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 Entered dateSiteRegionReactor typeEvent description
ENS 5329929 March 2018 16:42:00Quad CitiesNRC Region 3GE-3On March 29, 2018, Exelon Generation Company, LLC notified the Illinois Environmental Protection Agency (IEPA) and the Illinois Emergency Management Agency (IEMA) in accordance with state regulations of an unpermitted release of radionuclides at the Quad Cities Nuclear Power Station within the site boundary. There has been no detection of the liquid release beyond the site boundary. No impact to human health or the environment are anticipated. This notification is being made to satisfy 10CFR50.72(b)(2)(xi), notification of the NRC for any event related to the health and safety of the public for which a notification to other government agencies has been or will be made. The source of the Tritium release was from the Rad waste system. The spill was reported to be within the protected area which is within the site boundary. The quantity of the release is unknown at this point as the investigation and spill cleanup is in progress. The Licensee Notified the NRC Resident Inspector.
ENS 5328725 March 2018 23:43:00PilgrimNRC Region 1GE-3

On March 25, 2018 at 1616 hours (EDT), with the reactor in cold shutdown condition, two control rod drive piping lines were determined to be potentially inoperable in the event of a design basis earthquake due to support defects. The control rod drive piping forms a portion of the reactor coolant pressure boundary and primary containment boundary. The supports will be repaired prior to plant startup. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. The licensee will notify the Commonwealth of Massachusetts.

  • * * RETRACTION FROM JOE FRATTASIO TO HOWIE CROUCH AT 1500 EDT ON 4/13/18 * * *

The purpose of the notification is to retract ENS notification 53287 made on 03/25/18 for Pilgrim Nuclear Power Station. The previous notification reported that control rod drive (CRD) piping could be potentially inoperable in the event of a design basis earthquake, at the time of discovery, due to piping support defects. Subsequent evaluation has demonstrated that the piping was not inoperable. Specifically, after an engineering evaluation, it has been determined that the CRD Hydraulic System operability was never lost and the system was operable, although non-conforming, based on the support configuration not conforming to the pipe support drawings. The affected pipe supports have been restored or reworked to the proper design condition in accordance with the design drawings. The CRD System has subsequently been restored to a fully operable status. Notified R1DO (Jackson) and IRD MOC (Pham).

ENS 5328523 March 2018 21:07:00MonticelloNRC Region 3GE-3This report is made for a loss of Emergency Assessment Capability associated with Emergency Action Levels for Toxic and Flammable Gas and is reportable under 10 CFR 50.72(b)(3)(xiii). During an emergency equipment inventory, it was identified that methods were not available to detect levels of toxic or flammable gas at the IDLH (Immediately Dangerous to Life and Health) level for a number of substances due to the detector having an unsuitable range. The IDLH is used to assess the Emergency Action Level Alert Range. The ability of the Control Room Staff to detect and respond to the presence of toxic or flammable gas is unaffected. Because there have been no chemical spills or releases that would require sampling to be performed, the health and safety of the public was not affected. The resident NRC Inspector has been notified. The licensee will be notifying the state of Minnesota.
ENS 5325913 March 2018 15:54:00PilgrimNRC Region 1GE-3On March 13, 2018 at 1000 hours (EDT), with the reactor in Cold Shutdown condition, both 345kV incoming power lines and 23 kV Shutdown Transformer became unavailable during the Northeast winter storm. Per procedures, the emergency on-site emergency power supplies (Emergency Diesel Generators) were running and providing power to essential systems. In addition, the back-up Diesel Air Compressor was in service and one Reactor Protection System bus was on the back-up power supply prior to the loss. With both 345kV incoming power lines and 23 kV Shutdown Transformer unavailable, Pilgrim Nuclear Power Station procedures direct a report be made to the NRC per the requirements of Title 10 Code of Federal Regulations 50.72(b)(3)(v), any event that could have prevented the fulfillment of the safety function. No actual loss of safety function has occurred since the on-site emergency power supplies are maintaining the reactor in a safe shutdown condition and removing residual heat. The loss of incoming power is under investigation. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified.
ENS 532423 March 2018 02:19:00PilgrimNRC Region 1GE-3At 2315 EST on March 2, 2018, Pilgrim Nuclear Power Station (PNPS) determined, based on information received from the Commonwealth of Massachusetts, that there may be a potential loss of offsite response capabilities due to ongoing severe natural hazard conditions (i.e., major winter storm) along the coast of Massachusetts. According to information received by PNPS, towns within the 10 Mile EP Radius could be hampered in implementing some protective actions specified in the emergency plan in the unlikely event an emergency were to occur. There is no condition at the Station that would warrant implementation of any emergency plan at this time. PNPS continues to operate safely and is monitoring the weather conditions closely. The Station maintains emergency assessment, response, and communication capability. This report is being made conservatively in accordance with 10 CFR 50.72(b)(3)(xiii) which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. As stated previously, the Station maintains emergency assessment, response, and communication capability. The licensee notified the NRC Resident Inspector.
ENS 531987 February 2018 15:04:00DresdenNRC Region 3GE-3At approximately 1040 CST, seven (7) Dresden Nuclear Power Station Offsite Emergency Notification sirens (i.e., Siren Nos. DR1, DR4, DR5, DR6, DR9, DR10, and DR11) were inadvertently activated. The Kendall County, IL Emergency Management Agency notified the Exelon Generation Company, LLC. Emergency Response Organization that at 1040 CST, a contract individual inadvertently cut a wire that resulted in the actuation of these seven sirens for three minutes. The contract organization personnel are addressing the issue with the sirens. The Kendall and Will County Emergency Management Agency contacted Exelon Generation Company regarding this event. This notification is being made in accordance with 10 CFR 50.72(b)(2)(xi) as an event where other government agencies were notified. The NRC Resident Inspector has been informed of this notification. The sirens are operable.
ENS 5318931 January 2018 18:19:00Quad CitiesNRC Region 3GE-3At 1310 hours (CST) on January 31, 2018, the Unit 2B fuel pool radiation monitor spiked high due to an invalid actuation which caused the U1 and U2 reactor building ventilation system to isolate, B train standby gas treatment system (SBGTS) started, and the control room ventilation system also isolated as designed. Secondary containment vacuum was lost for approximately one minute, and then subsequently returned to an acceptable level in accordance with Technical Specification 3.6.4.1, 'Secondary Containment.' As a result of this transient, secondary containment was inoperable for approximately one minute. No emergency conditions were determined to exist. Troubleshooting of the radiation monitor spike is underway. Given the temporary loss of secondary containment vacuum, this event is reportable under 10CFR50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified.
ENS 531474 January 2018 17:57:00PilgrimNRC Region 1GE-3On January 4, 2018, at 1410 hours EST, with the reactor at approximately 100 percent power and steady state conditions, the winter storm across the Northeast caused the loss of offsite 345 kV Line 342. Reactor power was reduced to approximately 81 percent and a procedurally required manual reactor scram was initiated. All control rods fully inserted. As a result of the reactor scram, indicated reactor water level decreased, as expected, to less than +12 inches resulting in automatic actuation of the Primary Containment Isolation Systems for Group II - Primary Containment Isolation and Reactor Building Isolation System, and Group VI - Reactor Water Cleanup System. Reactor Water Level was restored to the normal operating band. The Primary Containment Isolation Systems have been reset. The Reactor Protection System signal has been reset. Following the reactor scram, the non-safety related Control Rod Drive Pump "B" tripped on low suction pressure. Control Rod Drive Pump "A" was placed in service. All other systems operated as expected, in accordance with design. This event is reportable per the requirements of Title 10, Code of Federal Regulations (CFR) 50.72 (b)(2)(iv)(B) - "RPS Actuation" and 10 CFR 50.72 (b)(3)(iv)(A) - "Specified System Actuation. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. The main steam isolation valves are open with decay heat being removed via steam to the main condenser. Offsite power is still available from 345kV line 355. As a contingency, emergency diesel generators are running and powering safety busses per licensee procedure. The licensee notified the Commonwealth of Massachusetts. The licensee will be notifying the town of Plymouth as part of their local notifications. The licensee will be issuing a press release.
ENS 5308923 November 2017 02:54:00Quad CitiesNRC Region 3GE-3

