Emergency Notification System

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[RSS - Reactor Events]

04-05-2019

Region 1 1 [Region 1 Events] [RSS]
Region 2 3 [Region 2 Events] [RSS]
Region 3 3 [Region 3 Events] [RSS]
Region 4 2 [Region 4 Events] [RSS]
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[list]

by company

Ameren 1 [Ameren events] [RSS]
American Electric Power 0 [American Electric Power events] [RSS]
Arizona Public Service 1 [Arizona Public Service events] [RSS]
Cleveland Electric 0 [Cleveland Electric events] [RSS]
Constellation 0 [Constellation events] [RSS]
DTE Energy 1 [DTE Energy events] [RSS]
Dominion 0 [Dominion events] [RSS]
Duke Energy 5 [Duke Energy events] [RSS]
EDF Energy 0 [EDF Energy events] [RSS]
Energy Northwest 0 [Energy Northwest events] [RSS]
Entergy 2 [Entergy events] [RSS]
Exelon 2 [Exelon events] [RSS]
FirstEnergy 0 [FirstEnergy events] [RSS]
GEH Hitachi 0 [GEH Hitachi events] [RSS]
Jiangsu Nuclear Power Corporation 0 [Jiangsu Nuclear Power Corporation events] [RSS]
Luminant 0 [Luminant events] [RSS]
NextEra Energy 0 [NextEra Energy events] [RSS]
Niagara Mohawk 0 [Niagara Mohawk events] [RSS]
Ontario Power Generation 0 [Ontario Power Generation events] [RSS]
PSEG 0 [PSEG events] [RSS]
Pacific Gas & Electric 0 [Pacific Gas & Electric events] [RSS]
Progress Energy 0 [Progress Energy events] [RSS]
STP Nuclear Operating Company 0 [STP Nuclear Operating Company events] [RSS]
South Carolina Electric & Gas Company 0 [South Carolina Electric & Gas Company events] [RSS]
Southern Nuclear 3 [Southern Nuclear events] [RSS]
Talen Energy 0 [Talen Energy events] [RSS]
Tennessee Valley Authority 2 [Tennessee Valley Authority events] [RSS]
Wolf Creek Nuclear Operating Corporation 0 [Wolf Creek Nuclear Operating Corporation events] [RSS]
Xcel Energy 2 [Xcel Energy events] [RSS]
Électricité de France 0 [Électricité de France events] [RSS]

by site

Site#CompanyEvent lists
Arkansas Nuclear 0 Entergy [Arkansas Nuclear events] [RSS]
Beaver Valley 0 FirstEnergy [Beaver Valley events] [RSS]
Big Rock Point 0 Consumers Power [Big Rock Point events] [RSS]
Braidwood 0 Exelon [Braidwood events] [RSS]
Browns Ferry 1 Tennessee Valley Authority [Browns Ferry events] [RSS]
Brunswick 4 Duke Energy [Brunswick events] [RSS]
Byron 0 Exelon [Byron events] [RSS]
Callaway 1 Ameren [Callaway events] [RSS]
Calvert Cliffs 0 Exelon [Calvert Cliffs events] [RSS]
Catawba 0 Duke Energy [Catawba events] [RSS]
Clinton 0 Exelon [Clinton events] [RSS]
Columbia 0 Energy Northwest [Columbia events] [RSS]
Comanche Peak 0 Luminant [Comanche Peak events] [RSS]
Cook 0 American Electric Power [Cook events] [RSS]
Cooper 0 Entergy [Cooper events] [RSS]
Crystal River 0 Duke Energy [Crystal River events] [RSS]
Davis Besse 0 Cleveland Electric [Davis Besse events] [RSS]
Diablo Canyon 0 Pacific Gas & Electric [Diablo Canyon events] [RSS]
Dresden 0 Exelon [Dresden events] [RSS]
Duane Arnold 0 NextEra Energy [Duane Arnold events] [RSS]
Farley 1 Southern Nuclear [Farley events] [RSS]
Fermi 1 DTE Energy [Fermi events] [RSS]
FitzPatrick 0 Exelon [FitzPatrick events] [RSS]
Fort Calhoun 0 Exelon [Fort Calhoun events] [RSS]
Fort Saint Vrain 0 Xcel Energy [Fort Saint Vrain events] [RSS]
Ginna 0 Exelon [Ginna events] [RSS]
Grand Gulf 0 Entergy [Grand Gulf events] [RSS]
Haddam Neck 0 Connecticut Light & Power Co [Haddam Neck events] [RSS]
Hallam 0 [Hallam events] [RSS]
Harris 1 Duke Energy [Harris events] [RSS]
Hatch 1 Southern Nuclear [Hatch events] [RSS]
Hope Creek 0 PSEG [Hope Creek events] [RSS]
Humboldt Bay 0 Pacific Gas and Electric [Humboldt Bay events] [RSS]
Indian Point 1 Entergy [Indian Point events] [RSS]
Kewaunee 0 Dominion [Kewaunee events] [RSS]
La Crosse 0 Dairyland Power Cooperative [La Crosse events] [RSS]
LaSalle 0 Exelon [LaSalle events] [RSS]
Limerick 0 Exelon [Limerick events] [RSS]
MIT Nuclear Research Reactor 0 [MIT Nuclear Research Reactor events] [RSS]
Maine Yankee 0 Maine Yankee Atomic Power Company [Maine Yankee events] [RSS]
McGuire 0 Duke Energy [McGuire events] [RSS]
Millstone 0 Dominion [Millstone events] [RSS]
Monticello 2 Xcel Energy [Monticello events] [RSS]
Nine Mile Point 1 Exelon [Nine Mile Point events] [RSS]
North Anna 0 Dominion [North Anna events] [RSS]
Oconee 0 Duke Energy [Oconee events] [RSS]
Oyster Creek 1 Exelon [Oyster Creek events] [RSS]
Palisades 0 Entergy [Palisades events] [RSS]
Palo Verde 1 Arizona Public Service [Palo Verde events] [RSS]
Peach Bottom 0 Exelon [Peach Bottom events] [RSS]
Perry 0 FirstEnergy [Perry events] [RSS]
Pilgrim 0 Entergy [Pilgrim events] [RSS]
Point Beach 0 NextEra Energy [Point Beach events] [RSS]
Portsmouth Gaseous Diffusion Plant 0 [Portsmouth Gaseous Diffusion Plant events] [RSS]
Prairie Island 0 Xcel Energy [Prairie Island events] [RSS]
Quad Cities 0 Exelon [Quad Cities events] [RSS]
Rancho Seco 0 Sacramento Municipal Utility District [Rancho Seco events] [RSS]
River Bend 0 Entergy [River Bend events] [RSS]
Robinson 0 Duke Energy [Robinson events] [RSS]
Saint Lucie 0 NextEra Energy [Saint Lucie events] [RSS]
Salem 0 PSEG [Salem events] [RSS]
San Onofre 0 Southern California Edison [San Onofre events] [RSS]
Seabrook 0 NextEra Energy [Seabrook events] [RSS]
Sequoyah 1 Tennessee Valley Authority [Sequoyah events] [RSS]
South Texas 0 STP Nuclear Operating Company [South Texas events] [RSS]
Summer 0 South Carolina Electric & Gas Company [Summer events] [RSS]
Surry 0 Dominion [Surry events] [RSS]
Susquehanna 0 Talen Energy [Susquehanna events] [RSS]
Three Mile Island 0 Exelon [Three Mile Island events] [RSS]
Turkey Point 0 NextEra Energy [Turkey Point events] [RSS]
Vallecitos 0 GEH Hitachi [Vallecitos events] [RSS]
Vermont Yankee 0 Entergy [Vermont Yankee events] [RSS]
Vogtle 1 Southern Nuclear [Vogtle events] [RSS]
Waterford 1 Entergy [Waterford events] [RSS]
Watts Bar 0 Tennessee Valley Authority [Watts Bar events] [RSS]
Wolf Creek 0 Wolf Creek Nuclear Operating Corporation [Wolf Creek events] [RSS]
Zion 0 Exelon [Zion events] [RSS]

