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 Entered dateSiteRegionReactor typeEvent description
ENS 5399111 April 2019 10:28:00WaterfordNRC Region 4On April 11, 2019, at 0200 CDT the shift operating crew declared the control room envelope inoperable in accordance with Technical Specification (TS) 3.7.6.1 due to the door handle for Door 86 (H&V Airlock Access Door) being detached. Operations entered TS 3.7.6.1 action b, which requires that with one or more control room emergency air filtration trains inoperable due to inoperable control room envelope boundary in MODES 1, 2, 3, or 4, then: 1. Immediately initiate action to implement mitigating actions; 2. Within 24 hours, verify mitigating actions ensure control room envelope occupant exposures to radiological, chemical, and smoke hazards will not exceed limits; and 3. Within 90 days, restore the control room envelope boundary to OPERABLE status. Action b.1 was completed by sealing the hole in Door 86 at 0232 CDT. This event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(D), 'event or condition that could have prevented fulfilment of a safety function of structures or systems that are needed to (D) mitigate the consequences of an accident,' due to the control room envelope being inoperable. The licensee notified the NRC Resident."
ENS 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 4At 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.
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.
ENS 536465 October 2018 09:52:00CooperNRC Region 4

EN Revision Text: MAIN STEAM ISOLATION VALVES EXCEEDED PRIMARY CONTAINMENT LOCAL LEAK RATE ACCEPTANCE CRITERIA At 0520 (CDT), on October 05, 2018, it was discovered that a Primary Containment local leak rate test performed on Main Steam Isolation Valves (MSIV) exceeded its acceptance criteria.

During Mode 1, 2, and 3, Surveillance Requirement 3.6.1.3.10 requires MSIV leakage for a single MSIV line to be less than or equal to 106 standard cubic feet per hour (scfh) when tested at 29 psig and Surveillance Requirement 3.6.1.3.12 requires the combined leakage rate for all MSIV leakage paths to be less than or equal to 212 scfh when tested at 29 psig.

As-found for the 'C' MSIV line leakage results were unquantifiable and gave a (minimum) path value greeter than 160 scfh. This leakage rate lead to Surveillance Requirement 3.6.1.3.10 and 3.6.1.3.12 limits to be exceeded. This event is being reported as a condition of the nuclear power plant, including its principal safety barriers, being seriously degraded per 10 CFR 50.72(b)(3)(ii)(A) since the Primary Containment Isolation Valves leakage limits for MSIVs were exceeded. The NRC Resident Inspector has been notified.

  • * * RETRACTION AT 2320 EDT ON 10/24/2018 FROM THOMAS FORLAND TO MARK ABRAMOVITZ * * *

CNS (Cooper Nuclear Station) is retracting the 8-hour non-emergency notification made on October 5, 2018 at 0520 CDT (EN# 53646). Subsequent evaluation concluded that overall as-found 'C' MSIV leakage rate was not at a level that exceeded the surveillance requirement 3.6.1.3.10 and 3.6.1.3.12 limits and thus the Primary Containment Isolation Valve leakage rate limits for the MSIVs were not exceeded. The NRC Senior Resident Inspector has been notified. Notified the R4DO (Drake).

