Emergency Notification System

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[RSS - Reactor Events]

05-11-2018

Region 1 1 [Region 1 Events] [RSS]
Region 2 1 [Region 2 Events] [RSS]
Region 3 0 [Region 3 Events] [RSS]
Region 4 5 [Region 4 Events] [RSS]
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[list]

by company

Ameren 2 [Ameren events] [RSS]
American Electric Power 1 [American Electric Power events] [RSS]
Arizona Public Service 1 [Arizona Public Service events] [RSS]
Cleveland Electric 1 [Cleveland Electric events] [RSS]
Constellation 0 [Constellation events] [RSS]
DTE Energy 1 [DTE Energy events] [RSS]
Dominion 0 [Dominion events] [RSS]
Duke Energy 2 [Duke Energy events] [RSS]
EDF Energy 0 [EDF Energy events] [RSS]
Energy Northwest 1 [Energy Northwest events] [RSS]
Entergy 5 [Entergy events] [RSS]
Exelon 3 [Exelon events] [RSS]
FirstEnergy 0 [FirstEnergy events] [RSS]
Jiangsu Nuclear Power Corporation 0 [Jiangsu Nuclear Power Corporation events] [RSS]
Luminant 2 [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 1 [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 2 [Southern Nuclear events] [RSS]
Talen Energy 0 [Talen Energy events] [RSS]
Tennessee Valley Authority 1 [Tennessee Valley Authority events] [RSS]
Wolf Creek Nuclear Operating Corporation 0 [Wolf Creek Nuclear Operating Corporation events] [RSS]
Xcel Energy 0 [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]
Bellefonte 0 Nuclear Development LLC [Bellefonte events] [RSS]
Big Rock Point 0 Consumers Power [Big Rock Point events] [RSS]
Braidwood 1 Exelon [Braidwood events] [RSS]
Browns Ferry 0 Tennessee Valley Authority [Browns Ferry events] [RSS]
Brunswick 0 Duke Energy [Brunswick events] [RSS]
Byron 0 Exelon [Byron events] [RSS]
Callaway 2 Ameren [Callaway events] [RSS]
Calvert Cliffs 1 Exelon [Calvert Cliffs events] [RSS]
Catawba 0 Duke Energy [Catawba events] [RSS]
Clinton 0 Exelon [Clinton events] [RSS]
Columbia 1 Energy Northwest [Columbia events] [RSS]
Comanche Peak 2 Luminant [Comanche Peak events] [RSS]
Cook 1 American Electric Power [Cook events] [RSS]
Cooper 0 Entergy [Cooper events] [RSS]
Crystal River 0 Duke Energy [Crystal River events] [RSS]
Davis Besse 1 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 2 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 2 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 0 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 0 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]
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 0 Xcel Energy [Monticello events] [RSS]
Nine Mile Point 1 Exelon [Nine Mile Point events] [RSS]
North Anna 0 Dominion [North Anna events] [RSS]
Oconee 1 Duke Energy [Oconee events] [RSS]
Oyster Creek 0 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]
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 2 Entergy [River Bend events] [RSS]
Robinson 0 Duke Energy [Robinson events] [RSS]
Saint Lucie 0 NextEra Energy [Saint Lucie events] [RSS]
Salem 1 PSEG [Salem events] [RSS]
San Onofre 0 Southern California Edison [San Onofre events] [RSS]
Seabrook 0 NextEra Energy [Seabrook events] [RSS]
Sequoyah 0 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]
Vermont Yankee 0 Entergy [Vermont Yankee events] [RSS]
Vogtle 0 Southern Nuclear [Vogtle events] [RSS]
Waterford 1 Entergy [Waterford events] [RSS]
Watts Bar 1 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 1 [B&W-L-LP events] [RSS]
B&W-R-LP 1 [B&W-R-LP events] [RSS]
CANDU 0 [CANDU events] [RSS]
CANDU-6 0 [CANDU-6 events] [RSS]
CE 3 [CE events] [RSS]
GE-2 0 [GE-2 events] [RSS]
GE-3 0 [GE-3 events] [RSS]
GE-4 1 [GE-4 events] [RSS]
GE-5 2 [GE-5 events] [RSS]
GE-6 4 [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 3 [Westinghouse PWR 3-Loop events] [RSS]
Westinghouse PWR 4-Loop 8 [Westinghouse PWR 4-Loop events] [RSS]

Recent Notifications

 Entered dateSiteRegionReactor typeSystemScramEvent description
ENS 5342623 May 2018 18:32:00CallawayNRC Region 4Westinghouse PWR 4-LoopOn May 23, 2018 Callaway determined that a violation of one provision of the site's Fitness for Duty (FFD) policy occurred. FFD pre-access testing confirmed a test failure for alcohol. The violation was committed by a non-licensed supervisory employee. The individual did not hold unescorted access to the plant but did perform behavioral observation program (BOP) duties. The BOP qualification has been removed. The NRC Resident Inspector has been notified by the licensee.
ENS 5342423 May 2018 17:37:00Palo VerdeNRC Region 4CEReactor Protection System
Reactor Coolant System
Automatic ScramThe following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.

On May 23, 2018, at approximately 1128 Mountain Standard Time (MST), the Palo Verde Nuclear Generating Station (PVNGS) Unit 2 control room received reactor protection system alarms for low departure from nucleate boiling ratio and an automatic reactor trip occurred. Prior to the reactor trip, Unit 2 was operating normally at 100 percent power. Plant operators entered the reactor trip procedures and diagnosed an uncomplicated reactor trip. All CEAs (control element assemblies) fully inserted into the core. No emergency classification was required per the PVNGS Emergency Plan. The Unit 2 safety-related electrical buses remained energized from normal offsite power during the event. There was no impact to the required circuits between the offsite transmission network and the onsite Class 1E Electrical Power Distribution System; the offsite power grid is stable. No major equipment was inoperable prior to the event that contributed to the event or complicated operator response. Unit 2 is currently stable in Mode 3 with the reactor coolant system at normal operating temperature and pressure. The cause of the reactor trip is under investigation. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public.