by Reactor type

B&W-L-LP 0 [B&W-L-LP events] [RSS]
B&W-R-LP 0 [B&W-R-LP events] [RSS]
CANDU 0 [CANDU events] [RSS]
CANDU-6 0 [CANDU-6 events] [RSS]
CE 0 [CE events] [RSS]
GE-1 0 [GE-1 events] [RSS]
GE-2 0 [GE-2 events] [RSS]
GE-3 0 [GE-3 events] [RSS]
GE-4 0 [GE-4 events] [RSS]
GE-5 0 [GE-5 events] [RSS]
GE-6 0 [GE-6 events] [RSS]
W-AP1000 0 [W-AP1000 events] [RSS]
Westinghouse PWR 2-Loop 0 [Westinghouse PWR 2-Loop events] [RSS]
Westinghouse PWR 3-Loop 0 [Westinghouse PWR 3-Loop events] [RSS]
Westinghouse PWR 4-Loop 0 [Westinghouse PWR 4-Loop events] [RSS]

Recent Notifications

 Entered dateSiteRegionReactor typeSystemScramEvent description
ENS 5400817 April 2019 15:21:00HarrisNRC Region 2At 0812 EDT on 4/17/2019, it was discovered that both sets of turbine trip solenoids were previously unable to actuate within the allowable time frames; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). At the time of discovery, one set of turbine trip solenoids had been restored. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."
ENS 5400517 April 2019 06:25:00CallawayNRC Region 4Service waterAt approximately 0137 CDT, with the Plant (Callaway) in No Mode (Defueled) the "B" Switchyard Bus cleared resulting in a loss of normal power to "A" Train Safety Related Transformer XNB01. This resulted in an under voltage condition on Safety Related Bus NB01. The "A" Emergency Diesel started per design and re-energized Bus NB01. This actuated the shutdown sequencer which first sheds loads including the "A" Spent Fuel Pool Cooling Pump and started "A" Essential Service Water Pump, "A" Component Cooling Water Pump, "A" Control Room A/C and other design loads. No complications were identified. The "A" Switchyard Bus remained energized at all times. The "A" Spent Fuel Pool Cooling Pump was restarted per off normal procedure response at 0149 CDT. Spent Fuel Pool water temperature started at 102 F and rose to 103 F prior to restart. There was no movement of irradiated fuel in progress in the Fuel Building during this time. The plant remains stable in No Mode (Defueled).

At the time of the loss of "B" Switchyard Bus, the plant was closing Generator Output breaker MDV53 to establish a backfeed alignment. Further investigation is in progress. This event is reportable per 10 CFR 50.72(b)(3)(iv)(A).

The NRC Senior Resident has been notified."
ENS 5400216 April 2019 02:59:00FarleyNRC Region 2At 2355 CDT on 4/15/19, life-saving activities by offsite medical personnel for a Farley employee were terminated. The coroner declared the individual deceased at the plant site at 0130 CDT.

The fatality is not believed to be work-related and the individual was inside of the Radiological Controlled Area. This is a four-hour notification, non-emergency for a notification of other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. The licensee will be notifying the Occupational Safety and Health Administration due to the on-site fatality.

The licensee will perform a radiological survey of the individual prior to transportation offsite.
ENS 5400015 April 2019 11:36:00MonticelloNRC Region 3At 0511 CDT on 4/15/2019, transport of a potentially radiologically contaminated person from the Monticello Nuclear Plant to a local hospital was performed prior to conducting a radiological survey as a prudent measure to ensure timely medical support. At 0658 CDT a radiological survey determined that the individual and their clothing were not contaminated. This is reportable under 10 CFR 50.72(b)(3)(xii). The NRC Resident Inspector has been notified."
ENS 5399914 April 2019 06:44:00SequoyahNRC Region 2Feedwater
Steam Generator
Auxiliary Feedwater
Reactor Coolant System
Reactor Protection System
Automatic ScramAt 0320 EDT, April 14, 2019, Sequoyah Unit 1 experienced an automatic reactor trip. The event was initiated by the trip of the 1A main feedwater pump. During the automatic unit runback, an automatic reactor trip was initiated due to low-low level in Steam Generator number 3.

The Auxiliary Feedwater System (AFWS) automatically actuated as required when the expected post-trip feedwater isolation actuated. Reactor Coolant System temperature is being maintained by the AFWS and the steam dump system. During this operational cycle, one control Rod Position Indicator (RPI) for core position E-5 in shutdown bank 'A' has been inoperable, and the appropriate Condition and Required Actions of (Technical Specification Limiting Condition of Operation) 3.1.7 were complied with. Due to this inoperable RPI, the associated shutdown rod is conservatively assumed to be full out and untrippable. Consequently, boration was required to establish adequate shutdown margin. All other Control and Shutdown rods fully inserted. All safety systems responded as designed. No primary or secondary safety valves actuated during or after the reactor trip. The unit is currently stable in Mode 3. Unit 1 is in a normal shutdown electrical alignment. There was no impact on Unit 2. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. There was no impact on the health and safety of the public or plant personnel.

The NRC Senior Resident Inspector has been notified."
ENS 5399814 April 2019 03:21:00Nine Mile PointNRC Region 1High Pressure Coolant Injection
Reactor Pressure Vessel
Feedwater
Automatic ScramOn April 14, 2019 at 0003 (EDT), Nine Mile Point Unit 1 experienced an automatic reactor scram during reactor startup. The cause of the automatic scram was due to high (Reactor Pressure Vessel) pressure following closure of the turbine stop valves. All control rods fully inserted and all plant systems responded per design following the scram.

