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 Entered dateSiteRegionReactor typeEvent description
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 5423220 August 2019 18:28:00CooperNRC Region 4At 0939 CDT, on 8/19/19, the National Weather Service reported to Cooper Nuclear Station that the National Warning System (NAWAS) Radio would neither transmit nor receive. The system has been intermittently available since then, but never declared fully functional. The backup notification system has been verified to be available throughout this period. Additional information from the National Weather Service received 8/20/19 at 1414 determined that the Shubert Tower transmitter is non-functional and would not be repaired until 8/21/19. The transmission outage is conservatively assumed to have begun at the first notification on 8/19/19 at 0939. The Shubert Tower transmitter activates the (EMERGENCY ALERT SYSTEM) EAS/Tone Alert Radios used for public notification. This is considered to be a major loss of the Public Prompt Notification System capability and is reportable under 10 CFR 50.72(b)(3)(xiii) when the primary notification system is or will be unavailable for greater than 24 hours with the backup system available. The NRC Senior Resident has been informed."
ENS 542058 August 2019 13:26:00WaterfordNRC 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 June 25, 2019, at Waterford 3, while performing an emergent replacement of relays on the Engineered Safety Features Actuation System Train A that affected Shield Building Ventilation Train A and HVAC Equipment Room Supply Fan AH-1 3A, unintentional contact was made between two contacts on the relay, resulting in an inadvertent initiation of other relays in the sequencer circuit. This caused the starting of Low Pressure Safety Injection Pump A, Switchgear Ventilation Fan A, and Boric Acid Makeup pumps. This was a partial actuation of Engineered Safety Features Actuation System Train A. Affected plant systems started and functioned successfully. 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 the Waterford 3 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)(1), a telephone notification is being made in lieu of submitting a written Licensee Event Report. The NRC Senior Resident Inspector has been notified."
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 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 5407320 May 2019 00:02:00Arkansas NuclearNRC Region 4On May 19, 2019, at 1809 CDT, the Safety Parameter Display System (SPDS) was lost to both Arkansas Nuclear One Units 1 and 2 due to the SPDS Inverter (2Y-26) failure. The SPDS Inverter is the power supply to both units' SPDS. The Unit 2 Control Room dispatched operators in response to a smoke alarm received from the 2Y-26 Inverter room. Upon arrival, smoke was reported emanating from 2Y-26. There was no report of fire at any time. Field operators de-energized 2Y-26 and the smoke ceased. The loss of SPDS also caused the Power Operating Limits (POL) function of the Unit 2 Core Operating Limits Supervisory System (COLSS) to be lost, so Unit 2 reduced power to 91 (percent) in accordance with Technical Specifications. Both units are at power and stable. The NRC Resident has been notified. This is reportable per 10 CFR 50.72(b)(3)(xiii)."
ENS 5407118 May 2019 02:09:00PilgrimNRC Region 1

On Friday, May 17, 2019 at 2303 (EDT), with the reactor at 70 (percent) core thermal power, Pilgrim Nuclear Power Station initiated a manual reactor scram due to degrading condenser vacuum as a result of the trip of Seawater Pump B. All control rods inserted as designed. The plant is in hot shutdown. Plant safety systems responded as designed. Pressure is being controlled using the Mechanical Hydraulic Control System and Main Condenser. Reactor water level is being maintained with the feedwater and condensate system. During the manual reactor scram, the plant experienced the following isolation signals as designed:

"Group 2 Isolation: Miscellaneous containment isolation valves
Group 6 Isolation: Reactor Water Clean-up
Reactor Building Isolation Actuation

Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B), 'any event that results in actuation of the reactor protection system (RPS) when the reactor is critical...' This notification is also being made in accordance with 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section...' (B)(2) 'General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).' This event has no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The licensee will notify the Massachusetts Emergency Management Agency."

