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 Entered dateSiteRegionReactor typeEvent description
ENS 5330425 February 2020 10:00:00Grand GulfNRC Region 4GE-6At 0206 (CDT) on March 31, 2018, with the plant in Mode 1 at 100% rated core thermal power, Grand Gulf Nuclear Station experienced a loss of Secondary Containment. During the performance of a Standby Gas Treatment System (SGTS) drawn down test with Auxiliary Building train bay door (1A319A) as the secondary containment boundary, Grand Gulf was unable to maintain secondary containment pressure, as required by SR (surveillance requirement) 3.6.4.1.4, greater than or equal to 0.266 inches of water vacuum for 1 hour. Following initial vacuum draw down, secondary containment pressure degraded to 0.225 inches of water vacuum with operators in the field reporting air leakage from door 1A319A. The test was secured and Secondary Containment was declared inoperable and Technical Specification 3.6.1.4 A.1 was entered. Following completion of the failed surveillance test, Secondary Containment was returned to an operable status at 0315 hours on March 31, 2018, by returning the system to a previously known operable configuration by closing doors 1A310, 1A312 and 1A319. This is being report under 10 CFR 50.72(b)(3)(v)(C). The licensee has notified the NRC Resident Inspector.
ENS 5186225 February 2020 09:41:00PilgrimNRC Region 1GE-3On April 12, 2016, with the reactor at 100 percent power and the mode switch in RUN, Pilgrim Nuclear Power Station entered an unplanned 24-hour Limiting Condition for Operation (LCO) action statement due to both emergency diesel generators (EDG) being inoperable (Technical Specification 3.5.F.1). At 0050 (EDT) this morning, with EDG B out of service for a planned LCO maintenance window, EDG A was declared inoperable due to a 130 drop per minute leak on a line to a jacket water pressure indicator. Repairs to EDG A are underway at this time. The following plant equipment has been verified operable: both 345 Kv transmission lines; 23kV transmission line; Station Blackout EDG. This condition is reportable to the NRC Staff as an Event or Condition that Could Have prevented Fulfillment of a Safety Function (Mitigate the consequences of an accident) under 10 CFR 50.72(b)(3)(v)(D), and requires an 8-hour notification. The licensee has notified the NRC Senior Resident Inspector. The licensee will notify the Commonwealth of Massachusetts.
ENS 5079525 February 2020 09:04:00Grand GulfNRC Region 4GE-6A reactor SCRAM occurred at 1856 CST on 2/7/15 from 100 percent core thermal power. The cause of the SCRAM appears to be a Generator/Turbine trip, but it is still under investigation. Appropriate off-normal event procedures were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF power occurred. No ECCS initiation signals were reached, and no ECCS or Emergency Diesel Generator initiations occurred. Main Steam Isolation Valves remained open and Safety Relief Valves lifted and reseated as designed. Currently, reactor water level is being maintained by the Condensate and Feedwater system in normal band, and reactor pressure is being controlled via turbine bypass valves to the main condenser. Following the reactor SCRAM, all rods fully inserted and all systems functioned as expected. The plant is in a normal electrical lineup. The licensee has notified the NRC Resident Inspector.
ENS 5077125 February 2020 08:52:00PilgrimNRC Region 1GE-3On Tuesday, January 27, 2015, at 0948 EST, with the Reactor Mode Select Switch (RMSS) in the Shutdown position and Pilgrim Nuclear Power Station (PNPS) at 0% core thermal power, the High Pressure Coolant Injection (HPCI) system was isolated by the main control room operating crew and declared INOPERABLE. HPCI had been in service for reactor pressure control following the automatic reactor scram experienced during winter storm 'Juno' reported in EN# 50769. It appears there was a malfunction of the HPCI turbine gland seal condenser blower or associated condensate pump. Reactor pressure control was transitioned to the safety relief valves and the reactor cooldown was continued. The plant is stable. The Emergency Diesel Generators are powering the safety related 4KV buses and reactor water level is being maintained by the Reactor Core Isolation Cooling (RCIC) system. HPCI is required to be OPERABLE in accordance with Technical Specification 3.5.C.1. Since HPCI is a single train system, the INOPERABILITY is reportable in accordance with 10CFR50.72(b)(3)(v)(D). The cause of the HPCI malfunction is not known at this time and troubleshooting continues. This event had no impact on the health and/or safety of the public. The USNRC Senior Resident Inspector has been notified. Shutdown cooling is in service.
ENS 5134025 February 2020 08:52:00Arkansas NuclearNRC Region 4CE
B&W-L-LP
This notification is being made in accordance with 10 CFR 50 72(b)(3)(xiii) as an event that will result in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g. a significant portion of control room indication, Emergency Notification System or offsite notification system.) The emergency preparedness plan requires seismic monitoring instruments to diagnose an earthquake for emergency actions levels (EAL) HU6 (Natural or destructive phenomena affecting protected area) and HA6 (Natural or destructive phenomena affecting vital areas). At 1020 CDT on August 24, 2015 the Semi-Annual Seismic System Functional Test commenced. While this test is in progress, seismic alarm capability is not available for EAL declaration purposes. ANO procedures provide compensatory measures of using offsite sources to obtain seismic data. It should be noted that seismic data will still remain capable of being recorded, only alarm capability is lost. The licensee notified the NRC Resident Inspector.
ENS 5103325 February 2020 08:52:00PalisadesNRC Region 3CE

At 1241 EDT, Operations staff at Palisades declared an Unusual Event under EAL HU1.1 due to seismic activity felt on site. No seismic alarms were initiated. No plant equipment was affected. The epicenter of the 4.2 magnitude earthquake was located south of Galesburg, MI. Palisades continues to operate at 100% power. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM JC RANEY TO DANIEL MILLS AT 1601 EDT ON 5/2/15 * * *

The licensee terminated the Unusual Event at 1541 EDT on 5/2/15. The licensee has notified the NRC Resident Inspector and the state and local government. Notified R3DO (Orlikowski), IRD MOC (Stapleton), NRR EO (Morris), NRR ET (Dean), and R3RA (Pederson). Notified other Federal Agencies (DHS SWO, FEMA Ops, FEMA NWC, NICC Watch Officer and NuclearSSA).