Following the automatic scram, the High Pressure Coolant Injection (HPCI) System automatically initiated as expected. At Nine Mile Point Unit 1, a HPCI System actuation signal on low Reactor Pressure Vessel (RPV) level is normally received following a reactor scram, due to level shrink. HPCI is a flow control mode of the normal feedwater systems, and is not an Emergency Core Cooling System. At 0004, RPV level was restored above the HPCI System low level actuation set point and the HPCI System initiation signal was reset. Pressure control was established on the Turbine Bypass Valves, the preferred system. No Electromatic Relief Valves actuated due to this scram. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor water level and pressure maintained within normal bands. The offsite grid is stable with no grid restrictions or warnings in effect. The unit is currently implementing post scram recovery procedures.

The NRC Resident Inspector has been notified. The Licensee will notify the State of New York.
ENS 5399713 April 2019 02:04:00MonticelloNRC Region 3Core SprayAt approximately 1815 CDT on April 12, 2019, High Energy Line Break (HELB) Door-410A in the Reactor Building was discovered in the closed position. HELB Door-410B was previously closed for maintenance. Either Door-410A or Door-410B must be open to support the current HELB analyses. With both doors closed, this is considered an unanalyzed condition resulting in the loss of a post-HELB safe shutdown path.

With Door-410A and Door-410B closed, LPCI (Low Pressure Coolant Injection) and Core Spray injection valves in both divisions are no longer considered available. This condition is being reported under 10 CFR 50.72(b)(3)(ii) as an unanalyzed condition that significantly degrades plant safety and 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented the fulfillment of a safety function. The condition was resolved at approximately 1845 CDT on April 12, 2019 when Door-410A was blocked open. The health and safety of the public was not affected by this condition.

The NRC Resident has been notified."
ENS 5399111 April 2019 10:28:00WaterfordNRC Region 4On April 11, 2019, at 0200 CDT the shift operating crew declared the control room envelope inoperable in accordance with Technical Specification (TS) 3.7.6.1 due to the door handle for Door 86 (H&V Airlock Access Door) being detached. Operations entered TS 3.7.6.1 action b, which requires that with one or more control room emergency air filtration trains inoperable due to inoperable control room envelope boundary in MODES 1, 2, 3, or 4, then: 1. Immediately initiate action to implement mitigating actions; 2. Within 24 hours, verify mitigating actions ensure control room envelope occupant exposures to radiological, chemical, and smoke hazards will not exceed limits; and 3. Within 90 days, restore the control room envelope boundary to OPERABLE status. Action b.1 was completed by sealing the hole in Door 86 at 0232 CDT. This event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(D), 'event or condition that could have prevented fulfilment of a safety function of structures or systems that are needed to (D) mitigate the consequences of an accident,' due to the control room envelope being inoperable. The licensee notified the NRC Resident."
ENS 539825 April 2019 12:19:00FermiNRC Region 3A non-licensed employee falsified pre-employment information. The employee's access to the plant has been terminated.
ENS 539774 April 2019 15:16:00Oyster CreekNRC Region 1Oyster Creek NGS (Nuclear Generating Station) Tech Support Center (TSC) ventilation is not functional due to a broken belt on exhaust fan FN-843-14. The TSC ventilation system will remain non-functional until fan belt replacement can be completed.

If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature, radiological, or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff to an alternate location in accordance with applicable site procedures.

This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to partial loss of the TSC. An update will be provided once the TSC ventilation has been restored to normal operation. The NRC Resident Inspector has been notified."
ENS 539681 April 2019 06:40:00Palo VerdeNRC Region 4At 2006 (MST), on 3/31/2019, the Palo Verde Nuclear Generating Station Unit 1 Shift Manager was informed that leakage was measured from the Train A Emergency Core Cooling System (ECCS) piping at approximately 100 ml/minute through a High Pressure Safety Injection (HPSI) A drain valve. This value exceeds the assumed 3000 ml/hour ECCS leakage for a large break loss of coolant accident analysis. At 0230 (MST) on April 1, 2019, the valve was flushed and the leakage reduced to 10 ml/minute (600 ml/hour) and was no longer above the limit of the safety analysis. This condition is being reported as an unanalyzed condition per 10 CFR 50.72(b)3)(ii)(B) and a condition that could have prevented the fulfillment of a safety function to the control the release of radioactive material per 10 CFR

50.72(b)(3)(v)(C). This event did not result in an abnormal release of radioactive material. Notification received by Caty Nolan and emailed to HOO.HOC@NRC.GOV The NRC asked a followup question: Why was the criterion for Control of Radioactive Material selected? per the PVNGS Unit 1 Shift Manager, this criterion was selected due to the potential of exceeding offsite dose projections, post recirculation, following a Design Basis Accident.

The resident inspector has been notified.
ENS 5396731 March 2019 00:17:00VogtleNRC Region 2Main Steam Isolation Valve
Auxiliary Feedwater
Manual ScramAt 2130 (EDT) on March 30, 2019, with Unit 2 in Mode 1 at 30 percent reactor power, the reactor was manually tripped due to a main steam isolation valve failing closed. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam lines through the steam dumps and into the condenser.

The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). Unit 1 was not affected. There was no impact to the health and safety of the public or plant personnel.

The NRC Resident Inspectors have been notified."
ENS 5396630 March 2019 21:06:00BrunswickNRC Region 2Main Turbine
Shutdown Cooling
Main Steam Isolation Valve
High Pressure Coolant Injection
Primary Containment Isolation System
Feedwater
Manual ScramAt 17:47 Eastern Daylight Time (EDT) on March 30, 2019, with Unit 2 in Mode 1 at approximately 23 percent reactor power and main turbine startup in progress coming out of a refuel outage, a high temperature was sensed at main turbine bearing #9. As a result of and to arrest the high temperature condition, the main control room inserted a manual reactor scram.

All control rods inserted as expected during the scram. When the scram was inserted, reactor water level dropped below the Low Level 1 actuation setpoint. Per design, the Low Level 1 signal resulted in 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 main control room manually closed all Main Steam Isolation Valves (MSIVs), in anticipation of a low vacuum prior to the Group 1 automatic closure signal being received. High Pressure Coolant Injection (HPCI) was aligned for pressure control and Reactor Coolant Isolation System (RCIC) was aligned for level control. The Reactor Coolant Sample Line Isolation valves closed as expected on low main condenser vacuum. All systems responded as designed. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation and 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuations of the Primary Containment Isolation System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

At the time of notification, decay heat was being removed by the condenser through one open MSIV and a feedwater pump running.
ENS 5396228 March 2019 20:55:00BrunswickNRC Region 2Primary Containment Isolation System
Main Steam Isolation Valve
Reactor Protection System
At 1654 EDT on March 28, 2019, with Unit 1 in Mode 3 at 0 percent power, an actuation of the Primary Containment Isolation System occurred, closing the outboard Main Steam Isolation Valves (MSIVs) due to a low condenser vacuum signal. The MSIVs had been manually closed, per procedure, during the shutdown evolution to address drywell leakage. The inboard MSIVs had not been reopened when the isolation occurred.

