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	Discovered date	Reporting criterion	Title	Event description
ENS 56980	19 February 2024 15:45:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Loss of Reactor Building Ventilation	The following information was provided by the licensee via email: At 1045 EST, on 2/19/2024, during a maintenance activity, a loss of all reactor building ventilation occurred on Unit 2. With no flow past the ventilation radiation monitors, the radiation monitors were inoperable to support their ability to perform primary and secondary containment isolation functions or start the standby gas treatment system. Reactor building ventilation was restored within 15 minutes. Due to this inoperability, the radiation monitor system was in a condition that could have prevented fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector will be notified. * * RETRACTION ON 3/15/24 AT 1315 EDT FROM BILL LINNELL TO ADAM KOZIOL * * * Upon further investigation, it was verified that the reactor building and the refueling floor radiation monitors are not needed to control the release of radiation for events described in chapter 14 of the updated Final Safety Analysis Report. For the analyzed loss of coolant accident (LOCA), the primary and secondary signals for this purpose were available and unaffected by this event. The radiation monitors provide a tertiary redundant method that is not credited within the station analysis. For all other analyzed accidents, the signal provided by the radiation monitors is not needed, as the secondary containment isolation function and start of the standby gas treatment system are not credited. Additionally, the fuel handling accident was not credible during the time of the event because no activities were in progress on the refueling floor. Therefore, the threshold for reporting the issue as an event or condition that could have prevented the fulfillment of a safety function was not met. The NRC Resident Inspector has been notified. Notified R1DO (Jackson)
ENS 56584	28 April 2023 09:02:00	10 CFR 50.73(a)(1), Submit an LER	60 Day Notification for an Invalid Actuation of Primary Containment Isolation Logic	The following information was provided by the licensee email: This telephone notification is provided in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to report an invalid actuation of Division 2 Primary Containment Isolation logic at the Monticello Nuclear Generating Plant (MNGP) that occurred while in a refueling outage. At approximately 0402 Central Daylight Time (CDT) on April 28, 2023 and at approximately 1611 and 2143 CDT on May 4, 2023, momentary losses of 'Y80 Division 2 Uninterruptible 120VAC Class 1E Distribution Panel', which provides power to Division 2 Primary Containment Isolation logic, resulted in a partial Primary Containment Group 2 Isolation (gas systems), initiation of the Standby Gas Treatment system, and the shift of Control Room ventilation to the high radiation mode. The momentary losses of 'Y80' were due to an intermittent, age-related degradation issue with the 'Uninterruptible Power Supply Y81, Division 2 120VAC Class 1E Inverter', which resulted in a temporary loss of output plus a lack of static switch transfer from the inverter supply to the alternate source as designed. The actuations were not initiated in response to actual plant conditions, these were not intentional manual initiations, and there were no parameters satisfying the requirements for initiation. Therefore, these events have been determined to be invalid actuations that were attributed to the same cause. All systems responded as designed to the actuation signal. Operations reset the partial Primary Containment Group 2 Isolation signal, shutdown the Standby Gas Treatment system, and restored Control Room ventilation per the procedure. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 55976	4 July 2022 06:30:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Both Trains of Standby Gas Treatment Inoperable	The following information was received from the licensee via email: At 0130 CDT on July 4 2022, it was discovered both trains of Standby Gas Treatment System were simultaneously inoperable due to failure to reach required flow rates. Both trains were capable of starting but failed to reach the required flow of 4000 SCFM. Secondary Containment differential pressure was not able to be maintained at greater than or equal to 0.25 inches of vacuum water gauge, causing Secondary Containment to also be inoperable. Due to this inoperability, the system was in a condition that could have prevented the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 55708	29 November 2021 17:28:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of Secondary Containment Relays	The following information was provided by the licensee via email: This telephone notification is provided in accordance with 10 CFR 50.73(a)(1) to report an invalid actuation of secondary containment relays in accordance with 10 CFR 50.73(a)(2)(iv)(A). On November 29, 2021, the `B' Fuel Pool radiation monitor spiked high during restoration following the performance of the 0068 procedure `Spent Fuel Pool & Reactor Building Exhaust Plenum Monitor Calibration' due to cable to radiation monitor connector degradation from handling. This resulted in a Partial Primary Containment Group II isolation (gas systems), initiation of Standby Gas Treatment system, and isolation of the Reactor Building Ventilation system. All systems responded as designed to the actuation signal. Operations reset the Partial Primary Containment Group II isolation signal, shutdown Standby Gas Treatment System, and restored Reactor Building Ventilation system per procedures. At the time of the occurrence, the `A' Fuel Pool radiation monitor was reading normal at approximately 1.5 mr/hr. The `B' Fuel Pool radiation monitor spiked above the 50 mr/hr setpoint and continued to read erratically. Work was performed to clean and reconnect the connector and testing per 0068 procedure verified the condition was corrected. The `B' Fuel Pool radiation monitor returned to service. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 55345	13 May 2021 11:00:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of Containment Isolation Signal	This 60-Day telephone notification is being made per the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of containment isolation signal affecting more than one system. On May 13, 2021, during the restoration of the Unit 2 Refuel Floor High Radiation Isolation Logic an invalid isolation signal was received. The condition requiring an isolation signal was verified not to be present prior to restoring the logic; however, it was not recognized that a previous isolation signal was latched in and had not been reset. When the isolation logic was restored, the Primary Containment Isolation System (PCIS) isolated on the invalid signal. The systems successfully completed the isolation per the plant design and plant configuration. The following systems actuated due to the Unit 2 PCIS Group 6C Isolation: - Isolation of Containment Hydrogen and Oxygen Sampling Valves, - Start of the 2A Reactor Enclosure Recirculation System, - Trip of the Units 1 and 2 Refuel Floor HVAC, - Start of the A and B Trains of Standby Gas Treatment Systems. The NRC Resident Inspector was notified.
ENS 54452	18 December 2019 14:08:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Pressure Degraded	On December 18, 2019, at 0908 EST, with the East and Center Reactor Building HVAC (RBHVAC) trains in service, secondary containment pressure degraded to the point where the Technical Specification (TS) requirement for secondary containment pressure was not met and secondary containment was declared inoperable. Secondary containment pressure did not meet the TS required limit for approximately four minutes. The maximum secondary containment pressure observed during that time was approximately 0.064 inches of vacuum water gauge. Secondary containment pressure was returned to within the TS operability limit of greater than or equal to 0.125 inches of vacuum water gauge (TS SR 3.6.4.1.1) by starting Division 1 of the Standby Gas Treatment System (SGTS). Secondary containment was declared Operable at 0912 EST. A modulating damper associated with the Center train of RBHVAC was identified as not properly controlling; an investigation is in progress. RBHVAC was manually secured to support problem identification and resolution. Secondary containment pressure is currently stable with Division 1 SGTS in service. There were no radiological releases associated with this event. Declaring secondary containment inoperable is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The NRC Resident Inspector has been notified.
ENS 54239	25 August 2019 16:02:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Scram Due to Main Generator Ground Fault Relay	On August 25, 2019, at 1102 (CDT), Quad Cities Unit 1 experienced an automatic scram from 100 percent power. All rods fully inserted and there were no complications. The trip was initiated from a main generator ground fault relay. Troubleshooting of the fault is in progress. All systems responded as designed. There were no systems inoperable and no TS (Technical Specification) action statements were in progress prior to the Reactor Scram. Reactor water level dropped below the Group 2 and Group 3 Reactor Water Level Isolation set-points as expected, and recovered via the Feedwater system. Standby Gas Treatment System auto started and Reactor Building Ventilation Isolation occurred as expected. Unit 1 remains in Mode 3. Decay heat is being removed using the steam bypass valves to the condenser and the safety relief valves did not lift as a result of the trip. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. Unit 2 was not affected.
ENS 54197	3 August 2019 07:26:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	En Revision Imported Date 8/7/2019	EN Revision Text: AUTOMATIC REACTOR SCRAM ON LOW REACTOR WATER LEVEL At 0226 (CDT), an automatic scram on low reactor water level occurred due to a trip of the 'B' Reactor Feed pump. All control rods fully inserted. Reactor water level 2 was reached and the High Pressure Core Spray system, Reactor Core Isolation Cooling system, Division 3 diesel generator, Standby Gas Treatment Systems 'A' and 'B' and all shutdown safety related service water pumps started as expected. Reactor Core Isolation Cooling and High Pressure Core Spray injected as expected. All level 2 containment isolation signals occurred as expected and all level 2 containment valves closed as expected. Reactor water level is currently being controlled in band by condensate. Reactor pressure is being maintained by main turbine Bypass Valves. This event is being reported under 10 CFR 50.72(b)(2)(iv)(A), for ECCS discharge to RCS; 10 CFR 50.72(b)(2)(iv)(B), for RPS actuation, and 10 CFR 50.72(b)(3)(iv)(A), for specified system actuation. The NRC Senior Resident Inspector has been notified. No safety relief valves lifted during the transient. The plant is in a normal shutdown electrical lineup with all safety equipment available. The licensee notified the Illinois Emergency Management Agency per their communications protocol. * * UPDATE FROM DAVID LIVINGSTON TO HOWIE CROUCH AT 0321 EDT ON 8/4/19 * * * Following automatic initiation of the High Pressure Core Spray (HPCS) System as described above, the HPCS System was manually secured following station procedures after verification that additional RPV (reactor pressure vessel) injection was no longer required. Securing HPCS injection in this manner prevents automatic restart of the system in the event of a subsequent low RPV level condition, rendering it inoperable. As the HPCS system is considered a single train safety system, this meets the reportability requirements of 10 CFR 50.72(b)(3)(v)(D). This reportable condition was identified following review of post-scram actions. The HPCS system has been restored to a Standby lineup. The licensee will be notifying the NRC Resident Inspector. Notified R3DO (Pelke). * * UPDATE FROM JAMES FORMAN TO KERBY SCALES AT 1545 EDT ON 8/6/19 * * * Following the scram, the Primary Containment to Secondary Containment and the Drywell to Primary Containment differential pressure limits were exceeded. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.1.4, Primary Containment Pressure, and 3.6.5.4, Drywell Pressure, Actions A.1, B.1, and B.2 were entered. Primary Containment to Secondary Containment differential pressure and Drywell to Primary Containment differential pressure were restored to within the LCO limits at 1505 on 8/3/19 and the associated TS Actions were exited. This event is reportable under 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that could have prevented the fulfillment of the primary containment function due to being outside the initial conditions to ensure that drywell and containment pressures remain within design values during a loss of coolant accident. This event is also reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of the drywell and primary containment functions to control the release of radioactive material for the same reason. The licensee notified the NRC Resident Inspector. Notified R3DO (Pelke).