ENS 4477525 February 2020 04:51:00Arkansas NuclearNRC Region 4CEUnit 2 performed a controlled shutdown from 60% power to make repairs to a low pressure feedwater heater. The unit was taken off line at 1148 today. The unit is stable in Mode 3, Hot Standby. A press release was made at 1430 today. The licensee notified the NRC Resident Inspector.
ENS 4467225 February 2020 04:35:00PilgrimNRC Region 1GE-3The High Pressure Coolant Injection (HPCI) system was declared inoperable on 11/20/08 at 1657 EST due to a Group 4 isolation signal generated during scheduled surveillance testing in accordance with PNPS (Pilgrim Nuclear Power Station) Procedure 8.M.2-2.5.3, Attachment 1. The HPCI testing was stopped to determine the cause of the isolation which was not part of the planned evolution. HPCI isolation was reset and HPCI was restored to standby lineup at 1804 EST. This event is an eight-hour notification. Efforts are ongoing to determine the cause of the error during testing. This event had no adverse effect to the health and/or safety of the public. The licensee has notified the NRC Resident Inspector of this event. This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(D) due to the loss of a single train system required to mitigate the consequences of an accident.
ENS 544255 December 2019 16:03:00CooperNRC Region 4The following was received via email from Cooper Nuclear Station: At 0810 (CST), on 12/5/19, Operations personnel discovered BLDG-DOOR-R209, FIRE DOOR BETWEEN CRITICAL SWITCHGEAR ROOMS F & G, was unlatched. The door was immediately latched upon discovery. Based on door logs, the door separating the two critical switchgear rooms was inadvertently left unlatched for approximately 5 minutes. This door is a Steam Exclusion Boundary (SEB) door. It is required to be closed and latched when the Auxiliary Steam Boiler is in service due to Auxiliary Steam piping passing through Critical Switchgear Room 'G'. If a steam line break was to occur with the door unlatched, steam could render both Critical Switchgear busses inoperable. This is being reported under 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition, and 10 CFR 50.72(b)(3)(v), 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 (B) remove residual heat and to (D) mitigate consequences of an accident. There was no impact on the health and safety of the public or plant personnel. The door closes automatically and appeared to have been left unlatched by the last person passing through. The door was tested and latches as required. The licensee notified the NRC Resident Inspector.
ENS 543716 November 2019 01:17:00Grand GulfNRC Region 4

On November 5, 2019 at 1811 CST, station service water A and the Division 1 diesel generator (DG) were declared inoperable based on the results of an engineering evaluation of a Class 3 piping leak. This was determined to be a potential inability to fulfill a safety function due to concurrent inoperability of two emergency diesel generators. Division 3 DG was inoperable due to planned maintenance on November 4, 2019 at 0000 CST. This event is being reported an 8-hour non-emergency notification per 10 CFR 50.72 (b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of a safety function (Accident Mitigation). Division 3 DG and high pressure core spray have been restored, and the fulfillment of the accident mitigation safety function has been restored. The NRC Senior Resident Inspector has been notified.

  • * * RETRACTION ON 11/11/19 AT 1739 EST FROM GABRIEL HARGROVE TO BETHANY CECERE * * *

This was initially reported under 10 CFR 50.72(b)(3)(v)(D). However, subsequent engineering evaluation determined that the condition did not affect safety system operability. The evaluation determined that the leakage was within allowable limits and piping structural integrity was not challenged at this time nor in the past three years. The Division 1 DG and SSW A were at the time of discovery OPERABLE and EN54371 is being retracted. The licensee will notify the NRC Resident Inspector. Notified R4DO (Drake).

ENS 5434824 October 2019 16:25:00River BendNRC Region 4At 1035 CDT the Automatic Depressurization System (ADS) was rendered inoperable due to the failure of the 'A' Safety Vent Valve (SVV) Compressor (SVV-C4A) to manually start with SVV-C4B tagged out. System pressure slowly dropped below 131 psig (normal pressure is 165 psig). This caused the ADS safety relief valves to be declared inoperable. The station entered Technical Specification 3.5.1 Condition G. The Required Action was to be in Mode 3 in 12 hours. As a result, the station was in a condition that could have prevented the fulfillment of a safety function. The breaker for SVV-C4B was reset and the clearance for SVV-V4B was released. System pressure was restored to greater than 131 psig at 1116 CDT which allowed exit of the action statement to be in Mode 3 in 12 hours. System parameters are currently stable in the normal pressure range. Investigation for the cause of the system failure is ongoing. No radiological releases have occurred due to this event from the unit. The licensee notified the NRC Resident Inspector.
ENS 5433920 October 2019 19:21:00Arkansas NuclearNRC Region 4

At 1030 CDT, it was discovered that the loop seal on the condensate drain was empty for VUC-9 Control Room AC Unit. This creates a breach in the Control Room envelope. Unit 2 entered (Technical Specification) T.S. 3.7.6.1 Action D. Unit 1 is in Mode 6; therefore, not in a mode of applicability. Compensatory action were being performed and the licensee was in the process of sealing the loop. The licensee will notify the NRC Resident Inspector.

