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 Entered dateSiteScramRegionReactor typeEvent description
ENS 5333614 April 2018 14:34:00FermiAutomatic ScramNRC Region 3GE-4

At 1040 EDT, Fermi 2 automatically scrammed on RPV (Reactor Pressure Vessel) Level 3 following a loss of the Division 1 Station System Transformer (SST) #64. All control rods fully inserted. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) automatically started as designed on Reactor Water Level (RWL) 2 and restored RWL. The lowest RWL reached was 101.8 inches (above Top of Active Fuel). HPCI injected for approximately 35 seconds. RWL is currently being maintained in the normal level band with RCIC. No Safety Relief Valves (SRVs) actuated. All isolations and actuations for RWL 3 and 2 occurred as expected. Investigation into loss of SST #64 continues. At the time of the scram, all Emergency Core Cooling Systems (ECCS) and Emergency Diesel Generators (EDGs) were operable, and no safety related equipment was out of service. This report is being made in accordance with 10CFR50.72(b)(2)(iv)(A), any event that results in ECCS discharge into the reactor coolant system as a result of a valid signal and 10CFR50.72(b)(2)(iv)(B), any event that results in the actuation of the Reactor Protection System (RPS) when the reactor is critical. Following the loss of power and reactor scram, the Division 2 EECW (Emergency Equipment Cooling Water) Temperature Control Valve (TCV) controller was in Emergency Manual and maintaining max cooling. Operators placed the controller in Auto and the TCV is controlling normally. The NRC Senior Resident has been notified. Decay heat is being removed via Division 2 steam dumps to the condenser. The plant is in a modified shutdown electric lineup with offsite power available and stable. Emergency diesel generators did automatically start and load.

  • * * UPDATE ON 4/14/2018 AT 1838 EDT FROM JEFF MYERS TO HOWIE CROUCH * * *

This update provides additional clarification of the applicable reporting criteria for this event associated with Primary Containment Isolation Actuations. All isolations and actuations for RWL (Groups 4, 13, and 15) and RWL 2 (Groups 2, 10, 11, 12, 14, 16, 17, and 18) occurred as expected. This report is also being made in accordance with 10CFR50.72(b)(3)(iv)(A), any event or condition that results in valid actuation of any systems listed in paragraph (b)(3)(iv)(B): RPS, HPCI, and RCIC. RPV pressure is being maintained by the bypass valves to the main condenser. All actuations that occurred were fully completed and restored. The licensee notified the NRC Resident Inspector. Notified R3DO (Stone).

  • * * UPDATE ON 4/15/2018 AT 1950 EDT FROM KELLEY BELENKY TO DAVID AIRD * * *

This update provides additional information regarding the specified system actuations and an additional applicable reporting criteria. The loss of Division 1 Station System Transformer (SST) #64 at 1040 EDT on 4/14/2018 resulted in the automatic initiation of Emergency Diesel Generators (EDG) 11 and 12. The EDGs started as expected and continue to supply their associated busses. This is reportable pursuant to 10CFR50.72(b)(3)(iv)(A), as an event or condition that resulted in a valid actuation of any system listed in paragraph (b)(3)(iv)(B), including EDGs. In addition, the loss of the Division 1 SST #64 resulted in the expected transfer from the normal to alternate power source for the Low Pressure Coolant Injection (LPCI) swing bus, rendering LPCI loop select inoperable. The alternate power source continued to energize the LPCI swing bus throughout the event until the system was realigned to the normal power source at 1239 EDT on 4/14/2018. This condition is reportable pursuant to 10CFR50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Notified R3DO (Stone).

ENS 533197 April 2018 12:10:00BrunswickAutomatic ScramNRC Region 2GE-4

On April 7, 2018, at 0836 EDT, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped during testing of the stator cooling system. The trip was uncomplicated with all systems responding normally. No safety-related equipment was inoperable at the time of the event. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B).

Operations responded using Emergency Operating Procedures and stabilized the plant in Mode 3. Reactor water level being maintained via normal feedwater system. Decay heat is being removed through the bypass valves.

Reactor water level reached low level 1 (LL1) as a result of the reactor trip. The LL1 signal causes a Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The LL1 isolations occurred as designed; the Group 8 valves were closed at the time of the event. Due to the Primary Containment Isolation System (PCIS) actuation, this event is also being reported as an eight-hour, non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the PCIS. Unit 2 was not affected. There was no impact on the health and safety of the public or plant personnel. The safety significance of this event is minimal. The automatic reactor trip was not complicated and all safety-related systems operated as designed. Investigation of the cause of the Reactor Protection System actuation is in progress. The licensee notified the NRC Resident Inspector.