ENS 53851	30 January 2019 15:10:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Inoperable	At 0910 (CST) on January 30, 2019, the Dresden Station Heater Boiler 'B' tripped while placing the station Heater Boiler 'A' in service. With colder temperatures, the density of the supply air increased and contributed to a greater quantity of air entering the Reactor Building than what was previously being supplied with heating steam in service. The Reactor Building differential pressure (DP) degraded and dropped below 0.25 inches water column vacuum. This condition represents a failure to meet Technical Specification (TS) Surveillance Requirement 3.6.4.1.1. Entry into TS 3.6.4.1 Condition A was made due to Secondary Containment becoming inoperable. Standby Gas Treatment System was initiated to assist with Reactor Building DP control. Reactor Building DP was restored to greater than 0.25 inches water column vacuum. TS 3.6.4.1 Condition A was exited. This event is being reported under 10 CFR 50.72(b)(3)(v)(C), 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to ... control the release of radioactive material.' The NRC Resident Inspector has been notified.
ENS 53828	16 January 2019 05:00:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Differential Pressure Exceeded Technical Specification Allowed Value	On January 16, 2019, with James A. Fitzpatrick Nuclear Power Plant operating at 100 percent power, the Emergency and Plant Information Computer (EPIC) indicated that Secondary Containment differential pressure exceeded the Technical Specification Surveillance Requirement of greater than or equal to 0.25 inches of vacuum water gauge while isolating Reactor Building Ventilation. The Secondary Containment differential pressure was less than 0.25 inches of vacuum water gauge for approximately ten (10) seconds, and then immediately returned to greater than or equal to 0.25 inches of vacuum water gauge. This condition did not impact the leak tightness of Secondary Containment or the ability of the Standby Gas Treatment system to establish and maintain the required differential pressure. When Secondary Containment did not meet the Technical Specification Surveillance Requirement 3.6.4.1.1 for differential pressure, the Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was inoperable. This event is being reported under 10 CFR 50.72(b)(3)(v)(C). The licensee has notified the NRC Resident Inspector.
ENS 53778	5 December 2018 05:00:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Three Minute Loss of Secondary Containment Vacuum	At 1010 (EST) on December 5, 2018, Secondary Containment differential pressure exceeded the Technical Specification Surveillance Requirement of greater than or equal to 0.25 inches of vacuum water gauge. This condition existed for approximately 3 minutes before the differential pressure was restored to normal when the Standby Gas Treatment system was manually initiated. This event was caused by a trip of the service air compressor 39AC-2A. The loss of instrument air pressure caused Reactor Building ventilation to isolate and raise Secondary Containment differential pressure. The instrument air pressure was restored when 39AC-2A was isolated and the two backup air compressors started. This condition did not impact the leak tightness of Secondary Containment or the ability of the Standby Gas Treatment system to establish and maintain the required differential pressure. When Secondary Containment did not meet the Technical Specification Surveillance Requirement 3.6.4.1.1 for differential pressure, the Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was inoperable. This event is being reported under 10 CFR 50.72(b)(3)(v)(C). The licensee notified the NRC Resident Inspector.
ENS 53776	13 October 2018 05:00:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification of Invalid Specified System Actuation	This 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 53304	31 March 2018 07:06:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Inoperable Due to Failed Surveillance Test	At 0206 (CDT) on March 31, 2018, with the plant in Mode 1 at 100% rated core thermal power, Grand Gulf Nuclear Station experienced a loss of Secondary Containment. During the performance of a Standby Gas Treatment System (SGTS) drawn down test with Auxiliary Building train bay door (1A319A) as the secondary containment boundary, Grand Gulf was unable to maintain secondary containment pressure, as required by SR (surveillance requirement) 3.6.4.1.4, greater than or equal to 0.266 inches of water vacuum for 1 hour. Following initial vacuum draw down, secondary containment pressure degraded to 0.225 inches of water vacuum with operators in the field reporting air leakage from door 1A319A. The test was secured and Secondary Containment was declared inoperable and Technical Specification 3.6.1.4 A.1 was entered. Following completion of the failed surveillance test, Secondary Containment was returned to an operable status at 0315 hours on March 31, 2018, by returning the system to a previously known operable configuration by closing doors 1A310, 1A312 and 1A319. This is being report under 10 CFR 50.72(b)(3)(v)(C). The licensee has notified the NRC Resident Inspector.
ENS 53189	31 January 2018 19:10:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Momentary Loss of Secondary Containment	At 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 53165	11 January 2018 15:41:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Momentary Low Pressure	On January 11, 2018, at 1041 EST, a planned train swap of the Reactor Building Heating Ventilation and Air Conditioning (RBHVAC) system resulted in the Technical Specification (TS) for secondary containment pressure boundary not being met for less than one minute. The maximum secondary containment pressure observed during that time was approximately 0.117 inches of vacuum water gauge. Secondary containment pressure was returned to within the TS operability limit of greater than or equal to 0.125 inches of vacuum water gauge per TS Surveillance Requirement (SR) 3.6.4.1.1 by starting Division 1 of the Standby Gas Treatment System (SGTS) in addition to the RBHVAC system already in operation. Secondary containment pressure is currently stable. Secondary containment was declared Operable at 1045 EST. There were no radiological releases associated with this event. Declaring secondary containment inoperable as a result of not meeting TS SR 3.6.4.1.1 is reportable under 10CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The licensee has notified the NRC Resident Inspector.
ENS 53110	9 December 2017 19:48:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Manual Reactor Scram Due to Loss of Division 1 Ac Power to Numerous Components	At approximately 1347 (CST) on 12/09/17, the Main Control Room received annunciators that indicated a trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1 breaker. Numerous Division 1 components lost power (powered from unit subs 1A and A1). The Division 1 containment Instrument Air isolation valves had failed closed by design due to the loss of power. Due to the loss of containment instrument air, several control rods began to drift into the core as expected and, by procedure, the reactor mode switch was placed in the shutdown position at 1353 (CST). All control rods fully inserted. Also due to the loss of power, the Fuel Building ventilation dampers failed closed by design. With the normal ventilation system secured, secondary containment differential pressure rose to slightly greater than 0 inches water gauge which exceeded the Technical Specification requirement of greater than 0.25 inches vacuum water gauge at 1348 (CST). The Control Room entered EOP-8, Secondary Containment Control. Secondary Containment differential pressure was restored within Technical Specification requirements at 1351 (CST) by starting the Division 2 Standby Gas Treatment system. This event is being reported as a manual actuation of the Reactor Protection System (RPS) and as a Condition that Could Have Prevented Fulfillment of a Safety Function. The cause is currently under investigation. The NRC Resident has been notified. The licensee informed the NRC Resident Inspector. * * UPDATE FROM DALE SHELTON TO VINCE KLCO AT 1658 EST ON 12/10/2017 * * * During a review of plant logs it was identified that the primary to secondary containment differential pressure was identified to be outside of Technical Specification 3.6.1.4 limits of 0 plus or minus 0.25 psid at 2009 on 12/9/17 due to the primary containment ventilation system dampers closing as a result of the loss of power. This parameter is an initial safety analysis assumption to ensure that primary containment pressures remain within the design values during a Loss of Coolant Accident (LOCA). As a result, this condition is reportable as an unanalyzed condition that significantly degrades plant safety. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone). * * UPDATE FROM MICHAEL ANTONELLI TO VINCE KLCO ON 12/11/17 AT 1805 EST * * * During the post transient review of the trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1, it was identified that the unplanned INOPERABILITY of the Low Pressure Core Spray (LPCS) system due to the loss of power to the injection valve constitutes an event or condition that could have prevented fulfillment of a safety function and is reportable under 10CFR50.72(b)(3)(v)(D) for Accident Mitigation. The High Pressure Core Spray (HPCS) remained available to perform the core spray function, if necessary, during a design basis Loss of Coolant Accident (LOCA), however HPCS and LPCS are each considered single train safety systems. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone).
ENS 52996	30 September 2017 05:34:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Declared Inoperable	At 0134 (EDT) on September 30, 2017, Nine Mile Point Unit 2 entered Tech Spec 3.6.4.1 when secondary containment was declared inoperable due to secondary containment differential pressure being above the Tech Spec Surveillance Requirement of -0.25 inches vacuum water gauge. The Division II Standby Gas Treatment System was started to restore differential pressure at 0135 (EDT) on September 30, 2017 the differential pressure was restored, the secondary containment was declared operable and the Tech Spec3.6.4.1 exited. Secondary containment being inoperable is a 8-hour report for 10 CFR 50.72(b)(3)(v)(C), Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. The cause of this condition is being investigated. The NRC Resident Inspector has been notified. The licensee will also inform the State of New York.
