 11/19/17 for the NOAA/NWS tower. Further discussions with the National Weather Service determined that the tower did experience an outage on 11/19/17 which affected the ability to activate EAS (Emergency Alert System)/Tone Alert Radios. Final determination that the EAS/Tone Alert Radios were affected during this outage was made at 1559 (CST), which was the time that the National Weather Service sent the e-mail to the EP (Emergency Planning) Manager and EP Offsite Coordinator with notification that activation of the EAS/Tone Alert Radios was affected during the outage. This is considered to be a major loss of the Public Prompt Notification System capability, and is reportable under 10CFR 50.72(b)(3)(xiii). The transmission outage was on 11/19/2017 0853 until 1100 but CNS (Cooper Nuclear Station) was not notified until 1559 on 12/20/2017. The NRC Senior Resident has been informed
ENS 5312819 December 2017 17:17:00CooperNRC Region 4GE-4

During regular power operations at 100% power, DG#1 and DG#2 were declared inoperable due to a common issue associated with indicating lights and the associated sockets installed in various control and auxiliary circuits for both DG's. The indicating lights in question are incandescent 120V AC style 120MB bulbs in a socket with a 550 ohm resistor. Style 120MB light bulbs have a failure mechanism where the bulb can cause a short circuit rather than the more common open circuit that is expected when an incandescent bulb filament fails. Cooper originally believed that the socket's integral resistor was sufficient to protect the circuit. In testing performed by an outside laboratory and confirmed on-site using warehouse stock, it was determined that the integral resistor may not have the power dissipation capability to protect the circuit ln which the light and socket are installed if a bulb fails in short circuit. This condition resulted in both DG's being declared inoperable at 1340 (CST) due to a loss of reasonable expectation that they would meet their safety function required action to start, load and run to support loads required to mitigate the consequences of an accident. This is a loss of safety function under 10CFR 50.72(b)(3)(v)(D) subject to an 8 hour report. As a result of both DG's being inoperable, the Control Room Emergency Filtration System is also inoperable. This is also a loss of safety function subject to an 8 hour report for the same criterion. The Senior Resident has been notified.

  • * * RETRACTION AT 0942 EST ON 02/14/2018 FROM DAVID VAN DERKAMP TO JEFF HERRERA * * *

CNS is retracting the 8-hour non-emergency notification made on December 19, 2017 at 1340 CST (EN# 53128). Subsequent evaluation concluded a postulated lamp short circuit failure in any of the affected circuits would not impact the ability of the Diesel Generators to perform their safety function and therefore, were operable. With DG operability not affected, the Control Room Emergency Filtration System also remained operable. The NRC Resident Inspector has been notified. Notified the R4DO (Werner).

ENS 5312317 December 2017 05:48:00BrunswickNRC Region 2GE-4

On December 17, 2017 at 0316 EST, the Unit 2 HPCI system was isolated and declared inoperable due to a packing failure of the HPCI Turbine Steam Supply Valve (i.e., 2-E41-F001). Isolation of the HPCI system due to the packing failure prevents the HPCI system from performing its design safety function. As such, this event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system that is needed to mitigate the consequences of an accident. Unit 2 HPCI system has been isolated and depressurized. The HPCI system will remain inoperable until the valve can be repaired. The safety significance of this condition is minimal. All other Emergency Core Cooling Systems (ECCS) and the Reactor Core Isolation Cooling (RCIC) system remain operable. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 1/29/18 AT 1514 EST FROM MARK TURKAL TO DONG PARK * * *

Based upon further evaluation, Duke Energy is retracting Event Notification 53123. Engineering has determined that the packing failure of the HPCI Turbine Steam Supply Valve did not prevent the HPCI system from performing its safety function. Environmental conditions resulting from the steam leak would not have caused automatic HPCI isolation or otherwise have degraded HPCI operation. Additionally, the amount of steam diverted through the packing leak was negligible with respect to total steam flow and did not affect HPCI system performance. HPCI would have remained operable throughout its entire mission time. Therefore, this condition does not represent an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident and is not reportable in accordance with 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector was notified of this retraction. Notified R2DO (Heisserer).