Subsequently, at 1658 EDT a Reactor Protection System (RPS) actuation occurred due to reactor water level dropping below the actuation setpoint. All control rods were inserted at the time of the actuation. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuations of the Primary Containment Isolation System and the Reactor Protection System. There was no impact on the health and safety of the public or plant personnel. The safety function of both the MSIVs and the RPS had already been completed at the time of the event.

The NRC Resident Inspector has been notified."
ENS 5396128 March 2019 15:07:00BrunswickNRC Region 2Reactor Coolant SystemAt 1450 EDT on March 28, 2019, the licensee observed that the Unit 1 unidentified Reactor Coolant System (RCS) leakage was greater than 10 gallons per minute (gpm) for greater than or equal to 15 minutes. The licensee declared an Unusual Event in accordance with their EAL SU 5.1. The licensee initiated a unit shutdown in accordance with their procedures and the unit was approximately 58 percent reactor power at 1507 EDT, with unit shutdown in progress. The licensee also received an alarm due to increasing Drywell Pressure at 1.7 pounds drywell pressure.

At 1600 EDT the licensee called with an update. Unit 1 was still in an Unusual Event with the unit at 37 percent power with the shutdown continuing. Drywell Pressure had decreased to 0.8 pounds. At 1603 the licensee scrammed Unit 1. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * UPDATE ON 3/28/2019 AT 1808 EDT FROM MARK TURKAL TO THOMAS KENDZIA * * *

At 1437 EDT on March 28, 2019, with Unit 1 in Mode 1 at approximately 100 percent power, a Technical Specification-required shutdown was initiated due to indication of a leak in the drywell. Technical Specification Action 3.4.4.A, Unidentified Reactor Coolant System (RCS) leakage increase not within limit, requires RCS leakage to be reduced to within limits within 8 hours. It is expected that the leakage would not have been reduced to within limits within the required Technical Specification completion time; therefore, this event is being reported in accordance with 10 CFR 50.72(b)(2)(i). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 03/29/19 AT 0302 EDT FROM TOM FIENO TO BETHANY CECERE * * *

At 0259 EDT on March 29, 2019, the Unusual Event was terminated because RCS leakage was reduced to less than 10 gallons per minute. The most recent leakage rate measured at 0225 EDT was 3.9 gpm. The source of the leak will be identified when plant conditions allow containment entry. No elevated radiation levels were observed during this event. Drywell pressure is currently 0.0 psig. Unit 1 is in Mode 4. The licensee notified the NRC Resident Inspector.

Notified R2DO (Bonser), NRR EO (Miller), IRD MOC (Grant), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
ENS 5395926 March 2019 16:08:00Browns FerryNRC Region 2On 3/26/2019 at 1030 CDT Engineering evaluation determined that Traversing lncore Probe (TIP) System test results related to Leak Rate Testing of 2-CKV-76-653, TIP Purge Header Check Valve, during the Unit 2 Refueling Outage resulted in a reportable condition. On 3/24/2019 at 1558 CDT, Leak Rate Testing identified a (local leak rate test) LLRT failure of 2-CKV-76-653. The gross leakage Leak Rate value exceeded the Technical Specification allowable value for Type C valves of less than 0.6 (allowable leakage) La.

This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

The short-term corrective actions include repairing the valve such that it passes the test. The valve needs to be repaired before the unit can change modes.
ENS 5395525 March 2019 11:14:00BrunswickNRC Region 2Emergency Diesel GeneratorAt 0402 Eastern Daylight Time (EDT) on March 25, 2019, an actuation of the four Emergency Diesel Generators (EDGs) occurred. At the time of the event, Unit 1 was in Mode 1 at approximately 100% power and Unit 2 was in Mode 4 at 0% power. Unit 2 was in the process of aligning the electrical distribution system to power the emergency buses via the Unit Auxiliary Transformer (UAT) in accordance with plant procedures. It was determined that a fault occurred on the power path between the 230 KV switchyard and the UAT. This caused a main generator differential lockout relay to actuate; thereby starting the EDGs. All emergency buses remained energized from offsite power via the Startup Auxiliary Transformer and, therefore, the EDGs did not tie to their respective buses. The EDGs responded per design to this event. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuation of the EDGs. Due to the shared configuration of the Brunswick electrical system, both Unit 1 and Unit 2 are affected.

The Unit 2 main generator lockout was reset and the EDGs have been restored to standby condition. Troubleshooting activities to determine the cause of the fault are in progress. There was no impact on the health and safety of the public or plant personnel.

The NRC Resident Inspector has been notified."
ENS 5395424 March 2019 17:40:00Indian PointNRC Region 1Auxiliary Feedwater
Steam Generator
Automatic ScramOn March 24, 2019, at 1445 EDT, Indian Point Unit 2 automatically tripped on a turbine trip due to a loss of excitation. All control rods fully inserted and plant equipment responded normally to the unit trip. This RPS (reactor protection system) actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B).

The auxiliary feedwater system actuated following the automatic trip as expected. This specified system actuation is reportable under 10 CFR 50.72(b)(3)(iv)(A). During the event offsite power remained available and stable. No primary or secondary reliefs lifted. Unit 2 is in Mode 3 at normal operating temperature and pressure. Decay heat removal is via the steam generators to the atmospheric steam dumps. No radiation was released. Indian Point Unit 3 was unaffected by this event and remains defueled in a scheduled refueling outage. A post trip investigation is in progress. The licensee has notified the NRC Resident Inspector

The New York State Public Service Commission, Consolidated Edison System Operator, and New York State Independent System Operator were also notified.
ENS 5395324 March 2019 05:23:00HatchNRC Region 2Feedwater
Primary Containment Isolation System
Reactor Protection System
Manual ScramAt 0159 (EDT), with Unit 2 in Mode 1 at 25 percent power, the reactor was manually tripped due to degrading condenser vacuum. After the turbine was tripped, the station service electrical buses did not transfer to alternate supply resulting in loss of the condensate feedwater system and level being controlled by the RCIC system.

Operators responded and stabilized the plant. Reactor water level is being maintained via the RCIC system. Pressure is being controlled and decay heat is being removed by the HPCI system in pressure control mode. Unit 1 is not affected. Additionally, an actuation of the primary containment isolation system occurred during the reactor scram. The reason for the actuation was a group II isolation signal was received on reactor water level and a group I isolation was received on decreasing vacuum. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non- emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, this event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the primary containment isolation system.