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 5406212 May 2019 15:28:00Grand GulfNRC Region 4

At 1039 CDT the reactor was manually (scrammed) due to a partial loss of plant service water. The loss of plant service water was caused by a loss of (balance of plant) BOP transformer 23. Reactor power was reduced in an attempt to restore pressure to plant service water. Reactor level is being maintained with condensate and feedwater. Reactor pressure is being maintained with bypass control valves. Standby Service Water A and B were manually initiated to supply cooling to Control Room A/C and (Engineered Safety Feature) ESF switchgear room coolers. The cause is under investigation. The NRC Resident Inspector has been notified. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B) as any event or condition that results in actuation of the Reactor Protection System (RPS), when the reactor is critical and also reported under 10 CFR 50.72(b)(3)(iv)(A), as any event or condition that results in actuation of RPS and Standby Service Water. The plant is currently in a normal electrical lineup.

  • * * UPDATE ON 5/12/19 AT 1846 EDT FROM GERRY ELLIS TO JEFFREY WHITED * * *

This is an update to the original notification. The Drywell and Containment exceeded the technical specification (TS) temperature limits of 135 degrees F (TS Limiting Condition of Operation (LCO) 3.6.5.5) and 95 degrees F (TS LCO 3.6.1.5), respectively. An 8-hour notification is being added 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 control the release of radioactive material per 10 CFR 50.72(b)(3)(v)(C). Notified R4DO (Alexander).

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 5393415 March 2019 07:08:00CooperNRC Region 4

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

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

The licensee terminated the Unusual Event at 1601 CDT due to lowering Missouri River water levels. River water level is currently at 896.0 feet MSL and lowering. A press release will be issued to inform the public of Cooper Nuclear Station's exit from the Notification of Unusual Event regarding high Missouri River level. The initial entry into the Notification of Unusual Event occurred on 03/15/2019 and was exited on 03/24/2019 at 1601 CDT. The press release is reportable per 10 CFR 50.72(b)(2)(xi). The licensee notified the NRC Resident Inspector. Notified R4DO (O'Keefe), NRR EO (Miller), IRD MOC (Gott), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

ENS 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 5389423 February 2019 19:05:00Grand GulfNRC Region 4Actuation of RPS (Reactor Protection System) with the reactor critical. Reactor scram occurred at 1458 (CST) on 2/23/2019 from 100% power. The cause of the scram was due to Turbine Control Valve Fast Closure. All control rods are fully inserted. Currently reactor water level is being maintained by the Condensate Feedwater System in normal band and reactor pressure is being controlled via Main Turbine Bypass valves to the main condenser. No ECCS (Emergency Core Cooling System) initiation signals were reached and no ECCS or Diesel Generator initiation occurred. The Low-Low Set function of the Safety Relief Valves actuated upon turbine trip. This was reset when pressure was established on main turbine bypass valves. The cause of the turbine trip is still under investigation. There were no complications with scram response. The licensee notified the NRC Resident Inspector. There was no maintenance occurring on the main turbine at the time of the scram.