ENS 531088 December 2017 17:25:00LimerickNRC Region 1GE-4U/2 HPCI (Unit 2/High Pressure Coolant Injection) was declared inoperable due to leak by of the pump discharge check valve after pump shutdown from flow testing. This resulted in cycling of the minimum flow valve. The discharge valve was closed to prevent the continued cycling of the minimum flow valve. This condition was identified during normal surveillance testing. The licensee notified the NRC Resident Inspector.
ENS 5310130 November 2017 17:17:00Hope CreekNRC Region 1GE-4

An Unusual Event was declared at 1657 EST due to an earthquake detected onsite. The Unusual Event was declared under EAL HU1.1. There is no release in progress due to this event. There are no protective actions recommended at this time. The Licensee will notify the NRC Resident Inspector. Note: See also EN #53099 for Salem Unusual Event.

  • * * UPDATE FROM JOSHUA MYERS TO DONALD NORWOOD AT 1742 EST ON 11/30/2017 * * *

An earthquake was felt onsite at time 1645 EST. Multiple phone calls were made to the Control Room confirming the earthquake. It was verified there was an earthquake felt in Delaware with a magnitude of 4.4. Neither seismic monitor at Salem Unit 1, Salem Unit 2, and Hope Creek actuated. There is no indication of any damage to any systems or plant structures. Plant walk-downs have been initiated in accordance with plant operating procedures for a seismic event. No injuries have been reported to the Control Room. The licensee will notify the NRC Resident Inspector and State and local government agencies. Notified R1DO (Gray), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), and DHS Nuclear SSA (email).

  • * * UPDATE FROM THOMAS CLARK TO DAVID AIRD AT 2137 EST ON 11/30/2017 * * *

The licensee terminated the Unusual Event at 2125 EST on 11/30/2017 following plant walkdowns that revealed no damage to plant structures, systems, or components. The NRC Resident Inspector has been notified. Notified R1DO (Gray), IRD (Grant), and NRR EO (Miller), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), and DHS Nuclear SSA (email).

ENS 5309830 November 2017 16:02:00SusquehannaNRC Region 1GE-4On November 30, 2017 at 1026 EST, the Susquehanna Control Room received indication that a loss of Secondary Containment Zone 2 differential pressure had occurred. Control Room operators noted a differential pressure of 0.0 inch WC (water column) for several seconds, followed by a high DP of 0.5 inch WC. System DP was restored to normal in 3 minutes. The cause of the pressure swings is under investigation. Zone 2 differential pressures being less than 0.25 inch WC constitutes a loss of Secondary Containment based on not meeting requirements of SR 3.6.4.1.1. This event is being reported under 10 CFR 50.72(b)(3)(v)(c) 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.
ENS 5307416 November 2017 08:17:00CooperNRC Region 4GE-4At 0008 CST on 11/16/2017, Cooper Nuclear Station (CNS) was notified by Omaha Weather that the NOAA broadcast and the Shubert radio tower for this area is off. This affects the tone alert radios used to notify the public in event of an emergency condition. This is considered to be a major loss of the Public Prompt Notification System capability, and is reportable under 10CFR50.72(b)(3)(xiii). The transmission outage actually began at 2007 (CST), 11/15/2017, but CNS was not notified until 0008 (CST), 11/16/2017. Backup notification methods remained available throughout the period. At time 0447 CST on 11/16/2017, Cooper Nuclear Station was notified that the NOAA broadcast and Shubert radio transmission tower was returned to service. Nemaha County, NE, Richardson County, NE, and Atchison County, MO authorities within the 10 mile EPZ were notified by Cooper Nuclear Station of the condition and the effect on the tone alert radios at 0642 (CST), 11/16/2017. This is reportable under 10CFR50.72(b)(2)(xi) as a 4 hour report. The NRC Senior Resident has been informed.
ENS 5307014 November 2017 15:13:00Browns FerryNRC Region 2GE-4This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On September 15, 2017, during a TVA (Tennessee Valley Authority) review of Operations logs, it was determined that a reportable condition occurred in January 2017 but no NRC report had been made. On January 10, 2017, at 0300 Central Standard Time (CST), Browns Ferry Nuclear Plant, Unit 3, received Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolation signals. The Group 2, 3, 6, and 8 isolations caused the initiation of all three trains of the Standby Gas Treatment (SBGT) system and Control Room Emergency Ventilation (CREV) subsystem 'A.' At 0311 CST, Operations personnel discovered that the 3A1 RPS circuit protector had tripped on undervoltage. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywall Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywall Pressure. At the time of the event, these conditions did not exist; therefore the actuation of the PCIS was invalid. All affected equipment responded as designed. This condition was the result of an undervoltage condition on the 3A1 circuit protector. During trouble shooting, the undervoltage setpoints were found to be 116 VAC and 115 VAC, when the normal as left acceptance band is 109.7 VAC to 111.3 VAC. The 3A RPS protective relays had been previously replaced in September 2016. The most likely cause of the undervoltage condition in these relays is infant mortality. The NRC Resident Inspector has been notified of this event.
ENS 5306914 November 2017 12:42:00CooperNRC Region 4GE-4