ENS 5326918 March 2018 16:16:00Browns FerryAutomatic ScramNRC Region 2GE-4At 1158 CDT on March 18, 2018, the Unit 1 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from High Reactor Steam Dome Pressure in response to Turbine Control Valve Closure. The reactor had been operating at 100 percent power. Investigation is in progress. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Steam Relief Valves (MSRVs) operating on the initial transient as expected. Main Turbine Bypass Valves are currently controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. All safety system operated as expected. At no time was public health and safety at risk. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation' and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.
ENS 5316210 January 2018 13:53:00Browns FerryAutomatic ScramNRC Region 2GE-4At 0928 CST on January 10, 2018, the Unit 3 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from Turbine Control Valve Emergency Trip System pressure low. The reactor had been operating near 73 percent power for an emergent issue for Turbine Control Valve (TCV) No. 3. With TCV No. 3 out of service and closed, the unit was operating with RPS in a half scram condition. A subsequent failure of the TCV No. 2 sensing line resulted in RPS coincidence logic being met for TCV fast closure SCRAM. The investigation of the TCV No. 2 sensing line failure continues. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Turbine Bypass Valves controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. Neither High Pressure Coolant Injection nor Reactor Core Isolation Cooling initiation signals were received. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical except when the actuation results from and is part of a preplanned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident inspector has been notified.
ENS 527958 June 2017 19:10:00SusquehannaAutomatic ScramNRC Region 1GE-4At 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 5269620 April 2017 05:57:00HatchAutomatic ScramNRC Region 2GE-4On 04/20/2017 at 0302 EST during a reactor startup, a reactor scram resulted from upscale spike on two Intermediate Range Monitors (IRMs), 1C51K601A and 1C51K601B. IRM A, 1C51K601A is in Reactor Protection System Channel A and IRM B, 1C51K601B is in Reactor Protection System Channel B. All control rods fully inserted. No PCIS (Primary Containment Isolation System) actuations occurred and none were expected to occur based upon plant condition following the reactor scram. Investigation is in progress. Condition was not due to a true flux event. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B) as an event or condition that resulted in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. CR 10356172 The NRC Resident has been notified. The reactor was at 0.5% (percent) power at the time of the event and will remain in Hot Shutdown pending the results of the root cause investigation.
ENS 5264829 March 2017 23:36:00Browns FerryManual ScramNRC Region 2GE-4At 1844 CDT on 3/29/2017, Unit 2 initiated a manual scram due to multiple rods inserting. At 1842 during Unit 2 start-up, Intermediate Range Monitor (IRM) 'G' drifted low. The operator adjusted the range down one position with no immediate reaction. At 1844, a spike on IRM 'G' caused a half scram on Reactor Protection System (RPS) 'A' trip system. The half scram was being reset after evaluating no trip condition was present. As the operator reset groups 2 and 3, a trip signal from IRM 'F' was received on the RPS 'B' trip system, resulting in rod insertion for groups 1 and 4. When the operator identified multiple rods inserting, the actions of procedure 2-AOI-100-1 were followed and a manual scram was inserted. Investigation is ongoing. All safety systems remained in standby readiness configuration. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. Primary Containment Isolations Systems did not receive an actuation signal and performed as designed. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the RPS when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of systems listed in paragraph (b)(3)(iv)(B) Reactor Protection System(RPS) including reactor scram and reactor trip'. This event requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.
ENS 5204224 June 2016 16:06:00FitzPatrickManual ScramNRC Region 1GE-4At 1215 (EDT) on 6/24/2016, James A. FitzPatrick (JAF) was at 100% power when Breaker 710340 tripped and power was lost to L-gears L13, L23, L33, and L43. These provide non-vital power to Reactor Building Ventilation (RBV), portions of Reactor Building Closed Loop Cooling (RBCLC), and 'A' Recirculation pump lube oil systems. Off-site AC power remains available to vital systems and Emergency Diesel Generators (EDG) are available. Due to the loss of RBV, Secondary Containment differential pressure increased. At 1215 (EDT), Secondary Containment differential pressure exceeded the Technical Specifications (TS) Surveillance Requirement SR-3.6.4.1.1 of greater than or equal to 0.25 inches of vacuum water gauge. The Standby Gas Treatment (SBGT) system was manually initiated and Secondary Containment differential pressure was restored by 1219 (EDT). The 'A' Recirculation pump tripped at 1215 (EDT) and reactor power decreased to approximately 50%. 'B' Recirculation pump temperature began to rise due to the degraded RBCLC system. At 1236 (EDT), a manual scram was initiated. Reactor Pressure Vessel (RPV) water level shrink during the scram resulted in a successful Group 2 isolation. All control rods have been inserted. The RPV water level is being maintained with the Feedwater System and pressure is being maintained by main steam line bypass valves. A cooldown is in progress and JAF will proceed to cold shutdown (Mode 4). Due to complete loss of RBCLC system, the Spent Fuel Pool (SFP) cooling capability is degraded but the Decay Heat Removal system remains available. SFP temperature is slowly rising and it is being monitored. The time (duration) to 200 degrees is approximately 117 hours. The initiation of reactor protection systems (RPS) due to the manual scram at critical power is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The general containment Group 2 isolations are reportable per 10 CFR 50.72(b)(3)(iv)(A). In addition, the temporary differential pressure change in Secondary Containment is reportable per 10 CFR 50.72(b)(3)(v)(C), as an event that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector and the State of New York.
ENS 519681 June 2016 10:40:00LimerickManual ScramNRC Region 1GE-4Limerick Unit 2 was manually scrammed from 100 (percent) power at 0900 (EDT) on 6/1/2016 in accordance with plant procedure OT-112 'Unexpected/Unexplained change in core flow' when both 2A and 2B Recirculation Pump Adjustable Speed Drives (ASDs) tripped due to an electrical fault. The shutdown was normal and the plant is stable in Hot Shutdown with normal pressure control via the Main Steam Bypass valves to the Main Condenser and normal level control using Feedwater. The Manual RPS actuation is reportable under 10 CFR 50.72 (b)(2). All rods inserted fully on manual scram and the plant is in a normal shutdown electrical line up. Unit 1 was not affected by this event. The licensee plans to issue a press release. The licensee notified local counties and Pennsylvania Emergency Management Agency (PEMA). The NRC Resident Inspector has been notified.
ENS 5192513 May 2016 05:00:00SusquehannaManual ScramNRC Region 1GE-4At approximately 0110 hours (EDT) on May 13, 2016, Susquehanna Steam Electric Station Unit Two reactor was manually scrammed by plant operators due to a sustained loss of AC power to essential plant loads. Power to MCC 2B246 was lost at 2355 on May 12, 2016, resulting in a loss of Drywell cooling. Drywell pressure increased to 1.3 psig when operators placed the mode switch to the shutdown position to manually SCRAM the reactor. All rods inserted as expected. Reactor water level lowered to -27 inches and was immediately restored by normal feedwater level control. Level 3 (+13 inch) PCIS isolations occurred, along with an initiation of the RCIC system (-30 inches). Once adequate level was verified, RCIC was overridden. Pressure was controlled with turbine bypass valves, and subsequently main steam line drains. All safety systems functioned as expected. The power loss also tripped Reactor Building HVAC, causing a loss of secondary containment differential pressure resulting in a loss of safety function. Due to the loss of drywell cooling, high drywell pressure actuations and a second reactor SCRAM signal, this signal was automatic, occurred at 0314 hours. HPCI (which automatically initiated on high drywell pressure) was subsequently overridden and declared inoperable, resulting in a loss of safety function. (HPCI did not inject into the vessel). The reactor is currently stable in Mode 3. Initial reports from the field indicate a phase to phase fault on the MCC 2B246 bus bars. The licensee has notified the NRC Resident Inspector and will be issuing a press release.
ENS 517157 February 2016 13:46:00BrunswickManual ScramNRC Region 2GE-4