ENS 5362225 September 2018 20:25:00River BendNRC Region 4At 1200 CDT on September 25, 2018, while the plant was in MODE 1 at 90 percent power, it was identified that an additional condition existed which had not previously been considered in developing the compensatory measures implemented for design flaws and single point vulnerabilities associated with the Control Building Chilled Water System. Specifically, a 20 minute 'quick restart timer' on Control Building Chillers that have analog control systems (HVK-CHL1A & 1B) would prevent the chillers from starting in specific scenarios. The recommended compensatory actions to address the new condition were implemented at 1235 CDT on September 25, 2018. Currently the Chilled Water System is otherwise operating as designed. Operator actions are in place to ensure the plant meets all required design safety system functions. Work is currently underway to identify and correct all design vulnerabilities. The (NRC) Senior Resident Inspector has been notified. This was identified by engineering during an extended condition search.
ENS 5361118 September 2018 09:06:00Indian PointNRC Region 1Due to a steam leak on the reheater line to 36C Feedwater Heater, operators initiated a manual trip of the reactor, verified the reactor trip, and closed all Main Steam Isolation Valves. The plant is currently stable in Mode 3 with the steam leak isolated. The Auxiliary Feedwater System actuated following the trip as expected. This 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 atmospheric steam dumps. Unit 2 was unaffected and remains at 100 percent power. The licensee notified the NRC Resident Inspector, the State of New York, and the local transmission company.
ENS 5360814 September 2018 21:09:00Grand GulfNRC Region 4At 1644 (CDT) a manual reactor scram was inserted by placing the Reactor Mode Switch to Shutdown. At 1643 (CDT) the Condensate Booster Pump A tripped on low suction pressure. At 1644 (CDT) the Reactor Feed Pump A tripped on low suction pressure. A Recirculation Flow Control Valve runback occurred as designed. Reactor Water level was approaching the Automatic Low Water Level 3 (11.4 inches) scram set point and manual actions were taken by placing the Mode Switch to Shutdown before the low level set point was reached. All systems responded as expected following the manual scram. The plant is stable in mode 3. This event is being reported under 10CFR 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. The NRC Senior Resident Inspector has been notified. All control rods fully inserted, and decay heat is being removed through the turbine bypass valves to the main condenser. The licensee is investigating the cause of the event.
ENS 5360113 September 2018 08:24:00Indian PointNRC Region 1A hurricane is within 500 nautical miles from Indian Point Energy Center with wind speed in excess of 87 knots. The National Weather Service has issued a hurricane warning for a hurricane with wind in excess of 87 knots (approximately 100 mph) within 500 nautical miles of the facility. Per the Technical Requirement Manual a prompt report shall be made to the NRC Incident Response Center within 1 hour of receipt of that hurricane warning. The licensee has notified the NRC Resident Inspector."
ENS 535781 September 2018 13:52:00Indian PointNRC Region 1Twelve (12) cans of an alcoholic beverage were discovered inside the protected area. The 12 cans were removed from the protected area and taken to and secured by site station security. The licensee notified the NRC Resident Inspector.
ENS 5356627 August 2018 13:15:00Grand GulfNRC Region 4On 27 August 2018 at 0918 CDT Grand Gulf Control Room was informed that the onsite credit union silent alarm was actuated. The credit union is located outside of the Secure Owner Controlled Area (SOCA), but is located within the Owner Controlled Area. GGNS (Grand Gulf Nuclear Station) Security entered an elevated security position at 0920 (CDT) and requested assistance from local law enforcement. Claiborne County Sheriff's Department responded to the site. Investigation of the area conducted with the assistance of GGNS Security personnel determined that the cause of the notification was not valid. GGNS Security stood down from the elevated security positon at 0945 (CDT). Pursuant to 10 CFR 50.72(b)(2)(xi), this issue is being reported as any event or situation, related to the health and safety of the public or on-site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made."
ENS 5355923 August 2018 12:11:00WaterfordNRC Region 4This is a non-emergency notification from Waterford 3. 10 CFR Part 21 Notification - Defect of Westinghouse 7300 Process Analog Control System circuit cards On August 14, 2018, Entergy Operations, Inc. (Entergy) completed an evaluation of a deviation at Waterford Steam Electric Station, Unit 3 (Waterford 3) which concluded the condition constitutes a defect pursuant to 10 CFR Part 21. The Waterford 3 Site Vice President was notified of the result of this evaluation on August 21, 2018. An interim report stating that an evaluation of this deviation was in progress was submitted to the NRC on July 5, 2018 (Entergy letter W3F1-2018-0040, ADAMS Accession Number ML18186A694). Three Westinghouse 7300 Process Analog Control System (PAC) circuit cards were identified to be failed due to failed hex inverter chips. Some of these cards were installed in applications which support the Ultimate Heat Sink (UHS) at Waterford 3. These PAC cards use Texas Instruments Part Number SN74LS04N, W113 hex inverter chips. The circuit card types of concern are Analog Comparator model number 2838A32G01, Control Board model number 2838A30G011, and Prom Logic model number 2838A33G01. Entergy concluded that this condition could have prevented the UHS from performing its safety function and thus could have created a substantial safety hazard. The NRC Resident Inspector has been notified."
ENS 5347626 June 2018 23:29:00CooperNRC Region 4

EN Revision Text: CONTROL ROOM EMERGENCY FILTRATION SYSTEM DECLARED INOPERABLE On June 26, 2018, at 1630 CDT, the Control Room Emergency Filtration System (CREFS) was declared inoperable when Main Control Room Supply Fan SF-C-1B was discovered to have elevated vibrations that brought into question the ability to meet its mission time. CREFS is a single train safety system. Per 10 CFR 50.72(b)(3)(v)(D), an 8 hour report is required due to the fact that at the time of discovery this condition could have prevented the fulfillment of a safety function of an SSC (System Structure or Component) that is required to mitigate the consequences of an accident. The licensee has notified the NRC Senior Resident Inspector.