At 1118 CST on 11/14/17, Cooper Nuclear Generating Station declared an Unusual Event due to a hydrogen leak on a main generator purge line. The leak was reported to be caused by Maintenance cutting into a one inch line. The total size of the leak is unknown, however, it is estimated to be depressurizing in the main generator at approximately 1lb per hour. The current pressure is 52 to 53 lbs. pressure and is stable. The operations staff have entered their abnormal procedure and are taking actions to isolate the leak. Operators have isolated the source of hydrogen and have opened the exterior roll up doors to increase the airflow and minimize the concentration of hydrogen in the area. The area has been evacuated and hot work has been stopped. The NRC Resident Inspector has been notified. Notified DHS, FEMA, NICC and NNSA (via email).

  • * * UPDATE AT 1433 EST ON 11/14/2017 FROM ROY GILES TO MARK ABRAMOVITZ * * *

On 11/14/2017, Nebraska Public Power District will issue a press release concerning the declaration of a Notification of Unusual Event (EN#53069) declared today at 1118 (CST) for a small hydrogen leak in the turbine building. This is a four hour report per 10CFR50.72(b)(2)(xi) for any event or situation for which a news release is planned or notification to other government agencies has been or will be made which is related to heightened public or government concern. Notified the R4DO (Kozal).

  • * * UPDATE AT 1904 EST ON 11/14/17 FROM TRENT SYDOW TO JEFF HERRERA * * *

At 1744 CST the licensee exited from the Unusual Event. The leak was patched under a temporary repair. The patch was tested to verify the leak has stopped. The NRC Resident Inspector was notified. Notified the R4DO (Kozal), IRDMOC (Gott), NRR EO (Miller), DHS, FEMA, NICC and NNSA (via email).