At 1346 EST the licensee reported that at 1326, Brunswick Unit 1 declared an Alert under EAL HA 2.1 due to an explosion/fire in the Balance of Plant 4 kV switchgear bus area. Prior to the Alert declaration, the operators initiated a manual SCRAM due to an unexpected power decrease from 88% to 40%. The licensee has visually verified that there is no ongoing fire and is investigating the initial cause of the event. Offsite power is available to the site, but EDGs 1 and 2 are running and supplying Unit 1 loads. The MSIVs shut and HPCI/RCIC are being used to maintain vessel level. At 1412 EST, NRC decided to remain in Normal Mode. At 1704 EST the licensee reported the following: At 1313 hours Eastern Standard Time (EST) a manual reactor scram was initiated due to loss of both recirculation system variable speed drives as a result of an electrical fault. At this time, a Startup Auxiliary Transformer (SAT) experienced a lockout fault; interrupting offsite power to emergency buses 1 and 2. Emergency Diesel Generators (EDGs) 1, 2, 3, and 4 automatically started and EDGs 1 and 2 synchronized to emergency buses 1 and 2 per design. The power interruption resulted in closure of the Main Steam Isolation Valves, per design. The manual scram also resulted in closure of Group 2, 6, and 6 Containment Isolation Valves. The Reactor Core Isolation Cooling (RCIC) system was manually started and is being used to control reactor water level. The High Pressure Coolant Injection (HPCI) system was manually started and is being used for pressure control. The Plant response to the event was per design. Unit 2 is not directly affected by the event, however, due to the shared electrical distribution system is in a Technical Specification Action statement due to the Inoperable Unit 1 SAT. The public health and safety is not impacted by this event. At 1751 EST, the licensee reported that the emergency declaration had been downgraded to an Unusual Event at 1730 because the plant no longer meets the criteria for an Alert, but does meet the criteria for an Unusual Event due to a "loss of all offsite power to Emergency 4 kV buses E1 (E3) and E2 (E4) for greater than or equal to 15 minutes." The NRC Resident Inspector has been notified. The licensee has notified the State and Local governments. Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).

  • * * UPDATE FROM MARTY IRWIN TO DANIEL MILLS AT 1825 ON 2/07/16 * * *

At 1814 EST the emergency declaration was terminated because offsite power was restored. The NRC Resident Inspector has been notified. The licensee has notified the State and Local governments. Notified R2DO (Musser), NRR EO (Morris), IRD MOC (Stapleton), R2RA (Haney), NRR ET (Lubinski), NRR ET (Dean), DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).