The NRC Resident Inspector has been informed of the Unit 2 reactor trip.

Decay is being removed via steam dumps to condenser. Units 1 and 3 at Palo Verde were unaffected by the transient and continue to operate at 100 percent power.
ENS 5342323 May 2018 16:10:00Comanche PeakNRC Region 4Westinghouse PWR 4-LoopAt time 0848 (CDT), Main Steamline Radiation Monitor 2-RE-2328 (Main Steamline 2-04) lost communications and was declared non-functional.

With this radiation monitor non-functional, all of the emergency action levels for a steam generator tube rupture in steam generator 2-04 could neither be evaluated nor monitored. This unplanned condition is reportable as a loss of assessment capability per 10CFR50.72(b)(3)(xiii). Comanche Peak Nuclear Power Plant (CPNPP) has assurance of steam generator integrity and fuel cladding integrity and there is a negligible safety significance to the current condition from a public health and safety perspective. Additionally, compensatory measures are in place to assure adequate monitoring capability is available to implement the CPNPP emergency plan in the unlikely event of challenges to the steam generator or fuel cladding. The N16 radiation monitor serves as a backup with alarm function and Radiation Protection technicians have been briefed on taking local readings with a Geiger-Mueller tube on MSL 2-04. Corrective actions are being pursued to restore 2-RE-2328 to a functional status.

The NRC Resident Inspector has been notified.
ENS 5341018 May 2018 13:27:00ColumbiaNRC Region 4GE-5Reactor Pressure Vessel
Reactor Protection System
Automatic ScramAt 0651 (PDT) on May 18th, 2018, Columbia Generating station experienced a Main Transformer trip, that caused a Reactor Scram. Reactor Power, Pressure and Level were maintained as expected for this condition. MS-RV-1A (Safety Relief Valve) and MS-RV-1B (Safety Relief Valve) opened on reactor high pressure during the initial transient. MS-RV-1B appeared to remain open after pressure lowered below the reset point. The operating crew removed power supply fuses for MS-RV-1B and it currently indicates intermediate position. SRV (Safety Relief Valve) tail pipe temperatures indicate all valves are closed. Suppression pool level and temperature have remained steady within normal operating levels.

All control rods inserted and reactor power is being maintained subcritical. RPV (Reactor Pressure Vessel) water level is being maintained with condensate and feed system with startup flow control valves in automatic. Reactor Pressure is being maintained with the Turbine Bypass valves controlling in automatic. The main condenser is the heat sink. No ECCS (Emergency Core Cooling Systems) systems actuated or injected; the EOC-RPT (End of Cycle-Recirculation Pump Trip) and RPS (Reactor Protection System) systems actuated causing a trip of the RRC pumps and a reactor scram. Core recirculation is being maintained with RRC-P-1A (Reactor Recirculation Pump) running. No release has occurred. At this time there will be no notifications to state, local or other public agencies. The NRC Senior Resident has been notified. The cause of the event is currently under investigation. Plant conditions are stable.

The plant is in its normal electrical alignment and offsite power is available to the site.
ENS 5340114 May 2018 13:17:00Calvert CliffsNRC Region 1CEOn May 14, 2018, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of a tornado generated missile, Calvert Cliffs identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from a tornado generated missile. A tornado could generate a missile that could strike the Unit 1 Saltwater system header and associated piping. This could result in damage to the unit 1 Saltwater system header which could affect the ability of the Unit 1 Saltwater subsystems to perform their design function if such a tornado would occur.

This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v)(D) for 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 condition is being addressed in accordance with NRC enforcement guidance provided in EGM 15-002 and DSS-ISG-2016-01. Compensatory measures have been implemented in accordance with these documents.

The NRC Resident Inspector has been informed of this notification.
ENS 5339912 May 2018 06:58:00Grand GulfNRC Region 4GE-6Emergency Diesel Generator
Shutdown Cooling
On 5/11/2018, at 2327 hours CDT, with the plant in Mode 5, Grand Gulf Nuclear Station was making preparations for surveillance test 06-OP-1P75-R-0003, Standby Diesel Generator 1 Functional Test. The Grand Gulf Nuclear Station experienced an auto-start of the Division 1 (Emergency) Diesel Generator (EDG) when the 15AA Bus Potential Transformer (PT) fuse drawer was racked out instead of the line PT fuse drawer for Bus 15AA feeder breaker 152-1514. This resulted in the 15AA Incoming Feeder Breaker 152-1511 from Engineered Safety Features Transformer 12 opening, de-energizing the 15AA Bus. The Division 1 EDG started and energized Bus 15AA. The Division 1 LSS SYSTEM FAIL annunciator was received and Standby Service Water A failed to start due to the 15AA Bus PT fuse drawer being racked out. Standby Gas Treatment Train B was manually initiated per the Loss Of AC Power Off Normal Emergency Procedure. Station equipment operated as expected based on the PT fuse drawer that was racked out.

The Division 1 EDG was manually tripped from the Control Room because cooling from the Standby Service Water A was not available. RHR (residual heat removal) B was in Shutdown Cooling (mode) and was verified not affected

The licensee has notified the NRC Resident Inspector.
ENS 5339811 May 2018 15:19:00Watts BarNRC Region 2Westinghouse PWR 4-LoopAt 1011 EDT on May 11, 2018, Containment Shield Building Annulus differential pressure exceeded the required limit. The Shield Building was declared inoperable requiring entry into Technical Specification (TS) 3.6.15 Conditions A and B. The event was initiated by failure of the operating annulus vacuum fan. Main Control Room Operators manually started the stand-by annulus vacuum fan to recover pressure. Shield Building Annulus differential pressure was restored to the required value at 1016 EDT and TS 3.6.15 Condition A and B were exited on May 11, 2018 at 1016 EDT.