ENS 530491 November 2017 22:26:00Browns FerryNRC Region 2GE-4At 1425 (CDT) on November 1, 2017, Operations was notified of a condition affecting Unit 3 4kV Shutdown Boards 3EA, 3EB, 3EC, and 3ED. It was discovered that multiple potential transformer (PT) primary fuses are GE type EJ1 size 0.5 AMP which does not coordinate with the PT's secondary fuses. A fault on the associated cable could clear the primary PT primary fuses for the 4kV Shutdown Board. This would result in the board tripping 4kV motor loads, disconnecting from Off-site power and connecting to the Emergency Diesel Generator. However, since the PT fuse is cleared, the under-voltage trips on the 4kV motors would remain in if there is no Common Accident Signal (CAS) present. The 4kV motor loads include Residual Heat Removal (RHR) Pumps, Core Spray (CS) Pumps, Residual Heat Removal Service Water (RHRSW) Pumps, and Emergency Equipment Cooling Water (EECW) pumps. Review of NFPA 805 analyses show the cables for all four U3 4kV Shutdown Boards are routed in Fire Area 03-03 and Fire Area 16. Therefore a fire in either area could result in a loss of all four U3 4kV Shutdown Boards motor loads. Cables for 4kV Shutdown Board 3EA and 3EB are both routed in Fire Area 21 which could result in a loss of both Division I Shutdown Board motor loads. Compensatory fire watch measures have been established. This event requires an 8 hour report in accordance with 50.72(b)(3)(ii)(B), 'Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' The NRC Resident Inspector has been notified. CR 1354129 was initiated in the Corrective Action Program.
ENS 5303123 October 2017 09:35:00Peach BottomNRC Region 1GE-4On Monday, October 23, 2017, with PBAPS (Peach Bottom Atomic Power Station) Unit 3 in Mode 3 at the beginning of a refueling outage, personnel entered the drywell to perform an inspection. At approximately 0400 (EDT), leakage was identified on a one-inch diameter instrument line socket weld for the 'B' recirculation pump. Because the leak was misting, the leakage rate could not be quantified. However, Unit 3 reactor coolant unidentified leakage prior to plant shutdown was 0.18 gpm. This line is considered part of the primary coolant pressure boundary. This event is being reported as an occurrence of an event or condition that results in the condition of the nuclear power plant, including its principal safety barriers being seriously degraded under 10 CFR 50.72(b)(3)(ii). The Station is preparing an evaluation and repair plan at this time. The NRC Resident Inspector has been notified.
ENS 5302218 October 2017 05:27:00CooperNRC Region 4GE-4

Eight hour report due to HPCl (High Pressure Coolant Injection) inoperability. HPCl valve operability testing was performed on October 18, 2017. Following satisfactory completion of opening stroke timing, the control switch for HPCI-MOV-MO19, HPCI Injection Valve, was taken to close. The valve indicates that it moved to an intermediate position, but it has not indicated that it has fully closed. This resulted in the valve being declared inoperable. This valve is normally closed and automatically opens on a HPCI initiation signal. HPCl was previously declared inoperable at time 0136 (CDT) on October 18 for surveillance testing. Entry was made into Tech Spec LCO 3.5.1 Condition C - HPCI System Inoperable at that time. Required Actions for Condition C are to verify by administrative means RCIC System is operable within 1 hour and restore HPCI System to operable status within 14 days. RClC was verified operable by administrative means concurrent with declaration of HPCI inoperable. Troubleshooting activities for HPCI are being planned. HPCI is a single train safety system. This report is submitted as a condition that at time of discovery could prevent the fulfillment of the safety function of an SSC (structures, systems, and components) needed to mitigate the consequences of an accident. This condition has been entered into the CNS Corrective Action Program. The licensee notified the NRC Resident Inspector.

  • * * RETRACTION ON 11/14/17 AT 0849 EST FROM DAVID VAN DER KAMP TO BETHANY CECERE * * *

CNS is retracting the 8-hour non-emergency notification made on October 18, 2017 at 0209 CDT (EN# 53022). Subsequent evaluation concluded HPCI-MOV-MO19 was still capable of performing its safety function with the failed torque switch identified during troubleshooting and would have supported the operability of the HPCI system. HPCI-MOV-MO19 only has a safety function to open to support HPCI safety function. The failed torque switch only affects the close function of the valve; therefore the HPCI system remained fully capable of performing its required safety function and was operable with the identified condition. The NRC Resident Inspector has been notified. Notified R4DO (Haire).