There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."
ENS 5394419 March 2019 07:41:00HatchNRC Region 2High Pressure Coolant InjectionAt 0140 (EDT) on 03/19/2019, while the unit was at approximately 1% power and 154 psig pressure in MODE 2, it was discovered that Unit 2 High Pressure Coolant Injection (HPCI) was INOPERABLE. HPCI does not have a redundant system, therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v).

There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Unit 2 entered LCO 3.5.1.c for the HPCI being inoperable.

There is no effect on Unit 1.
ENS 5394217 March 2019 14:10:00Browns FerryNRC Region 2High Pressure Coolant Injection
Reactor Core Isolation Cooling
At 0735 CDT on March 17, 2019, the High Pressure Coolant Injection (HPCI) system was isolated due to a water-side leak from the HPCI Gland Seal Condenser. Unit 3 declared the HPCI system Inoperable and entered Technical Specification LCO 3.5.1 Condition C with required actions to verify the Reactor Core Isolation Cooling system is Operable, and to restore the HPCI system to Operable status within 14 days. All other Unit 3 Emergency Core Cooling Systems (ECCS) remain Operable.

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 5394116 March 2019 13:42:00CooperNRC Region 4At approximately 1100 CDT on March 15, 2019, Cooper Nuclear Station was notified by the National Weather Service that the Shubert radio transmission tower was not functioning due to evacuating their office in Omaha as a result of local flooding. This affects the tone alert radios used to notify the public in event of an emergency condition. Loss of function of this tower is reportable at 1100 CDT on March 16, 2019, when the tower could not be restored within 24 hours of the loss. This condition is reportable under 10 CFR 50.72(b)(3)(xiii). A backup notification method is available and will be utilized for notifications if needed.

A return to service time for the Shubert tower is not currently available.

The NRC Senior Resident Inspector has been informed."
ENS 5393715 March 2019 13:39:00Indian PointNRC Region 1Auxiliary Feedwater
Steam Generator
Automatic ScramOn March 15, 2019 at 1300 EDT, Indian Point Unit 2 automatically tripped offline from mode 1 - 100% power operations. Reactor Operators verified the reactor trip and the plant is currently stable in mode 3. All automatic systems functioned as required. The auxiliary feedwater system actuated following the trip, as expected. All control rods fully inserted upon the trip, as expected.

This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(2)(iv)(B). The unit remains on offsite power in hot standby at normal operating temperature and pressure. Decay heat is being removed from the steam generators via the auxiliary feedwater system and the condensate steam dump valves. Unit 3 remains in mode 6 for a scheduled refueling outage. The licensee notified the NRC Resident Inspector, the local transmission company, and New York State Independent System Operator.

The Indian Point Unit 2 automatic trip was caused by the trip of the main generator. The cause of the generator trip is unknown at this time.
ENS 5393415 March 2019 07:08:00CooperNRC Region 4EN Revision Text: UNUSUAL EVENT DECLARED DUE TO HIGH RIVER LEVEL

At 0546 CDT, Cooper Nuclear Station declared an Unusual Event due to the Missouri River level reaching 899.05 feet above mean sea level (MSL), which is above the Emergency Action Level (EAL) HU 1.5 elevation of 899 feet above MSL. The river is expected to crest above 901 feet above MSL within the next day, and remain above 899 feet above MSL for the next several days. Declaration of an Unusual Event is a 1 hour report, and is reportable under 10 CFR 50.72.a.1.1. Actions are in progress in accordance with site flooding procedure, including strategic placement of barriers at building entrances and important facilities. There is no major plant equipment out of service at this time. Personnel access to the site is not presently impeded and emergency evacuation routes remain available. A press release is planned for this event, which is a four hour report, reportable under 10 CFR 50.72.b.2.11. If the Missouri River were to reach 901.5 feet above MSL, Cooper would initiate a unit shutdown in accordance with their procedures. If the Missouri River were to rise greater than 902 feet above MSL, Cooper will declare an Alert. 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).

  • * * UPDATE AT 1742 EDT ON 3/24/2019 FROM KLINTON BEHRENDS AND CURTIS MARTIN TO JEFFREY WHITED * * *

The licensee terminated the Unusual Event at 1601 CDT due to lowering Missouri River water levels. River water level is currently at 896.0 feet MSL and lowering. A press release will be issued to inform the public of Cooper Nuclear Station's exit from the Notification of Unusual Event regarding high Missouri River level. The initial entry into the Notification of Unusual Event occurred on 03/15/2019 and was exited on 03/24/2019 at 1601 CDT. The press release is reportable per 10 CFR 50.72(b)(2)(xi). The licensee notified the NRC Resident Inspector.

Notified R4DO (O'Keefe), NRR EO (Miller), IRD MOC (Gott), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
ENS 5393314 March 2019 17:08:00Peach BottomNRC Region 1A licensed employee was determined to be under the influence of alcohol during a random test. The employee's access to the plant has been suspended pending an investigation. The licensee notified the NRC Resident Inspector.
ENS 5392912 March 2019 11:21:00Indian PointNRC Region 1A contract employee failed to report for a random fitness for duty test. The contractor's access to the plant has been terminated. The licensee notified the NRC Resident Inspector and the NY Public Service Commission.
ENS 5392411 March 2019 10:00:00Beaver ValleyNRC Region 1Emergency Response Data SystemOn 03/11/2019 planned maintenance activities will be performed on the Beaver Valley Power Station (BVPS) Unit 2 Digital Radiation Monitoring System's (DRMS) Communications System.

The work includes upgrades to the DRMS hardware, software and computer peripherals. Components to be upgraded under this planned maintenance include: Redundant Servers, Operator Console, Health Physics Office Console/Workstation, Printers and Portable Mass Storage/Backup and Computer peripherals necessary to interface with the computer system. This planned upgrade on the Unit 2 DRMS Communications System will result in the loss of Unit 2 radiological monitoring capability in the Control Room and on the Plant Computer System (PCS). Neither the Emergency Response Facilities nor the Emergency Response Data System (ERDS) will receive radiological data. No actual radiation monitors are affected. The scheduled work duration is approximately three weeks. When the DRMS is out of service for the upgrade, compensatory actions will be in place. Radiation monitors will be continuously monitored for any increases in radiation levels. The Unit 2 Shift Manager will be notified of any increase in radiation monitor readings, including exceeding Emergency Action Levels (EAL). This is an eight-hour, non-emergency notification for a Loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the upgrade to the DRMS Communications System will result in the loss of the Unit 2 radiological monitoring capability in the Control Room and to the Unit 2 Plant Computer System that affects the functionality of an Emergency Response Facility. There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

The licensee is currently in a Technical Specification Action Statement, Limiting Condition for Operation for spent fuel pool radiation monitor and leakage rate monitor.
ENS 5392310 March 2019 04:38:00Browns FerryNRC Region 2Primary Containment Isolation System
High Pressure Coolant Injection
Reactor Core Isolation Cooling
Residual Heat Removal
Reactor Protection System
Main Steam Isolation Valve
Core Spray
Feedwater
Emergency Diesel Generator
Automatic ScramAt 2259 CST on 3/9/2019, Browns Ferry Unit-3 received an automatic SCRAM on Main Generator Breaker Failure and Turbine Load Reject. Unit-3 declared a Notification of Unusual Event SU1 for loss of offsite AC power to Unit-3 specific 4kV Shutdown Boards for greater than 15 minutes.