ENS 52795	8 June 2017 19:27:00	10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	Automatic Reactor Scram After Main Turbine Control Logic Loss of Power	At 1527 hrs (EDT) on June 8, 2017, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed due to a loss of Main Turbine Electro-Hydraulic Control (EHC) logic power causing a High Flux Reactor Power RPS (Reactor Protection System) trip. All control rods (fully) inserted and both reactor recirculation pumps tripped due to reaching reactor water level 2. Reactor water level lowered to -49 inches causing Level 3 (+13 inches) and Level 2 (-38 inches) isolations. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) automatically initiated and were overridden by control room operators after RPV (Reactor Pressure Vessel) water level was restored to the normal band with feedwater. HPCI and RCIC injected to the Reactor Coolant System during reactor level stabilization. All isolations and initiations occurred as expected. No main steam relief valves opened. Pressure was controlled via main turbine bypass valve operation. All safety systems operated as expected. Secondary Containment Zone 1, 2, and 3 differential pressure lowered to 0 inch WG (Water Gauge) due to a trip of the Reactor Building Ventilation system that resulted from Unit 1 Level 2 isolation. Differential pressure was restored to Zones 1, 2, and 3 by the initiation of Standby Gas Treatment System on the Unit 1 Level 2 initiation. Unit 1 reactor is currently stable in Mode 3. Investigation into the loss of Main Turbine EHC logic power is underway. The NRC Resident Inspector has been notified. A voluntary notification to PEMA and press release will occur. The suspected cause of the loss of power to the EHC logic circuit is ongoing maintenance on the system.
ENS 52651	30 March 2017 21:10:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Technical Specification Not Met	On March 30, 2017 at 1710 EDT, with Reactor Building HVAC in service maintaining normal building pressure, Reactor Building pressure began to rise for an unknown reason. The Technical Specification (TS) for secondary containment pressure boundary was not met for approximately 50 seconds. Division 1 Standby Gas Treatment System was started and returned Secondary Containment pressure to the TS operability limit of 0.125 inches of vacuum water gauge (TS SR 3.6.4.1.1). The highest pressure observed on the Main Control Room indications was 0.105 inches of vacuum water gauge. During the event, Operations with the Potential to Drain the Reactor Vessel (OPDRV) were in progress. Actions to immediately suspend OPDRVs were taken. Investigation of the cause of the event is in progress. There were no radiological releases associated with this event. Declaring secondary containment inoperable is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The licensee has notified the NRC Resident Inspector. * * RETRACTION AT 1652 EST ON 5/15/17 FROM JEFF YEAGER TO JEFF HERRERA * * * The purpose of this notification is to retract a previous report made on March 30, 2017 (EN 52651). The notification to the NRC involved an event where secondary containment momentarily exceeded the Technical Specification (TS) requirements during refueling activities which had been designated as operations with the potential to drain the reactor vessel (OPDRVs). The notification was made under 10 CFR 50.72(b)(3)(v)(C) as an 'event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to ... control the release of radioactive material.' Subsequent to the initial notification, the event and the NRC guidance in NUREG-1022 pertaining to 10 CFR 50.72(b)(3)(v) were reviewed further. At the time of the event, Fermi 2 was shutdown (Mode 5. Refueling). In Mode 5, the pressures and temperatures that could cause a loss of coolant accident (LOCA) are not present. No movement of fuel was in progress such that the fuel handling accident (FHA) was also not applicable. Thus secondary containment was only required per TS 3.6.4.1 due to the ongoing OPDRVs. The Fermi 2 UFSAR does not describe OPDRVs as an accident that secondary containment is required to mitigate. Based on this information, secondary containment was not required to mitigate the consequences of an accident as described in the UFSAR during the event on March 30, 2017. Under these circumstances, the momentary exceedance of TS requirements for secondary containment is not considered a loss of safety function under 10 CFR 50.72(b)(3)(v) per the guidance in NUREG-1022. Therefore, EN 52651 is retracted and no Licensee Event Report (LER) under 10 CFR 50.73(a)(2)(v) is required to be submitted. The NRC Resident Inspector has been notified. Notified the R3DO (Cameron).
ENS 52437	15 December 2016 15:10:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Momentary Low Pressure	On December 15, 2016, at 1010 EST, the startup of the Reactor Building HVAC (Heating Ventilation and Air Conditioning) system resulted in the Technical Specification (TS) for secondary containment pressure boundary not being met for approximately 1 second. The maximum secondary containment pressure observed during that time was approximately 0.044 inches of vacuum water gauge. Secondary containment pressure was returned to within the TS operability limit of 0.125 inches of vacuum water gauge (TS SR 3.6.4. 1.1) by Reactor Building HVAC and Standby Gas Treatment System already in operation. There were no radiological releases associated with this event. Declaring secondary containment inoperable is reportable under 10CFR50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The licensee has notified the NRC Resident Inspector. * * RETRACTION AT 1922 EDT ON 3/17/2017 FROM DEREK ETUE TO BETHANY CECERE * * * In this event notification, DTE Electric Company (DTE) reported conditions whereby the Fermi 2 secondary containment was believed to have exceeded Technical Specification Surveillance Requirements due to high winds. DTE hereby retracts this event notification as the Fermi 2 secondary containment has been determined to have been operable during this event as described below. The Fermi 2 secondary containment pressure is maintained at a pressure less than the external pressure to contain, dilute, hold up, and reduce the activity level of fission products prior to release to the environment, and to isolate and contain fission products that are released during a Design Basis Accident or certain operations. Secondary containment pressure is monitored by a number of differential pressure (dP) sensors. High wind gusts have resulted in momentary negative pressure on the leeward side of the building, causing a more positive pressure indication from one or more dP sensors. The secondary containment building pressure remains relatively constant during these 'wind events.' In December 2016, DTE implemented a software design change to display a 120-second rolling average for secondary containment dP indication. A 120-second rolling average recorded every second provides the operator a more accurate report of actual secondary containment conditions, while mitigating the signal noise and wind gust effects. The conditions associated with the subject event notification were re-reviewed in light of the improved secondary containment dP indication and it was determined that the Fermi 2 secondary containment was operable during this event. Specifically, the secondary containment pressure did not exceed Technical Specification Surveillance Requirements during this event. In summary, the above event notification is retracted because the Fermi 2 secondary containment was determined to have been fully operable during the conditions identified in the subject report. The licensee notified the NRC Resident Inspector. Notified R3DO (Stoedter).
ENS 52432	14 December 2016 05:00:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Pressure Boundary Out of Specification	On December 14, 2016, at 1314 EST, the startup of the Reactor Building HVAC (Heating, Ventilation and Air Conditioning) system resulted in the Technical Specification (TS) for secondary containment pressure boundary not being met for approximately 1 second. The maximum secondary containment pressure observed during that time was approximately 0.07 inches of vacuum water gauge. Secondary containment pressure was returned to within the TS operability limit of 0.125 inches of vacuum water gauge (TS SR 3.6.4.1.1) by Reactor Building HVAC and Standby Gas Treatment System already in operation. There were no radiological releases associated with this event. Declaring secondary containment inoperable is reportable under 10CFR50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The licensee has notified the NRC Resident Inspector. * * RETRACTION AT 1922 EDT ON 3/17/2017 FROM DEREK ETUE TO BETHANY CECERE * * * In this event notification, DTE Electric Company (DTE) reported conditions whereby the Fermi 2 secondary containment was believed to have exceeded Technical Specification Surveillance Requirements due to high winds. DTE hereby retracts this event notification as the Fermi 2 secondary containment has been determined to have been operable during this event as described below. The Fermi 2 secondary containment pressure is maintained at a pressure less than the external pressure to contain, dilute, hold up, and reduce the activity level of fission products prior to release to the environment, and to isolate and contain fission products that are released during a Design Basis Accident or certain operations. Secondary containment pressure is monitored by a number of differential pressure (dP) sensors. High wind gusts have resulted in momentary negative pressure on the leeward side of the building, causing a more positive pressure indication from one or more dP sensors. The secondary containment building pressure remains relatively constant during these 'wind events.' In December 2016, DTE implemented a software design change to display a 120-second rolling average for secondary containment dP indication. A 120-second rolling average recorded every second provides the operator a more accurate report of actual secondary containment conditions, while mitigating the signal noise and wind gust effects. The conditions associated with the subject event notification were re-reviewed in light of the improved secondary containment dP indication and it was determined that the Fermi 2 secondary containment was operable during this event. Specifically, the secondary containment pressure did not exceed Technical Specification Surveillance Requirements during this event. In summary, the above event notification is retracted because the Fermi 2 secondary containment was determined to have been fully operable during the conditions identified in the subject report. The licensee notified the NRC Resident Inspector. Notified R3DO (Stoedter)
ENS 52146	2 August 2016 14:15:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Technical Specification Not Met	On August 2, 2016 at 1015 EDT, while restoring the east train of Reactor Building HVAC (RBHVAC) after a surveillance test on Division 2 Standby Gas Treatment System (SGTS), the Technical Specification (TS) for the secondary containment pressure boundary was not met for a duration time of approximately 1 second. The maximum secondary containment pressure observed during that time was approximately 0.120 inches of vacuum water gauge. Secondary containment pressure was returned to within the TS operability limit by RBHVAC and SGTS already in operation. There were no radiological releases associated with this event. The cause of the event is under investigation. The TS requirement is to maintain secondary containment vacuum greater than or equal to 0.125 inches of vacuum water gauge (TS SR 3.6.4.1.1) for secondary containment operability. Declaring secondary containment inoperable is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The licensee has notified the NRC Resident lnspector.