  • * * RETRACTION FROM DONNA BOYD TO DONALD NORWOOD AT 1336 EDT ON 10/24/2019 * * *

This report is being retracted. The Control Room Envelope (CRE) provides a safety function which limits radiological dose to occupants to no more than 5 rem for 30 days post-accident. The dose limitation assumes the occupants are stationed within the CRE 24 hours a day for the entire 30-day period. The CRE also functions to protect occupants from potential hazards such as smoke or toxic chemicals. The CRE is declared inoperable when a potential breach is identified, regardless of the ability to seal the breach. With respect to the event of October 20, 2019, the water level in a loop seal could not be maintained at the desired level. Subsequent evaluation determined that sufficient water was maintained in the loop seal to prevent a breach of the CRE. The subject reporting criterion is based on the assumption that safety-related systems, structures, and components (SSCs) may no longer be capable of mitigating the consequences of an accident. In accordance with NUREG 1022, 'Event Report Guidelines 10 CFR 50.72 and 50.73,' a report may be retracted based on a revised operability determination. The CRE remained operable; therefore, this report may be retracted. The licensee notified the NRC Resident Inspector. Notified R4DO (Young).

ENS 5433818 October 2019 10:45:00River BendNRC Region 4

EN Revision Text: INADVERTENT OPENING OF MAIN TURBINE BYPASS VALVES POTENTIONALLY AFFECTED SAFE SHUTDOWN CAPABILITY At 0207 (CDT), the Bypass Electro-Hydraulic Control (EHC) system was secured for planned maintenance. When the Bypass EHC pumps were secured, both of the Main Turbine Bypass Valves unexpectedly opened to approximately 4.5 percent. Plant parameters indicated no impact to Turbine Control Valve position, Reactor Pressure, Turbine First Stage Pressure, or Main Steam Line flows. There were no other abnormal indications noted. With the Turbine Bypass Valves partially open, there is a potential to affect instrumentation that trips on high Turbine First Stage Pressure. Therefore, this event is being reported as a potential loss of Safety Function. At 0256, the Bypass EHC system pumps were restored and the Turbine Bypass Valves Closed. No radiological releases have occurred due to this event from the unit. The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION FROM THONG LE TO HOWIE CROUCH AT 1019 EST ON 11/19/19 * * *

This Event Notification was contingent on the Main Turbine Bypass Valves opening which resulted in the inoperability of Turbine First Stage Pressure monitoring instrumentation. A detailed review of system design and plant parameter trends has confirmed that the Main Turbine Bypass Valves remained closed for the duration of the event, permitting the instrumentation systems dependent on accurate Turbine First Stage Pressure to perform their respective design and licensing basis functions. Valve drift in the open direction was observed by position indication when hydraulic control pressure was removed. However, the valves were at an over-travel closed position prior to the event allowing the valves to settle at a position where an internal spring could provide closing force to the valve disc. Multiple plant parameter trends including Turbine First Stage Pressure, Reactor Pressure, Main Steam Line flows, and Main Turbine Bypass Valve discharge line temperatures indicate that the Main Turbine Bypass Valves remained closed for the duration of the event. The licensee has notified the NRC Resident Inspector. Notified R4DO (O'Keefe).