Primary Containment Isolation Systems (PCIS) Groups 1, 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all required components actuated as required. Main steam relief valves lifted on the initial transient. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiated on low reactor water level. HPCI remains in service for reactor level and pressure control. RCIC is not in service at this time, the station is investigating low flow from the pump. All four Unit-3 Diesel Generators started and loaded as expected. Residual Heat Removal System is in service for suppression pool cooling. 4kV Station Unit Boards have been restored from the 161kV system. Actions are in progress to restore 4kV Shutdown Boards to offsite power. This event is reportable within 1 hour in accordance with 10 CFR 50.72(a)(1)(i) for declaration of the Licensees Emergency Plan. Complete as documented on EN 53922. This event requires a 4 hour report 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.' This event also requires an 8 hour report 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), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs), (4) ECCS (Emergency Core Cooling System) for boiling water reactors (BWRs) including: core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system, (5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system, and (8) Emergency AC electrical power systems, including: Emergency diesel generators (EDGs).' The NRC resident inspector has been notified.

As of the event report, the MSIVs were opened and decay heat was being removed via the bypass valves to the condenser.
ENS 5392210 March 2019 00:48:00Browns FerryNRC Region 2Main Steam Isolation Valve
Shutdown Cooling
At 0012 EST on 3/10/2019, Browns Ferry Unit-3 declared an Unusual Event due to a spurious trip of the generator breaker, resulting in a loss of AC power to the 4 kV shutdown boards greater than 15 minutes. All diesel generators started and loaded to supply onsite power. The reactor auto-scrammed, with all rods fully inserting. The Main Steam Isolation Valves opened and shutdown cooling was being conducted via the condenser. The licensee will exit the emergency declaration once offsite power is restored. There is no estimated restart date.

Browns Ferry Unit 1 remains in Mode-1 (100%), Unit 2 remains in Mode-5 for a refueling outage. The NRC Resident Inspector has been notified. This event is related to EN 53923. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * UPDATE ON 3/10/19 AT 1419 EDT FROM JOHN HOLLIDAY TO BETHANY CECERE * * *

At 1310 CDT, Browns Ferry Unit-3 exited the Unusual Event when 161 kV lines were made available. The licensee is executing procedures for securing the diesel generators while alternate offsite power methods are utilized. Switchyard damage evaluation is in progress. The licensee will notify the NRC Resident Inspector. Notified R2DO (Desai), R2RA (Haney), DNRR (Nieh), NRR EO (Miller), and IRD (Grant).

Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
ENS 539219 March 2019 13:59:00Calvert CliffsNRC Region 1A sewage line on the south end of the plant backed up causing sanitary wastewater to flow into storm drains and out to the Chesapeake Bay. This is a required notification of the Maryland Department of the Environment under COMAR (Code of Maryland Regulations) 26.08 for discharge of a pollutant into navigable waters or the adjoining shoreline. The amount has been estimated at less than 1000 gallons and the source has been isolated and storm drains have been covered to stop any flow into them and subsequently to the Chesapeake Bay. This notification is made in accordance with 10CFR50.72(b)(2)(xi) due to notification of a state agency. The licensee notified the NRC Resident Inspector.
ENS 539188 March 2019 15:51:00Browns FerryNRC Region 2Browns Ferry Nuclear Plant (BFN) is notifying state and local agencies of the presence of an oil sheen in the cold water channel. Water from the cold water channel was running into a tunnel that connects to the waters of the US.

BFN Procedure RWI-007, Spill Prevention Control and Countermeasure Plan requires the National Response Center as well as other state and local agencies be notified of any oil sheen on the water. This oil spill is reportable to the EPA (National Response Center) under 40 CFR 112. The notification was made to the National Response Center at 1113 CST under notification number 1239580. The Alabama Emergency Management Agency (AEMA) and Alabama Department of Environmental Management (ADEM) were notified at 1120 CST. This event is reportable as a 4-hour Non-Emergency Notification report in accordance with 10 CFR 50.72(b)(2)(xi) 'Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made.' The licensee has notified the NRC Resident Inspector.

The oil is believed to come from the number one cooling tower basin due to heavy rainfall.
ENS 539115 March 2019 12:46:00BrunswickNRC Region 2Primary Containment Isolation System
Shutdown Cooling
Residual Heat Removal
At 05:35 Eastern Standard Time (EST) on March 5, 2019, with Unit 2 in Mode 5 at 0% power, an actuation of the Primary Containment Isolation System occurred during hydrolazing of the reactor water level variable leg instrumentation line nozzle N011B in the reactor cavity. The hydrolazing activity caused low reactor water level to be sensed on Division II of the shutdown range level instrumentation. Per design, the low level 1 signal resulted in 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.

This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Primary Containment Isolation System. There was no impact on the health and safety of the public or plant personnel. The Group 8 was reset and shutdown cooling was restored at approximately 05:45 EST. The safety significance of this event was minimal. Although there was a brief interruption of the shutdown cooling, the Residual Heat Removal (RHR) shutdown cooling system operation was restored in approximately 10 minutes without extensive troubleshooting or maintenance, and remained operable. The RHR shutdown cooling system is not credited in any Updated Final Safety Analysis Report Chapter 6 or 15 accidents or transients.

The NRC Resident Inspector has been notified."
ENS 539093 March 2019 09:54:00North AnnaNRC Region 2At 0916 EST on March 3, 2019, North Anna Unit 2 declared a Notice of Unusual Event under Emergency Action Level HU 2.1 (fire in/or restricting access to any table H-1 area not extinguished within 15 minutes of control room notification or verification of a control room alarm). At 0906 the control room received a heat sensor alarm for the Unit 2, Reactor Coolant Pump motor cube. The fire brigade was dispatched to the scene where they found no indication of fire, no smoke and no fire damage. There were no actuations associated with the alarm and no redundant indications of fire. There was no effect on plant equipment and no indications of RCS leaks. The site determined that the alarm was invalid and terminated the NOUE. Unit 2 is in a stable condition and in a normal electrical lineup. Offsite support was not requested.