The failure mechanism for the annulus vacuum fan is being investigated. The Containment Shield Building ensures the release of radioactive material from the containment atmosphere is restricted to those leakage paths and associated leakage rates assumed in the accident analysis during a Loss of Coolant Accident (LOCA). The Emergency Gas Treatment System (EGTS) would have automatically started and performed its design function to maintain the Shield Building Annulus differential pressure within required limits.

The event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D). The NRC Resident has been notified.
ENS 5339610 May 2018 17:03:00OconeeNRC Region 2B&W-L-LPUnit 3 experienced a loss of AC power while in Mode 6. Power was regained automatically from Keowee via the underground path.

Decay heat removal has been restored. Spent fuel cooling has been restored. Emergency procedures (are) in progress. The Licensee notified the senior NRC resident inspector, State of South Carolina and local authorities. The total loss of 4160 volt AC power was for approximately 30 seconds. The unit is refueled and reactor reassembly complete up to bolting on the reactor head. There was no effect on Units 1 and 2. Notified DHS SWO, FEMA Ops Center, FEMA NWC, DHS NICC, and NuclearSSA

  • * * UPDATE FROM SCOTT HAWKESWORTH TO HOWIE CROUCH AT 0554 EDT ON 5/11/18 * * *

At 0530 EDT, Oconee terminated the notification of unusual event on Unit 3. The basis for termination was that offsite power was restored and the plant is now in its normal shutdown electrical lineup. The licensee has notified Oconee and Pickens counties and will be notifying the NRC Resident Inspector.

Notified R2DO (Ehrhardt), NRR EO (Miller), IRD MOC (Gott), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email) and NuclearSSA (email).
ENS 5339510 May 2018 08:59:00Nine Mile PointNRC Region 1GE-5Emergency Diesel GeneratorAt 0248 (EDT), with the plant shutdown in Mode 4, Nine Mile Point Unit 2 experienced a partial loss of off-site power during relay testing that resulted in an automatic start of the Division 2 Emergency Diesel Generator. All systems responded as expected for the event. The cause is being investigated. The station responded in accordance with appropriate Special Operating Procedures and restored impacted systems.

This event is being reported in accordance with 10CFR50.72(b)(3)(iv)(A) At the time of the report, the emergency diesel generators are loaded and supplying plant safety equipment.

The licensee has notified the state of New York Emergency Management Agency and the NRC Resident Inspector.
ENS 533938 May 2018 10:38:00FarleyNRC Region 2Westinghouse PWR 3-LoopOn May 8, 2018 at 0139 Central Daylight Time, Farley Nuclear Plant Unit 1 declared containment inoperable due to total containment leak rate greater than technical specifications. The 1B containment cooler had seat leakage of approximately 30 gallons per minute from a service water drain valve.

Though the containment cooler service water supply is not tested per the Appendix J program, a loss of the containment barrier is possible under accident conditions. The service water flow path to the 1B containment cooler has been isolated to exit the condition.

The licensee will notify the NRC resident inspector.
ENS 533928 May 2018 01:39:00FarleyNRC Region 2Westinghouse PWR 3-LoopReactor Coolant System
Control Room Emergency Filtration System
On May 7, 2018 at 1041 CDT, Unit 1 performed an RCS (reactor coolant system) leakrate procedure that calculated an unidentified RCS leakrate of 0.202 gpm. The leak source investigation concluded at 2150 that the packing for the charging flow control valve (FCV) was the source of the RCS leakage when it was bypassed, which isolated the leakage. A second RCS leakrate calculation was performed after the charging flow control valve was isolated which calculated an acceptable leakrate of 0.00 gpm.

The packing leakage from the charging flow control valve represented leakage external to containment which would result in a greater that 5 Rem dose projection to control room personnel during accident conditions which does not satisfy the GDC19 criteria described in Technical Specification Bases 3.7.10. Therefore the control room emergency filtration system would not be able to fulfill its design function resulting in an unanalyzed condition. This condition is being reported pursuant to 10CFR50.72(b)(3)(ii) for a 'condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety'. The packing leak from the charging flow control valve will remain isolated until repaired under work order SNC944374.

The NRC Resident Inspector has been notified.
ENS 533897 May 2018 17:40:00WaterfordNRC Region 4CEA non-licensed supervisor had a confirmed positive result for alcohol during a random fitness for duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 533887 May 2018 16:31:00CallawayNRC Region 4Westinghouse PWR 4-LoopReactor Coolant System
Residual Heat Removal
Auxiliary Feedwater
On May 7, 2018, during an engineering review of mission time requirements for Technical Specification related equipment, a deficiency was discovered regarding the Emergency Operating Procedure (EOP) guidance for natural circulation cooldown with a stagnant loop. This condition could be the result of a postulated Main Steam Line Break with a loss of offsite power.

During a natural circulation cooldown with a faulted steam generator, flow in the stagnant reactor coolant system (RCS) loop associated with the isolated faulted steam generator (SG) could stagnate and result in elevated temperatures in that loop. This becomes an issue when RCS depressurization to residual heat removal system (RHR) entry conditions is attempted. The liquid in the stagnant loop will flash to steam and prevent RCS depressurization. In this condition, the time required to complete the cooldown would be sufficiently long that the nitrogen accumulators associated with Callaway's atmospheric steam dumps and turbine driven auxiliary feedwater pump flow control valves would be exhausted. The atmospheric steam dumps and turbine driven auxiliary feedwater pump would not be capable of performing their specified safety functions of cooling the plant to entry conditions for RHR operation. This issue has been analyzed by Westinghouse in WCAP-16632-P. This WCAP determined that to prevent loop stagnation, the RCS cooldown rate in these conditions should be limited to a rate dependent on the temperature differential present in the active loops. The WCAP analysis was used to support a revision to the generic Emergency Response Guideline (ERG) for ES-0.2 "Natural Circulation Cooldown." Figure 1 in ES-0.2 provides a curve of the maximum allowable cooldown rate as a function of active loop temperature differential which is directly proportional to the level of core decay heat. At the time of discovery of this condition, Callaway's EOP structure did not ensure that the ES-0.2 guidance would be implemented for a natural circulation cooldown with a stagnant loop. Callaway has issued interim guidance to the on-shift personnel regarding this concern and is in the process of revising the applicable EOPs. This condition is reportable per 10 CFR 50.72(b)(3)(v) for 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 (A) Shutdown the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) mitigate the consequences of an accident."