ENS 5427112 September 2019 00:49:00Grand GulfNRC Region 4On September 11, 2019 at 1719 CDT, plant personnel identified a condition in which the 208 foot elevation inner primary containment airlock door was not in its fully seated and latched position while the 208 foot elevation outer primary containment airlock door was opened. The 208 foot elevation outer containment airlock door was subsequently closed by the individual exiting the area. The time that both 208 foot elevation containment airlock doors were not in their fully seated and latched positions was less than 1 minute. Following this occurrence, maintenance personnel inspected the 208 foot elevation inner containment airlock door and re-positioned this door to its fully seated and latched position. There was no radioactive release as a result of this event. This condition requires an 8-hour non-emergency notification in accordance with 10CFR50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material. The NRC Resident Inspector has been notified.
ENS 5424428 August 2019 19:10:00Grand GulfNRC Region 4On Wednesday, August 28, 2019, at 1316 CDT, Grand Gulf Nuclear Station experienced a power loss to the Control Room High Pressure Core Spray (HPCS) Instrumentation Panel due to an internal inverter failure. The power loss caused the loss of the HPCS System (a single train system). The minimum flow valve (a Primary Containment Isolation Valve) for HPCS opened due to this power loss as well. This valve was manually closed in response to this, and the outboard isolation requirement for the associated penetration (which) is closed (for the) system remained intact throughout this event. No other accident mitigation systems were affected by this event. The cause of this event is under investigation at this time. The NRC Resident Inspectors were notified. This Condition is an 8-hour reportable condition as an event or condition that could have prevented the fulfillment of a safety function, in accordance with 10 CFR 50.72(b)(3)(v)(D).
ENS 542015 August 2019 17:06:00Grand GulfNRC Region 4On August 5, 2019, at 0936 CDT, Grand Gulf entered Technical Specification (TS) 3.6.4.1 due to a Secondary Containment personnel door, 1A401B, not being able to meet its design function. Door 1A401B was unable to be closed and latched. This condition is being reported as a loss of safety function. The station also entered 05-S-01-EP-4, Auxiliary Building Control (Secondary Containment) to address Auxiliary Building differential pressure due to the opened Secondary Containment penetration. Actions were taken to close and latch Door 1A401B. Secondary Containment has been declared operable. TS 3.6.4.1 and 05-S-01-EP-4 were exited. The NRC Resident Inspector was notified of the condition.
ENS 5419131 July 2019 16:20:00WaterfordNRC Region 4On July 31, 2019, at 1206 CDT, Waterford 3 commenced initiation of a plant shutdown as required by Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.3. Prior to this, on July 31, 2019, at 1108 CDT, the boron injection flow paths were declared inoperable in accordance with LCO 3.1.2.2, 'Flow Paths - Operating,' and the charging pumps were declared inoperable in accordance with LCO 3.1.2.4, 'Charging Pumps-Operating.' This was due to visual examination identifying that propagation had progressed on a previously identified flaw on piping upstream of the header supplying the charging pumps. TS LCO 3.0.3 was entered due to the action statements of LCOs 3.1.2.2 and 3.1.2.4 not being met. LCO 3.0.3 requires that action shall be initiated within one hour to place the unit in a mode in which the specification does not apply by placing it in hot standby within the next 6 hours and cold shutdown within the next 30 hours. At 1206 CDT, Waterford 3 commenced direct boration to the reactor coolant system. This condition meets the reporting criteria of 10 CFR 50.72(b)(2)(i) due to the initiation of plant shutdown required by Technical Specifications and 10 CFR 50.72(b)(3)(v)(A) and (D) due to an event or condition that could have prevented fulfillment of a safety function of structures or systems that are needed to (A) shutdown the reactor and maintain it in a safe shutdown condition and (D) mitigate the consequences of an accident.
ENS 541528 July 2019 18:40:00Arkansas NuclearNRC Region 4A non-licensed contract supervisor had a confirmed positive for a controlled substance during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 541473 July 2019 18:32:00Arkansas NuclearNRC Region 4This 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to provide information pertaining to an invalid Engineered Safety Feature actuation signal. On May 9, 2019, at Arkansas Nuclear One (ANO) Unit 1, while performing an Emergency Feedwater Initiation and Control (EFIC) Channel B monthly test, a test pushbutton was mispositioned, resulting in an inadvertent initiation of the Emergency Feedwater (EFW) System. In accordance with the Engineered Safeguards Actuation System (ESAS) Trip Test portion of the surveillance, the first technician placed EFIC Train B in the tripped condition. The second technician then went to the front of the control room to verify Remote Switch Matrix (RSM) indications. The first technician recalls thinking he was given the order to reset Train B EFW Bus 1 Trip. Therefore, the first technician performed the step using three-part communication, but there is uncertainty about what was said. Due to the amount of time the second technician spent in front of the control room, the first technician assumed Operations reset the RSM to complete the Train B reset. The second technician returned to the ESAS cabinet and directed the first technician to perform the reset of Train B EFW Bus 1 Trip. The first technician, expecting his next action to be the trip of Train B EFW Bus 2, placed Bus 2 in the tripped condition. This put both buses of Train B EFW in trip and caused the actuation of P-7A EFW Pump. This inadvertent actuation was caused by human error and was not a valid signal resulting from parameter inputs. The 1992 Statements of Consideration define an invalid signal to include human error. Therefore, this actuation is considered invalid. This event was entered into ANO's corrective action program for resolution. This event did not result in any adverse impact to the health and safety of the public. The plant responded as expected. In accordance with 10 CFR 50.73(a)(i) a telephone notification is being made in lieu of submitting a written Licensee Event Report. The licensee has notified the NRC Resident Inspector.
ENS 5413125 June 2019 11:09:00WaterfordNRC Region 4On June 25, 2019, at 0428 CDT, the Waterford 3 shift operating crew declared the control room envelope inoperable in accordance with Technical Specification (TS) 3.7.6.1 due to both Broad Range Gas Monitors being inoperable. 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 placing the control room in isolate mode at time 0441 CDT. This event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(A) and 10 CFR 50.72(b)(3)(v)(D), event or condition that could have prevented fulfillment of a safety function of structures or systems that are needed to (A) shutdown the reactor and maintain it in shutdown condition and (D) mitigate the consequences of an accident, due to the control room envelope being inoperable. The NRC Senior Resident Inspector has been notified.
ENS 5412117 June 2019 12:56:00River BendNRC Region 4This 60-day telephone notification is being made in accordance with the reporting requirements specified by 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of a general containment isolation signal affecting multiple systems. On April 30, 2019, at approximately 0650 CDT, a level 2 containment isolation signal was introduced when a fuse for the Nuclear Steam Supply Shutoff System was removed for a maintenance clearance. The level 2 containment isolation signal caused a trip of the Division I DC bus back-up charger, leaving only the Division I battery to carry the DC bus. At 0707 CDT the bus was de-energized when another unrelated clearance opened the battery supply breaker to the DC bus causing another containment isolation signal. This event did not affect Shutdown Cooling or any other protected Safety Related Equipment. The containment isolation signals caused an isolation of the systems listed below. All components that were not removed from service, gagged in position, already in the expected position due to plant conditions, or de-energized due to plant condition performed as designed. Containment Isolation valves for the following systems isolated as expected: Drywell and Containment Floor Drains, Drywell and Containment Equipment Drains, Condensate Makeup, Fire Protection Water, Service Air, Instrument Air, Reactor Water Cleanup, Spent Fuel Cooling and Cleanup, Reactor Plant Component Cooling Water, Chilled Water, Reactor Recirculation, Main Steam Drains, Reactor Building Ventilation, and Fuel Building Ventilation. The licensee notified the NRC Resident Inspector.
ENS 5410811 June 2019 16:57:00Arkansas NuclearNRC Region 4A non-licensed contract employee supervisor had a confirmed positive for a controlled substance during a pre-access fitness for duty test. The individual's unescorted access to the plant has been terminated and the badge removed.
ENS 540994 June 2019 11:39:00CooperNRC Region 4