ENS 5168024 January 2016 01:10:00FitzPatrickManual ScramNRC Region 1GE-4At 2241 (EST) on 1/23/2016, James A FitzPatrick inserted a manual scram from 89 percent power due to lowering intake level. Following the successful scram, a residual transfer occurred, resulting in a loss of the non-vital busses, loss of all Circulating Water Pumps, and a manual closure of the Main Steam Isolation Valves (MSIVs). The cause of the residual transfer is unknown. RPV (Reactor Pressure Vessel) level shrink during the scram resulted in a successful Group 2 isolation. Reactor Vessel (level) and pressure are being maintained with the High Pressure Coolant Injection System which was manually started. A cooldown is in progress. FitzPatrick will proceed to Mode 5 until the cause is identified and corrected. The Emergency Diesel generators auto started as a result of the loss of power to the non-vital busses. Offsite power remained available throughout the event. Operators are controlling pressure manually via the relief valves. FitzPatrick will notify the Public Service Commission of the event. The NRC Resident Inspector was notified.
ENS 5161419 December 2015 10:38:00LimerickAutomatic ScramNRC Region 1GE-4At 0702 EST on 12/19/2015, the Unit 2 reactor automatically scrammed on a valid reactor low level signal (12.5 inches). The reactor low level signal was caused by the trip of the in-service 2A Reactor Feedwater Pump, causing reactor level to lower, exceeding the low reactor level setpoint of 12.5 inches. The shutdown was normal. The plant is stable in Hot Shutdown with normal reactor level, pressure control via the Main Steam Bypass Valves to the Main Condenser and normal level control using the Condensate System. High level trip of the 2A Reactor Feedwater Pump was caused by high reactor level of +54 inches following opening of bypass valves during reactor start-up and pressurization. The cause is still being investigated. All systems functioned as expected following the reactor scram. The licensee notified the NRC Resident Inspector.
ENS 5143028 September 2015 22:49:00Hope CreekAutomatic ScramNRC Region 1GE-4On September 28, 2015 at 2046 EDT, the Hope Creek reactor scrammed following a trip of both reactor recirculation pumps. All control rods fully inserted into the core. All safety systems responded as designed and expected. There was no radiological release. The unit is stable in Mode 3 with decay heat being removed via the turbine bypass valves rejecting steam to the main condenser. Normal feedwater level control is providing makeup to the reactor vessel. No personnel injuries resulted from the event. The Outage Control Center has been staffed to determine the cause of the reactor scram. The Hope Creek NRC Resident Inspector has been notified. The licensee notified Lower Alloways Creek township of the event.
ENS 5139114 September 2015 02:46:00FermiManual ScramNRC Region 3GE-4

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 5098113 April 2015 09:56:00LimerickManual ScramNRC Region 1GE-4Following a pre-planned scram for entry into the U2 Refueling Outage, the 'B' Reactor Protection system (RPS) was unable to be reset due to a deficiency with the '2H' Intermediate Range Power Monitor. Due to this failure to reset RPS, a manual full scram was initiated as required by plant procedures. The licensee notified the NRC Resident Inspector.
ENS 5097311 April 2015 01:52:00SusquehannaManual Scram
Automatic Scram
NRC Region 1GE-4At 2346 EDT on April 10, 2015, Susquehanna Steam Electric Station Unit 2 reactor automatically scrammed due to a main turbine trip caused by loss of turbine steam seals and degrading main condenser vacuum. Unit 2 reactor was being shutdown for a refueling outage. At approximately 37 percent power, turbine steam seals were lost resulting in a degrading vacuum. The vacuum degraded quickly, resulting in a main turbine trip before the reactor operator could insert a manual scram. At 37 percent power, the turbine trip caused an automatic scram. This occurred during a transfer from normal steam seal supply to the auxiliary boiler supply. All control rods (fully) inserted. Reactor water level lowered to +2 inches causing Level 3 (+13 inches) isolation. No ECCS actuations occurred. The Operations crew subsequently maintained reactor water level at the normal operating band using RCIC (Reactor Core Isolation Cooling). No steam relief valves opened. The reactor recirculation pumps tripped on EOC-RPT due to the turbine trip at power. The reactor is currently stable in Mode 3. Investigation into the cause of the loss of turbine steam seals is underway. The NRC Resident Inspector was notified. A voluntary notification to PEMA (Pennsylvania Emergency Management Agency) and press release will occur. Unit 2 is in a normal shutdown electrical lineup. Turbine steam seals were restored to the normal steam supply and condenser vacuum was restored. Decay heat is being removed via the steam bypass valves to the condenser. Unit 2 is proceeding with their cooldown to support the scheduled refueling outage.
ENS 5090319 March 2015 10:51:00FermiAutomatic ScramNRC Region 3GE-4At 0702 EDT on March 19, 2015, Fermi 2 received an automatic scram due to actuation of the Reactor Protection System (RPS) function of Oscillation Power Range Monitor (OPRM) Upscale. The plant had recently transitioned to Single Loop Operation after securing the 'A' Reactor Recirculation Pump due to loss of normal and emergency cooling water supply. The lowest reactor water level was 134 inches above top of active fuel. Reactor water level is being maintained in the normal band by the Feedwater and Control Rod Drive Systems. No Safety Relief Valves (SRV) actuated. Reactor pressure is being maintained via the Main Turbine Bypass Valves and Main Condenser. Reactor Pressure Vessel Level 3 isolation occurred. No additional safety system actuations occurred. All off-site power sources were available throughout the event. The plant is currently in Mode 3 and in a stable condition. Investigation into the cause of the event is ongoing. This event is being reported under the four hour Non-Emergency reporting criteria of 10CFR50.72(b)(2)(iv)(B). The NRC Resident Inspector has been notified.
ENS 5084724 February 2015 01:02:00LimerickAutomatic ScramNRC Region 1GE-4At 2140 EST on 02/23/2015, Unit 1 reactor automatically scrammed on a valid reactor high pressure signal (1096#). The reactor high pressure signal was caused by the closure of the 1C inboard main steam isolation valve (MSIV), causing reactor pressure to rise, exceeding the reactor protection system (RPS) setpoint of 1096# pressure. The shutdown was normal and the plant is stable in Hot Shutdown with normal pressure control via the main steam bypass valves to the main condenser and normal level control using the feedwater system. The closure of the 1C inboard MSIV appears to have been caused by a loss of primary containment instrument gas (PCIG) pneumatic supply to the valve. Instrument air was aligned to the remaining MSIV's. Limerick Unit 1 will remain in Hot Shutdown until repairs can be made. All rods inserted into the core during the scram. No relief or safety valves actuated during the transient. The electric grid is stable and supplying all plant loads. There was no affect on Unit 2. The licensee notified the NRC Resident Inspector, Pennsylvania Emergency Management Agency, and Berks, Chester and Montgomery counties.
ENS 5040426 August 2014 21:24:00Browns FerryAutomatic ScramNRC Region 2GE-4At 1730 CDT on August 26, 2014, Browns Ferry Unit 1 experienced a turbine trip resulting in an automatic reactor scram. The cause of the turbine trip was a control valve fast closure signal that was generated by a turbine trip on generator neutral over voltage signal. The Main Steam Isolation Valves (MSIVs) remained open with the main turbine bypass valves controlling reactor pressure. The Reactor Feedwater Pumps are in service to control reactor water level. Primary Containment Isolation Systems (PCIS) Groups 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all required components actuated as required with the exception of Standby Gas Treatment (SBGT) train A, which is under a clearance for planned maintenance. Neither High Pressure Coolant Injection (HPCI) nor Reactor Core Isolation Cooling (RCIC) initiation signals were received. Initially, three Main Steam Relief Valves (MSRVs) opened to control the pressure surge and subsequently reclosed. This event requires a 4 hour report per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an 8 hour report per 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including reactor scram or reactor trip, and (2) General containment Isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).' The NRC Resident Inspector has been notified. Service Request 926468 was initiated in the Corrective Action Program. The plant is in its normal shutdown electrical lineup. The licensee is investigating the cause of the generator neutral overvoltage signal. There was no impact on units 2 and 3.
ENS 500906 May 2014 13:27:00Browns FerryAutomatic ScramNRC Region 2GE-4