ENS 51928	13 May 2016 17:00:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Existing Design Inadequacy Could Prevent Standby Gas Treatment System Operablity	At 1200 (CDT) May 13, 2016, while the plant was operating at 100% power, it was brought to the attention of the River Bend Station Main Control Room staff that an existing design inadequacy could prevent both trains of the Standby Gas Treatment System (GTS) from performing its design function. Under certain specific conditions, the installed Masterpact breakers may not close to allow energization of the filter train exhaust fans. A start signal (reactor level 2, drywell pressure 1.68 psid, annulus high radiation, annulus low flow) combined with a trip signal within a certain time differential, could result in a failure of the breakers to close. As a result of this condition, both Standby Gas Trains were declared inoperable, which required entry into LCO 3.6.4.3 Condition C (requires entering Mode 3 in 12 hours). Declaring both trains of Standby Gas Treatment System inoperable resulted in loss of the safety function since a system that has been declared inoperable is one in which the capability has degraded to the point where it cannot perform with reasonable expectation or reliability. The Standby Gas Treatment System (GTS) limits release to the environment of radioisotopes, which may leak from the primary containment, ECCS systems, and other potential radioactive sources to the secondary containment under accident conditions. At 1240 (CDT) May 13, 2016, one division of GTS, GTS 'A', was manually started from the Main Control Room. This action prevents the breaker failure mode, restored the operability of one train and restored the safety function of the GTS system. LCO 3.6.4.3 Condition A (restore Operability in 7 days) is currently entered for Standby Gas Train 'B'. During the 40 minutes of inoperability, both trains of Standby Gas remained available. At no time was the health or safety of the public impacted. This condition is being reported in accordance with 10CFR50.72(b)(3)(v)(C) as an event that could have caused a loss of safety function to control the release of radioactive material. The Senior NRC Resident was notified. * * UPDATED AT 1341 EDT ON 05/17/16 FROM DAN PIPKIN TO RICHARD SMITH * * * Further review has determined that the design inadequacy discussed in EN #51928 could adversely effect the ability of the main control building heating, ventilation, and air conditioning (HVAC) system to perform its design safety function, based upon a particular sequence of events occurring within a short window of time (approximately 75 milliseconds). River Bend has implemented compensatory actions to ensure operability of the main control building HVAC system. The Resident Inspector has been notified by the licensee. Notified the R4DO (Miller).
ENS 51732	13 February 2016 08:06:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Inoperable	On 02/13/2016 at 0206 CST, an unexpected trip of a Fuel Building ventilation exhaust fan occurred and secondary containment differential pressure became positive. Secondary containment was declared INOPERABLE when Technical Specification-required differential pressure was not being maintained and entered LCO 3.6.4.1 Action A.1 At 0256 (CST), the standby gas treatment system was started and secondary containment differential pressure was restored to Technical Specification requirements at 0257 CST. This loss of secondary containment is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material. The cause of the fuel building exhaust fan trip is unknown at this time. The NRC Resident Inspector has been notified.
ENS 51701	29 January 2016 21:18:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Specified System Actuation After Loss of One Offsite Power Source	On January 29, 2016, at 1518 CST, with the plant in cold shutdown, power was lost on reserve station service (RSS) line no. 1. This is one of two sources of offsite power required by Technical Specifications. The power loss de-energized the Division 1 onsite AC safety-related switchgear, causing an automatic start of the Division 1 emergency diesel generator (EDG). The Division 1 reactor protection system (RPS) bus was also de-energized, causing a half-scram signal. Approximately 8 minutes later, a full actuation of the RPS occurred due to a high water level condition in the control rod drive hydraulic system scram discharge volume header. All reactor control rods were already fully inserted. The loss of Division 1 RPS also caused the actuation of the Division 1 primary containment isolation logic. The Division 1 isolation valves in the balance-of-plant systems closed as designed. Both trains of the standby gas treatment system actuated. The loss of RSS no. 1 occurred during post-modification testing on relays at the local 230kV switchyard. The exact cause of the event is under investigation. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). The unit remains in cold shutdown with 1 source of offsite power and all 3 (EDG) available. The (NRC) Resident Inspector has been notified.
ENS 51659	16 January 2016 02:38:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Momentary Loss of Secondary Containment	On January 15, 2016 at 2038 CST, an alarm was received indicating Secondary Containment Differential Pressure rose unexpectedly above the Technical Specification Surveillance Requirement, SR 3.6.4.1.1, limit of 0.10 inch of vacuum water gauge. This loss of differential pressure occurred when Operations had entered the 2A Reactor Water Cleanup Pump room. The pump room door was closed and Secondary Containment Differential Pressure returned to Technical Specification limits in approximately 4 minutes. The Standby Gas Treatment System remained in standby and fully operable. 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 momentarily due to secondary containment being inoperable. Given the temporary loss of secondary containment, 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 licensee also notified the State of Illinois Emergency Management Agency.
ENS 51650	12 January 2016 19:40:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Momentary Loss of Secondary Containment	On January 12, 2016, at 1340 CST, an alarm was received indicating secondary containment differential pressure rose unexpectedly above the Technical Specification Surveillance Requirement, SR 3.6.4.1.1, limit of 0.10 inch of vacuum water gauge. A reactor building supply fan was immediately secured to restore differential pressure below the Technical Specification limit. The secondary containment differential pressure returned to below the Technical Specification limit within one minute. The Standby Gas Treatment System remained in standby, and fully operable. At 1341 CST, the secured reactor building supply fan was restarted to restore the normal ventilation lineup. The secondary containment differential pressure remained below the Technical Specification limit. Troubleshooting is in progress. 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 momentarily due to secondary containment being inoperable. Given the temporary loss of secondary containment, 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.
ENS 51526	10 November 2015 04:40:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Reactor Building Vacuum Less than Technical Specifications Requirement	At 2040 PST on 11/9/2015, Reactor Building (Secondary Containment) pressure increased to above the Technical Specification Surveillance requirement of 0.25 inches vacuum water gauge for approximately seven minutes. Operators took action to manually start Standby Gas Treatment System to restore Reactor Building pressure. This event is reportable as an event that could have prevented fulfillment of a safety function needed to control the release of radiation and mitigate the consequences of an accident. The cause of the event is under investigation. There were no radiological releases associated with the event. No safety system actuations or isolations occurred. The licensee notified the NRC Resident Inspector.
ENS 51512	22 September 2015 20:09:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Differential Pressure Exceeded Technical Specification Allowed Value	On September 22, 2015, with James A. Fitzpatrick Nuclear Power Plant operating at 100 percent power, the Emergency and Plant Information Computer (EPIC) indicated a spike in Secondary Containment differential pressure during performance of a surveillance test associated with automatic initiation of the Standby Gas Treatment System. Plant data systems recorded Secondary Containment differential pressure exceeding the Technical Specification allowed value. The Secondary Containment differential pressure was at or above zero inches of water for approximately ten (10) seconds, and then immediately trended negative following auto-start of one of the trains of Standby Gas Treatment. An operator was subsequently dispatched to the ventilation control panel, and verified that Secondary Containment differential pressure was more negative than the Technical Specification allowed value. This condition was entered into the Corrective Action Program, and subsequently, it was determined that the approximate ten second duration that Secondary Containment differential pressure was greater than the Technical Specification allowed value was reportable pursuant to 10 CFR 50.72(b)(3)(v)(C), as an event or condition that could have prevented fulfillment of a safety function. Secondary Containment was Operable following reestablishment of greater than or equal to 0.25 inches of water vacuum, and remains Operable. The licensee has notified the NRC Resident Inspector.
ENS 51480	23 August 2015 16:42:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification of Invalid Primary Containment Isolation Signal	On 8/23/2015 at 1242 (EDT), with the reactor at 100% power, an invalid RPS MG (Reactor Protection System Motor-Generator) set 'A' trip resulting in a loss of RPS bus 'A'; this occurred during testing of the RPS instrument channels. All equipment operated as designed as a result of the loss of power to the 'A' RPS bus. The invalid trip was determined to be a result of the overvoltage relay being set too low. The above event meets the reporting criteria of 10CFR50.73(a)(2)(iv)(A) since the loss of RPS bus resulted in primary containment isolation signals affecting containment valves in more than one system. The following systems isolated as a result of the loss of 'A' RPS bus: Reactor Water Cleanup, Reactor Building ventilation, 'A' Containment Atmosphere Dilution, Torus Vent and Purge, Drywell Equipment and Floor Drain Sumps, 'A' Drywell Containment Atmospheric Monitors, Recirculation System Sample Line, Main Steam Line Drains and Residual Heat Removal drain valve to radwaste. 'A' Standby Gas Treatment System started as designed. This notification is being made in accordance with 10CFR50.73(a)(2)(iv)(A) to provide information pertaining to an invalid 'A' Reactor Protection System actuation. Completed actions were the replacement of overvoltage relay and voltage setpoint change, completed on 9/11/2015. In accordance with 10CFR50.73(a)(i) a telephone notification is being made instead of submitting a written Licensee Event Report.