On November 22, 2017, at 2043 (CST), Unit I MCC (Motor Control Center) 18/19-5 overvoltage relay target was found actuated and would not reset. MCC 18/19-5 was powered from the normal feed, Bus 19. Bus 19 voltages were verified to be normal. The overvoltage relay actuation would result in MCC 18/19-5 being de-energized in the event of a DBA LOCA (Design Basis Accident Loss of Coolant Accident) in which the 1/2 Emergency Diesel Generator fails to energize Bus 18, therefore rendering both divisions of the Low Pressure Cooling Injection (LPCI) mode of Residual Heat Removal (RHR) system inoperable. Technical Specification 3.5.1 Condition E was entered, requiring restoration of LPCI in 72 hours. The overvoltage target was subsequently able to be reset at 2114 (CST), restoring the LPCI function of RHR. Technical Specification 3.5.1 Condition E was exited at that time. This event is reportable under 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector.

  • * * RETRACTION FROM RONALD SNOOK TO STEVEN VITTO ON 01/11/18 AT 1913 EST * * *

The purpose of this notification today (01/11/18) is to retract the ENS Report made on November 23, 2017 at 0248 hours EST (ENS Report #53089). Upon further review, it was determined that the Unit 1 MCC 18/19-5 overvoltage relay target that was found actuated and would initially not reset was caused only by intermittent degraded DC control power. During this event, MCC 18/19-5 remained powered from the normal feed Bus 19, and Bus 19 voltages were verified to be normal. It was further determined from plant drawings that under this condition the degraded DC control power to the Unit 1 MCC 18/19-5 overvoltage relay has no impact to the Technical Specification 3.5.1 required capability to auto transfer power from the normal Bus 19 to the alternate Bus 18 should Bus 19 lose power such as during a DBA LOCA. This overvoltage relay was installed in the early 1990's only to support enhanced reliability of the power supply to the LPCI injection valves, and its actuation due to degraded DC control power would not impact the ability to auto transfer to alternate Bus 18. Therefore, both divisions of the Low Pressure Cooling Injection (LPCI) mode of Residual Heat Removal (RHR) system would have remained fully operable under the as-found relay condition, and Technical Specification 3.5.1 Condition E was not required to be entered. On December 6, 2017, it was determined that a loose fuse clip terminal had caused the DC control power to the overvoltage relay to become degraded which in turn caused the relay target and its reset to become erratic. This fuse clip terminal was repaired on December 6, 2017. Based on the subsequent reviews of this event, the LPCI system was not required to be declared inoperable in accordance with Technical Specifications 3.5.1 during the period of the MCC 18/19-5 overvoltage relay actuation (i.e., 31 minutes on 11/22/17), and hence was not required to be reported under 10CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented fulfillment of a safety function. Therefore, based on this information, ENS Report #53089 is being retracted. The NRC Resident Inspector has been notified. R3DO(Jeffers) has been notified.

ENS 530597 November 2017 22:09:00Quad CitiesNRC Region 3GE-3On November 7, 2017 at 1810 (CST), Unit 1 High Pressure Coolant Injection (HPCI), was manually isolated following failure of the remote turbine trip pushbutton to function. Unit 1 HPCI Operability Testing was in progress to the point of securing the HPCI turbine with the remote manual pushbutton. The pushbutton failed to trip the turbine resulting in operator action to lower the flow controller setpoint and isolating the HPCI steam line. HPCI remains isolated and is Inoperable pending resolution of the Turbine Trip circuitry. This event is being reported as a condition that could have prevented fulfillment of a safety function in accordance with 10CFR50.72(b)(3)(v)(D). The HPCI system is a single train system and the loss of HPCI could impact the plant ability to mitigate the consequences of an accident. The Reactor Core Isolation Cooling (RCIC) system was confirmed operable. The NRC Senior Resident Inspector has been notified.
ENS 530481 November 2017 16:33:00DresdenNRC Region 3GE-3On November 1, 2017 at 1225 CDT, both the 2-220-58A Feed Water Inboard Check Valve and the 2-220-62A Feed Water Outboard Check Valve failed Local Leak Rate Testing (LLRT) acceptance criteria due to excessive leakage. These valves are considered primary containment isolation valves, and as such, are required to ensure that an adequate primary containment boundary is maintained. Technical Specification (TS) 5.5.12, 'Primary Containment Leakage Rate Testing Program,' establishes limits for Primary Containment leakage. Based upon the results of the LLRT, Dresden, Unit 2, may not have met the limits for primary containment leakage during the last operating cycle as specified in TS 5.5.12.C. Dresden Unit 2 is currently in Mode 5 for a refueling outage and per Dresden TS 3.6.1.1, 'Primary Containment,' Primary Containment is not required in the current mode of operation (i.e., Mode 5). However, in accordance with 10 CFR 50.72(b)(3)(ii)(A), this event is reportable as a condition that resulted in a principal safety barrier being seriously degraded. The NRC Resident Inspector has been notified.
ENS 5304130 October 2017 10:38:00PilgrimNRC Region 1GE-3There was a loss of power from the local grid which did not affect the power block. The support buildings lost power and a UPS failed which affects computers, switching, and telephones. This includes a loss of the Emergency Response Data System (ERDS). The Joint Information Center and Emergency Operations Facility were not affected. Though this is a major loss of communications ability, alternate communications methods are available. The licensee notified the NRC Resident Inspector.
ENS 5298421 September 2017 21:07:00Quad CitiesNRC Region 3GE-3On September 21, 2017 at 1730 (CDT) the Control Room Emergency Ventilation Air Condition (CREV AC) system was declared inoperable due a refrigerant leak from the air conditioning compressor. As a result, Technical Specification 3.7.5 Condition A was entered for Units One and Two. The CREV AC system maintains a habitable control room environment and ensures the operability of components in the control room emergency zone during accident conditions. This notification is being made in accordance with 10CFR50.72(b)(3)(v)(D) because the CREV system is a single train system. The loss of CREV AC could impact the plant's ability to mitigate the consequences of an accident. The NRC Resident Inspector has been notified.
ENS 5296412 September 2017 14:20:00DresdenNRC Region 3GE-3On September 12, 2017 at 1131 CDT, both Unit 3 Standby Liquid Control system (SLC) subsystems were declared inoperable for a through wall leak on the common discharge piping. With both subsystems inoperable, the SLC system was unable to fulfill its safety function. This event is reportable under 10 CFR 50.72(b)(3)(v)(A) for an event or condition that could have prevented the fulfillment of a safety function of a system that is needed to shut down the reactor and maintain it in a safe shutdown condition and under 10 CFR 50.72(b)(3)(v)(D) for a system that was unavailable for accident mitigation. The NRC Resident Inspector has been notified. With both trains of SLC inoperable, the licensee entered an 8-hr. action statement to restore at least one train to operability. If unable to do so, then the plant will enter a 12-hr. shutdown action statement.
ENS 529558 September 2017 16:45:00Quad CitiesNRC Region 3GE-3