At 0830 (CDT) on 05/06/2014, the Unit 3 reactor automatically scrammed due to low reactor water level as a result of a trip of both recirculation pumps. Main Steam Isolation Valves remained open with main turbine bypass valves controlling reactor pressure. Reactor feedwater pumps are in service to control reactor water level. Primary Containment Isolation System Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. The Reactor Feedwater System controlled and maintained water level above the level 2 initiation setpoint. Prior to the Scram, the reactor was operating at 100% power. A Core and Containment Cooling Systems Analog Trip Unit Functional Test was in progress. The cause of the recirculation pump trip is under investigation. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified. U1 and U2 remained at 100% power and were unaffected.

  • * * UPDATE AT 1302 EDT ON 05/09/14 FROM TODD BOHANAN TO DONG PARK * * *

Investigation revealed that a failed power supply caused an Anticipated Transient Without Scram/Alternate Rod Insertion (ATWS/ARI) signal to be generated when a level 2 Reactor Water Level was simulated on one instrument. All systems responded to the ATWS/ARI signal as designed. This signal opened the Recirc Pump Trip breakers for both Recirculation Pumps and opened the ARI valves to bleed air from the Reactor Protection System (RPS) scram air header. The resulting transient caused reactor water level to dip below the RPS trip setpoint (level 3 Reactor Water Level), a normal plant response, and the automatic scram signal occurred. At the time of the RPS scram signal, all rods were inserting and reactor power was approximately 2-3% and lowering. The NRC Resident Inspector has been notified. Notified R2DO (Bonser).

ENS 498715 March 2014 02:26:00LimerickManual ScramNRC Region 1GE-4At 2334 EST on 3/4/14 Unit 1 was manually scammed during a Rapid Plant Shutdown. The Rapid Plant Shutdown was initiated due to an Electro Hydraulic (Control) (EHC) System failure resulting in all Low Pressure Turbine lntercept Valves failing closed. The shutdown was normal and the plant is stable in Hot Shutdown with normal pressure control via the Main Steam Bypass Valves to the Main Condenser and normal level control using Feedwater. The licensee informed both State and local agencies and the NRC Resident Inspector. A press release will be issued by the licensee.
ENS 496085 December 2013 05:40:00Hope CreekAutomatic ScramNRC Region 1GE-4

While operating at 76% power on 12/5/13 at 0325 EST, the main turbine tripped on moisture separator hi level. The reactor scrammed along with the main turbine trip. All safety systems responded as designed and expected. There was no radiological release. There were no injuries. During the scram, all rods inserted into the core. Plant is stable in Mode 3 in its normal S/D (shutdown) electrical line up. Decay heat is being removed via the turbine bypass valves dumping steam to the main condenser. At 0505 EST while securing from cooldown in an attempt to start a recirc pump, BPVs (Bypass Valve) opened causing reactor level swell and subsequent shrink. During this time, RPV (Reactor Pressure Vessel) level lowered to below RPV level 3 and caused a RPS (Reactor Protection System) actuation. RPV level was recovered and is now stable in normal band. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE ON 12/5/13 AT 1000 EST FROM LINDSAY KOBERLEIN TO DONG PARK * * *

This update to ENS #49608 adds reporting criterion 10CFR50.72(b)(3)(iv)(A) for the RPS actuation at 0505 EST during post-scram recovery.