ENS 51432	29 September 2015 14:30:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Both Trains of the Standby Gas Treatment System Declared Inoperable	On 9/29/15 at 1020 EDT, the 'B' train of Standby Gas Treatment System was declared inoperable for planned testing. On 9/29/15 at 1030 EDT, during performance of a surveillance on Unit 1 Reactor Pressure Vessel water level instrumentation, one channel was found to not meet acceptance criteria. The failed level channel is part of the initiation logic for the 'A' train of Standby Gas Treatment. This resulted in a loss of safety function for the Standby Gas Treatment System. On 9/29/15 at 1145 EDT, the 'B' train of Standby Gas Treatment was restored to operable by restoring from the planned testing. This event is being reported under 10 CFR 50.72(b)(3)(v)(c) and per the guidance of NUREG 1022 Rev 3 section 3.2.7 as a loss of a Safety Function. The NRC Resident Inspector has been informed.
ENS 51391	14 September 2015 03:05:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Manual Scram Due to Loss of Turbine Building Closed Cooling Water	At 2305 EDT on September 13, 2015, a manual scram was initiated in response to a loss of all Turbine Building Closed Cooling Water (TBCCW). All control rods fully inserted. The lowest Reactor Water Level (RWL) reached was 137 inches. All isolations and actuations for RWL 3 occurred as expected. Decay heat was initially being removed through the Main Turbine Bypass System to the Main Condenser, however, as a result of the loss of TBCCW, the Main Feed Pumps lost cooling and had to be secured. At 2310, Standby Feedwater was initiated and Main Feedwater was secured. The loss of TBCCW also caused all Station Air Compressors (SACs) to trip on loss of cooling. The loss of SACs caused the Instrument Air header pressure to degrade to the point at which the Secondary Containment isolation dampers drifted closed. This resulted in the Reactor Building vacuum exceeding the Technical Specification limit. At 2325, operators started the Standby Gas Treatment system and manually initiated a Secondary Containment isolation signal. Secondary Containment vacuum was promptly restored to within Technical Specification limits. Additionally, Operators were monitoring for expected MSIV drift due to the degraded Instrument Air header pressure. When outboard MSIVs were observed to be drifting, Operators closed the outboard and inboard MSIVs at 2345. At 2352, Safety Relief Valves (SRVs) reached the Low-Low Setpoint and began cycling to control reactor pressure. RWL is currently being maintained in the normal level band with the Standby Feedwater and Control Rod Drive systems. Reactor Pressure is being controlled with Safety Relief Valves. Operators are currently in the Emergency Operating Procedure for Reactor Pressure Vessel control. Investigation into the loss of TBCCW continues. No safety-related equipment was out of service at the time of the event. All offsite power sources were adequate and available throughout the duration of the event. The NRC resident inspector has been notified. * * UPDATE AT 0555 EDT AT 09/14/15 FROM CHRIS ROBINSON TO JEFF HERRERA * * * At 0409 EDT the Reactor Core Isolation Cooling (RCIC) system was placed in service due to identification of an unisolable leak in the Standby Feedwater System. Reactor water level and pressure is now being controlled though the RCIC system and Safety Relief Valves. This event update is reportable as a valid manual initiation of a specified safety system under 10CFR50.72(b)(3)(iv)(A). The NRC resident inspector has been notified. The leak rate was reported as approximately 5-10 gallons per minute from a weld on the standby feedwater pump header drain valve F326. The licensee reported the leak stopped once RCIC was placed into service. The licensee is still investigating the issue. Notified the R3DO (Pelke), IRD Manager (Grant), NRR EO (Morris). * * UPDATE PROVIDED BY CHRIS ROBINSON TO JEFF ROTTON AT 2135 EDT ON 09/14/2015 * * * At 1847 EDT on September 14, 2015, a valid automatic Reactor Protection System (RPS) actuation occurred due to Reactor Water Level 3 while shutdown in MODE 3. Operators were manually controlling Reactor Pressure Vessel (RPV) level and pressure with Reactor Core Isolation Cooling (RCIC) and Safety Relief Valves (SRV). While operators were cycling SRVs, the RPV level went below the Level 3 setpoint. Operators promptly restored RPV level by manual operation of RCIC. The Level 3 actuation and associated isolations were verified to operate properly. The scram signal has been reset. Fermi 2 remains in MODE 3 controlling RPV Level and Pressure through manual operation of RCIC and SRVs. This is the second occurrence of a valid specified safety system actuation reportable under 10CFR50.72(b)(3)(iv)(A) for this ongoing event. The NRC Resident Inspector has been notified. Notified R3DO (Riemer), IRD Manager (Grant), and NRR EO (Morris) * * UPDATE FROM BRETT JEBBIA TO JOHN SHOEMAKER AT 1446 EST ON 2/27/16 * * * This update provides clarification of the applicable reporting criteria for this Event associated with primary containment isolation actuations. Upon the manual reactor scram at 2305 EDT on September 13, 2015, Reactor Protection System (RPS) Level 3 actuated and Primary Containment Isolation System (PCIS) Groups 4, 13 and 15 actuated as expected. The applicable reporting criterion for these actuations is 10 CFR 50.72(b)(3)(iv)(A). The applicable reporting criterion for the manual closure of the inboard and outboard main steam isolation valves at 2345 EDT on September 13, 2015, is also 10 CFR 50.72(b)(3)(iv)(A). In addition, the manual closures of all MSIV lead to a loss of condenser vacuum which resulted in the actuation of PCIS Group 1 at 0001 EDT on September 14, 2015, as expected. The applicable reporting criterion for this actuation is also 10 CFR 50.72(b)(3)(iv)(A). Upon reaching Level 3 at 1847 EDT on September 14, 2015, PCIS Groups 4, 13 and 15 actuated as expected. The applicable reporting criterion for this actuation is 10 CFR 50.72(b)(3)(iv)(A). The licensee informed the NRC Resident Inspector. Notified the R3DO (Stone).
ENS 51313	12 August 2015 14:07:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Technical Specifications Not Met	At 1007 (EDT) on August 12, 2015, while restoring Reactor Building (RB) HVAC (RBHVAC) after surveillance testing, an equipment malfunction resulted in improper damper alignment resulting in Secondary Containment Technical Specifications (TS) to not be met. The plant TS require Secondary Containment pressure be maintained greater than or equal to -0.125 inches of vacuum water gauge (TS SR 3.6.4.1.1). This specification was not maintained for five seconds and the highest pressure observed was -0.095 inches of vacuum water gauge. This value was observed on only one of two installed recorders, of the Secondary Containment pressure recorders. The highest observed pressure on the other recorder was -0.14 inches of vacuum water gauge. Secondary Containment was restored by the Standby Gas Treatment System (SGTS) already in operation and shutting down the affected train of RBHVAC. The technical specification requirement is to maintain secondary containment at -0.125 inches of vacuum water gauge for secondary containment operability. Declaring secondary containment inoperable is reportable under 10 CFR50.72(b)(3)(v)(c) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The licensee notified the NRC Resident Inspector. * * UPDATE AT 1159 EDT ON 08/13/15 FROM BRETT JEBBIA TO S. SANDIN * * * The licensee is updating this report to delete the minus sign for all references to inches of vacuum water gauge. FOLLOW UP - CORRECTED INFORMATION: At 1007 (EDT) on August 12, 2015, while restoring Reactor Building (RB) HVAC (RBHVAC) after surveillance testing, an equipment malfunction resulted in improper damper alignment resulting in Secondary Containment Technical Specifications (TS) to not be met. The plant TS require Secondary Containment pressure be maintained greater than or equal to .125 inches of vacuum water gauge (TS SR 3.6.4.1.1). This specification was not maintained for five seconds and the highest pressure observed was .095 inches of vacuum water gauge. This value was observed on only one, of two installed recorders, of the Secondary Containment pressure recorders. The highest observed pressure on the other recorder was .14 inches of vacuum water gauge. Secondary Containment was restored by the Standby Gas Treatment System (SGTS) already in operation and shutting down the affected train of RBHV AC. The technical specification requirement is to maintain secondary containment greater than or equal to .125 inches of vacuum water gauge for secondary containment operability. Declaring secondary containment inoperable is reportable under 10CFR50.72(b)(3)(v)c as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. the licensee informed the NRC Resident Inspector. Notified R1DO (Powell).
ENS 51242	20 July 2015 11:40:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Temporary Loss of Differential Pressure in Secondary Containment	On the morning of July 20, 2015 at 0740 EDT, with James A. FitzPatrick Nuclear Power Plant (JAF) operating at 100 percent power, the Secondary Containment differential pressure decreased below the JAF Technical Specification (TS) Surveillance Requirement (SR-3.6.4.1.1) value of greater than or equal to 0.25 inch of vacuum water gauge. Both trains of the Standby Gas Treatment System were placed in service and the Reactor Building was isolated. The decrease in Secondary Containment differential pressure was caused by Reactor Building roof maintenance creating multiple openings. Maintenance workers were immediately ordered to stop work and address the condition. Secondary Containment differential pressure was restored to within the TS SR value at 0915 EDT, and remains greater than 0.25 inch of vacuum water gauge. The secondary containment is a structure that surrounds the primary containment and is designed to provide secondary containment for postulated loss-of-coolant accidents inside the primary containment. To prevent exfiltration the secondary containment requires the control volume pressure at less than the external pressure. The differential pressure requirement of TS SR-3.6.4.1.1 ensures that the secondary containment boundary is sufficiently leak tight to preclude exfiltration. During this period there were no unmonitored radioactive releases; however, this event could have prevented the fulfillment of a safety function to control the release of radioactive material and it is reported pursuant to 10 CFR 50.72(b)(3)(v)(C). The NRC Resident Inspector has been informed.