The NRC Resident Inspector, State, and local authorities have been notified by the licensee.

Notified R2RA (Haney), DNRR (Evans), IRD MOC (Grant), R4RDO (Rose), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
ENS 539083 March 2019 00:13:00North AnnaNRC Region 2Reactor Protection System
Steam Generator
Manual ScramOn March 2, 2019 at 2237 EST, North Anna Unit 2 reactor was manually tripped, while operating at approximately 12 percent power, due to degrading vacuum in the main condenser. The unit was in the process of a planned shutdown for refueling when condenser vacuum degraded to greater than 3.5 inches of mercury absolute. The operations crew entered the reactor trip procedure and stabilized the unit in Mode 3 at normal operating pressure and temperature.

All control rods fully inserted into the core following the reactor trip. The reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). There were no ESF system actuations. Decay heat is being removed by the Steam Generator Pressure Operated Relief valves. Unit 2 is in a normal shutdown electrical lineup.

The NRC Resident Inspectors have been notified. The Louisa County Administrator will be notified."
ENS 539062 March 2019 06:12:00Comanche PeakNRC Region 4Feedwater
Auxiliary Feedwater
Steam Generator
At 0317 CST, the Unit 2 Reactor tripped due to Feedwater Isolation Valve (FWIV) 2-04 going closed. All Auxiliary Feedwater Pumps started due to steam generator Lo-Lo levels.

Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IPO-007B. The Emergency Response Guideline Procedure Network has been exited. Decay heat is being rejected to the Main Condenser via the Steam Dump Valves. The cause of the FWIV going closed is currently under investigation. All control rods fully inserted and the reactor trip was uncomplicated. Unit 2 is in a normal post-trip electrical line-up. There was no impact on Unit 1 due to the Unit 2 reactor trip.

The licensee notified the NRC Resident Inspector.
ENS 539031 March 2019 04:03:00LaSalleNRC Region 3Emergency Diesel Generator
Reactor Protection System
On February 28, 2019, at 2217 CST, LaSalle Unit 2 experienced a trip of the 241Y Safety Related Bus during surveillance testing resulting in a valid undervoltage actuation signal to the Common Emergency Diesel Generator ('O' EDG), causing it to start and load to Bus 241Y. The purpose of the surveillance testing was to demonstrate the operability of the breakers necessary to provide the second off site source to Unit 2. This event is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A), as an event that results in a valid actuation of the emergency AC electrical power system. In addition to the 241Y bus trip and 'O' EDG actuation signal, the following plant responses occurred as designed due to the momentary loss of this AC Bus: "A" RPS de-energized due to the loss of the 2A Reactor Protection System Motor-Generator Set, and the running Unit 2 Fuel Pool Cooling pump tripped. The Non-Safety Related Bus 241X de-energized resulting in a trip of the Unit 2 Station Air Compressor. All systems have been restored and troubleshooting is currently in progress. Unit 1 remained in MODE 1 during this event. The NRC Senior Resident Inspector has been notified."
ENS 5389625 February 2019 03:44:00PerryNRC Region 3Reactor Protection SystemAutomatic ScramAt 0024 EST on 2/25/19, with Unit 1 in Mode 1 at 74 percent power, the reactor automatically tripped due to a generator trip. The trip was not complex, with all systems responding normally post-trip.

Operations responded and stabilized the plant. Reactor water level is being maintained via the feed system. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

The generator trip is under investigation, but is believed to be due to grid perturbations.
ENS 5389423 February 2019 19:05:00Grand GulfNRC Region 4Feedwater
Main Turbine
Automatic ScramActuation of RPS (Reactor Protection System) with the reactor critical. Reactor scram occurred at 1458 (CST) on 2/23/2019 from 100% power. The cause of the scram was due to Turbine Control Valve Fast Closure.

All control rods are fully inserted. Currently reactor water level is being maintained by the Condensate Feedwater System in normal band and reactor pressure is being controlled via Main Turbine Bypass valves to the main condenser. No ECCS (Emergency Core Cooling System) initiation signals were reached and no ECCS or Diesel Generator initiation occurred. The Low-Low Set function of the Safety Relief Valves actuated upon turbine trip. This was reset when pressure was established on main turbine bypass valves. The cause of the turbine trip is still under investigation. There were no complications with scram response. The licensee notified the NRC Resident Inspector.

There was no maintenance occurring on the main turbine at the time of the scram.
ENS 5389323 February 2019 09:13:00HatchNRC Region 2Emergency Diesel GeneratorAt 0212 EST on February 23, 2019, with Unit 2 in Mode 5, an actuation signal for the 2C Emergency Diesel Generator (EDG) was received during the Loss of Coolant Accident / Loss of Offsite Power logic system functional test. The 2C EDG was running and tied onto the 2G 4160 emergency bus when the alternate supply breaker was closed as required per the test procedure. Immediately upon closing the alternate supply breaker, both the alternate supply breaker and 2C EDG output breaker tripped open. The 2C EDG output breaker reclosed once the 2G 4160 bus undervoltage relays sensed a deenergized bus. When the 2C EDG tied to the 2G 4160 bus, the bus voltage was noted as being high, and the 2C EDG was secured. Investigation is ongoing to determine the cause of the initial bus undervoltage and the subsequent bus excessive voltage.

This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the emergency AC power system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

This event puts Unit 1 in a 72 hour Limiting Condition for Operation for the 1C Startup Transformer being out of service.
ENS 5388319 February 2019 15:19:00FermiNRC Region 3Secondary containmentOn February 19, 2019, at 1307 EST, with the reactor at 100 percent Core Thermal Power and steady state conditions, plant personnel notified the Main Control Room that both doors in the Secondary Containment Airlock on the Reactor Building Fifth Floor were opened simultaneously for a period of approximately five minutes (i.e., from 1253 to 1258 EST). The failure of this interlock, which is intended to prevent both doors from being opened simultaneously, resulted in the Technical Specification (TS) Surveillance Requirement (SR) 3.6.4.1.3 not being met. The maximum Secondary Containment pressure observed during that time remained within TS limits. There were no radiological releases associated with this event.

Declaring Secondary Containment inoperable as a result of not meeting TS SR 3.6.4.1.3 is reportable under 10 CFR 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.