The licensee notified the NRC Resident Inspector and the Lower Alloways Creek township. The licensee will be making a press release. Notified R1DO (Cook).

ENS 495921 December 2013 10:02:00Hope CreekAutomatic ScramNRC Region 1GE-4While operating at 100% power on 12/01/2013, at 0613 EST, the main turbine tripped on moisture separator hi level. The reactor scrammed along with the main turbine trip. All safety systems responded as designed and expected. There was no radiological release. There were no injuries. During the scram, all rods inserted into the core. The plant is stable in mode 3 in its normal shutdown electrical line up. Decay heat is being removed via the turbine bypass valves dumping steam to the main condenser. The licensee notified the NRC Resident Inspector and will be notifying Lower Alloways Creek township.
ENS 4934415 September 2013 11:52:00SusquehannaManual ScramNRC Region 1GE-4

At 1123 EDT, (on 9/15/13), Susquehanna Unit 2 received a Division 2 RHR (Residual Heat Removal) room Flooded alarm. Plant operators reported 3 inches of water in the room with water coming from the 2B RHR pump suction relief valve. The Suppression Pool Isolation valve was closed to isolate the leaking relief valve from the Suppression Pool. This action stopped Suppression Pool level from lowering. The Leak is isolated at this time. Further water is being added to the room as the system piping is being drained. The cause of the relief valve lifting is unknown at this time and is under investigation. At the time of the event, Division 2 RHR system had been declared inoperable for unrelated equipment issues. At 1222 EDT, the licensee confirmed that leakage from the 2B RHR pump suction relief valve had stopped. Susquehanna Unit 2 remains stable and there was no impact on Susquehanna Unit 1. The licensee notified the NRC Resident Inspector, State, and Local authorities. Notified DHS SWO, FEMA, DHS NICC, and Nuclear SSA (email only).

  • * * UPDATE FROM MARTIN LICHTNER TO JOHN SHOEMAKER AT 1641 EDT ON 9/15/13 * * *

U2 SSES (Susquehanna Steam Electric Station) has exited the Unusual Event for Division 2 RHR room flooded as of 1552 EDT. The leak has been isolated, water removed from the room and Division 1 RHR is operating in shutdown cooling proceeding to Mode 4 (Cold Shutdown). A press release for this event was authorized at 1613 and issued at 1628 (on 9/15/13). The licensee has notified the NRC Resident Inspector. Notified R1DO (Cahill), NRR (Dorman), IRD (Gott), DHS SWO, FEMA, DHS NICC, and Nuclear SSA (email only).