On September 8, 2017 at 1130 hours CDT, Unit Two High Pressure Coolant Injection (HPCI) Minimum Flow Valve MO 2-2301-14 flow indicating switch (FIS 2-2354) failed to meet the Technical Specification Allowable Value during calibration testing. Technical Specification Table 3.3.5.1-1 Allowable Value (3.f) requires greater than or equal to 634 gpm (3.14 inches water column as required by procedure). HPCI was subsequently declared inoperable. This event is being reported as a condition that could have prevented fulfillment of a safety function in accordance with 10CFR50.72(b)(3)(v)(D). The HPCI system is a single train system and the loss of HPCI could impact the plant's ability to mitigate the consequences of an accident. The Reactor Core Isolation Cooling (RCIC) system was confirmed operable. Note: On September 8, 2017 at 1140 hours CDT, the HPCI Minimum Flow Valve MO 2-2301-14 flow indicating switch (FIS 2- 2354) was successfully recalibrated and HPCI was returned to Operable status. The NRC Senior Resident Inspector has been notified.

  • * * RETRACTION AT 1216 EDT ON 10/19/17 FROM RYAN DECKER TO DONG PARK * * *

The purpose of this notification today (10/19/17) is to retract the ENS Report made on September 8, 2017 at 1545 hours CDT (ENS Report #52955). Upon further investigation, it was determined that a surveillance procedure contained an overly restrictive statement that directed operators to immediately declare the High Pressure Coolant Injection (HPCI) system inoperable when the HPCI Minimum Flow Valve MO 2-2301-14 flow indicating switch (FIS 2-2354) fails. This statement was in conflict with existing Technical Specification (TS) 3.3.5.1, Condition E, that allows seven days to restore the HPCI FIS (instrument channel only) to an operable status prior to entry into TS 3.3.5.1, Condition H, which requires declaring HPCI inoperable immediately. Hence, during the period of FIS inoperability (i.e., 10 minutes), the HPCI system was not required to be declared inoperable in accordance with Technical Specifications. Therefore, based on this information, ENS Report # 52955 is being retracted. Note: On September 8, 2017 at 1140 hours CDT, the HPCI Minimum Flow Valve MO 2-2301-14 flow indicating switch (FIS 2-2354) was successfully recalibrated and HPCI was returned to Operable status. The NRC Resident Inspector has been notified. Notified R3DO (Daley).