ENS 5301413 October 2017 22:46:00Browns FerryNRC Region 2GE-4On October 13, 2017 at 1700 CDT, Unit 1 High Pressure Coolant Injection (HPCI) was declared Inoperable due to discovery of a leak on a sensing line to 1-PCV-073-0043, Lube Oil Cooler & Gland Seal Condenser Pressure Control Valve. The leak is a steady stream located where the sense line connects to the valve. This constitutes an unplanned HPCI System inoperability and requires an 8 hour ENS notification in accordance with 10 CFR 50.72(b)(3)(v)(D), due to the failure of a single train system affecting accident mitigation and a 60 day written report in accordance with 10 CFR 50.73(a)(2)(v)(D). The NRC Resident Inspector has been notified by the Licensee.
ENS 530036 October 2017 21:46:00SusquehannaNRC Region 1GE-4On October 6, 2017 at 1945 EDT, a loss of Control Room Habitability Envelope (CRE) was declared due to failing to meet the requirements of SR 3.7.3.4 during 72 month surveillance testing. Measured in-leakage exceeded the SR acceptance value. The CRE is required to be maintained such that occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The station remains in compliance with Technical Specification required action statements. This event is being reported under 10 CFR 50.72(b)(3)(v)(D) 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 Control Room Habitability Envelope. The licensee notified the NRC Resident Inspector.
ENS 5297417 September 2017 16:49:00BrunswickNRC Region 2GE-4On September 17, 2017, during planned surveillance activities involving Emergency Diesel Generator (EDG) 4, unexpected voltage and frequency indications were noted when EDG 4 was synchronized to Emergency Bus E4. With EDG 4 in manual mode, the Operator responded by lowering load to reopen the EDG 4 output breaker. Opening of the EDG 4 output breaker with the breakers from Balance of Plant (BOP) Bus 2C, which normally feeds the Emergency Bus E4, opened; resulted in de-energizing Emergency Bus E4. The EDG 4 voltage regulator and governor automatically reverted to auto control, and EDG 4 reconnected to Emergency Bus E4. Normal frequency and voltage were restored with EDG 4 in auto control. The momentary power interruption to Emergency Bus E4 resulted in Unit 2 Primary Containment Isolation System (PCIS) Group 2 (i.e., Drywell Equipment and Floor Drain, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems), and Group 10 (i.e., air isolation to the drywell) isolations. The actuations of Primary Containment Isolation Valves (PCIVS) were completed and the affected equipment responded as designed. Per design, no Unit 1 safety system group isolations or actuations occurred. These actuations are being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). Additional Unit 2 actuations included PCIS Group 3 (i.e., Reactor Water Cleanup), Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), and the automatic start of Standby Gas Treatment (SGT) System subsystems A and B. These systems functioned as designed. This event did not impact public health and safety. The NRC Resident Inspector has been notified. The safety significance of this event is minimal. Safety systems functioned as designed following the power perturbation on E4. Plant systems responded as designed. The cause of the event is under investigation.
ENS 5297316 September 2017 15:53:00SusquehannaNRC Region 1GE-4On September 16th, 2017 at 1330 hrs. (EDT), a loss of secondary containment differential pressure (D/P) occurred due to an equipment failure. This caused a reduction in Reactor Building Zone 2 (Unit 2) D/P to less than the required 0.25 inches WC per SR (Surveillance Requirement) 3.6.4.1.1. 2V206B, Reactor Building Zone 2 Equipment Compartment Exhaust Fan, was manually started and Reactor Building Zone 2 D/P was restored to greater than 0.25 inches WC by 1333 hrs. Reactor Building Zone 1 (Unit 1) and Zone 3 (Units 1&2) ventilation remained in service and stable. This event is being reported under 10 CFR 50.72(b)(3)(v)(c) 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 5296111 September 2017 15:00:00Peach BottomNRC Region 1GE-4This report is being made as required by 10 CFR 50.73(a)(2)(iv)(A) to describe an automatic actuation of containment isolation valves in more than one system. Because the actuation was invalid, this 60-day telephone notification is being made instead of a written LER (Licensee Event Report), in accordance with 10 CFR 50.73(a)(1). On 07/14/17, at approximately 1453 hours (EDT), an electrical transient occurred due to an off-site lightning strike that de-energized one of the station's two qualified off-site power sources. This resulted in an automatic fast transfer of four 4 kV electrical buses to the alternate off-site source. The fast transfer occurred as designed without complications. The loss of power had numerous impacts on plant equipment that occurred in accordance with plant design, including a Group 2 primary containment isolation on both units. The Group 2 isolation affected multiple systems, including Reactor Water Cleanup, Instrument Nitrogen, and the Drywell Floor Drain. The fault on the off-site transmission line immediately cleared after the lightning strike and at 1457 hours (EDT) the transmission system operator gave the station permission to reclose the breaker to the off-site source. Following system restorations and equipment walkdowns, plant operators re-established normal connections to the off-site source on 7/14/17 at 2322 hours (EDT) in accordance with station procedures. The containment isolation occurred as a result of the loss of an off-site power source and was not due to actual plant conditions or parameters meeting design criteria for containment isolation. Therefore, this is considered to be an invalid actuation. The NRC Resident Inspector has been informed of this notification.