On 06/04/2019, Nebraska Public Power District will issue a press release concerning the spurious actuation of emergency sirens near Cooper Nuclear Station and Indian Cave State Park. This is a four hour report per 10 CFR 50.72(b)(2)(xi) for any event or situation for which a news release is planned or notification to other government agencies has been or will be made which is related to heightened public or government concern. The cause of the siren actuation is still under investigation. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM TERRELL HIGGINS TO HOWIE CROUCH AT 1301 EDT ON 6/4/19 * * *

During this event, State & local government agencies (Nemaha County, Atchison County, Richardson County, and Indian Cave State Park) were contacted regarding the spurious actuation of emergency sirens. This is an update to the original Event Notification # 54099. Notified R4DO (Kellar).

ENS 540961 June 2019 03:15:00River BendNRC Region 4

At 2345 CDT at River Bend Station (RBS) Unit 1, a manual Reactor scram was inserted in anticipation of receiving an automatic Reactor Water Level 3 (9.7") scram due to the isolation of the 'B' Heater String with the 'A' Heater String already isolated. The 'B' heater string isolation caused loss of suction and subsequent trip of the running Feed Water Pumps 'A' and 'C'. All control rods fully inserted with no issues. Subsequently Reactor level was controlled by the Reactor Core Isolation Cooling (RCIC) system. Feed Water Pump 'C' was restored 4 minutes after the initial trip and the RCIC system secured. Currently RBS-1 is stable and is being cooled down using Turbine Bypass Valves. No radiological releases have occurred due to this event from the unit. The plant is currently under a normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector.

  • * * UPDATE AT 1603 EDT ON 6/10/19 FROM ALFONSO CROEZE TO JEFF HERRERA * * *

This amended event notification is being made to provide additional information that was not included in the original notification made on 6/1/19 at 0315 EDT. This event was reportable under 10 CFR 50.72(b)(3)(iv)(A) which was not annotated or described in the original report. Forty-two minutes after the Feed Water Pump 'C' was started, the pump tripped causing a Reactor Water Level 3 (9.7") RPS actuation. Feed Water was restored five minutes later using the Feed Water Pump 'A'. The NRC Resident Inspector has been notified. Notified the R4DO (Warnick).

ENS 5409126 May 2019 09:25:00Arkansas NuclearNRC Region 4This is a 4-hour Non-Emergency 10 CFR 50.72(b)(2)(iv)(B) notification due to a Plant Protection System (PPS) actuation. Arkansas Nuclear One, Unit 2, automatically tripped from 100 percent power at 0512 CDT. The reactor automatically tripped due to 2P-32B Reactor Coolant Pump tripping as a result of grounding. No additional equipment issues were noted. All control rods fully inserted. Emergency Feedwater (EFW) actuated and was utilized to maintain Steam Generator (SG) levels. The EFW actuation meets the 8-hour Non-Emergency Immediate Notification Criteria of 10 CFR 50.72(b)(3)(iv)(A). No Primary safety valves lifted. Main Steam Safety Valves (MSSVs) did lift initially after the trip. The NRC Resident Inspector has been notified. Decay heat is being removed via the steam dump valves to the main condenser. Unit 2 is in a normal shutdown electrical lineup. Unit 1 was not affected by the transient on Unit 2. The licensee notified the State of Arkansas.
ENS 5406816 May 2019 18:07:00WaterfordNRC Region 4This is a non-emergency notification from Waterford 3. On May 16, 2019, at 1348 CDT, Waterford 3 experienced an automatic reactor trip due to Steam Generator number 1 high level, which was the result of a Main Turbine trip and subsequent reactor power cutback which had occurred at 1345 CDT. The cause of the Main Turbine trip is currently under investigation. Subsequent to the Reactor trip, Main Feedwater Isolation Valves number 1 and number 2 closed on high Steam Generator levels. Emergency Feedwater automatically actuated for Steam Generator number 2 at 1419 CDT and Steam Generator number 1 at 1425 CDT. Main Feedwater was restored to both Steam Generators by 1629 CDT. The plant entered the Emergency Operating Procedure for an uncomplicated reactor trip and is in the process of transitioning to the normal operating shutdown procedure. The plant is currently in Mode 3 and stable with Main Feedwater feeding and maintaining both Steam Generators. The NRC Senior Resident Inspector has been notified. All control rods fully inserted. Decay heat is being removed through the main condenser. The plant is in a normal shutdown electrical lineup.
ENS 540495 May 2019 20:41:00CooperNRC Region 4