ENS 51240	18 July 2015 17:14:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Standby Gas Treatment System Declared Inoperable Due to Inoperable Sump Pumps	At 1214 CDT on 7/18/2015, during normal surveillance testing of Z sump, Z2 sump pump run time was found to exceed its upper Augmented IST limit, rendering it non-functional. Operators continued with the surveillance, opening the power supply breaker to Z1 sump pump, rendering it non-functional also. When this was identified, station personnel backed out of testing and restored power to Z1 sump pump. Z sump functions to limit condensation buildup in the common Standby Gas Treatment System (SGT) discharge line to support SGT operability. One sump pump is required to be functional to support SGT operability. With both Z sump pumps non-functional, operability of both trains of SGT is not assured. Loss of both trains of SGT constitutes a loss of safety function for control of rad release and accident mitigation per 10 CFR 50.72(b)(3)(v) functions (C) and (D). The breaker for Z1 sump pump was reclosed at 1225 CDT, restoring functionality to Z1 sump pump and operability to both trains of SGT. Both Z sump pumps were considered non-functional and unable to support SGT operability for a period of 11 minutes. The licensee notified the NRC Resident Inspector. * * RETRACTION ON 8/18/2015 AT 1347 EDT FROM DAVID VAN DER KAMP TO MARK ABRAMOVITZ * * * Cooper Nuclear Station (CNS) is retracting the 8-hour non-emergency notification made on July 18, 2015 at 1839 EDT (EN# 51240). The notification on July 18, 2015 reported a condition where the two Z Sump Pumps were considered non-functional and unable to support the operability of the Standby Gas Treatment System (SGT). Subsequent evaluation concluded that Z2 Sump Pump was functional with the run time identified and would have supported the operability of SGT during the time the Z1 Sump Pump breaker was open for surveillance testing. A loss of safety function did not exist. A modification had been previously installed at CNS that prevents the buildup of significant amounts of water in a hold-up line. This volume of water was the basis for the original Z Sump Pump IST (in-service-testing) run times. With the buildup of water previously resolved, the calculated Z Sump Pump IST run times are much longer than measured on July 18, 2015. Notified the R4DO (Hagar).
ENS 51202	7 July 2015 18:35:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Pressure Boundary Vacuum Not Maintained within Specifications	On 7/7/2015 at approximately 1435 EDT, the Technical Specification for Secondary Containment Pressure Boundary was not met when vacuum could not be maintained greater than or equal to -0.125 inches of water gauge for approximately 41 seconds. As part of post-maintenance testing for the non-safety related Reactor Building HVAC Center Exhaust Fan, the fan was started while the safety-related Standby Gas Treatment system was also in operation. Shortly after the fan was started, operators observed degrading vacuum in secondary containment and subsequently secured the center exhaust and supply fans. Vacuum continued to degrade momentarily after the fans were secured, and then returned to a Technical Specification allowable value. Subsequent inspections discovered that the affected fan was operating in the reverse direction. This is believed to have caused Secondary Containment pressure to increase. Since vacuum could not be maintained with the safety-related Standby Gas Treatment system operating, the plant operated in an unanalyzed condition. The cause of the reverse rotation is under investigation. There were no radiological releases associated with this event. The NRC Senior Resident Inspector has been notified. * * UPDATE FROM CHRIS ROBINSON TO VINCE KLCO ON 7/7/2015 AT 2153 EDT* * * Based on plant configuration at the time of the event and further review of the Fermi 2 UFSAR, the plant did not operate in an unanalyzed condition. The Reactor Building HVAC fans would have tripped, as designed, upon receipt of a safety-related Standby Gas Treatment actuation signal during the time of the event. Therefore, the fans' pressurizing effect on secondary containment would have ceased within the time limits assumed in the existing accident analysis. The reporting criteria of 10CFR50.72(b)(3)(v)(C) remains valid. The licensee notified the NRC Resident Inspector. Notified the R3DO (Stone).
ENS 51168	21 June 2015 20:25:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Secondary Containment Isolation	In response to a trip of Unit 2 reactor enclosure HVAC (Heating, Ventilation and Air Conditioning), subsequent loss of reactor enclosure delta-p and rising reactor enclosure room temperatures, a Unit 2 manual secondary containment isolation was initiated per station procedures. This manual isolation also resulted in an isolation signal to containment atmosphere control (CAC) system valves and primary containment instrument gas (PCIG) system valves. System responses were as expected. Unit 2 secondary containment delta-p and room temperatures were restored via the standby gas treatment system (SGTS), and Unit 2 secondary containment integrity remains intact and operable. Investigation of the trip of Unit 2 reactor enclosure HVAC is ongoing. This is being reported under 50.72(b)(3)(iv)(A) for containment isolation signal affecting containment isolation valves in more than one system. The licensee informed the NRC Resident Inspector.
ENS 51038	4 May 2015 11:15:00	10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness	Loss of Power to the Area Radiation and Process Radiation Local Area Network	At approximately 0615 (CDT) on 5/4/2015, the Area Radiation (AR)/Process Radiation (PR) Local Area Network (LAN) lost power following preparation for a planned Unit Sub 1M outage during the current refueling outage. Preparations for this planned outage included de-energizing the plant process computer data diode uninterruptible power supply that in turn caused a loss of power to the AR/PR LAN. As a result of this power loss, indication was lost in the Main Control Room for the main HVAC and Standby Gas Treatment System effluent radiation monitors without any viable compensatory measure to determine Total Noble Gas Release Rates. The station has determined that this constitutes a major loss of assessment capability per 10 CFR 50.72(b)(3)(xiii). Local radiation monitors continue to function properly. Power to the AR/PR LAN has been restored. Time of restoration was 1159 CDT. There is no impact to current plant operation. The NRC Resident Inspector has been notified. * * RETRACTION PROVIDED BY MARK CONSTABLE TO JEFF ROTTON AT 1420 EST ON 12/02/2015 * * * This event has been reviewed and it was determined that the radioactive release rates displayed on the Safety Parameter Display System screens are obtained directly from the associated radiation monitors (0RIX-PR008 and 0RIX-PR012) and HVAC stack and Standby Gas Treatment System stack flow monitors (0UIX-PR050 and 0UIX-PR051) and are not processed through the AR/PR LAN. As a result, there was no loss of radiation release Emergency Action Level assessment capability with a loss of the AR/PR LAN. Therefore, there was no major loss of emergency assessment capability and this event is not reportable. The NRC Resident Inspector has been notified. Notified R3DO (Lipa)
ENS 50831	19 February 2015 08:04:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Building Declared Inoperable Due to Ventilation System Trip	At 0304 EST on February 19, 2015, Fermi 2 experienced a trip of the Reactor Building Ventilation (RB) (HVAC) during plant operations associated with very cold temperatures outside. At the time of the trip, outside air temperature was -1 degrees Fahrenheit and RB HVAC tripped due to a Freeze-Stat actuation (a freeze protection feature). The plant Technical Specifications require that Secondary Containment pressure be maintained greater than or equal to -0.125 inches of vacuum water gauge (TS SR 3.6.4.1.1). This specification was not maintained and the highest pressure observed was -0.11 inches of vacuum water gauge. Subsequently, at 0450, during restoration activities, RB pressure degraded again to higher than -0.125 inches of vacuum water gauge for 38 seconds. The lowest observed pressure was -0.11 inches of vacuum water gauge. RB HVAC has been restored by resetting the Freeze-Stat and the Standby Gas Treatment System (SGTS) has been placed back in a standby condition. The technical specification requirement is to maintain secondary containment at -0.125 inches of vacuum water gauge for secondary containment operability. Declaring secondary containment inoperable is reportable under 10CFR50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The licensee has notified the NRC Resident Inspector. * * RETRACTION FROM WARREN PAUL TO DANIEL MILLS AT 1035 ON 4/8/2015 * * * After reviewing the events that occurred on February 19, 2015 against the accident analyses in Chapter 15 of the UFSAR and design functions of the Standby Gas Treatment System and Secondary Containment structure, it is concluded that a condition that could have prevented the fulfillment of a safety function to control the release of radioactive material did not occur as a result of momentarily exceeding the Technical Specification for Secondary Containment vacuum after a loss of the normal Reactor Building Ventilation System. The Fermi 2 accident analysis for a LOCA does not assume that secondary containment is under vacuum throughout the duration of an accident and contains conservative leakage assumptions to bound the effects of a postulated ground level release. The accident analysis credits the operation of the Standby Gas Treatment System (SGTS); both divisions of SGTS were operable at the time of the event. Although secondary containment was declared inoperable due to exceeding the Technical Specification value for secondary containment vacuum, the structural integrity of the secondary containment was not degraded at the time. Upon receipt of an accident signal, SGTS would have automatically started and restored secondary containment vacuum to within the bounding analyses of Chapter 15 of the UFSAR. Secondary containment was capable of performing its design function of minimizing any ground level release of radioactive material by maintaining boundary integrity so that the SGTS may draw a vacuum in the Reactor Building and filter radioactive material at all times. The event reported in EN # 50831 did not result in a condition that could have prevented the fulfillment of a safety function to control the release of radioactive material. This event report is being retracted. The licensee informed the NRC Resident Inspector. Notified R3DO (Skokowski). * * UPDATE FROM WARREN PAUL TO CHARLES TEAL ON 4/15/15 AT 1348 EDT * * * Upon further review of NUREG-1022 section 3.2.7, the original Non-Emergency Event Notification, 50831, remains valid. The NRC Resident Inspect has been informed. Notified R3DO (McCraw).