The repair to the failed interlock is in progress. As a compensatory measure signs are posted on the doors to notify personnel to not access the Reactor Building via those doors.
ENS 5387013 February 2019 09:42:00Grand GulfNRC Region 4On 1/17/2019 at 0619 CST, a non-licensed employee supervisor failed to report to perform a fitness for duty test. The individual's access to the site was terminated. The NRC Resident Inspector will be notified.
ENS 538647 February 2019 17:05:00CallawayNRC Region 4A non-licensed and non-supervisory employee inadvertently possessed and consumed alcohol within the protected area. The employee's access to the plant has been suspended. The Senior Resident Inspector has been notified by the licensee.
ENS 538636 February 2019 12:46:00Indian PointNRC Region 1On February 05, 2019 at approximately 1800 EST, candy that contained alcohol was discovered in the plant protected area. The candy was removed from the protected area by station security management. The licensee notified the NRC Resident Inspector and the State of New York Public Service Commission.
ENS 538616 February 2019 02:22:00Quad CitiesNRC Region 3High Pressure Coolant Injection
Reactor Core Isolation Cooling
On February 5, 2019, at 1804 (CST), during a Unit 1 High Pressure Coolant Injection (HPCI) operability surveillance, a fuse blew in the logic for the motor speed changer for the turbine. The Unit 1 HPCI system was taken out of service for planned maintenance earlier in the day. The fuse issue was not related to any maintenance activities. Had HPCI been demanded, this fuse failure would not have allowed HPCI to reach its required speed. HPCI remains inoperable pending resolution of the issue. The Reactor Core Isolation Cooling (RCIC) system was confirmed operable. There were no other systems inoperable at the time of the event. HPCI had been last successfully tested on November 6, 2018.

This event is being reported as a condition that could have prevented fulfillment of a safety function in accordance with 10 CFR 50.72(b)(3)(v)(D). The HPCI system is a single train system and the loss of HPCI could impact the plant's ability to mitigate the consequences of an accident. The NRC Senior Resident Inspector has been notified.

Inoperable HPCI places the unit in a 14 day Technical Specification Limiting Condition of Operability.
ENS 538604 February 2019 18:40:00ColumbiaNRC Region 4A licensed reactor operator had a confirmed positive random fitness-for-duty drug test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 5385431 January 2019 17:06:00SeabrookNRC Region 1On 1/31/2019, the Medical Review Officer for Seabrook was notified of an error made by the HHS (Health and Human Services) certified lab resulting in a false negative Fitness for Duty quality assurance test result. The test lab, Quest Diagnostics-Norristown, was provided an adulterated sample specimen that was part of a blind performance test, which tested negative. This false negative test result will be investigated and the results reported as required.

This event is being reported in accordance with 10 CFR 26.719(c)(3).

NRC Resident Inspector has been notified."
ENS 5385331 January 2019 10:48:00Prairie IslandNRC Region 3Emergency Diesel GeneratorEN Revision Text: BOTH EMERGENCY DIESEL GENERATORS INOPERABLE DUE TO LOW AIR TEMPERATURE

At 0743 (CST) on 1/31/2019, both trains of Unit 2 Diesel Generators were declared INOPERABLE due to outside air temperature exceeding the low temperature design limit for the diesel engines; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v) for an event or condition that could have prevented the fulfillment of a safety function. The Unit 2 Diesel Generators are still able to start if necessary to provide power. Additionally, multiple layers of defense in depth measures are in place to ensure safety. Prairie Island has five sources of offsite power; all of which are currently available. The Unit 1 Diesel Generators are OPERABLE and capable of being cross-connected to Unit 2. Additional equipment capable of responding to beyond design basis events is available on site providing another layer of defense in depth. Both Unit 2 Diesel Generators were returned to an OPERABLE status at 0810 on 1/31/2019 based on outside air temperature rising above the low temperature design limit with forecasted temperatures to remain above the low temperature design limit. There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The air temperature limit was -30 degrees Fahrenheit. Unit 1 was not affected. The EDGs were supplied by a different manufacturer with different air temperature limits.

  • * * RETRACTION AT 1340 EDT ON 03/22/2019 FROM BRIAN JOHNSON TO JEFFREY WHITED * * *

Engineering analysis performed subsequent to the event notification has determined that both Unit 2 Diesel Generators would have been able to fulfill their safety function during the period of time when the outside air temperature had exceeded the low temperature design limit. Therefore, EN# 53853 is being retracted. The NRC Resident Inspector has been notified of the event notification retraction.

Notified R3DO (McCraw).
ENS 5385231 January 2019 04:23:00SalemNRC Region 1Auxiliary Feedwater
Steam Generator
Reactor Protection System
Manual ScramAt 0301 (EST) on 1/31/19, with Unit 2 in Mode 1 at 100% power, the reactor was manually tripped due to icing conditions requiring the removal of 4 Circulating Water Pumps from service. The trip was not complex, with all systems responding normally post-trip. 21 CFCU (Containment Fan Cooler Unit) was inoperable prior to the event for a planned maintenance window and did not contribute to the cause of the event and did not adversely impact the plant response to the trip. An actuation of the Auxiliary Feedwater System occurred following the manual reactor trip. The reason for the Auxiliary Feed Water System auto-start was due to low level in a steam generator. Operations responded and stabilized the plant. Decay heat is being removed by the Main Steam Dumps and Auxiliary Feedwater System.

Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported 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 Auxiliary Feed Water System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The icing condition was described as frazil ice.

Unit-1 reduced power to 88% because one circulating water pump was shutdown.
ENS 5385130 January 2019 17:41:00DresdenNRC Region 3Secondary containment
Standby Gas Treatment System
At 0910 (CST) on January 30, 2019, the Dresden Station Heater Boiler 'B' tripped while placing the station Heater Boiler 'A' in service. With colder temperatures, the density of the supply air increased and contributed to a greater quantity of air entering the Reactor Building than what was previously being supplied with heating steam in service. The Reactor Building differential pressure (DP) degraded and dropped below 0.25 inches water column vacuum. This condition represents a failure to meet Technical Specification (TS) Surveillance Requirement 3.6.4.1.1. Entry into TS 3.6.4.1 Condition A was made due to Secondary Containment becoming inoperable.

Standby Gas Treatment System was initiated to assist with Reactor Building DP control. Reactor Building DP was restored to greater than 0.25 inches water column vacuum. TS 3.6.4.1 Condition A was exited. This event is being reported under 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 NRC Resident Inspector has been notified."
ENS 5384222 January 2019 14:40:00Wolf CreekNRC Region 4On January 22, 2019, 0723 CST, the Contracts group called and reported to FFD (fitness for duty), that a box of candy was received from a supplier vendor that may contain alcohol. Upon investigation of the candy with the manufacturer, pieces of the candy contain 20-30mg of Alcohol after the chocolate is cooked. The Contract offices are located in the Protected Area and this incident is being reported under 10 CFR 26.719.

No for-cause testing was performed based on the consumption of the candy. No safety related work was performed by the individuals who may have consumed the candy. The box of candy was removed from the site.

The licensee notified the NRC Resident Inspector.
... further results