  • * * RETRACTION FROM THOMAS FORLAND TO VINCE KLCO ON 8/13/18 AT 1024 EDT * * *

The following retraction was received from Cooper Nuclear Station (CNS) via facsimile and phone call: CNS is retracting the 8-hour non-emergency notification made on June 26, 2018 at 1630 CDT (EN# 53476). Subsequent evaluation concluded that overall vibration levels were not at a level that would impact the ability of the Main Control Room Supply Fan SF-C-18 to perform its safety function for its required mission time and the CREFS therefore, was operable. The NRC Senior Resident Inspector has been notified. Notified the R4DO (Deese).

ENS 5345916 June 2018 15:56:00Arkansas NuclearNRC Region 4At 1121 CDT on June 16, 2018, Arkansas Nuclear One, Unit 1 (ANO-1) performed a manual reactor trip due to a Turbine Bypass valve failing open on reactor startup. At the time, ANO-1 was in Mode 2 at approximately 2 percent power. The failed Turbine Bypass valve resulted in an overcooling event and the Overcooling Emergency Operating Procedure (EOP) was entered. Main Steam Line Isolation (MSLI) automatic actuation occurred on 2 of the 4 channels of Emergency Feedwater Initiation and Control during the overcooling event in the 'B' Steam Generator. The remaining channels of MSLI were manually actuated by the control room staff from the control room. Overcooling was terminated after the closure of the Main Steam Isolation Valve (MSIV) and reactor coolant parameters were stabilized as directed by the Overcooling EOP. Additionally, Gland Sealing Steam was lost to the main turbine due to the closure of the 'B' Steam Generator MSIV and Loss of Condenser Vacuum Abnormal Operating Procedure was entered. This is a 4-hour non-emergency 10 CFR 50.72 (b)(2)(iv)(B) notification due to a Reactor Protection System actuation (scram) and an 8-hour non-emergency 10 CFR 50.72 (b)(3)(iv)(A) notification for safety system actuation." All control rods fully inserted into the core during the trip. Heat removal is via the Atmospheric Dump Control valves to atmosphere. The NRC Resident Inspector has been notified. The licensee also notified the State of Arkansas.
ENS 5345612 June 2018 22:11:00Arkansas NuclearNRC Region 4On June 12, 2018, at 1500 CDT, a Reactor Coolant System (RCS) Pressure Boundary leak was identified during a Mode 3, hot shutdown walkdown on a High Pressure Injection Line (HPI) to Reactor Coolant Pump (P32C) drain line weld near MU-1066A HPI Line Drain Valve and MU-1066B HPI Line Drain Valve. The 3/4 inch drain line containing drain valves MU-1066A and MU-1066B on the 'C' HPI header (CCA-5 pipe class) has a through-wall defect on the pipe stub or welds between the sockolet and valve MU-1066A. The leak location is in the ASME Class I RCS Pressure Boundary. The hot shutdown walkdown was being performed as part of a planned outage to investigate excessive Reactor Building Sump inleakage. Total unidentified RCS leakage prior to the investigation was determined to be at 0.165 gpm. After the initial investigation of the leakage, the following Tech Specs (TS) were determined be applicable: TS 3.4.5 - RCS Loops Mode 3, TS 3.4.13 - RCS Leakage, TS 3.5.2 - ECCS. Unit 1 is currently in Mode 3 and in progress of an RCS cooldown to comply with Tech Spec requirements. The licensee notified the NRC Resident Inspector.
ENS 5339912 May 2018 06:58:00Grand GulfNRC Region 4GE-6On 5/11/2018, at 2327 hours CDT, with the plant in Mode 5, Grand Gulf Nuclear Station was making preparations for surveillance test 06-OP-1P75-R-0003, Standby Diesel Generator 1 Functional Test. The Grand Gulf Nuclear Station experienced an auto-start of the Division 1 (Emergency) Diesel Generator (EDG) when the 15AA Bus Potential Transformer (PT) fuse drawer was racked out instead of the line PT fuse drawer for Bus 15AA feeder breaker 152-1514. This resulted in the 15AA Incoming Feeder Breaker 152-1511 from Engineered Safety Features Transformer 12 opening, de-energizing the 15AA Bus. The Division 1 EDG started and energized Bus 15AA. The Division 1 LSS SYSTEM FAIL annunciator was received and Standby Service Water A failed to start due to the 15AA Bus PT fuse drawer being racked out. Standby Gas Treatment Train B was manually initiated per the Loss Of AC Power Off Normal Emergency Procedure. Station equipment operated as expected based on the PT fuse drawer that was racked out. The Division 1 EDG was manually tripped from the Control Room because cooling from the Standby Service Water A was not available. RHR (residual heat removal) B was in Shutdown Cooling (mode) and was verified not affected The licensee has notified the NRC Resident Inspector.