ENS 50658	7 October 2014 15:35:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Specified System Actuation	This 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 general containment isolation signals affecting containment isolation valves in more than one system. On October 7, 2014, at 2135 (CDT), while in a refueling outage with the reactor non-critical (Mode 5), work activities were in progress that included replacement of an excess flow check valve and execution of a Technical Specification Surveillance Procedure on the Automatic Depressurization System. Subsequent to valving in a level transmitter (LT), water levels in both the variable and reference legs of the LT were disturbed resulting in a Unit 1 full scram and Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolation signals due to receipt of an invalid low reactor water level signal. The PCIS Groups 2, 3, 6, and 8 isolations caused the initiation of Trains A, B, and C of the Standby Gas Treatment System and Control Room Emergency Ventilation Subsystem 'A'. The Reactor and Refuel Zone ventilation fans tripped and the secondary containment dampers isolated. Operations personnel responded to the PCIS initiation, ensured all equipment operated as designed, and placed affected systems back in service. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Problem Evaluation Report 943038. The NRC Resident Inspector has been notified of this event.
ENS 50579	28 October 2014 21:08:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Reactor Building Vacuum Below Technical Specification Limit	On the evening of October 28, 2014 at 1708 EDT, with James A. FitzPatrick (JAF) Nuclear Power Plant operating at 100 percent power, the Reactor Building differential pressure decreased below the JAF Technical Specification (TS) Surveillance Requirement (SR) value of at least 0.25 inches water vacuum for a period of thirty-four (34) seconds. This occurred during restoration of the Reactor Building Ventilation System (RBVS) following planned maintenance. The Reactor Building differential pressure was 0.50 inches water vacuum with the 'A' RBVS fans in-service in conjunction with the Standby Gas Treatment System (SGTS). The Reactor Building differential pressure decreased to 0.19 inches water vacuum when the SGTS was secured. The Reactor Building Vent was subsequently isolated, and the alternate 'B' RBVS fans were placed in-service; the differential pressure increased to within the required 0.25 inches water vacuum value. The JAF TS bases associated with Secondary Containment state that, 'for Secondary Containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.' Troubleshooting activities indicated that the transient was due to a non-safety related, non-TS damper downstream of one of the 'A' RBVS fans that did not fully stroke open. The subject damper is not part of Secondary Containment, and has no safety related function. This condition did not impact the leak tightness of Secondary Containment or the ability of the associated equipment to establish and maintain the required differential pressure. Secondary Containment would have fulfilled its safety function. However, because the JAF TS SR value of 0.25 inches water vacuum was not met, Secondary Containment was considered Technical Specification INOPERABLE for a period of thirty-four (34) seconds. The Secondary Containment is considered a single-train system; therefore, this condition is reportable pursuant to 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector.
ENS 50375	15 August 2014 17:18:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Standby Gas Treatment System Inoperable	On 8/15/2014 at 1218 CDT, the 'B' Standby Gas Treatment (SBGT) System was undergoing its monthly surveillance testing. With the 'B' fan running, as part of the surveillance, the 'A' Standby Gas Treatment Mode Select Switch was taken to Manual. This renders the 'A' SBGT subsystem inoperable. Almost simultaneously the 'B' fan Flow Indicating Controller went blank and flashed an error message although indicated flow through the 'B' train remained at 4073 SCFM. Based on the indication seen on 'B' controller, regardless of flow, the 'B' SBGT subsystem was also declared inoperable. In accordance with the surveillance the 'A' SBGT mode switch was placed back in the AUTO position on 8/15/2014 at 1220 CDT, restoring that train to operability. The 'B' SBGT was still considered inoperable based on its flow indicating controller being blank and flashing an error message. For a period of two minutes both SBGT subsystems were considered inoperable which is a condition that could have prevented the fulfillment of the safety function of SBGT system to control the release of radioactive material. This is considered a 8-hour reportable event per 50.72(b)(3)(v)(C) '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 control the release of radioactive material.' During the 2 minutes that 'A' SBGT was in Manual, the 'B' SBGT train maintained 4073 scfm train flow which is at the required flow rate per STP 3.6.4.3-01B. In addition, the 'A' train could have been initiated manually at any time during that 2 minutes by the operator who was stationed at the panel performing the surveillance. The licensee notified the NRC Resident Inspector. * * RETRACTION PROVIDED BY BOB MURRELL TO DANIEL MILLS AT 0835 EDT ON 09/16/2014 * * * The purpose of this notification is to retract a previous report made on 08/15/2014 at 1802 (EDT) (EN 50375). Notification of the event to the NRC was initially made as a result of declaring both trains of the Standby Gas Treatment (SBGT) System inoperable. Specifically, during performance of planned surveillance testing required by Technical Specification (TS) Surveillance Requirement (SR) 3.6.4.3.2, with the 'B' train of SBGT running as part of the testing, the 'A' SBGT train's mode switch was taken to manual. This action renders the 'A' train inoperable. Simultaneously with this action, the 'B' SBGT Flow Indicator Controller went blank and flashed an error message. This resulted in the 'B' train being declared inoperable. Subsequent to the initial report, NextEra Energy Duane Arnold (NextEra) has determined that TS SR 3.6.4.3.2 contains a note that states, 'When a SBGT subsystem is placed in an inoperable status solely for the performance of VFTP testing required by this Surveillance on the other subsystem, entry into associated Conditions and Required Actions may be delayed for up to 1 hour.' The 'A' SBGT train mode switch was in manual for approximately 2 minutes; therefore, entry into the associated conditions and actions was not required. Therefore, this event is not considered a Safety System Functional Failure and is not reportable to the NRC as a Licensee Event Report (LER) per 10 CFR 50.73. The NRC Senior Resident Inspector has been notified. Notified R3DO (Dickson)
ENS 50207	17 June 2014 22:29:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Technical Specification Shutdown Due to Loss of a and B Control Bay Chiller Trains	At 1729 (CDT) on June 17, 2014 Browns Ferry Units 1, 2, and 3 initiated actions to commence a reactor shutdown to comply with TS LCO (Technical Specifications Limiting Condition of Operation) 3.0.3. TS LCO 3.0.3 was entered at 1632 and was required due to the concurrent loss of 'A' and 'B' Control Bay Chiller units which resulted in loss of cooling to the U1 and U2 4kV Shutdown Board Rooms. Required actions for the loss of cooling to the U1 and U2 4kV Shutdown Board Rooms is to declare the electrical equipment in the 4kV Shutdown Board Rooms inoperable. The declaration of inoperability of the equipment supported by the U1 and U2 4kV Shutdown Boards resulted in TS LCO 3.0.3 for Units 1, 2 and 3. TS LCO 3.0.3 requires actions to be initiated within one hour to place the units in MODE 2 within 10 hours; MODE 3 within 13 hours; and MODE 4 within 37 hours. This event requires a 4 hour report IAW 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The NRC Resident Inspector has been notified. Service Request 899734 was initiated in the Corrective Action Program. The 'B' Control Bay Chiller train was operating when the control room entered the Loss of Chilling procedure. The 'A' Control Bay Chiller train was started to supply Control Bay cooling, and the 'A' train unit tripped during start. The failure affected unit-3 because each unit has a 50% capacity common standby gas treatment system. The loss of the the U1 and U2 4kV shutdown boards failed two of the three systems. * * UPDATE PROVIDED BY TODD BOHANAN TO JEFF ROTTON AT 2154 EDT ON 06/17/2014 * * * "'A' Control Bay Chiller declared operable at 2035 CDT. TS LCO 3.0.3 exited and plant shutdowns have been secured." The licensee will be notifying the NRC Resident Inspector. Notified R2DO (Haag)
ENS 50164	5 April 2014 11:00:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Primary Containment Isolation System Actuation	This notification is being made in accordance with 10 CFR 50.73 (a)(2)(iv)(A) to provide information pertaining to an invalid Primary Containment Isolation System (PCIS) Group 3 actuation signal that affected containment valves in more than one system. On April 5, 2014, with the reactor at 100% power, an invalid PCIS Group 3 actuation occurred from a momentary spike of the 'A' Refuel Floor radiation monitor which reached the instrument's high radiation trip set point. A radiation protection technician was dispatched to the refuel floor and dose rates in the vicinity of the 'A' radiation monitor detector were verified to be normal and below the alarm set points. The radiation monitor was verified to be indicating normal expected radiation levels. The detector was replaced, a functional check and calibration of the radiation monitor was completed satisfactory and the instrument channel was returned to service. The issue has been entered into the station's corrective action program. Both trains of Standby Gas Treatment System started as designed and Reactor Building ventilation isolated as a result of the invalid PCIS actuation. The PCIS functioned successfully, providing a complete Group 3 isolation. PCIS Group 3 involves the following system isolation valves: Drywell and Suppression Chamber Air and Vent: V16-19-6, 6A, 6B, 7, 7A, 7B, 8, 9, 10, 23 Containment Makeup: V-16-20-20, 22A, 22B Containment Air Sampling: VG-23, 26, V109-76A, 76B Containment Air Compressor Suction: V72-38A, 38B Containment Air Dilution: VG-9A, 9B, 22A, 22B, NG-11A, 11B, 12A, 12B, 13A, 13B Since no actual high radiation condition existed which required the PCIS Group 3 isolation, and the actuation was not in response to actual plant conditions satisfying the requirements for isolation, this event has been classified as an invalid actuation. This event did not result in any adverse impact to the health and safety of the public. In accordance with 10 CFR 50.73(a)(1) a telephone notification is being made instead of submitting a written Licensee Event Report. The licensee has notified the NRC resident inspector.