The licensee notified the NRC Resident Inspector of this condition.
ENS 533877 May 2018 06:42:00CookNRC Region 3Westinghouse PWR 4-LoopReactor Protection System
Auxiliary Feedwater
Feedwater
Manual ScramOn May 7, 2018 at 0336 (EDT), DC Cook Unit 2 Reactor was manually tripped due to a high-high level experienced in the East Moisture Separator Drain Tank (MSDT) of the Moisture Separator Reheater (MSR).

This notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation as a four (4) hour report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System, as an eight (8) hour report. The NRC Resident Inspector has been notified.

Unit 2 is being supplied by offsite power. All control rods fully inserted. All Aux Feedwater Pumps started properly. Decay heat is being removed via the Steam Generator Power Operated Relief Valves following Main Steam Stop Valve closure at 0431 due to a slow RCS (Reactor Coolant System) cooldown. Preliminary evaluation indicates all plant systems functioned normally following the Reactor Trip. DC Cook Unit 2 remains stable in Mode 3 while conducting the Post Trip Review. No radioactive release is in progress as a result of this event.
ENS 533867 May 2018 05:23:00SalemNRC Region 1Westinghouse PWR 4-LoopAuxiliary Feedwater
Reactor Coolant System
Manual ScramThis 4 and 8 hour notification is being made to report that Salem Unit 2 initiated a manual reactor trip and subsequent automatic Auxiliary Feedwater system actuation. The trip was initiated due to a 21 Reactor Coolant Pump reaching its procedural limit for motor winding temperature of 302F.

Salem Unit 2 is currently stable in Mode 3. Reactor Coolant system pressure is 2235 PSIG and Reactor Coolant System temperature is 547 F with decay heat removal via the Main Steam Dump and Auxiliary Feedwater Systems. Unit 2 has no active shutdown technical specification action statements in effect. All control rods inserted on the reactor trip. All ECCS (emergency core cooling systems) and ESF (emergency safety function) systems functioned as expected. No safety related equipment or major secondary equipment was tagged for maintenance prior to this event. No personnel were injured during this event.

The NRC Resident Inspector was notified. The Lower Alloways Creek Township will be 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 533824 May 2018 13:50:00River BendNRC Region 4GE-6During performance of an extent of condition evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, River Bend Station identified non-conforming conditions in the plant design such that specific TS equipment is considered to not be adequately protected from tornado missiles. The reportable condition is postulated by tornado missiles entering the Diesel Generator Building through conduit and pipe penetrations. A tornado could generate multiple missiles capable of striking Division 1, Division 2, and Division 3 Diesel Generator support equipment rendering all Safety Related Diesel Generators inoperable.

This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v) for 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 (A) Shut down the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) Mitigate the consequences of an accident. This condition was identified as part of an on-going extent of condition review of potential tornado missile related site impacts. Enforcement discretion per Enforcement Guidance Memorandum EGM 15-002 has been implemented and required actions taken. Corrective actions will be documented in a follow-on licensee event report.

The licensee has notified the NRC Resident Inspector.
ENS 533814 May 2018 12:50:00Davis BesseNRC Region 3B&W-R-LPEmergency Diesel GeneratorOn March 8, 2018, an invalid system actuation occurred while preparations were underway to perform Safety Features Actuation System (SFAS) integrated response time surveillance testing during the recent Davis Besse Nuclear Power Station refueling outage. Several minutes after connecting a data recorder to monitor the Emergency Diesel Generator (EDG) 1 start signal, at 1323 hours (EST), the EDG started with no valid actuation signals or test inputs present. The EDG successfully came up to speed and voltage as expected. The associated essential 4160 volt electrical bus remained energized from the normal power supply, therefore, the EDG output breaker did not close to supply power to the bus. Troubleshooting determined the inadvertent actuation was due to a short in the test lead wires at the recorder connection caused by a faulty test lead. The test lead was replaced and the SFAS surveillance testing completed satisfactorily.

This event is being reported as an invalid system actuation per 10 CFR 50.73(a)(2)(iv)(A); this 60-day optional telephone notification is being made per 10 CFR 50.73(a)(i) in lieu of submitting a written Licensee Event Report.

The NRC Resident Inspector was notified of the inadvertent EDG start at the time of the event and has been notified of this invalid specified system actuation notification.
ENS 533803 May 2018 18:40:00Comanche PeakNRC Region 4Westinghouse PWR 4-LoopDuring planned maintenance on Unit 2 Radiation Monitor 2-RE-4270 (Service Water Train B to Discharge Canal Rad Monitor), at 1220 CDT, several other Unit 2 Radiation Monitors that are used for Emergency Action Level evaluation became nonfunctional for about 1 hour. With these radiation monitors non-functional, all of the Emergency Action Levels associated with these monitors could neither be evaluated nor monitored. This unplanned condition is reportable as a loss of assessment capability per 10 CFR 50.72(b)(3)(xiii). A PC11 computer reboot restored the affected radiation monitors to a functional status. The NRC Resident Inspector has been notified.
ENS 533741 May 2018 20:42:00Grand GulfNRC Region 4GE-6Reactor Pressure Vessel
Residual Heat Removal
Shutdown Cooling
At 1551 hrs (CDT) on 5/1/2018, with the plant in Mode 5, a division one Reactor Pressure Vessel (RPV) Level 1 signal was received; however there was no actual change in RPV level. RPV Level remained at High Water Level supporting refuel operations. This caused an actuation of division one Load Shed and Sequencing system that shed and then re-energized the 15 bus. Division one diesel generator started from standby. Residual Heat Removal pump 'A', which was in shutdown cooling mode, was lost during the bus shed, and was re-sequenced upon re-energization of the 15 bus. Upon restoration of shutdown cooling, the RHR pump discharged into the RPV. RCS temperature increased approximately 5 degrees Fahrenheit as a result of the loss of shutdown cooling. The cause of the actuation signal is under investigation.