ENS 5282022 June 2017 20:33:00PilgrimNRC Region 1GE-3On June 20, 2017, at 1444 hours (EDT), with the reactor at 100% core thermal power and steady state conditions, plant personnel identified that both doors in one of the secondary containment airlocks (Door #58 and Door #85) were open briefly as part of normal passage of personnel. The Technical Specification definition of SECONDARY CONTAINMENT INTEGRITY states 'At least one door in each access opening is closed.' Actions were taken to immediately close both doors and restore operability of secondary containment. PNPS (Pilgrim Nuclear Power Station) is providing an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(v)(C), 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. The licensee notified the Commonwealth of Massachusetts.
ENS 5281420 June 2017 08:18:00MonticelloNRC Region 3GE-3At 2353 CDT on 6/19/2017, while performing the High Pressure Coolant Injection (HPCI) quarterly surveillance following planned maintenance, the HPCI turbine did not start as expected due to the HPCI turbine stop valve failing to open. This issue is being reported under 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function at the time of discovery. Investigation into the failure of the HPCI system to start is in progress. The unit remains at 100% power. The health and safety of the public was not affected. The NRC Resident Inspector has been notified.
ENS 527811 June 2017 23:41:00MonticelloNRC Region 3GE-3Planned maintenance to restore normal power to Plant Computer Systems resulted in an unexpected loss of all Meteorological (MET) Tower Data (at 1645 CDT). As a result, this represents a Loss of Emergency Assessment Capability and is reportable under 10CFR 50.72 (b)(3)(xiii). The isolation was restored and MET Tower Data was restored at 1845. The health and safety of the public was not affected as the plant is operating in a normal condition with no severe weather or storms in the area. Additionally meteorological data was available from the National Weather Service should this data had been necessary. The NRC Resident Inspector has been notified." The licensee will be notifying the State of Minnesota.
ENS 5275816 May 2017 00:27:00Quad CitiesNRC Region 3GE-3On May 15, 2017 at 1918 hours (CDT), Unit Two High Pressure Coolant Injection (HPCI) Minimum Flow Valve MO 2-2301-14 failed to open as required by procedure and HPCI was declared inoperable. When the HPCI Turbine was tripped, the Minimum Flow Valve did not open when system flow reduced to the low flow setpoint. This event is being reported as a condition that could have prevented fulfillment of a safety function in accordance with 10CFR50.72(b)(3)(v)(D). The HPCI system is a single train system and the loss of HPCI could impact the plant's ability to mitigate the consequences of an accident. In accordance with Technical Specification 3.5.1 Condition G, the Reactor Core Isolation Cooling (RCIC) system was confirmed operable. This places the plant in a 14-day LCO action statement. The licensee has notified the NRC Resident Inspector.
ENS 5274911 May 2017 18:11:00MonticelloNRC Region 3GE-3A can of alcohol (16.9 ounce foreign beer) was discovered unopened in an administration building refrigerator. Site security took possession of the can of alcohol. The owner of the can of alcohol is unknown. This licensee is making this 24 hour notification in accordance with 10CFR26.719(b)(1). The licensee notified the NRC Resident Inspector.
ENS 5274410 May 2017 21:40:00PilgrimNRC Region 1GE-3On Wednesday May 10, 2017, at 1411 EDT, with the reactor at 0 percent core thermal power (CTP), Pilgrim Nuclear Power Station (PNPS) was in a Refueling Outage, performing a review of Local Leak Rate Testing results, when it was concluded that PNPS had exceeded its Title 10 Code of Federal Regulations Part 50, Appendix J, Option B, Type B and C Local Leak Rate Test (LLRT) leakage criteria. Previously, on April 22, 2017, when PNPS was performing LLRT of the High Pressure Coolant Injection (HPCI) steam exhaust line check valves, both valves failed to meet their LLRT acceptance criteria specified in plant procedures. Neither of the check valves seated acceptably. Based on the ongoing evaluation of these test exceedances, it was concluded that these test results cause the plant to exceed the overall as-found minimum path Appendix J acceptance criteria of 0.6 La (126.3 SLM (Standard Liters per Minute)). Further investigation is ongoing. This event has no impact on the health and safety of the public. The licensee has notified the NRC Senior Resident Inspector. This notification is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A), any event or condition that resulted in the condition of the nuclear power plant, including its principle safety barriers, being seriously degraded. The licensee will notify the Commonwealth of Massachusetts Emergency Management Agency.
ENS 5271528 April 2017 21:21:00MonticelloNRC Region 3GE-3This report is being made pursuant to 10 CFR 50.72(b)(2)(xi), as an event where notification to other government agencies has been made. On April 28, 2017, notification to the Minnesota State Duty Office was made due to a non-compliance with release of wastewater requirements in the Monticello Nuclear Generating Plant's National Pollutant Discharge Elimination System permit. There were no consequences to the health and safety of the public. The licensee notified the NRC Resident Inspector.
ENS 5268215 April 2017 13:03:00MonticelloNRC Region 3GE-3During shutdown activities with the reactor subcritical, actions were being taken to remove 11 Reactor Feed Pump from service in support of a scheduled refueling outage. Reactor Water Level on Safeguards level instrumentation dropped below +9 inches, which resulted in a valid Reactor Protection System (RPS) Scram signal and Partial Group 2 Primary Containment Isolation System (PCIS) signal. All systems functioned as required. Reactor Water Level on Safeguards instrumentation was restored to greater than +9 inches immediately. RPS and PCIS logic was reset. There was no impact to the health and safety of the public as a result of this event. This actuation of these systems is being reported per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of any of the systems listed in 10 CFR 50.72(b)(3)(iv)(B). The NRC Resident Inspector has been notified.
ENS 5265531 March 2017 19:14:00PilgrimNRC Region 1GE-3On March 31, 2017 at 1155 hours (EDT), with the reactor at 97% core thermal power and steady state conditions, operators inadvertently caused water level to rise in the Pressure Suppression Pool (TORUS). Pilgrim Nuclear Power Station (PNPS) was restoring normal system valve line-ups after performing flushing of the suction piping of the Core Spray system in accordance with station procedures. During the process of restoring the appropriate valve line-ups, water was inadvertently transferred to the TORUS from the Condensate Storage Tank. The cause of the event is understood. The Technical Specification (TS) Limiting Condition for Operation (LCO) Action Statement (AS) 3.7.A.5 was entered. The LCO AS was exited at 1540 when TORUS water level was restored to the limits specified in LCO's 3.7.A.1.b and 3.7.A.1.m. Because the TORUS was declared inoperable, PNPS is providing an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(v)(D), an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident. This was a case of the water level in the TORUS being above the TS limit. The TORUS was potentially available to provide cooling to the reactor if required. The NRC Resident Inspector has been notified. The licensee notified the Commonwealth of Massachusetts and Plymouth County.
ENS 5264327 March 2017 21:54:00PilgrimNRC Region 1GE-3On March 27, 2017, at 1825 hours EDT, with the reactor at 100 percent core thermal power and steady state conditions, technicians inadvertently caused a High Pressure Coolant Injection (HPCI) System isolation, by testing the incorrect temperature switches in the TIP (Traversing In-core Probe) room. Pilgrim Nuclear Power Station (PNPS) was performing testing on the temperature switches for Reactor Core Isolation Cooling (RCIC), but the HPCI temperature switches were inadvertently actuated causing HPCI to isolate. The Limiting Condition for Operation (LCO) Action Statement 3.5.c.2 has been entered and the planned testing has been secured pending further investigation. PNPS is providing an 8-hour non-emergency notification that the HPCI System was declared inoperable in accordance with 10 CFR 50.72(b)(3)(v)(D), an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident. HPCI was returned to Operable within 40 minutes. The licensee notified the NRC Resident Inspector and the Commonwealth of Massachusetts.
ENS 5258228 February 2017 21:34:00DresdenNRC Region 3GE-3

At 1825 (CST) on 02/28/2017, Dresden Station received unexpected alarm 923-5 C-1, RX BLDG DP LO (Reactor Building Differential Pressure Low). Reactor Building differential pressure was observed to briefly lose vacuum and return to a normal reading of 0.6 inches vacuum water gauge. At the time of the transient, Grundy County was under a Severe Weather Warning and gusts of wind were being monitored from the Main Control Room up to approximately 57 mph. The Reactor Building differential pressure returned to (greater than or equal to) 0.25 inches vacuum water gauge at 18:25 after 18 seconds with no operator action. Operators observed differential pressure reading to lose vacuum, below 0 inches vacuum water gauge, for approximately 3 seconds. This condition represents a failure to meet Surveillance Requirement 3.6.4.1.1. As a result, entry into Technical Specifications 3.6.4.1 condition A was made due to Secondary Containment being inoperable. This event is being reported in accordance with 10CFR50.72(b)(3)(v)(C) as a condition that could have prevented the fulfillment of a safety function. An issue report has been initiated and a 60-day Licensee Event Report will be submitted in accordance with 10CFR50.73(a)(2)(v)(C). The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 4/14/17 AT 1429 EDT FROM BOBBY SHORT TO DONG PARK * * *

The purpose of this notification is to retract ENS notification 52582 made on 2/28/17 for Dresden Nuclear Power Station. The previous notification reported a potential loss of Reactor Building differential pressure due to high wind conditions and associated entry into Technical Specification 3.6.4.1 Condition A for failure to meet Surveillance Requirement 3.6.4.1.1 to maintain differential pressure above 0.25 inches vacuum water gauge. After further evaluation, it has been determined that the high winds caused a momentary low pressure pocket on the leeward side of the Reactor Building causing the differential pressure reading seen in the Main Control Room, but it did not challenge Reactor Building differential pressure or Secondary Containment. Reactor Building differential pressure indication utilizes four transmitters, one on each wall of the Reactor Building, and the most conservative reading is transmitted to the indicator in the Main Control Room. Wind conditions impacting a single transmitter would result in indication of low Reactor Building differential pressure in the Main Control Room. Procedures direct action to obtain local readings from all four Reactor Building differential pressure transmitters. After the low indication in the Main Control Room, Equipment Operators were dispatched to obtain local indication and all four transmitters were found to be indicating 0.6 inches vacuum water gauge. This was a short duration transient with no indications of an equipment failure that could impact Secondary Containment. The entire transient occurred within an 18 second window where differential pressure indication began at 0.6 inches vacuum water gauge, dropped to below 0 inches vacuum water gauge, and subsequently restored to 0.6 inches vacuum water gauge with no operator intervention. Furthermore, a significant change in Reactor Building differential pressure would impact readings on Drywell pressure because the Reactor Building pressure is used as a reference leg. Trends of Drywell pressure during the event indicated no adverse conditions implying that Reactor Building differential pressure was stable. Thus, it has been concluded that this was an indication issue and at no point during the transient would Secondary Containment have been unable to perform its safety function. Therefore, this event does not meet the criteria of 10 CFR 50.72(b)(3)(v)(C) as a condition that could have prevented the fulfillment of a safety function, and the ENS notification is being retracted. There is no longer a requirement for an associated 60-day Licensee Event Report in accordance with 10 CFR 50.73(a)(2)(v)(C). The NRC Resident Inspector has been notified. Notified R3DO (Jeffers).