EN Revision Text: SECONDARY CONTAINMENT DECLARED INOPERABLE DUE TO POTENTIAL EQUIPMENT FAILURE At 1405 CDT, Secondary Containment differential pressure exceeded the Technical Specification limit due to a potential equipment failure. This required entry into (Limiting Condition of Operation) LCO 3.6.4.1 Condition A for Secondary Containment inoperability. An event or condition that could have prevented the fulfillment of a safety function requires an 8 hour report per 10 CFR 50.72(b)(3)(v)(C) for Control of Rad Release. Secondary Containment differential pressure was restored to greater than or equal to 0.25 inches vacuum, water gauge in accordance with plant procedures. Secondary Containment was declared operable at 1600 CDT. The issue has been entered in the Corrective Action Program and investigation of the cause is in progress. The NRC Senior Resident Inspector has been informed of this condition.

  • * * RETRACTION AT 1759 EDT ON 5/30/2019 FROM ROY GILES TO JEFF HERRERA * * *

CNS (Cooper Nuclear Station) is retracting the 8-hour notification made for event 54049 which occurred on May 5, 2019 at 1405 CDT. Subsequent evaluation determined that no equipment failure occurred. In addition, there were no procedure inadequacies or human performance issues identified. The indications observed were expected and part of a pre-planned evolution which included entry into a planned LCO for the Secondary Containment. The NRC Resident Inspector has been notified. Notified the R4DO (Kozal).

ENS 5403730 April 2019 07:37:00Indian PointNRC Region 1A non-licensed employee supervisor had a confirmed positive test for a prohibited substance during a follow-up fitness-for-duty test. The individual's unescorted access to the plant has been terminated. The NRC Senior Resident Inspector was notified by the licensee.
ENS 5403126 April 2019 20:19:00River BendNRC Region 4At 1147 (CDT) on 4/26/19, a through wall leak (reported as 1 drop every 1 to 2 minutes) was identified and confirmed by operation and NDE (Non-Destructive Examination) personnel on the Standby Liquid Control injection line during pressure testing activities. The line is 1.5 inch in diameter and classified as an ASME Section Ill, Class 1 line. The leak is currently isolated from the reactor vessel by a danger tagged manual valve. The licensee notified the NRC Resident Inspector.
ENS 5399111 April 2019 10:28:00WaterfordNRC Region 4

On 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.

  • * * RETRACTION ON 5/17/19 AT 1620 EDT FROM MARIA ZAMBER TO BETHANY CECERE * * *

This is a Non-Emergency Notification from Waterford 3. This is a retraction of EN 53991. This event was evaluated in accordance with the corrective action process. The original operability determination of inoperable was made based on a conservative evaluation that with the door handle for Door 86 (Heating and Ventilation Airlock Access Door) being detached, the control room envelope boundary could not perform its safety function. A more detailed engineering evaluation was subsequently performed. This shows that the condition of the door handle being detached is bounded by the most recently performed non-pressurized radiological tracer gas test, as the control room envelope differential pressure was maintained more positive with the detached door handle as compared to that observed during the test. Additionally, the control room envelope differential pressure trends showed no discernable change between the two conditions of the door handle detached or with the opening taped over (resulting in an air tight seal). This information supports the conclusion that with the door handle for Door 86 being detached, the control room envelope boundary remained operable and did not constitute a condition that could have prevented fulfillment of a safety function of structures or systems that are needed to mitigate the consequences of an accident; therefore, this event is not reportable per 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Notified R4DO (Proulx).

ENS 5395424 March 2019 17:40:00Indian PointNRC Region 1On 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 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 1On 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 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 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 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 5383718 January 2019 17:03:00WaterfordNRC Region 4This is a non-emergency notification from Waterford 3. On January 18, 2019, a relevant indication was detected in the performance of Phased Array Ultrasonic Examinations of A600 Dissimilar Metal Piping Welds during planned inspections. The indication was observed during the analysis of data recorded of the Reactor Coolant System (RCS) Loop 2A Reactor Coolant Pump Suction Drain Nozzle to Safe-End Butt Weld (11-007). This indication does not meet applicable acceptance criteria under American Society of Mechanical Engineers (ASME) Section XI. The plant was in Mode 6 (Refueling) at 0 percent power for a planned refueling outage at the time of discovery. The condition will be resolved prior to plant startup. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50 .72(b)(3)(ii)(A), 'Any event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded,' because an indication was found that did not meet acceptance criteria referenced in ASME Section XI , IWB-3514-2 and Code Case N-770-2, 3132. The NRC Resident Inspector has been notified. Reference: CR-WF3-2019-01041
ENS 5383417 January 2019 23:07:00WaterfordNRC Region 4This is a non-emergency notification from Waterford 3. On January 17, 2019, a relevant indication was detected in the performance of Phased Array Ultrasonic Examinations of A600 Dissimilar Metal Piping Welds during planned inspections. The indication was observed during the analysis of data recorded of the Reactor Coolant System (RCS) Loop 1A Reactor Coolant Pump Suction Drain Nozzle to Safe-End Butt Weld (07-009). This indication does not meet applicable acceptance criteria under American Society of Mechanical Engineers (ASME) Section Xl. The plant was in Mode 6 (Refueling) at 0 percent power for a planned refueling outage at the time of discovery. The condition will be resolved prior to plant startup. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A), 'Any event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded,' because an indication was found that did not meet acceptance criteria referenced in ASME Section Xl. IWB-3514-2 and Code Case N-770-2, 3132. The NRC Resident Inspector has been notified. Reference: CR-WF3-2019-0967
ENS 538188 January 2019 15:45:00PilgrimNRC Region 1On January 8, 2019, at 0945 EST Pilgrim Nuclear Power Station discovered that the Reactor Core Isolation Cooling (RCIC) system failed to meet its surveillance test requirements and was declared inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v), 'event or condition that could have prevented the fulfillment of a safety function: (D), mitigate the consequences of an accident.' There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 538155 January 2019 17:30:00PilgrimNRC Region 1