In accordance with NUREG 1022, Event Reporting Guidelines, this event is conservatively reported under 10 CFR 50.72(b)(2)(iv)(A) as an event that results in emergency core cooling system discharge into the RCS as a result of a valid signal, under 10 CFR 50.72(b)(3)(iv)(B)(8) as an event that results in the actuation of emergency ac electrical power systems, and under 10 CFR 50.72(b)(3)(v)(B) as an event or condition that at the time of discovery could have prevented the fulfillment of a safety function (remove residual heat).

The licensee notified the NRC Resident Inspector.
ENS 5337130 April 2018 14:53:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopAuxiliary Feedwater
Feedwater
Automatic ScramAt 1124 CDT, Braidwood Unit 1 experienced an automatic Reactor Trip. The cause of the Reactor Trip was a Turbine Trip with reactor power greater than P-8. The turbine trip was actuated as a result of a Turbine Motoring Generator Trip. The cause of the generator trip is unknown at this time and is under investigation.

After the Reactor Trip occurred, the 1A Auxiliary Feedwater pump was manually started to provide feedwater flow to all four steam generators. The 1A Auxiliary Feedwater pump was subsequently secured and placed in standby when the Startup Feedwater pump was placed in service. Train A Main Control Room Ventilation Filtration system shifted to Makeup Mode due to a spurious actuation signal. No secondary relief valves lifted and no secondary steam was released as a result of the Reactor Trip. The Main Steam dump valves are in service to the Main Condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by Offsite Power with the Diesel Generators in standby and all safety systems available. There is no impact to Unit 2. This report is being made per 10 CFR 50.72(b)(2)(iv)(B) for a RPS actuation, 4-hr notification, and per 10 CFR 50.72(b)(3)(iv)(A) for a manual actuation of the Auxiliary Feedwater system, 8-hr notification.

The licensee notified the NRC Resident Inspector and Illinois Emergency Management Agency.
ENS 5336626 April 2018 20:23:00HarrisNRC Region 2Westinghouse PWR 3-LoopThis 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 planned maintenance activities were performed on April 23rd through April 25th on the seismic monitoring system without viable compensatory measures established. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5336526 April 2018 18:50:00River BendNRC Region 4GE-6High Pressure Core Spray
Feedwater
River Bend Station experienced an inadvertent initiation and injection of High Pressure Core Spray (HPCS) at 1531 (CDT) on 4/26/2018 while operating at 100 percent power.

During replacement of Level Transmitter B21-LTN081C 'Reactor Vessel Low Water Level 1', Main Control Room received an inadvertent initiation and injection of High Pressure Core Spray. The HPCS injection valve was open for approximately 40 seconds before the operators manually closed the valve. Feedwater Level Control responded per design and maintained Reactor Water Level nominal values. The Division 3 Diesel Generator (DG) also automatically started in response to the actuation signal. The DG did not automatically connect to the Division 3 switchgear since there was not a low voltage condition on the bus. The manual closure of the injection isolation valve caused the system to be incapable of responding to an automatic actuation signal. The manual override of the injection isolation valve was reset approximately 16 minutes after the event, restoring the system to its standby condition. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(A) as a condition that caused ECCS (Emergency Core Cooling System) discharge to RCS (Reactor Coolant System) and 10 CFR 50.72(b)(3)(v)(D) as a condition that caused the loss of function of the HPCS System.

The Senior NRC Resident inspector has been notified.
ENS 5336124 April 2018 14:44:00Diablo CanyonNRC Region 4Westinghouse PWR 4-LoopAt 0357 (PDT), Unit 2 Containment High Range Radiation Monitor RM-31 was declared inoperable due to erratic indication. At this time, Containment High Range Radiation Monitor RM-30 was out of service for routine calibration.

With both containment high range radiation monitors inoperable, this impacted DCPP's (Diablo Canyon Power Plant's) ability to evaluate containment radiation data for an unmonitored release in the event of an emergency. Compensatory measures were promptly put in place with the use of a portable radiation monitor as required by emergency preparedness procedures. This condition is being reported as a loss of assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). Actions are in progress to restore RM-30 and RM-31 to operable status. The NRC Senior Resident Inspector has been notified.

  • * * UPDATE ON 4/24/18 AT 1716 EDT FROM ERIC THOMAS TO DONG PARK * * *

RM-30 was restored to service. Portable radiation monitoring is not required.

The licensee will notify the NRC Resident Inspector. Notified R4DO (Vasquez).
ENS 5335822 April 2018 22:40:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopShutdown Cooling
Residual Heat Removal
Emergency Diesel Generator
On Sunday, April 22, 2018 at 1646 CDT, a valid actuation of Engineered Safety Feature (ESF) Bus 141 Undervoltage (UV) Relay occurred. At the time, Braidwood Station Unit 1 was performing a pre-planned 1A Diesel Generator (DG) Emergency Core Cooling System (ECCS) Actuation Surveillance, initiating the 1A DG to emergency start and sequence loads on a safety injection signal. Following the 1A DG solely supplying electrical power to Bus 141, the 1A DG lost voltage, resulting in an unplanned UV actuation of ESF Bus 141. The 1A DG output breaker was manually opened and local emergency stop of the 1A DG was attempted. The 1A DG continued to run at idle. Fuel supply was secured to the 1A DG and the engine stopped. Subsequently, operators restored power to ESF Bus 141 from the Unit 1 Offsite Power Source. Shutdown cooling was maintained throughout the event as the 1B Residual Heat Removal train was unaffected by the actuation.

This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) for 'Any event or condition that results in valid actuation of any of the systems listed...', specifically 10 CFR 50.72(b)(3)(iv)(B)(8) for the 'Emergency ac electrical power systems, including: emergency diesel generators (EDGs)...'.