ENS 525428 February 2017 17:06:00DresdenNRC Region 3GE-3

At 0851 CST on Wednesday, February 8th, 2017, the Dresden Nuclear Power Station (DNPS) Technical Support Center (TSC) Emergency Ventilation System was emergently declared inoperable due to a failure of the outside air damper to reposition. This resulted in the inability for the TSC ventilation to maintain the required air flow to support habitability during emergency conditions. Actions are being taken to repair damper to restore functionality of the TSC ventilation system. In the interim, station procedures provide guidance to relocate the TSC to an alternate facility. This event is being reported under 10 CFR 72(b)(3)(xiii), Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system). The NRC Resident Inspector has been notified.

  • * * UPDATE PROVIDED BY RYAN CHAMBERLAIN TO JEFF ROTTON AT 0418 EST ON 02/10/2017 * * *

At 0108 CST on February 10, 2017, Dresden TSC ventilation has been restored and is now functional. The NRC Resident Inspector has been notified. Notified R3DO (Kunowski).

ENS 525271 February 2017 21:22:00Quad CitiesNRC Region 3GE-3

On February 1, 2017, at 1929 hours (CST), a fire was discovered on the Unit 2 Main Control Room panel 902-3 in the 3E ERV/ADS valve switch. A reactor SCRAM was not required. No automatic isolations/actuations occurred. The fire was extinguished at 1932 and the reactor remained at 100% power. An Alert was declared at 1938 (CST). The initiation of the event was attempting to change a light bulb. The cause of the event is under investigation. The Senior Resident Inspector has been notified of the event. The licensee entered a 14 day Technical Specification Action statement as a result of the damage to the switch. Notified DHS SWO, DOE, FEMA, HHS, NICC, USDA, EPA, FDA, NNSA (e-mail), and NRCC SASC (e-mail).

  • * * UPDATE AT 0040 EST ON 2/2/17 FROM DAVID KNEPPER TO JEFF HERRERA * * *

Notified that the Alert was terminated at 2336 CST on 2/1/17. The licensee stated that the fire was extinguished and an extent of condition walkdown did not identify any additional equipment damage as a result of the fire. The licensee will be issuing a press release. The licensee will be notifying the NRC Resident Inspector. Notified the R3DO (Duncan), IRD MOC (Gott), NRR EO (Miller), DHS SWO, DOE, FEMA, HHS, NICC, USDA, EPA, FDA, NNSA (e-mail), and NRCC SASC (e-mail).

ENS 5250824 January 2017 14:00:00Quad CitiesNRC Region 3GE-3On January 24, 2017, at 1000 hours (CST), Operations was notified that two Secondary Containment interlock doors (between the Unit 2 Reactor Building and Unit 2 Turbine Building) were open simultaneously. The doors were immediately closed and Secondary Containment pressure remained negative. Unit 1 and Unit 2 share secondary containment. This condition represents a failure to meet Surveillance Requirement 3.6.4.1.2 given two doors in a single access opening were open. As a result, entry into Technical Specification 3.6.4.1, Condition A. was made momentarily due to Secondary Containment being inoperable. This event is reportable under 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function. The NRC Senior Resident Inspector has been notified. The cause of this event was due to an equipment interlock (solenoid) failure and the doors are currently blocked closed.
ENS 5249216 January 2017 16:39:00PilgrimNRC Region 1GE-3On January 16, 2017, with the reactor at 100 percent and the mode switch in RUN, Pilgrim Station was performing preventative maintenance of secondary containment isolation dampers when dampers AO-N-82 and AO-N-83, refueling floor supply isolation dampers, failed to fully close when the control switches were taken to close. This 8-hour non-emergency notification is being made in accordance with 10 CFR 50.72(b)(3)(v), Any event or condition that at the time discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. The reactor building isolation dampers were cleaned and lubricated and post-work tested and timed in accordance with station procedures to verify that they had satisfactory closing times. Pilgrim Nuclear Power Station has returned the dampers to Operable status. The licensee will notify the Commonwealth of Massachusetts. The licensee has notified the NRC Resident Inspector.
ENS 5245421 December 2016 18:00:00MonticelloNRC Region 3GE-3

At 0935 (CST) on 12/21/2016, while performing the High Pressure Coolant Injection (HPCI) Comprehensive Pump and Valve Tests for post-maintenance testing following scheduled maintenance, the HPCI turbine did not start as expected due to the HPCI turbine stop valve failing to open. This issue is being reported under 10CFR50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function at the time of discovery. Investigation into the failure of the HPCI system to start is in progress. The plant remains at 100% power with no challenges to the health and safety of the public. The NRC Resident Inspector has been notified. The plant is in a 14-day action statement under LCO 3.5.1, 'ECCS - Operating' due to the HPCI turbine stop valve failure. The licensee notified the Minnesota State Duty Officer.

  • * * RETRACTION FROM KIM HOFFMAN TO JOHN SHOEMAKER AT 1303 EST ON 1/17/18/17 * * *

On December 21, 2016, the NRC Operations Center was notified of Event Number 52454 that described a failure of the High Pressure Coolant Injection (HPCI) turbine stop valve to open during post maintenance testing prior to being declared operable. The condition was reported in accordance with 10 CFR 50.72 (b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function at the time of discovery. At the time, it was not readily apparent that the failure was due to the maintenance activities. Subsequent return-to-service testing showed the oil system vent and fill had been inadequate following the maintenance. This event occurred as a result of the maintenance process and would not have occurred during normal operation of the system. NUREG-1022, Revision 3 states, 'reports are not required when systems are declared inoperable as part of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS (unless a condition is discovered that would have resulted in the system being declared inoperable).' There was no discovered condition that would have resulted in the safety function of the system being declared inoperable under normal, non-maintenance conditions. Based on the above additional information, Monticello Nuclear Generating Plant is retracting this report. The plant was in a planned evolution and did not discover a condition that could have prevented performing a safety function. The licensee has notified the NRC Resident Inspector. Notified R3DO (McCraw).