ENS 5295910 September 2017 20:45:00Browns FerryNRC Region 2GE-4At 1151 (CDT) on September 10, 2017 Browns Ferry Units 1 and 2 declared 'B' Control Bay chiller inoperable. 'A' Control Bay chiller was previously declared inoperable. This resulted in inoperability of the equipment in the U1 and U2 4kV Shutdown Board Rooms. The declarations of the equipment in the Shutdown Board Rooms is a loss of safely function for electrical components (4kv Shutdown Boards and 480V Shutdown Boards) required for shut down of the U1 and U2 reactors and maintaining them in a safe shutdown condition, as well as RHR capability and Accident Mitigation. lnoperability of these boards also requires declaring two trains of Standby Gas Treatment inoperable resulting in a loss of safety function for Units 1, 2 and 3 for systems needed to control the release of radioactive material. This event requires an 8 hour report IAW 50.72(b)(3)(v), 'Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to (A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. The NRC Resident Inspector has been notified. CR 1336821 was initiated in the Browns Ferry Corrective Action Program.
ENS 529589 September 2017 15:04:00FermiNRC Region 3GE-4At 1000 EDT on September 9, 2017, the Division 2 Mechanical Draft Cooling Tower (MDCT) fans were declared inoperable due to failure of the over speed fan brake inverter. The brakes prevent fan over speed from a design basis tornado. The MDCT fans are required to support operability of the Ultimate Heat Sink (UHS). The UHS is required to support operability of the Division 2 Emergency Equipment Cooling Water (EECW) system. The EECW system cools various safety related components, including the High Pressure Coolant Injection (HPCI) system room cooler. An unplanned HPCI inoperability occurred based on a loss of the HPCI Room Cooler. Investigation into why the Division 2 MDCT fan over speed brake inverter failed is in progress. This report is being made pursuant to 10CFR50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of a safety function needed to mitigate the consequences of an accident based on a loss of a single train safety system. The licensee entered two (2) LCO Action Statements (AS); 14-day LCO AS 3.5.1 for ECCS (HPCI Inoperable) and 72-hour AS 3.7.2 for UHS. The licensee has two spare inverters on-site. After replacement and successful post-maintenance testing the licensee expects to exit both AS before 72-hours. The NRC Resident Inspector has been notified.
ENS 5293428 August 2017 14:47:00CooperNRC Region 4GE-4Pursuant to 10 CFR 21, this is a non-emergency notification by Nebraska Public Power District (NPPD) concerning a defect in an Allen-Bradley 700DC-P220Z2 relay received at Cooper Nuclear Station. On August 28, 2017, NPPD completed a 10 CFR 21 evaluation of a condition that was identified on April 25, 2016, associated with an Allen-Bradley 700DC-P220Z2 relay delivered by NuTherm. The evaluation was performed to determine the applications where the relays were approved for installation, where they were installed in the plant, and determine if the failure of the relays could result in a Substantial Safety Hazard as defined in 10 CFR 21. A model 700DC-P220Z2 relay failed after 133 hours of service. An independent laboratory determined the relay contained a wound wire fault at the beginning of the spool near the spool edge. The wire that faulted was connected to the right coil terminal. The fault was at a stress point where the wire came out of the insulating material channel and started onto the spool. Continuity testing indicated that the wire was open between the fault area and the terminal, within the potting material. The relay failure was a case of component infant mortality. It appeared the fault was caused by a manufacturing flaw that likely occurred due to a tensioning issue at the start of the coil wire winding process. This deviation presents a Substantial Safety Hazard as defined in 10 CFR 21, as this relay model was approved for use in safety related applications. The relay that failed was installed in the starter for the HPCI auxiliary lube oil pump and caused the HPCI system to be inoperable. This was reported to the NRC in ENS Notification 51882 at 0154 EDT on April 26, 2016 as a loss of safety function, but was not characterized as a 10 CFR Part 21 issue at that time. The relay model was approved for use in numerous HPCI starters and the Recirculation Pump discharge valves. The licensee notified the NRC Resident Inspector.
ENS 5290916 August 2017 15:41:00Peach BottomNRC Region 1GE-4On 8/16/2017, at 1039 (EDT), an un-planned trip of the Peach Bottom Station Blackout Transformer 34.5 kV feeder breaker 1005 and a loss of the 191-00 line occurred causing a loss of power to Unit 1 and the TSC. Power was not restored to the TSC or the ventilation system within 1 hour. Power was subsequently restored to the TSC at 1207 hours (EDT) and the ventilation system was restored to available. This report is being submitted pursuant to 10CFR50.72(b)(3)(xiii) as a Major Loss of Emergency Preparedness Capabilities due to a reduction in the effectiveness of the Onsite Technical Support Center (TSC). The NRC Resident Inspector has been informed of this notification.
ENS 528928 August 2017 22:33:00SusquehannaNRC Region 1GE-4On August 8th, 2017 at 2044 hrs. (EDT) a loss of secondary containment differential pressure (D/P) occurred due to an apparent equipment failure. This caused a reduction in Reactor Building Zone II (Unit 2) D/P to less than the required 0.25 inch WC (water column) per SR (Surveillance Requirement) 3.6.4.1.1. Reactor Building Zone II Exhaust Fans were manually swapped and Reactor Building Zone II D/P was restored to greater than 0.25 inch WC by 2112 hrs. Reactor Building Zone 1 (Unit 1) and Zone 3 (Units 1&2) ventilation remained in service and stable. This event is being reported under 10 CFR 50.72(b)(3)(v)(c) 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.
ENS 528884 August 2017 17:25:00BrunswickNRC Region 2GE-4