ENS 533897 May 2018 17:40:00WaterfordNRC Region 4CEA non-licensed supervisor had a confirmed positive result for alcohol during a random fitness for duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 533824 May 2018 13:50:00River BendNRC Region 4GE-6During performance of an extent of condition evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, River Bend Station identified non-conforming conditions in the plant design such that specific TS equipment is considered to not be adequately protected from tornado missiles. The reportable condition is postulated by tornado missiles entering the Diesel Generator Building through conduit and pipe penetrations. A tornado could generate multiple missiles capable of striking Division 1, Division 2, and Division 3 Diesel Generator support equipment rendering all Safety Related Diesel Generators inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (A) Shut down the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) Mitigate the consequences of an accident. This condition was identified as part of an on-going extent of condition review of potential tornado missile related site impacts. Enforcement discretion per Enforcement Guidance Memorandum EGM 15-002 has been implemented and required actions taken. Corrective actions will be documented in a follow-on licensee event report. The licensee has notified the NRC Resident Inspector.
ENS 533741 May 2018 20:42:00Grand GulfNRC Region 4GE-6At 1551 hrs (CDT) on 5/1/2018, with the plant in Mode 5, a division one Reactor Pressure Vessel (RPV) Level 1 signal was received; however there was no actual change in RPV level. RPV Level remained at High Water Level supporting refuel operations. This caused an actuation of division one Load Shed and Sequencing system that shed and then re-energized the 15 bus. Division one diesel generator started from standby. Residual Heat Removal pump 'A', which was in shutdown cooling mode, was lost during the bus shed, and was re-sequenced upon re-energization of the 15 bus. Upon restoration of shutdown cooling, the RHR pump discharged into the RPV. RCS temperature increased approximately 5 degrees Fahrenheit as a result of the loss of shutdown cooling. The cause of the actuation signal is under investigation. In accordance with NUREG 1022, Event Reporting Guidelines, this event is conservatively reported under 10 CFR 50.72(b)(2)(iv)(A) as an event that results in emergency core cooling system discharge into the RCS as a result of a valid signal, under 10 CFR 50.72(b)(3)(iv)(B)(8) as an event that results in the actuation of emergency ac electrical power systems, and under 10 CFR 50.72(b)(3)(v)(B) as an event or condition that at the time of discovery could have prevented the fulfillment of a safety function (remove residual heat). The licensee notified the NRC Resident Inspector.
ENS 5336526 April 2018 18:50:00River BendNRC Region 4GE-6River Bend Station experienced an inadvertent initiation and injection of High Pressure Core Spray (HPCS) at 1531 (CDT) on 4/26/2018 while operating at 100 percent power. During replacement of Level Transmitter B21-LTN081C 'Reactor Vessel Low Water Level 1', Main Control Room received an inadvertent initiation and injection of High Pressure Core Spray. The HPCS injection valve was open for approximately 40 seconds before the operators manually closed the valve. Feedwater Level Control responded per design and maintained Reactor Water Level nominal values. The Division 3 Diesel Generator (DG) also automatically started in response to the actuation signal. The DG did not automatically connect to the Division 3 switchgear since there was not a low voltage condition on the bus. The manual closure of the injection isolation valve caused the system to be incapable of responding to an automatic actuation signal. The manual override of the injection isolation valve was reset approximately 16 minutes after the event, restoring the system to its standby condition. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(A) as a condition that caused ECCS (Emergency Core Cooling System) discharge to RCS (Reactor Coolant System) and 10 CFR 50.72(b)(3)(v)(D) as a condition that caused the loss of function of the HPCS System. The Senior NRC Resident inspector has been notified.
ENS 5334819 April 2018 23:41:00Indian PointNRC Region 1Westinghouse PWR 4-LoopWhile performing a turbine startup, a turbine control anomaly caused a steam generator level transient. The rise in steam generator level above the setpoint caused the turbine to automatically trip. The high steam generator level of 73 percent caused a feedwater isolation signal at 2107 EDT, which also tripped both Main Boiler Feed Pumps. The tripping of the Main Boiler Feed Pumps auto started the motor driven Aux Boiler Feed Pumps 21 and 23. The reactor was manually tripped at 2108 EDT in accordance with AOP-FW-1 Loss of Main Feedwater. All control rods inserted. Electrical power is being provided from offsite via the Station Aux Transformer. Decay heat removal is being provided via the Atmospheric Dump Valves. An investigation into the cause of the turbine control anomaly is underway. The NRC Resident Inspector has been notified. The event did not have an affect on Unit 3 and there is no primary to secondary leakage.
ENS 5333513 April 2018 21:04:00Grand GulfNRC Region 4GE-6