ENS 49870	5 March 2014 01:17:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Momentary Loss of Secondary Containment	At 1917 hours (CST) on March 4, 2014, the Unit 1B fuel pool radiation monitor spiked high due to an invalid actuation which caused the U1 and U2 reactor building ventilation system to isolate (the control room ventilation system also isolated as designed). The Standby Gas Treatment system was already in operation for a scheduled surveillance as of 1900 hours on March 4, 2014. During the ensuing pressure transient, the Reactor Building differential pressure momentarily went positive. As a result, Secondary Containment was declared inoperable. Given the temporary loss in secondary containment, 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. After the transient, the reactor building ventilation system was shutdown for scheduled maintenance and the control room ventilation system was returned to its normal configuration. The Standby Gas Treatment system was operating to support planned reactor building ventilation system maintenance. Troubleshooting of the radiation monitor spike is underway.
ENS 49868	4 March 2014 06:43:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness	Unit 2 Manual Reactor Scram Following Loss of a Uninterruptible Power Supply (Ups)	At 0137 EST Nine Mile Point Unit 2 experienced a loss of an uninterruptible power supply 2VBB-UPS3B which resulted in a half scram and half isolations. This caused a loss of cooling water to the Reactor Recirculation Pumps and other indications for the loss of power. At 0143 EST a Manual Reactor Scram was inserted due to the rise of temperatures on the Reactor Recirculation Pump seal cavity temperature and motor winding temperature. The reactor building ventilation radiation monitor went non-functional when the reactor building isolated on the loss of UPS power. The standby gas treatment system was started as required and restored the reactor building differential pressure. This is a 4-Hour report for 10CFR50.72(b)(2)(iv)(B) RPS Actuation and 8-Hour report for 10CFR50.72(b)(3)(xiii) Loss of Emergency Assessment Capability. The NRC Resident inspector has been notified. All systems functioned as required following the manual scram. All control rods fully inserted. The cause of the loss of the UPS is under investigation.
ENS 49825	15 December 2013 05:00:00	10 CFR 50.73(a)(1), Submit an LER	60 Day Optional Telephone Notification of Invalid Pcis System Actuation	This notification is being made in accordance with 10 CFR 50.73 (a)(2)(iv)(A) to provide information pertaining to an invalid Primary Containment Isolation System (PCIS) Group 3 actuation signal that affected containment valves in more than one system. On 12/15/2013, with the reactor at 100% power, invalid PCIS Group 3 actuation occurred from a momentary spike of the 'B' Refuel Floor radiation monitor which reached the instrument's high radiation trip set point. A radiation protection technician was dispatched to the refuel floor and dose rates in the vicinity of the 'B' radiation monitor detector were verified to be normal and below the alarm set points. The radiation monitor was verified to be indicating normal expected radiation levels. The detector and trip unit were replaced, a functional check and calibration of the radiation monitor was completed satisfactory and the instrument channel was returned to service. The issue has been entered into the station's corrective action program. Both trains of Standby Gas Treatment System started as designed and Reactor Building ventilation isolated as a result of the invalid PCIS actuation. The PCIS functioned successfully providing a complete Group 3 isolation. The PCIS Group 3 isolation involves the following systems: Drywell and Suppression Chamber Air and Vent: V16-19-6, 6A, 6B, 7, 7A, 7B, 8, 9, 10, 23 Containment Makeup: V-16-20-20, 22A, 22B Containment Air Sampling: VG-23, 26, V109-76A, 76B Containment Air Compressor Suction: V72-38A, 38B Containment Air Dilution: VG-9A, 9B, 22A, 22B, NG-11A, 11B, 12A, 12B, 13A, 13B Since no actual high radiation condition existed which required the PCIS Group 3 isolation, and the actuation was not in response to actual plant conditions satisfying the requirements for isolation, this event has been classified as an invalid actuation. This event did not result in any adverse impact to the health and safety of the public. In accordance with 10 CFR 50.73(a)(1) a telephone notification is being made instead of submitting a written Licensee Event Report. The licensee has notified the NRC Resident Inspector.
ENS 49575	24 November 2013 05:01:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Momentary Loss of Reactor Building Ventilation	On 11/24/2013 at 00:00:54 seconds (EST), Reactor Building ventilation tripped due to low outside air temperature. At 00:01:42 seconds, Secondary Containment Differential pressure went positive, with a maximum of +0.08 inches WC (water column). This is a loss of Secondary Containment function. At 00:02:30 seconds, Standby Gas Treatment System was started and Secondary Containment pressure then decreased to <0 inches WC at 00:03:18 seconds. All the above data parameters were taken from the Division 2 Reactor Building Differential pressure recorder. Secondary Containment pressure is stable with differential pressure negative <-0.30 inches WC. The loss of Secondary Containment function is reportable under 10CFR 50.72(b)(3)(v)(c) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. Technical Specification requirement are to maintain secondary containment pressure <-0.125 WC. No actual radiation release occurred during the event. The licensee notified the NRC Resident Inspector.
ENS 49531	13 November 2013 07:26:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Zone II Differential Pressure Lost During Recovery from Ventilation Drawdown Test	On November 13, 2013 at 0226 (EST), Secondary Containment Zone II (Unit 2 Reactor Building) differential pressure was lost during restoration of a ventilation drawdown test. During restoration Unit 2 'A' Train Reactor Building Ventilation fans tripped. The 'B' Train fans were placed in service and secondary containment was restored. Zone I (Unit 1 Reactor Building) and III (Common Refuel Floor Area) ventilation remained in service and stable. Zone II differential pressure recovered within a few minutes and was verified to be stable. LCO 3.6.4.1 was exited for both units at 0257 (EST). Tech Spec Secondary Containment Operability requires a negative pressure of at least 0.25 inches water gauge. There have been no further perturbations in differential pressure and secondary containment remains operable. This event is being reported under 10 CFR 50.72(b)(3)(v)(c) and per the guidance of NUREG 1022 Rev 3 section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment System. The cause of the Unit 2 "A" Train Reactor Building ventilation fans tripping is still under investigation. The licensee has notified the NRC Resident Inspector. * * RETRACTION ON 1/10/14 AT 1657 EST FROM DOUG LAMARCA TO NESTOR MAKRIS * * * 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).' The event reported in this event notification occurred during a pre-planned evolution for surveillance testing that was done in accordance with an approved procedure and the Susquehanna Technical Specifications. The loss of differential pressure occurred during restoration from the surveillance test and occurred prior to completing the planned evolution and declaring the system OPERABLE. Specifically, the trip of the fans occurred when restoring the Reactor Building normal ventilation after a Zone 1, 2, and 3 isolation (returning back to normal ventilation from the Standby Gas Treatment System). The secondary containment boundary and standby gas treatment system were unaffected. This event occurred as a result of the testing process and would not have occurred during normal operation of the system. There was no discovered condition that would have resulted in the safety function of the system being declared inoperable under normal, non-testing conditions. Based on the above additional information, PPL is retracting this report. Susquehanna 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 the R1DO (Schmidt).
ENS 49358	23 July 2013 04:00:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Primary Containment Isolation System Group 3 Actuation	This notification is being made in accordance with 10CFR50.73 (a)(2)(iv)(A) to provide information pertaining to an invalid Primary Containment Isolation System (PCIS) Group 3 actuation signal that affected containment valves in more than one system. On 7/23/2013, and again on 7/24/2013, with the reactor at 100% power, an invalid PCIS Group 3 actuation occurred from a momentary spike of the 'B' Reactor Building ventilation radiation monitor which reached the instrument's high radiation trip set point. A radiation protection technician was dispatched and radiation levels in the vicinity of the 'B' radiation monitor were verified to be normal and below the alarm set points. The radiation monitor was verified to be indicating normal expected radiation levels. Subsequent visual inspection and functional checks of the radiation monitors were completed satisfactory and the instrument channel was returned to service. At the recommendation from the vendor, the detectors were replaced in both of the Reactor Building ventilation monitors as well as the Refuel Floor radiation monitors. On 8/19/2013, 9/12/2013, and 9/13/2013, with the reactor at 100% power, invalid PCIS Group 3 actuations occurred from a momentary spike of the 'A' Refuel Floor radiation monitor which reached the instrument's high radiation trip set point. A radiation protection technician was dispatched to the refuel floor and dose rates in the vicinity of the 'A' radiation monitor detector were verified to be normal and below the alarm set points. The radiation monitor was verified to be indicating normal expected radiation levels. Subsequent visual inspection and functional checks of the radiation monitors were completed satisfactory and the instrument channel was returned to service. Both trains of Standby Gas Treatment System started as designed and Reactor Building ventilation isolated as a result of the invalid PCIS actuation. The PCIS functioned successfully providing a complete Group 3 isolation. The PCIS Group 3 isolation involves the following systems: Drywell and Suppression Chamber air and vent: V16-19-6A, 6B, 7, 7A, 7B, 8, 9, 10, 23 Containment Makeup: V-16-20-20, 22A, 22B Containment Air Sampling: VG-23, 26, V109-76A, 76B Containment Air compressor suction: V72-38A, 38B Containment Air Dilution: VG-9A, 9B, 22A, 22B, NG-11A, 11B, 12A, 12B, 13A, 13B Since no actual high radiation condition existed which required the PCIS Group 3 isolation, and the actuation was not in response to actual plant conditions satisfying the requirements for isolation, this event has been classified as an invalid actuation. This event did not result in any adverse impact to the health and safety of the public. In accordance with 10CFR50.73(a)(1) a telephone notification is being made instead of submitting a written Licensee Event Report. The licensee has notified the NRC resident inspector.

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