ENS 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 538199 January 2019 13:23:00PalisadesNRC Region 3At 1034 EST on January 9, 2019, with the reactor at 100% power, an automatic reactor trip was initiated. The trip occurred while Reactor Protection System testing was in progress. The trip was uncomplicated with all systems responding normally following the rip. Troubleshooting and investigation of the cause is ongoing. All full-length control rods inserted fully. Auxiliary Feedwater System actuated as designed in response to low steam generator water levels. Operations stabilized the plant in Mode 3 (hot standby). Decay heat is being removed by the turbine bypass valve. This condition has no impact to the health and safety of the public. The licensee notified the NRC Resident Inspector.
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 538133 January 2019 23:57:00PalisadesNRC Region 3At 2028 (EST) on January 3, 2019, with the reactor at 85% power, the reactor was manually tripped due to cycling of Turbine Governor Valve #4. The trip was uncomplicated with all systems responding normally following the trip. Investigation of the cause of the valve cycling is ongoing. All full-length control rods inserted fully. Auxiliary Feedwater System actuated as designed in response to low steam generator water levels. Operations stabilized the plant in Mode 3 (hot standby). Decay heat is being removed by atmospheric dump valves. This condition has no impact to the health and safety of the public. The licensee notified the NRC Senior Resident Inspector. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A)."
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 537765 December 2018 11:24:00CooperNRC Region 4This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a Primary Containment Isolation System (PCIS) Group 1 for Main Steam Isolation Valves (MSIVs), Group 3 for Reactor Water Cleanup (RWCU), Group 6 for Secondary Containment isolation, Group 7 for Reactor Water Sampling, Diesel Generator, Reactor Core Isolation Cooling (RCIC) System logic, and Residual Heat Removal (RHR) logic. Group 1, Group 6, Diesel Generator actuation, RCIC actuation and RHR actuation are within scope of 10 CFR 50.73(a)(2)(iv). Group 3 and Group 7 are not within scope as they affect only one system. Cooper Nuclear Station (CNS) was shut down in Mode 5 at the time of the event with the reactor cavity flooded. On October 13, 2018, at 0028 Central Daylight Time, CNS received full PCIS Groups 1, 3, and 6, and a half Group 7 on the Division 1 side. The MSIVs and RWCU isolation valves were already closed for maintenance. The Secondary Containment isolated. Control Room Emergency Filter and the Standby Gas Treatment Systems initiated. The inboard Reactor Water Sample valve isolated. Diesel Generator #1 started but was not required to connect to the critical bus. Reactor Core Isolation Cooling System logic actuated with no expected response due to being isolated for shutdown conditions. Division 1 RHR pump logic actuated. Division 1 RHR system was operating in shutdown cooling mode. The actuation caused the Division 1 RHR outboard injection and heat exchanger bypass valves to open. Shutdown cooling was unaffected and remained in service throughout the event. The plant systems responded as expected with no Emergency Core Cooling System injection. At the time of the event, an in-service inspection of welds inside the reactor vessel was taking place using a robot scanner that uses two vortex thrusters to hold the robot to the vessel wall. The robot inadvertently passed over an instrument penetration, drawing suction on the process leg, resulting in low reactor water level indications and the subsequent invalid Level 1 and 2 system actuations. Actual reactor vessel water level remained steady at cavity flooded conditions. The NRC Resident Inspector has been notified of this event."
ENS 5375628 November 2018 05:40:00River BendNRC Region 4