On August 4, 2017, at 1511 EDT, Unit 1 Secondary Containment was declared inoperable due to a small (i.e., approximately 0.75 inch diameter) hole in Service Water system piping which was found during ultrasonic testing activities. The affected portion of piping penetrates Secondary Containment and flow in the piping creates a vacuum condition; thus bypassing Secondary Containment. The identified hole is being evaluated with respect to its impact on operability of the Service Water system. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(v)(C), as an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material. This event did not result in any adverse impact to the health and safety of the public. Initial Safety Significance Evaluation: The initial safety significance of this event is minimal. At the time of discovery, Unit 1 was at 100% steady state conditions. Reactor Building Ventilation was in service in a normal alignment. No abnormal radioactivity conditions existed within Secondary Containment. Corrective Actions: Temporary repair of the affected Unit 1 Service Water piping has been completed. This repair was evaluated by Engineering and it has been determined that the repair meets the requirements to maintain Secondary Containment operable. Unit 1 Secondary Containment operability was restored at 1704 EDT on August 4, 2017. The licensee notified the NRC Resident Inspector.

  • * * RETRACTION FROM MIKE BRADEN TO RICHARD SMITH AT 1447 EDT ON 9/27/17 * * *

Based upon further evaluation, Duke Energy is retracting Event Notification 52888. The safety objective of Secondary Containment is to limit the release of radioactivity to the environment after an accident so that the resulting exposures are kept to a practical minimum and are within regulatory limits. A bounding engineering evaluation was performed which demonstrates that potential releases from Secondary Containment could not have resulted in offsite or control room doses exceeding regulatory limits. Furthermore, the condition did not impact Technical Specification operability of Secondary Containment in that the ability of Secondary Containment to maintain the required vacuum was not impacted. Therefore, this condition does not represent an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material and is not reportable in accordance with 10 CFR 50.72(b)(3)(v)(C), and the event notification is being retracted. The NRC Senior Resident was notified of this retraction. Notified R2DO (A. Masters).