The licensee notified the NRC resident inspector.
ENS 5335622 April 2018 04:28:00Watts BarNRC Region 2Westinghouse PWR 4-LoopResidual Heat RemovalOn April 22, 2018 at 0222 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 entered TS (Technical Specifications) LCO (Limiting Condition for Operation) 3.0.3 due to both trains of the Residual Heat Removal System (RHRS) becoming inoperable. During surveillance testing, the gas void values on Emergency Core Cooling System (ECCS) piping common to both trains did not meet acceptance criteria. This caused both RHRS trains to become inoperable. Operations subsequently vented the RHRS to meet the acceptance criteria and exited TS LCO 3.0.3 at 0227 EDT. More frequent surveillances will be conducted to monitor gas void volumes while additional analysis is being performed to determine corrective actions.

The NRC Resident Inspector has been notified.

  • * * RETRACTION FROM TONY PATE TO HOWIE CROUCH ON 5/4/18 AT 1455 EDT * * *

This event is being retracted. The initial report was based on a conservative acceptance criteria for gas accumulation adopted on April 19, 2018 when it was determined that the previously used acceptance criteria for gas accumulation in the ECCS was non-conservative. Additional analysis has subsequently been performed and determined that a higher gas accumulation acceptance criteria does not challenge operability. With a void of less than the acceptance criteria, in the event of ECCS actuation, the system piping support loads will remain within structural limits and the piping system will remain operable. Therefore, both trains of Unit 2 RHRS were operable and the previously reported 10 CFR 50.72(b)(3)(v)(B) event is being retracted. The NRC Resident Inspector staff has been informed of this event retraction.

Notified R2DO (Desai) of this retraction.
ENS 5335522 April 2018 02:34:00Watts BarNRC Region 2Westinghouse PWR 4-LoopResidual Heat RemovalOn April 21, 2018 at 2152 EDT, Watts Bar Nuclear Plant (WBN) Unit 1 entered TS (Technical Specifications) LCO (Limiting Condition for Operation) 3.0.3 due to both trains of the Residual Heat Removal System (RHRS) becoming inoperable. During surveillance testing, the gas void values on Emergency Core Cooling System (ECCS) piping common to both trains did not meet acceptance criteria. This caused both RHRS trains to become inoperable. Operations subsequently vented the RHRS to meet the acceptance criteria and exited TS LCO 3.0.3 at 2222 EDT. More frequent surveillances will be conducted to monitor gas void volumes while additional analysis is being performed to determine corrective actions.

The NRC Resident Inspector has been notified.

  • * * RETRACTION FROM ANTHONY PATE TO DONALD NORWOOD AT 1310 EDT ON 5/9/2018 * * *

This event is being retracted. The initial report was based on a conservative acceptance criteria for gas accumulation adopted on April 19, 2018 when it was determined that the previously used acceptance criteria for gas accumulation in the ECCS was non-conservative. Additional analysis has subsequently been performed and determined that a higher gas accumulation acceptance criteria does not challenge operability. With a void of less than the acceptance criteria, in the event of ECCS actuation, the system piping support loads will remain within structural limits and the piping system will remain operable. Therefore, both trains of Unit 1 RHRS were operable and the previously reported 10 CFR 50.72(b)(3)(v)(B) event is being retracted. The NRC Resident Inspector has been informed of this event retraction.

Notified R2DO (Ehrhardt).
ENS 5335420 April 2018 22:22:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn Friday, April 20, 2018 at 1730 CDT, during the Braidwood Station Unit 1 refueling outage (A1R20), a scheduled ultrasonic test (UT) was performed on the top head to upper center disc weld of the Unit 1 reactor head. The UT identified 19 indications, 9 of which are not acceptable per ASME Section XI, 2001 Edition, 2003 Addenda, Paragraph IWB-3510.

This event is reportable under 10 CFR 50.72(b)(3)(ii)(A) for 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded'.

The licensee notified the NRC Resident Inspector.
ENS 5335320 April 2018 17:57:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopResidual Heat RemovalOn Friday, April 20, 2018 at 1042 CDT, Braidwood Station Unit 1 was at 0 percent power in Mode 6. The 1A Diesel Generator (DG) was inoperable with troubleshooting in progress. The 1B DG was being run for a normal monthly run in accordance with 1 BwOSR 3.8.1.2-2, 'Unit One 1B Diesel Generator Operability Surveillance,' and subsequently tripped. The trip was due to a failure of the overspeed butterfly valve actuator and springs, and not an actual overspeed condition. The unit entered Technical Specification (TS) 3.8.2, 'AC Sources - Shutdown,' Condition B for required DG inoperable. All required TS actions were met at the time of the 1B DG inoperability. The offsite power source remains available. At no time was residual heat removal lost.

This event is reportable under 10 CFR 50.72(b)(3)(v)(B) for 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 remove residual heat.

The licensee has notified 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 5334920 April 2018 00:55:00Watts BarNRC Region 2Westinghouse PWR 4-LoopResidual Heat RemovalOn April 19, 2018 at 1944 EDT, Watts Bar Nuclear Plant (WBN) determined that a preliminary analysis shows current acceptance criteria for gas accumulation in the WBN Unit 1 and Unit 2 Safety Injection System (SIS) and Residual Heat Removal System (RHRS) discharge piping may be non-conservative. The surveillances that check void values and allow venting of the systems are to be performed utilizing conservative criteria at more frequent intervals to ensure gas void volumes remain under acceptable limits. Additional analysis is being performed to determine final actions. The NRC Resident Inspector has been notified.
ENS 5334819 April 2018 23:41:00Indian PointNRC Region 1Westinghouse PWR 4-LoopFeedwaterManual ScramWhile performing a turbine startup, a turbine control anomaly caused a steam generator level transient. The rise in steam generator level above the setpoint caused the turbine to automatically trip. The high steam generator level of 73 percent caused a feedwater isolation signal at 2107 EDT, which also tripped both Main Boiler Feed Pumps. The tripping of the Main Boiler Feed Pumps auto started the motor driven Aux Boiler Feed Pumps 21 and 23.