EN Revision Text: POTENTIAL LOSS OF MSIV SCRAM FUNCTION DURING MAIN STEAM LINE ISOLATION VALVE TESTING At approximately 1040 EST on January 5, 2019, during evaluation of test results for the 'C' Main Steam Isolation Valve (MSIV), it was determined that closure of three of four Main Steam Lines would not necessarily have resulted in a full scram during testing due to failure of a limit switch (LS-6) associated with MSIV-1C while in the test configuration. This condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v), 'Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down the reactor and maintain it in a safe shutdown condition.' The system was restored from the testing configuration at 1057 EST and the failed trip channel was placed in the tripped condition at 1326 EST thus restoring the design function. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * RETRACTION AT 1529 EST ON 02/11/19 FROM JOSEPH FRATTASIO TO JEFF HERRERA * * *

The purpose of the notification is to retract ENS Notification 53815 made on 01/05/19 for Pilgrim Nuclear Power Station. The previous notification reported that there was a potential loss of Main Steam Isolation Valve (MSIV) scram function during main steam line isolation valve testing, at the time of discovery, due to failure of a limit switch (LS-6) associated with MSIV-1C while in the test configuration. Subsequent evaluation has demonstrated that the scram function credited in the design basis was not lost. Specifically, after an Engineering Evaluation, it has been determined that the MSIV position RPS logic was not lost for those functions within the design basis and, as such, was capable of performing its intended safety function. The NRC Resident Inspector has been notified. Notified the R1DO (Cahill).

ENS 5380929 December 2018 10:27:00CooperNRC Region 4

At 0904 CST, on December 29, 2018, Cooper declared a Notice of Unusual Event under emergency action level HU 3.1. The emergency declaration was due to a toxic gas asphyxiant as a result of a fire. The fire is contained and the fire brigade continues to extinguishing the fire. Offsite support has not been requested. The licensee notified the NRC Resident Inspector. Additionally, State and Local government agencies were also notified. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * UPDATE ON 12/29/2018 AT 1655 EST FROM JIM FLORENCE TO JEFFREY WHITED * * *

At 1544 CST, on December 29, 2018, Cooper terminated the Notice of Unusual Event under emergency action level HU 3.1. The fire was verified to be extinguished and the flammable material was removed. The plant remained at 100% power for the duration of the event. The licensee issued a press release regarding the event at 1202 CST, on December 29, 2018. The license notified the NRC Resident Inspector. Notified R4DO (Taylor), NRR EO (Groom), IRD MOC (Gott), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

ENS 5379620 December 2018 05:32:00WaterfordNRC Region 4On December 19, 2018, at 2322 CST, the shift operating crew declared the control room envelope inoperable in accordance with Technical Specification (TS) 3.7.6.1 due to valve HVC-102 exceeding its maximum allowed closed stroke time of 2.0 seconds during performing of surveillance procedure OP-903-119. Actual closed stroke time was 2.1 seconds. Valve HVC-102 is part of the control room envelope. TS 3.7.6.1 requires that two control room emergency air filtration trains shall be OPERABLE. 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. Actions b.1 and b.2 were completed by placing the control room ventilation system in isolate mode at 2355. This event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(D), 'event or condition that could have prevented fulfillment 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 NRC Resident Inspector has been notified.
ENS 5379318 December 2018 15:40:00Arkansas NuclearNRC Region 4On December 18, 2018 at 1126 CST, Arkansas Nuclear One, Unit 1 (ANO-1) reactor automatically tripped due to a loss of the A-1, non-vital 4160V bus. All control rods fully inserted. Loss of the A-1 bus resulted in de-energizing A-3 vital 4160V bus. Emergency Diesel Generator #1, K-4A, started automatically and is currently powering A-3 vital bus. Non-vital buses A-2, H-1, and H-2 and vital bus A-4 transferred power automatically to the Startup Transformer #1. Off-site power remains energized and available for ANO-1. The reason for loss of A-1 bus is unknown at this time. Currently, ANO-1 has stabilized in Mode 3, Hot Standby. Decay heat is being removed by the main condenser using the turbine bypass valves. The loss of the A-1, non-vital bus, is under investigation. The licensee has notified the NRC Resident Inspector and the state.
ENS 5378812 December 2018 17:29:00Grand GulfNRC Region 4

EN Revision Text: MANUAL REACTOR SCRAM DUE TO FAILED OPEN TURBINE BYPASS VALVE At 1351 CST, the reactor was manually shutdown due to 'A' Turbine Bypass Valve opening. The Main Steam Line Isolation Valves were manually closed to facilitate reactor pressure control. Reactor level is being maintained through the use of Reactor Core Isolation Cooling System, Control Rod Drive System, and High Pressure Core Spray System. High Pressure Core Spray System was manually started to initially support reactor water level control. Reactor Pressure is being controlled through the use of the Safety Relief Valves and the Reactor Core Isolation Cooling System. The plant is stable in MODE 3. The cause of the 'A' Turbine Bypass Valve opening is under investigation at this time. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 12/14/18 AT 1140 EST FROM GERRY ELLIS TO TOM KENDZIA * * *

This is an update to EN # 53788 to correct an error on the event classification block of the form. The original notification did not have the block for 8 hour notification for Specified System Actuation checked. The actuation of Reactor Core Isolation Cooling System was discussed in original notification. The licensee notified the NRC Resident Inspector. Notified R4DO (Taylor).