ENS 5245121 December 2016 16:38:00MonticelloNRC Region 3GE-3This 60-day telephone notification is being submitted in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to report an invalid actuation of the 12 Emergency Diesel Generator Emergency Service Water pump (12 ESW pump). At 1745 (CST) on October 24, 2016, an unexpected auto-start of the 12 ESW pump occurred. The 12 Emergency Diesel Generator (12 EDG), was previously properly removed from service and isolated for scheduled maintenance. Upon investigation, is was determined that no valid start signal was present and actuation occurred during relay replacement activities on the 12 EDG in C-92 (12 EDG (G-38) electrical control panel) cabinet when electricians inadvertently bumped a 12 EDG start relay. During this period, the Control Room received annunciators indicating the 12 EDG engine was running/cranking and the 12 ESW pump started. Due to being isolated, the 12 EDG did not actually start. The licensee notified the NRC Resident Inspector.
ENS 5244920 December 2016 19:45:00PilgrimNRC Region 1GE-3At 1830 EST on 20 December, 2016 the Massachusetts Department of Environmental Protection and the Plymouth Massachusetts Fire Department were notified of a Hydrogen release in accordance with plant procedures and 310CMR40.300, Massachusetts Contingency Plan Notification for Oil and Hazardous Material; Identification and Listing of Oil and Hazardous Material, due to a release of hydrogen gas to the environment exceeding the reportable quantity of ten pounds. The release, which is an expected part of a routine plant start-up was from the generator hydrogen cooling system. This event posed no danger to the health and safety of plant personnel or members of the general public. The NRC Resident Inspector has been notified.
ENS 5239627 November 2016 21:22:00MonticelloNRC Region 3GE-3At 1447 (CST) on 11/27/2016 while troubleshooting a minor leak on the High Pressure Coolant Injection (HPCI) turbine, it was discovered that the HPCI turbine exhaust drain pot high level bypass switch was not functioning per design to support removal of condensate from the HPCI turbine casing. This resulted in some water accumulation within the HPCI turbine casing. Subsequently, HPCI was declared INOPERABLE and this issue is being reported under 10CFR50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function at the time of discovery. The plant remains at 100 percent power with no challenges to the health and safety of the public. The NRC Resident Inspector has been notified. Technical Specification limiting condition for operation requires HPCI to be Operable within 14 days. The licensee will be notifying the State of Minnesota regarding the event.
ENS 5237016 November 2016 18:22:00DresdenNRC Region 3GE-3

At 1105 CST on Wednesday, November 16th, 2016, the Dresden Nuclear Power Station (DNPS) Technical Support Center (TSC) Emergency Ventilation System was emergently declared inoperable due to an unplanned loss of the pneumatic air supply compressor. The loss of the air compressor resulted in the emergency air filtration unit flow control damper failing to the full open position. In this condition, the emergency air filtration unit could exceed the max flow rate of 1100 SCFM and the max differential pressure of 6 inches H2O at rated flow rate resulting in degraded performance. This results in a potential loss of protective action function provided by the emergency ventilation filtration system and could impede the ability to perform Emergency Assessments should a radiological emergency event occur requiring the system to be in service. Actions are being taken to repair the pneumatic air system to restore functionality of the TSC ventilation system. In the interim, contingency actions are being developed to manually control the emergency air filtration unit flow control damper in a degraded condition. In the event that ventilation cannot be restored, Station Procedures provide guidance to relocate the TSC to an alternate facility. This event is being reported under 10 CFR 50.72(b)(3)(xiii), 'Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).' The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM ADAM SCHUERMAN TO HOWIE CROUCH AT 0812 EST ON 11/17/16 * * *

The pneumatic air supply compressor has been repaired and tested satisfactorily. The Technical Support Center is now considered operable. The licensee will be notifying the NRC Resident Inspector. Notified R3DO (Peterson).

ENS 5236816 November 2016 14:32:00Quad CitiesNRC Region 3GE-3On November 16, 2016 at approximately 1010 CST, a local government agency (Whiteside County, Illinois) inadvertently activated emergency response sirens for less than one minute. The inadvertent actuation occurred during the scheduled Quad Cities Station emergency planning graded exercise while local government agencies were participating. A related news release and radio message was subsequently issued by Whiteside County to report that Whiteside County was participating in a drill in coordination with the Quad Cities Exelon Generating Station that the emergency response sirens were inadvertently activated, and that there is no emergency at this time and no action is required. There was no impact to the health and safety of the public as a result of this event as the offsite response capabilities remain functional. The site is operating normally with no emergency conditions present. This event is being reported under 10 CFR 50.72(b)(2)(xi), as an inadvertent activation of emergency response sirens and news release. The NRC Resident Inspector has been notified.
ENS 523558 November 2016 12:50:00DresdenNRC Region 3GE-3On 11/8/16, Operators were performing Division I Undervoltage Testing Surveillance on Unit 3, when a 2 (psi) drywell signal was inserted, Reactor Building Ventilation tripped and SBGT initiated as expected. At 0510 (CST), reactor building to atmosphere differential pressure dropped below the (negative) 0.25 inches water. This condition represents a failure to meet Surveillance Requirement 3.6.4.1.1. As a result, entry into Technical Specification 3.6.4.1 condition A was made due to Secondary Containment becoming inoperable. This event is being reported in accordance with 10CFR 50.72(b)(3)(v)(C) as a condition that could have prevented the fulfillment of a safety function. At 0532, the 2/3 Reactor Building material interlock inner door was closed and Reactor Building (differential pressure) was restored to greater than (negative).25 inches of water column. An issue report has been initiated. An investigation will be conducted and a 60 day Licensee Event report will be submitted in accordance with 10 CFR 50.73(a)(2)(v)(C). The NRC Resident Inspector has been notified.
ENS 523527 November 2016 20:44:00PilgrimNRC Region 1GE-3On November 7, 2016, at 1609 (EST), with the reactor at 100 percent core thermal power and steady state conditions, the High Pressure Coolant Injection (HPCI) system was declared inoperable. Pilgrim Nuclear Power Station (PNPS) was performing planned quarterly testing per Technical Specifications 4.13.A.1. During a review of the HPCI pump data taken during the test, it was determined that the recorded vibration reading on the Main Pump Outboard horizontal point (P4H) was 0.8335 in./sec which exceeds the IST required action range high limit of less than or equal to 0.830 in./sec. Accordingly, the HPCI pump was declared inoperable. The Limiting Condition for Operation Action Statement 3.5.C.2 has been entered and planned troubleshooting Into the cause of the high vibration is in progress. In accordance with 10 CFR 50.72(b)(3)(v)(D), PNPS is providing an 8-hour non-emergency notification that the HPCI System is inoperable. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. The licensee will be notifying the State of Massachusetts regarding the event.
ENS 5233431 October 2016 09:27:00Quad CitiesNRC Region 3GE-3

On October 31, 2016, at 0239 hours (CDT), a defect (minor audible through-wall leak) was identified on the steam line drain valve 1-2301-55, HPCI Steam Line Drain Line Steam Trap Outlet Valve. The defect was identified by Operations personnel traversing through the HPCI room as part of normal rounds. HPCI was declared inoperable under Tech Specs 3.5.1, Condition G. The Reactor Core Isolation Cooling (RCIC) system was verified operable. HPCI remains available (but not operable). The leak has been isolated. The 1-2301-55 is a manual valve downstream of the HPCI steam line drain trap. In a standby line-up, this line drains condensation from the HPCI steam supply line to the main condenser. During operation in an accident scenario, this line drains condensation from the HPCI steam supply line to the Torus via a drain pot. The location of the defect is in class 2 safety related piping. HPCI is a single train safety system and this notification is being made in accordance with 10CFR50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified. Technical Specification 3.5.1, condition G requires that HPCI be Operable within 14 days.