The reactor was manually tripped at 2108 EDT in accordance with AOP-FW-1 Loss of Main Feedwater. All control rods inserted. Electrical power is being provided from offsite via the Station Aux Transformer. Decay heat removal is being provided via the Atmospheric Dump Valves. An investigation into the cause of the turbine control anomaly is underway. The NRC Resident Inspector has been notified.

The event did not have an affect on Unit 3 and there is no primary to secondary leakage.
ENS 5334719 April 2018 20:00:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopEmergency Diesel Generator
Shutdown Cooling
Residual Heat Removal
On Thursday, April 19, 2018 at 1152 CDT, a valid actuation of Engineered Safety Feature (ESF) Bus 141 Undervoltage (UV) Relay occurred. At the time, Braidwood Station Unit 1 was performing a pre-planned Bus 141 Undervoltage Actuation Surveillance, initiating the 1A Emergency Diesel Generator (EDG) to emergency start and sequence loads on the UV signal. Following the 1A EDG solely supplying electrical power to Bus 141, the EDG lost voltage resulting in an unplanned UV actuation of the ESF Bus 141. Subsequently, operators restored power to ESF Bus 141 via crosstie of the Unit 2 offsite power source. Shutdown cooling was maintained throughout the event as the 1B Residual Heat Removal train was unaffected by the actuation.

This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) for 'Any event or condition that results in valid actuation of any of the systems listed...', specifically 10 CFR 50.72(b)(3)(iv)(B)(8) for the 'Emergency ac electrical power systems, including: emergency diesel generators (EDGs)...'.

The licensee notified the NRC Resident Inspector.
ENS 5334218 April 2018 09:29:00CookNRC Region 3Westinghouse PWR 4-LoopOn 4/18/2018, at approximately 0300 EDT, a contract cleaning employee notified her supervisor that she had found an oven mitt and a bottle containing a liquid that was possibly urine. The bottle had a temperature strip and heating element attached to it. These items were found in the trash in a bathroom in the training center located near the bathroom used for Fitness-for-Duty testing. The supervisor notified Security. Security responded and took possession of the objects. The licensee notified the NRC Resident Inspector.
ENS 5334117 April 2018 16:29:00LimerickNRC Region 1GE-4High Pressure Coolant InjectionUnit 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-4High Pressure Coolant InjectionAt 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-4Emergency Diesel Generator
Reactor Coolant System
Reactor Protection System
Primary containment
Automatic ScramAt 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 5333513 April 2018 21:04:00Grand GulfNRC Region 4GE-6Primary containmentAt 1208 CDT on April 13, 2018, GGNS (Grand Gulf Nuclear Station) identified cracks in the primary containment concrete penetration (outer wall) around feed water line 'B'. There are no available dimensions for crack width or depth until further inspections are performed.

In accordance with NUREG 1022, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73, Section 3.2.4, any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers being seriously degraded, requires that when a principal safety barrier is declared inoperable the condition must be reported under 10 CFR 50.72 (b)(3)(ii). The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION FROM GERRY ELLIS TO HOWIE CROUCH AT 2012 EDT ON 4/15/18 * * *

Grand Gulf Nuclear Station (GGNS) personnel performed an inspection of the wall around feed water line 'B'. This inspection included the protective coating in the identified area and a partial inspection of the underlying concrete. The inspection of the protective coating found a collection of non-linear anomalies, chipping, and flaking. The inspection found non-significant linear indications in the concrete. Grand Gulf Nuclear Station determined that the collection of non-significant coating imperfections and non-significant indications in the concrete do not constitute serious degradation of primary containment. The indications do not adversely impact the operability, mission time, or safety-function (as described per Technical Specification 3.6.1.1, Primary Containment) of the containment structure. The as-found conditions have been entered into the GGNS corrective action program for final disposition. The containment structure is operable, therefore, GGNS is retracting this event notification.

The licensee has notified the NRC Resident Inspector. Notified R4DO (Kellar).
ENS 5333413 April 2018 20:01:00CookNRC Region 3Westinghouse PWR 4-LoopEmergency Diesel GeneratorAt 1555 EDT, the Unit 2 'CD' Emergency Diesel Generator (EDG) automatically started and loaded to 4kV Safeguards bus T21C. Testing was in-progress and the start was unplanned. Unit 2 is currently defueled. Unit 1 remains stable at 100 percent power.

The South Spent Fuel Pit Cooling Train lost power due to a load shed. The South Spent Fuel Pit Cooling Pump was restarted on 2 'CD' EDG at 1614 EDT. The North Spent Fuel Pit Cooling Train remained in-service the entire time. There was no observable change in Spent Fuel Pool temperature. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of an emergency diesel generator, as an eight (8) hour report.

The NRC Resident Inspector has been notified.
ENS 5332913 April 2018 06:07:00OconeeNRC Region 2B&W-L-LPFeedwaterManual ScramOn 4/13/2018 at 0227 (EDT), the Oconee Unit 1 Reactor was manually tripped from 24 percent power due to the inability to control main feedwater flow through the Main Feedwater Control Valves using the Integrated Control System. Due to the RPS actuation while critical, this event is being reported as a 4-hour Non-Emergency per 10 CFR 50.72(b)(2)(iv)(B).

Following the reactor trip, multiple Main Steam Relief Valves failed to reseat at the expected pressure. Using procedure guidance, Main Steam Pressure was lowered by 115 psig, resulting in the closing of all Main Steam Relief Valves. All other post-trip conditions are normal and all other systems performed as expected. Unit 1 is currently in Mode 3 and stable. Decay heat is being removed by the steam generators discharging steam to the main condenser using the turbine bypass valves. Units 2 and 3 are not affected by the Unit 1 reactor trip.