ENS 537775 December 2018 14:54:00Arkansas NuclearNRC Region 4This 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to provide information pertaining to an invalid Engineered Safety Feature actuation signal. On October 9, 2018, Arkansas Nuclear One, Unit 2 was in refueling Mode 6, when a vital inverter failed while aligned from its alternate power source causing a loss of one of four vital instrument buses. The loss of the instrument bus resulted in one of the four engineered safety feature protection channels to enter a tripped state. Because one of the other four channels was already in a tripped state in support of a channel power supply replacement activity, two out of four protection channels were now in the tripped state resulting in a Safety Injection Actuation Signal, Containment Spray Actuation Signal, Containment Cooling Actuation Signal, Recirculation Actuation Signal, Emergency Feed Actuation Signal, and Containment Isolation Actuation Signal. In general, only one train of equipment is protected and assumed to be available during Mode 6 operations. Due to the defense-in-depth plant configuration in Mode 6, which is intended to avoid inadvertent start of emergency systems, the resulting actuations caused no adverse impact to Shutdown Cooling or Spent Fuel Pool cooling operations. At least one train of the following systems was aligned for automatic actuation: Service Water Emergency Diesel Generator Containment Penetration Room Exhaust Fan Other non-essential components which are shed or realigned upon safeguards actuation The few systems and components that were aligned for automatic operation responded as designed, including containment isolation valves and valves associated with the above systems (if aligned for automatic operation). The Service Water system was already in operation and, therefore, no Service Water pumps actuated. All systems and components which were capable of automatic operation performed as designed. The Emergency Diesel Generator started but did not synchronize to the bus. No safety injection occurred to the core. This actuation was caused by equipment failure and was not an actual signal resulting from parameter inputs. The affected actuation signals do not perform a safety function in Mode 6 and are not required to be available or operable. Therefore, this actuation is considered invalid. This event was entered into ANO's corrective action program for resolution. This event did not result in any adverse impact to the health and safety of the public. In accordance with 10 CFR 50.73(a)(i) a telephone notification is being made in lieu of submitting a written Licensee Event Report. The licensee has notified the NRC Resident Inspector.
ENS 5374921 November 2018 17:27:00PalisadesNRC Region 3On November 21, 2018, during an extent of condition review, after completion of ultrasonic testing, further interrogation of reactor vessel closure head (RVCH) penetration 36 was performed using eddy current testing. The testing detected three repairable indications. No indication of boric acid leakage was identified at this location during the bare metal visual inspection. Extent of condition review is complete on all RVCH penetrations. The plant was in cold shutdown at 0 percent power and in Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. The licensee notified the NRC Senior Resident Inspector.
ENS 5373411 November 2018 21:59:00PalisadesNRC Region 3

On November 11, 2018, during ultrasonic data analysis from reactor vessel closure head in-service inspections, signals that display characteristics consistent with primary water stress corrosion cracking in head penetration 33 were identified. No indications of boric acid leakage and no surface indications were detected at this location during bare metal visual inspection.

The plant was in cold shutdown at 0% power and in Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. The licensee notified the NRC Senior Resident Inspector.

ENS 5373310 November 2018 17:48:00PalisadesNRC Region 3On November 10, 2018, during a planned bare metal visual inspection of the reactor head, boric acid was discovered at a CRDM (Control Rod Drive Mechanism) nozzle to reactor head penetration. Investigation of the source of the boric acid is ongoing. The plant was in cold shutdown at 0% power and Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. All other reactor vessel head penetrations have had a bare metal visual inspection completed with no other indications identified. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. The licensee notified the NRC Senior Resident Inspector.
ENS 5373210 November 2018 04:35:00River BendNRC Region 4

At 0046 CST, River Bend Station experienced an automatic reactor scram on high reactor pressure. Initial indications are that the cause of the scram was an uncommanded closure of the #3 turbine control valve. The plant is stable with reactor water level in the normal level band of 10-51 inches being maintained with feedwater and condensate. Reactor pressure is in the prescribed band of 500-1090 psig, being maintained with turbine bypass valves and steam line drains. No injection systems were actuated either manually or automatically as a result of the event. The reactor scrammed on a Reactor Pressure High scram signal. A Reactor Level 3 signal resulted from the normal post-scram water level response. All systems responded as designed.

This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an automatic RPS actuation with the reactor critical. All control rods fully inserted. The Unit is in a normal shutdown electrical alignment. All control rods inserted properly and all systems functioned as designed. The licensee is investigating the cause of the event. The licensee notified the NRC resident inspector.

ENS 5366816 October 2018 00:21:00CooperNRC Region 4In accordance with 10 CFR 26.719(b)(2)(ii), this notification reports a licensed Reactor Operator tested positive for alcohol during a random fitness for duty test. The employee's access to the plant has been suspended. The NRC Senior Resident Inspector has been notified.