ENS 528811 August 2017 17:00:00Duane ArnoldNRC Region 3GE-4

At 0934 CDT, while appropriately removed from service for pre-planned testing, the High Pressure Coolant Injection (HPCI) inboard steam isolation valve, MO-2238, valve position indication was found to be inadequate. As a result, MO-2238 was declared inoperable in accordance with Technical Specification (TS) 3.6.1.3. This resulted in a condition that caused the HPCI system to be inoperable. The cause of the inadequate valve position indication is currently being investigated. This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector has been informed.

  • * * RETRACTION ON 09/11/17 AT 1144 EDT FROM BOB MURRELL TO BETHANY CECERE * * *

The purpose of this notification is to retract a previous report made on 8/1/17 at 1700 (EDT) (EN 52881). NRC notification was initially made as a result of the High Pressure Coolant Injection (HPCI) inboard steam isolation valve, MO-2238, valve position found to be inadequate. Subsequent to the initial report, NextEra Energy Duane Arnold (NextEra) has determined that the condition observed on 8/1/17 for MO-2238 did not exist prior to removing the system from service for pre-planned testing, but was observed during post maintenance testing. Consequently, the failure does not meet the reporting requirements of 10 CFR 50.72 or 10 CFR 50.73. The NRC Resident Inspector has been notified. Notified the R3DO (Riemer).

ENS 5287427 July 2017 18:54:00LimerickNRC Region 1GE-4(Unit 2) HPCI was declared inoperable due to improper valve alignment stemming from an incorrect sequence directed from a work order. (Unit 2) HPCI was inoperable for 20 minutes and was manually re-aligned to an operable status. The licensee notified the NRC Resident Inspector.
ENS 5287124 July 2017 23:50:00LimerickNRC Region 1GE-4At time (2044 EDT), Limerick Generating Station notified the Pennsylvania DEP ( The licensee notified the NRC Resident Inspector.
ENS 5285914 July 2017 15:45:00FermiNRC Region 3GE-4This telephone notification, as allowed by 10 CFR 50.73(a)(1), is being made pursuant to 10 CFR 50.73(a)(2)(iv)(A) to describe an unplanned, invalid actuation of containment isolation valves in more than one system which occurred during the most recent refueling outage at Fermi 2. On 3/24/2017, at approximately 1548 EDT, when synchronizing an emergency diesel generator (EDG) to the grid during testing, an electrical perturbation occurred. Further investigation found that the EDG was slightly out of phase when it was attempted to be synchronized to the grid. The electrical perturbation resulted in an unexpected half-scram of Reactor Protection System (RPS) A and actuation (closure) of some containment isolation valves. The actuations were invalid as they were not initiated in response to actual plant conditions or parameters satisfying the requirements for initiation. Fermi 2 was shut down for a refueling outage at the time, and therefore, the half-scram of RPS A occurred after the safety function had already been completed. Containment isolation valves actuated (closed) in Division 1 of the Torus Water Management, Drywell Pneumatics, and Drywell Floor and Equipment Drain Sumps systems. All valves operated as expected. Since containment isolation valves in more than one system were actuated by this perturbation, this event constitutes an event or condition that resulted in manual or automatic actuation of the system listed in paragraph 10 CFR 50.73 (a)(2)(iv)(B)(2) and is reportable under 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been informed of this notification.
ENS 5284510 July 2017 19:26:00BrunswickNRC Region 2GE-4At approximately 14:10 Eastern Daylight Time (EDT), the Control Room was notified of a contract employee experiencing a non-work related medical emergency within the protected area in the service building. First responders were immediately dispatched. Off-site assistance was requested. The individual was transported to the New Hanover Regional Medical Center. No radioactive material or contamination was involved. At 16:02 EDT, hospital officials notified plant personnel that the patient was declared deceased. This notification is being made in accordance with 10 CFR 50.72(b)(2)(xi) for a 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. The NRC Resident Inspector has been notified.