At 1208 CDT on April 13, 2018, GGNS (Grand Gulf Nuclear Station) identified cracks in the primary containment concrete penetration (outer wall) around feed water line 'B'. There are no available dimensions for crack width or depth until further inspections are performed. In accordance with NUREG 1022, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73, Section 3.2.4, any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers being seriously degraded, requires that when a principal safety barrier is declared inoperable the condition must be reported under 10 CFR 50.72 (b)(3)(ii). The licensee has notified the NRC Resident Inspector.

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

Grand Gulf Nuclear Station (GGNS) personnel performed an inspection of the wall around feed water line 'B'. This inspection included the protective coating in the identified area and a partial inspection of the underlying concrete. The inspection of the protective coating found a collection of non-linear anomalies, chipping, and flaking. The inspection found non-significant linear indications in the concrete. Grand Gulf Nuclear Station determined that the collection of non-significant coating imperfections and non-significant indications in the concrete do not constitute serious degradation of primary containment. The indications do not adversely impact the operability, mission time, or safety-function (as described per Technical Specification 3.6.1.1, Primary Containment) of the containment structure. The as-found conditions have been entered into the GGNS corrective action program for final disposition. The containment structure is operable, therefore, GGNS is retracting this event notification. The licensee has notified the NRC Resident Inspector. Notified R4DO (Kellar).

ENS 5332411 April 2018 10:14:00River BendNRC Region 4GE-6At time 0150 CDT on April 11, 2018, a condition was identified that could impair the ability of the Control Building Air Conditioning System to perform its design function. Engineering determined that the time delay relays HVKA11-80YB or HVKA11-80YD (Division II chilled water LOW FLOW relays) could fail in a manner that challenges the design safety function of the Control Building Chilled Water System during a Loss of Offsite Power (LOP) Event. A failure of the time delay relay HVKA11-80YB or HVKA11-80YD (Division II chilled water LOW FLOW relays) to provide the time delay function would cause both the Division I and Division II HVK chilled water pumps to start after a LOP, which in turn could hinder the auto start of either Division I or Division II chillers. Currently the Chilled Water System is otherwise operating as designed. All operator actions are in place to ensure the plant meets all required designed safety system functions. Work is currently underway to correct this design vulnerability. The NRC Resident Inspector has been notified of this condition.
ENS 533175 April 2018 18:23:00Grand GulfNRC Region 4GE-6On Thursday, April 5, 2018, at approximately 1117 hours Central Daylight Time, Entergy contract personnel opened the personnel hatch allowing access to the roof of the Secondary Containment Building for the purposes of performing an inspection of various items located on the roof. During the time period the individuals were on the roof, the hatch was left open. An individual was adjacent to the door with a radio and had constant communication link with the control room operator. Pursuant 10 CFR 50.72(b)(3)(v)(C), and 10 CFR 50.72(b)(3)(v)(D) this event is being reported as an event or condition that could have prevented the fulfillment of a safety function. Because the site had an individual briefed and at the door in constant communications with the control room to close the hatch if condition required such an action, this event is not viewed as an actual loss of safety function. The NRC Resident Inspector was notified.