EN Revision Text: INOPERABILITY OF EQUIPMENT FOR CONTROL OF RADIOLOGICAL RELEASE At 2130 CST on 11/27/2018, Division 1 Main Steam Positive Leakage Control System (MS-PLCS) was declared inoperable because of a leaking check valve that caused excessive cycling of the associated air compressor. Division 2 MS-PLCS had been declared inoperable on 11/27/2018 at 1400 CST when a pressure control valve in the system exceeded the maximum allowable stroke time. Because MS-PLCS supplements the isolation function of the main steam isolation valves (MSIVs) by processing fission products that could leak through the closed MSIVs, both divisions of MS-PLCS inoperable at the same time represents a condition that could prevent the fulfillment of a safety function of an SSC (Structures, Systems and Components) that is needed to control the release of radioactive material. The station diesel air compressor is available to supply backup air to the safety relief valves as required by the Technical Requirements Manual." (This is associated with operability of the safety relief valves, due to the inoperable MS-PLCS air compressor.) The unit is in a 7 day shutdown Limiting Condition for Operation (LCO), 1-TS1-18-Div 1 & 2 MSPLCS-685, for the two divisions of MS-PLCS being inoperable. The licensee notified the NRC Resident Inspector.

  • * * RETRACTION ON 12/03/18 AT 1551 EST FROM TIM GATES TO BETHANY CECERE * * *

This event was initially reported under 10 CFR 50.72(b)(3)(v)(C) as a condition that could have prevented the Main Steam Positive Leakage Control System (MS-PLCS) from fulfilling its safety function to control the release of radioactive material. Division I was declared inoperable due to a failed component. Division II was declared inoperable due to a pressure control valve in the system exceeding the maximum allowable time to close by 0.50 seconds. An engineering evaluation has since been performed and concluded that the 2 second maximum allowable time to close was based on the pressure control valve being classified as a rapid closure valve and was established from the original baseline data of 0.50 seconds. This baseline data is an administrative target value per the In-Service Testing Program. There are no technical specification requirements associated with the 2 second closure time. The engineering evaluation also determined that the volume of air supplied through the pressure control valve during the extra 0.50 seconds of valve closure would have an inconsequential effect on the pressure within the volume of leakage barrier between the Main Steam Isolation Valves associated with the MS-PLCS pressure control valve or have any effect on containment over-pressurization. Based on the information provided by the engineering evaluation, the Division II MS-PLCS has been declared operable-degraded non-conforming since time of initial discovery. Consequently, this event is not reportable as a condition that could have prevented the Main Steam Positive Leakage Control System (MS-PLCS) from fulfilling its safety function. The (NRC) Resident Inspector has been notified via e-mail. Notified the R4DO (Gaddy).

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 5369928 October 2018 15:10:00Indian PointNRC Region 1During the performance of Service Water Essential header swap, SWN-6 (Supply to Turbine Building Oil Coolers) valve stem became disconnected from its gear box at 85% open and could not be operated. Therefore, the non-essential service water system was inoperable. LCO 3.0.3 was entered at 0930 (EDT) with required actions to be in Mode 3 in 7 hours, Mode 4 in 13 hours and Mode 5 in 37 hours. Repair efforts were successful at shutting SWN-6, and LCO 3.0.3 was exited at 1305 (EDT) before adding any negative reactivity in support of shutdown. ('TS Required S/D' box not checked.) This condition constituted a loss of safety function which requires an 8 hour report (in accordance with) IAW 50.72(b)(3)(v)(B): Without the ability to close SWN-6, the non-seismic portion of the conventional Service Water System could not be isolated as required in the event of either a seismic event or as required in the EOPs. The nonessential service water system is required to support the recirculation phase post (Design Basis Accident) DBA for accident mitigation. The licensee notified the NRC Resident Inspector and the State of New York.
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."
ENS 536506 October 2018 05:56:00CooperNRC Region 4On 10/5/2018, at 2219 CDT, the Control Room Emergency Filter (CREF) System was determined to be inoperable during a required condition of applicability due to being aligned to a Division 2 power source with its associated emergency power supply (Diesel Generator #2) removed from service earlier in the day. The power supply alignment was not identified at the time Diesel Generator #2 was removed from service (Diesel Generator #2 was rendered inoperable on 10/5/2018 at 1728 CDT). Movement of lately irradiated fuel assemblies in the Secondary Containment was in progress at the time of discovery of this condition. This condition represents an unplanned loss of safety function for a single train system during its specified condition of applicability. Movement of irradiated fuel was suspended until the power supplies to CREFs could be realigned to Division 1 which was completed at 0004 CDT on 10/6/2018. This represents a condition that could have prevented the fulfillment of the safety function of CREFs needed to mitigate the consequences of a fuel handling accident. The NRC Resident Inspector has been notified.
ENS 536485 October 2018 15:39:00PilgrimNRC Region 1On Friday, October 5, 2018 at 1209 hours, with the reactor at 100 percent core thermal power, Pilgrim Nuclear Power Station (PNPS) automatically tripped due to reactor water level perturbation and receipt of a low reactor water level Reactor Protection System (RPS) signal. The cause of the low reactor water level is under investigation. The plant is in hot shutdown. All other plant systems responded as designed. Pressure is being controlled using the Mechanical Hydraulic Control System and Main Condenser. Reactor water level is being maintained with the feedwater and condensate system. During the automatic reactor scram the plant experienced the following isolation signals as designed: Group 2 Isolation: Miscellaneous containment isolation valves Group 6 Isolation: Reactor Water Clean-up Reactor Building Isolation System Actuation Due to the RPS actuation while critical, this event is being reported as a four-hour, non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B), 'any event that results in actuation of the reactor protection system (RPS) when the reactor is critical.' This notification is also being made in accordance with 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section ... ' (B)(2) 'General containment isolation signals affecting containment isolation valves in more than one system.' This event has no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The licensee will notify the Massachusetts Emergency Management Agency.