  • * * RETRACTION ON 12/05/2016 AT 1505 EST FROM MARK BRIDGES TO STEVEN VITTO * * *

The purpose of this notification is to retract the ENS Report made on October 31, 2016, at 0239 hours CDT (ENS Report #52334). Upon further investigation, a pinhole through-wall leak was discovered in the body of the 1-2301-55 valve (HPCI Steam Line Drain Line Steam Trap Outlet Valve). The defect was characterized as a 1/32-inch rounded hole due to a manufacturing defect in the casting located on the downstream side of the valve near the piping connection. A subsequent evaluation performed by Quad Cities Station considering the defect size, location, and characterization, confirmed the Unit 1 High Pressure Coolant Injection (HPCI) system would have performed its safety function when required. Based on this subsequent evaluation, ENS Report 52334 is being retracted. Note: On November 1, 2016, at 1624 hours CDT, the 1-2301-55 valve (HPCI Steam Line Drain Line Steam Trap Outlet Valve) was successfully repaired and HPCI was returned to Operable status. The NRC Resident Inspector has been notified. Notified R3DO (Stone).

ENS 5225319 September 2016 21:40:00DresdenNRC Region 3GE-3

At 1550 (CDT) on September 19, 2016, Dresden received the Methyl Iodide Penetration test results for the Control Room Emergency Ventilation (CREVS) charcoal. The test results did not meet technical specification acceptance criteria. This results in the inoperability of CREVS. CREVS is a single train system and therefore is reportable per 10CFR50.72(b)(3)(v)(D). The Air Filtration Unit (AFU) is required to operate during a design basis accident to maintain Main Control Room habitability. This places unit 2 and unit 3 in a 7 day LCORA (Limiting Condition of Operation Required Action) per Tech Spec 3.7.4 Required Action A.1. The licensee notified the NRC Resident Inspector.

  • * * UPDATE AT 1635 EDT ON 03/23/17 FROM HENRY WATERS TO S. SANDIN * * *

The licensee is retracting this report based on the following: The purpose of this notification is to retract ENS notification 52253 made on September 19th, 2016, for Dresden Nuclear Power Station. After further evaluation and testing, it has been determined that the Control Room Emergency Ventilation System (CREVS) charcoal would have fulfilled its safety function given the Methyl Iodide Penetration test results. The initial tests were performed with a 2 inch bed depth due to a difference in batches used in each charcoal filter, but testing at a 4 inch bed depth is the correct testing methodology for Dresden's configuration. At a 4 inch bed depth, the test results met the Technical Specification acceptance criteria with significant margin. Therefore, this event does not meet the criteria of 10 CFR 50.72(b)(3)(v)(D) and the ENS report is being retracted. The NRC Resident Inspector has been notified. Notified R3DO (Orlikowski).

ENS 522319 September 2016 20:01:00PilgrimNRC Region 1GE-3

At 1739 (EDT) on Friday September 9, 2016 the Massachusetts Department of Environmental Protection and the Plymouth Massachusetts Fire Department were notified of a Hydrogen release in accordance with plant procedures and 310CMR40.300, Massachusetts Contingency Plan Notification for Oil and Hazardous Material; Identification and Listing of Oil and Hazardous Material, due to a release of hydrogen gas to the environment exceeding the reportable quantity of ten pounds. The Massachusetts DEP Tracking Number is RTN4-26311. The release was from the generator hydrogen cooling system. There was no plant damage. Hydrogen system pressure has been restored to the normal operating band with the Main Generator secured and is stable. The cause of the event is under investigation. This event posed no danger to the health and safety of plant personnel or members of the general public. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM ROBERT O'NEILL TO STEVEN VITTO AT 1724 EDT ON 09/14/2016 * * *

The Plymouth Massachusetts Fire Department was notified on Monday, September 12, 2016, at 1411 EDT. This clarifies information applicable to the local notification as identified in the original notification. Notified R1DO(Noggle) and NSIR (Stapleton) via email.

ENS 522236 September 2016 11:24:00PilgrimNRC Region 1GE-3On Tuesday, September 6, 2016 at 0827 (EDT), with the reactor at 91% core thermal power (CTP), Pilgrim Nuclear Power Station (PNPS) operators initiated a manual reactor scram due to high reactor water level resulting from feedwater level control oscillation. Other than the feedwater level control oscillations, all other plant systems responded as designed. Plant cooldown is in progress using the High Pressure Coolant Injection System in the pressure control mode. The plant is in hot shutdown. The cause of the feed water level control oscillations is under investigation. This event has no impact on the health and safety of the public. Subsequent to the manual reactor scram the plant experienced the following isolation signals: Group 1 Isolation: Main Steam Isolation Valves Group 2 Isolation: Miscellaneous containment isolation valves Group 6 Isolation: Reactor Water Clean-up Reactor Building (Ventilation) Isolation Actuation The licensee has notified the NRC Senior Resident Inspector. This notification is being made 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).' All rods were inserted. The plant is stable with normal off-site power line-up. The licensee will notify the Commonwealth of Massachusetts.
ENS 5218115 August 2016 17:48:00PilgrimNRC Region 1GE-3On Monday, August 15, 2016 at 1552 (EDT), with the reactor at (about) 70 percent core thermal power (CTP), Pilgrim Nuclear Power Station (PNPS) entered a 24-hour shutdown Limiting Condition for Operation Action Statement (LCO-AS) for Salt Service Water (SSW) inlet temperature exceeding the Technical Specification (TS) limit in TS 3.5.B.4. The LCO-AS was subsequently exited at 1651 hours when the temperature of SSW trended to below the TS limit. Under certain design conditions, the SSW system is required to provide cooling water to various heat exchangers such as the Reactor Building Closed Cooling Water (RBCCW) and Turbine Building Closed Cooling Water (TBCCW) systems. When the inlet temperature to these supplied loads exceeds the 75 degrees F limit established in the TS, the SSW system is conservatively declared inoperable until the temperature trends below this value. This condition existed for approximately 60 minutes. The SSW temperature is being closely monitored and trended on a continuous basis. This event has no impact on the health and safety of the public. The licensee has notified the NRC Senior Resident Inspector. This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(B) and 10 CFR 50.72(b)(3)(v)(D) due to an event or condition that could have prevented fulfillment of a safety function. The licensee will be notifying the Commonwealth of Massachusetts Emergency Management Agency.
ENS 521565 August 2016 13:58:00MonticelloNRC Region 3GE-3On 8/5/2016 at 1014 (CDT), the Monticello Nuclear Generating Plant (MNGP) was notified by the Minnesota Department of Health (MDH) of a notice of violation for exceeding the drinking water limit for carbon tetrachloride in the drinking water well that supplies the Security Access Facility. Additionally the MDH will be notifying the Minnesota Pollution Control Agency regarding the violation. As a result, this issue is being reported under 10CFR50.72(b)(2)(xi) for notifications to other offsite government agencies. There was no impact to the health and safety of the general public as a result of this issue. The drinking fountains in the Security Access Facility have been isolated. The NRC Resident Inspector has been notified.
ENS 521545 August 2016 06:26:00MonticelloNRC Region 3GE-3At 2240 CDT on August 4, 2016, it was discovered that the floor between the cable spreading room and the plant administration building (PAB) basement is not a credited Appendix R fire barrier. Because the cable spreading room and the plant administration building are located in the same fire area, a fire in the PAB could spread to the cable spreading room requiring evacuation of the control room. The travel path used to access the Alternate Shutdown Panel following control room evacuation traverses the same fire area in the PAB. Therefore, this event is being reported under 10 CFR 50.72(b)(3)(ii) for Degraded or Unanalyzed Condition as a fire in the PAB could have the potential to impact Division 1 equipment as well as impede the Operators ability to access Division 2 safe shutdown equipment. Fire watches have been established. There is no impact to the health and safety of the public. The NRC Resident Inspector has been notified. The licensee will notify the State of Minnesota.
ENS 521534 August 2016 22:04:00MonticelloNRC Region 3GE-3