The licensee notified the NRC Resident Inspector.
ENS 5332812 April 2018 17:36:00MillstoneNRC Region 1Westinghouse PWR 4-LoopAt 1148 EDT on April 12, 2018, a 16.2 ounce bottle of Kombucha tea was found in a small refrigerator in the Administration Building inside the Protected Area. The bottle was found to have a small amount missing from the contents. Kombucha tea is a fermented tea containing trace amounts of alcohol, and is legally sold without restrictions. Dominion Energy Nuclear Connecticut had previously notified its workforce that Kombucha tea was prohibited from being consumed or carried onsite. The owner has not yet been determined. This is considered an alcoholic beverage and is being reported pursuant to the requirements of 10 CFR 26.719 as a 24 hour report. The NRC Resident Inspector, the State of Connecticut, and local authorities have been notified.
ENS 5332712 April 2018 12:14:00Watts BarNRC Region 2Westinghouse PWR 4-LoopAuxiliary Feedwater
Reactor Protection System
Automatic ScramAt 0920 EDT on April 12, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 100 percent power. All control and shutdown bank rods inserted properly in response to the automatic reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems.

The cause of the automatic reactor trip is being investigated. The automatic actuation of the Reactor Protection System (RPS) is being reported as a four-hour report under 10 CFR 50.72 (b)(2)(iv)(B). 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). The NRC Senior Resident Inspector has been notified for this event.

The plant is currently stable at normal operating temperature and pressure. The grid is stable and the plant is in its normal shutdown electrical lineup. Unit 1 was unaffected by the Unit 2 trip.
ENS 5332611 April 2018 20:56:00HarrisNRC Region 2Westinghouse PWR 3-LoopOn April 11, 2018, while the Harris Nuclear Plant was shut down for a scheduled refueling outage, the reactor vessel head penetrations were being examined in accordance with the lnservice Inspection Program. Ultrasonic examinations identified a flaw in the head penetration nozzle number 33. The unit is in a safe and stable condition. The flaw will be repaired prior to startup from the refueling outage. The flaw and repair have no impact on the health and safety of the public or station employees. The NRC Resident Inspector has been notified.
ENS 5332411 April 2018 10:14:00River BendNRC Region 4GE-6At time 0150 CDT on April 11, 2018, a condition was identified that could impair the ability of the Control Building Air Conditioning System to perform its design function.

Engineering determined that the time delay relays HVKA11-80YB or HVKA11-80YD (Division II chilled water LOW FLOW relays) could fail in a manner that challenges the design safety function of the Control Building Chilled Water System during a Loss of Offsite Power (LOP) Event. A failure of the time delay relay HVKA11-80YB or HVKA11-80YD (Division II chilled water LOW FLOW relays) to provide the time delay function would cause both the Division I and Division II HVK chilled water pumps to start after a LOP, which in turn could hinder the auto start of either Division I or Division II chillers. Currently the Chilled Water System is otherwise operating as designed. All operator actions are in place to ensure the plant meets all required designed safety system functions. Work is currently underway to correct this design vulnerability.

The NRC Resident Inspector has been notified of this condition.
ENS 5332311 April 2018 09:09:00Three Mile IslandNRC Region 1B&W-L-LPOn April 11, 2018, a hydrazine spill resulted in measurable hydrazine levels released at the station outfall to the Susquehanna River over an approximately 16 minute time period. The hydrazine levels exceeded the station NPDES (National Pollutant Discharge Elimination System) permit effluent limitations. The Industrial Waste Treatment system release to the river was secured and no further hydrazine was released to the river.

The concentrations released did not threaten the downstream users or the environment. Pennsylvania Department of Environmental Protection was notified of the NPDES non-compliance on April 11, 2018 at 0542 EDT. Pursuant to 10 CFR 50.72(b)(2)(xi), this notification satisfies the requirement to notify the NRC of the occurrence of any event or situation related to the health and safety of the public or onsite personnel, or protection of the environment, for which notification to other government agencies has been made.

The NRC Resident Inspector has been notified.
ENS 533197 April 2018 12:10:00BrunswickNRC Region 2GE-4Reactor Protection System
Feedwater
Shutdown Cooling
Primary Containment Isolation System
Automatic ScramOn 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 533187 April 2018 11:59:00HarrisNRC Region 2Westinghouse PWR 3-LoopAuxiliary Feedwater
Feedwater
On April 7, 2018 at 0451 EDT, with Unit 1 in Mode 3 at 0 percent power, an auto actuation of 'A' and 'B' Motor Driven Auxiliary Feedwater (MDAFW) pumps occurred during the shutdown of Unit 1 for Harris Nuclear Plant's refueling outage. Plant Operators successfully took control of the AFW flow and noted the 'B' Main Feed pump was still running with proper suction and discharge pressures of 430 lbs. and 1000 lbs.

The 'A' and 'B' Motor Driven Auxiliary Feedwater (MDAFW) pumps automatically started as designed when the 'Loss of Both Main Feedwater Pumps' signal was received. The cause of the actuation is still being evaluated. 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 Auxiliary Feedwater system.

There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 533175 April 2018 18:23:00Grand GulfNRC Region 4GE-6Secondary containmentOn Thursday, April 5, 2018, at approximately 1117 hours Central Daylight Time, Entergy contract personnel opened the personnel hatch allowing access to the roof of the Secondary Containment Building for the purposes of performing an inspection of various items located on the roof.

During the time period the individuals were on the roof, the hatch was left open. An individual was adjacent to the door with a radio and had constant communication link with the control room operator. Pursuant 10 CFR 50.72(b)(3)(v)(C), and 10 CFR 50.72(b)(3)(v)(D) this event is being reported as an event or condition that could have prevented the fulfillment of a safety function. Because the site had an individual briefed and at the door in constant communications with the control room to close the hatch if condition required such an action, this event is not viewed as an actual loss of safety function.

The NRC Resident Inspector was notified.
ENS 533103 April 2018 02:53:00SusquehannaNRC Region 1GE-4Secondary containmentOn 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 533092 April 2018 18:33:00VogtleNRC Region 2W-AP1000A contractor employee supervisor had a confirmed positive for illegal drugs during a fitness for duty test. The employee's access to the site has been terminated. The NRC Resident Inspector has been notified.
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