At 1415 CDT on August 4, 2016, while performing a scheduled fire protection surveillance, it was discovered that a component within fire panel FZCP-7, BATTERY ROOM FIRE DETECTION had failed resulting in the inability of the installed fire detectors to detect a fire within the Division 1 and Division 2, 125 VDC battery rooms as well as the Division 2, 250 VDC battery room. This is being reported under 10 CFR 50.72(b)(3)(xiii) for a Loss Of Emergency Assessment Capability as the Control Room would not receive automatic notification of a fire in these areas for evaluation of HU2.1 and HA2.1 for fire within impacted battery rooms which are located within the Protected Area. There is no impact to the health and safety of the public. A 15 minute fire watch has been established for the affected fire zones. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM MARTIN RAJKOWSKI TO DANIEL MILLS AT 1050 EDT ON 08/05/2016 * * *

Event Notification 52153 completed at 2204 EDT on 8/4/2016 shown above contains an error. The failure of FZCP-7, BATTERY ROOM FIRE DETECTION, resulted in the inability to detect a fire within the Division 1 and Division 2 125 VDC battery rooms as well as the Division 1 250 VDC battery room. The Division 2 250 VDC battery room was not affected by this issue. Additionally, the State of Minnesota was notified of this issue. The NRC Resident Inspector has been notified of this update. Notified R3DO (Skokowski)

ENS 5212726 July 2016 18:12:00Quad CitiesNRC Region 3GE-3On July 26, 2016 at 1252 hours (CDT), the Control Room Emergency Ventilation (CREV) system was declared inoperable due to a toxic gas analyzer spurious alarm which resulted in the 'B' Air Filtration Unit (AFU) being inadvertently isolated. In this condition, Control Room Emergency Ventilation (CREV) system cannot be guaranteed to achieve required design flow rate. Tech Spec 3.7.4, Condition A was entered which requires the CREV system to be restored to an operable status in seven (7) days. The CREV system maintains a habitable control room environment and ensures the operability of components in the control room emergency zone during accident conditions as well as protection of the operators from a high dose environment assumed during a design basis accident. This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(D), 'Event or Condition That Could Have Prevented Fulfillment of a Safety Function,' because the CREV system is a single train system required to mitigate the consequences of an accident. The NRC Resident Inspector has been notified.
ENS 5210720 July 2016 12:02:00DresdenNRC Region 3GE-3

At 1130 CDT on Wednesday, July 20, 2016, the Dresden Nuclear Power Station (DNPS) Technical Support Center (TSC) emergency ventilation system will be removed from service for planned maintenance activities. During the maintenance, the TSC ventilation will be shut down. The TSC air filtration fan and dampers will be non-functional, rendering the TSC HVAC accident mode non-functional. This maintenance is scheduled to minimize out-of-service time. The planned TSC ventilation outage is scheduled to be completed in approximately 14 hours. Contingency plans are in place so that if an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing Emergency Planning (EP) procedures and checklists. If radiological or environmental conditions require TSC facility evacuation during ventilation system restoration, the Station Emergency Director will relocate the TSC staff in accordance with station procedures. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM RYAN CHAMBERLAIN TO DANIEL MILLS ON 07/21/2016 AT 1757 EDT * * *

At 1619 CST on July 21, 2016, Dresden TSC ventilation has been restored and is now functional." The licensee has notified the NRC Resident Inspector Notified R3DO (Stone).

ENS 5209918 July 2016 22:21:00Quad CitiesNRC Region 3GE-3Testing of the Everbridge ERO (Emergency Response Organization) notification system identified that the system cannot notify all ERO individuals. This constitutes a loss of offsite communications capability. The issue has subsequently been reported resolved by the vendor and both site testing and common ERF (Emergency Response Facility) (EOF (Emergency Operations Facility) at Cantera) has verified resolution. The Everbridge system capability loss for Quad Cities was identified at approximately 1450 (CDT) hours on July 18, 2016, due to an undetermined loss of system communications, which is currently being investigated. Emergency Response Data System (ERDS) capability was not lost. The Everbridge system capability loss for the common ERF (EOF at Cantera) was identified at approximately 1500 (CDT) on July 18, 2016. This event is reportable under 10 CFR 50.72(b)(3)(xiii) as a loss of EP offsite communications capability. The NRC Resident inspector has been notified. The site was developing compensatory measures when the event was terminated.
ENS 5209818 July 2016 22:07:00DresdenNRC Region 3GE-3Testing of the Everbridge ERO (Emergency Response Organization) notification system identified that the system cannot notify all ERO individuals. This constitutes a loss of offsite communications capability. Compensatory measures (ERO phone lists) were put in place. The Everbridge system capability loss for Dresden Station was identified at 1500 CDT on July 18, 2016, due to an undetermined loss of system communications. Emergency Response Data System (EROS) capability was not lost. This event is reportable under 10 CFR 50.72(b)3(xiii) as a major loss of communication capability. On July 18, 2016 at 1957 CDT, an Everbridge ERO call in drill was initiated and verified successful at 2030 CDT. The NRC Resident Inspector has been notified. Some of the ERO personnel did not receive a test page. The requirement is to have all ERO personnel receive the page within ten minutes and to be fully staffed within one hour.