ENS 4934214 September 2013 05:15:00SusquehannaManual ScramNRC Region 1GE-4At approximately 0330 hours on September 14, 2013, Susquehanna Steam Electric Station Unit 2 reactor was manually scrammed while transitioning the 'A' reactor feed pump from flow control mode to discharge pressure mode. Reactor water level rose to +54 inches causing a trip of reactor feedpumps. Subsequently the mode switch was taken to shutdown to manually scram the Unit 2 reactor. All control rods inserted. Reactor water level lowered to approximately +18 inches. There were no automatic emergency core cooling system initiations. No steam relief valves opened during the event. No containment isolations occurred. All safety systems operated as expected. RCIC system was manually initiated for level control until a reactor feedpump was recovered, then RCIC was manually shutdown. The cause of the feedwater flow transient and trip of the reactor feedwater pumps is under investigation. This report is being made per 10CFR50.72(b)(2)(iv)(B) for a 4 hour report, and 10CFR50.72(b)(3)(iv)(A) for an 8 hour report. Decay heat is being removed via the turbine bypass valve to the condenser. Offsite power remains stable, and there was no impact on Unit 1. The licensee has notified the NRC Resident Inspector. The Pennsylvania Emergency Management Agency will be notified, and the licensee will be making a press release.
ENS 4910812 June 2013 16:59:00Hope CreekManual ScramNRC Region 1GE-4This is a report of a manual RPS actuation and manual RCIC actuation per 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). At 1332 (EDT), on 6/12/13, the 'B' Circulating Water Pump tripped with a stuck open discharge valve resulting in a vacuum transient. Operators lowered reactor power from 100% in an effort to stabilize condenser vacuum. When vacuum reached 6.5 inches, the operators inserted a manual reactor scram at 1333 (EDT). All control rods inserted as required. No automatic ECCS or RCIC initiations occurred. No primary or secondary containment isolations occurred. The plant is stable in OP CON 3 HOT SHUTDOWN with the condensate pumps in service. The Reactor Recirculation Pumps are in service. At the time of the event, a RCIC surveillance was in progress, but did not contribute to the event. The RCIC pump was secured and subsequently placed in service for inventory control. The only safety-related equipment out of service at the time of the scram was the C Service Water Pump, which was tagged for scheduled maintenance. No personnel injuries occurred. No radiation releases occurred. The NRC Resident Inspector has been informed.
ENS 490997 June 2013 13:50:00SusquehannaManual ScramNRC Region 1GE-4At approximately 1203 (EDT) on June 7, 2013, Susquehanna Steam Electric Station Unit One reactor was manually scrammed during reactor startup. Pressure setpoint was being adjusted to the normal operating setpoint, from 750 psig to 934 psig, when all turbine bypass valves unexpectedly opened. Reactor Feed Pumps, Main Turbine, HPCI and RCIC tripped on the high level setpoint Level 8 (+54 inches) due to the resultant reactor level swell. The reactor operator then inserted a manual scram. All control rods inserted. Reactor water level lowered to approximately -10 inches causing Level 3 (+13 inches) isolations. There were no automatic emergency core cooling system initiations. No steam relief valves opened during the event. All safety systems operated as expected. The cause of the Turbine Bypass valve opening is under investigation. Unit 2 was unaffected. The licensee notified the NRC Resident Inspector and will notify the Pennsylvania Emergency Management Agency.
ENS 4893617 April 2013 02:11:00LimerickManual ScramNRC Region 1GE-4During outage main turbine stop valve RPS logic surveillance testing, an invalid RPS actuation occurred due to an error in executing main turbine surveillance testing procedures. A Turbine Stop Valve closure RPS signal occurred due to an error in the restoration sequence of restoring the RPS bypass signal and a subsequent manual trip of the main turbine. This resulted in a full scram and a trip of both reactor recirculation pumps. The site post-scram response procedure was entered, which required that the mode switch be placed in the locked SHUTDOWN position. This caused an expected but valid RPS actuation. No control rod motion occurred due to all control rods were inserted at the time of the invalid RPS actuation and subsequent valid RPS actuation. The license has notified the NRC Resident Inspector.
ENS 4882919 March 2013 08:37:00Browns FerryManual ScramNRC Region 2GE-4At 0402 (CDT) on 03/19/2013, the Unit 1 reactor was manually scrammed due to lowering main condenser vacuum. The cause of the loss of vacuum was a significant leak on the 1C feedwater heater level control line. The leak appeared as a steam/water leak near the penetration to the main condenser. As extraction steam was isolated, condenser vacuum deteriorated and was approaching the turbine trip setpoint, at which time the reactor was manually scrammed. Condenser vacuum recovered following the scram. MSIVs (Main Steam Isolation Valves) are open, main turbine bypass valves are controlling reactor pressure and reactor feedwater pumps are being used to control reactor water level. All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated as required. In response to the scram, all plant equipment responded as designed. The reactor had been operating near 95% power for several hours due to the 1C3 heater isolating at 2334 (hrs. CDT) on 3/18/13. This event is reportable within 4 hours per 10CFR50.72(b)(2)(iv)(B) `any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR50.73(a)(2)(iv)(A). All rods inserted into the core during the scram. No safety relief valves lifted during the transient. The plant is in its normal shutdown electrical lineup. The licensee has notified the NRC Resident Inspector.
ENS 4878225 February 2013 17:49:00Browns FerryAutomatic ScramNRC Region 2GE-4At 1313 (CST) on 02/25/2013, the Unit 3 reactor automatically scrammed due to actuation of the Reactor Protection System from a turbine trip. Preliminary indications show the turbine tripped on low condenser vacuum. Cause of loss of condenser vacuum has been identified as Reactor Feedwater recirculation piping separation. Main Steam Isolation Valves (MSIVs) were manually closed to isolate the leak. None of the Safety Relief Valves (SRVs) automatically cycled during the transient, and one Safety Relief Valve (SRV) was manually operated to maintain Reactor Pressure due to the Main Turbine Bypass Valves unavailability because of loss of condenser vacuum. All systems responded as expected to the turbine trip. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC), reactor water level initiation set points were reached. Reactor water level is being controlled by the RCIC system and Reactor Pressure is being controlled with the High Pressure Coolant Injection (HPCI) system. All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated, with the exception of one valve in Group 6. Drywell Continuous Air Monitor (CAM) Inboard Return Isolation Valve 3-FSV-90-257 did not have indication following isolation signal and was not able to be verified locally. Indication was subsequently restored following restoration of containment isolation signals, and the Drywell CAM was manually isolated at 1422 (CST) with positive indication of isolation, and isolation valves deactivated at time 1514 (CST) to satisfy TS LCO 3.6.1.3 required actions. This event is reportable within 4 hours per 10CFR50.72(b)(2)( iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation'. It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). At 1415 (CST), Suppression Pool Water level exceeded -1 inch due to operation with HPCI in pressure control mode, and required entry into TS LCO 3.6.2.2 condition A to restore level within 2 hours. Efforts are being made to lower suppression pool water level within limits. At 1615 (CST), water level remains above -1 inch requiring entry into TS LCO 3.6.2.2 condition B requiring action to be in MODE 3 in 12 hours and MODE 4 within 36 hours. This event is reportable within 4 hours per 10CFR 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. All control rods fully inserted and electrical offsite power is in a normal shutdown configuration. Residual Heat Removal is aligned for suppression pool cooling. There was no impact on either Unit 1 or 2.
ENS 4873810 February 2013 09:11:00HatchManual ScramNRC Region 2GE-4During normal power operations, the crew observed condensate/feedwater conductivity begin to increase at approximately 0530 EST on 02/10/13. The crew responded to the associated alarm response procedures and entered abnormal operating procedure 34AB-N61-001-1 due to degrading reactor water chemistry parameters. A power reduction (from 100%) was initiated at 0555 EST in accordance with station procedures for responding to a suspected condenser tube leak. At 0700 EST, a manual reactor SCRAM (from approximately 47%) was inserted due to the elevated reactor water conductivity in accordance with station abnormal operating procedures. All rods inserted completely and no complications were encountered following the reactor SCRAM, normal feedwater injection remained available. Following the SCRAM, a Group 2 Primary Containment Isolation Signal (PCIS) was received as a result of reactor water level lowering below +3 inches. The lowest reactor water level observed was (minus) 2 inches and was restored to normal operating levels utilizing normal feedwater injection. Following restoration of reactor water level to the normal operating level, the Group 2 PCIS signal was reset. No ECCS injection systems actuated as a result of the reactor SCRAM. The SCRAM was uncomplicated and the plant is stable. Decay heat removal is to the main condenser via the turbine bypass valves. The plant is in a normal offsite electrical power shutdown alignment. Efforts are in progress to isolate the condenser in-leakage. There was no impact on Unit 2. The licensee has notified the NRC Resident Inspector.
ENS 4862322 December 2012 16:39:00Browns FerryAutomatic ScramNRC Region 2GE-4On 12/22/2012 at 1152 CST, the Unit 2 reactor automatically scrammed due to actuation of the Reactor Protection System (RPS) from loss of power to RPS. At 1134 CST, the D 4kV Shutdown Board unexpectedly de-energized during performance of post-maintenance testing for the 3D Diesel Generator paralleling circuitry, resulting in loss of power to the 2B RPS subsystem. Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations were received along with automatic initiation of A, B, and C Standby Gas Treatment subsystems and A Control Room Emergency Ventilation subsystem due to loss of power to the 2B RPS subsystem. While attempting to reenergize the 2B RPS subsystem, the 2A RPS subsystem was inadvertently de-energized resulting in Unit 2 reactor automatic scram. All affected safety systems responded as expected for the loss of RPS and reactor scram. Due to the loss of RPS, the Main Steam Isolation Valves (MSIVs) closed. Reactor pressure did not rise to the automatic initiation set point for Safety Relief Valve (SRV) actuation. Reactor Core Isolation Cooling System (RCIC) and High Pressure Coolant Injection System (HPCI) reactor water level initiation set point of -45" was reached and RCIC and HPCI automatically initiated as designed to restore water level above the initiation set point. Both Recirculation Pumps also tripped on reactor water level of -45". Reactor pressure control was established by manually operating one SRV and water level control established with RCIC. HPCI was returned to standby readiness. The scram was reset, MSIVs were opened, and the Main Condenser was established as a heat sink. The scram event from critical is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B). It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector was notified. The 2A and 2B RPS subsystems were returned to service. The electrical grid is stable and supplying shutdown loads on Unit 2. Unit 1 and Unit 3 were unaffected and continue to operate at 100% power.
ENS 4860719 December 2012 20:29:00SusquehannaAutomatic ScramNRC Region 1GE-4At approximately 17:31 hours on December 19, 2012, Susquehanna Steam Electric Station Unit Two reactor automatically scrammed on low RPV level (Level 3, +13 inches) while transitioning the 'A' reactor feed pump from discharge pressure mode to flow control mode. All control rods inserted and both reactor recirculation pumps tripped. Reactor water level lowered to approximately -29 inches causing Level 3 (+13 inches) isolations. An automatic trip of the reactor recirculation pumps occurred, but is not expected at this RPV level. There were no automatic emergency core cooling system initiations. No steam relief valves opened during the event. All safety systems operated as expected. The cause of the loss of feed water flow and trip of the reactor recirculation pumps is under investigation. This report is being made per 10CFR50.72(b)(2)(iv)(B) for a 4 hour report, and 10CFR50.72(b)(3)(iv)(A) for an 8 hour report. Decay heat is removed via steam to the main condenser using the bypass valves . On-site electrical power is in the normal configuration. The Unit 2 reactor is currently stable in Mode 3. Unit 1 was not affected and operates at 99% power. The licensee will inform the Commonwealth of Pennsylvania and make a press release. The NRC Resident Inspector was notified.
ENS 4850111 November 2012 06:08:00FitzPatrickAutomatic ScramNRC Region 1GE-4

An unplanned, automatic reactor scram occurred at 0355 EST due to a Main Turbine trip signal. All safety systems operated and actuated as expected. Both the Main Transformer, T-1A and normal station services transformer T-4 activated their respective deluge systems. On-site fire brigade and offsite fire assistance have successfully extinguished the T-1A transformer fire. There is still an active fire in the T-1A bus ductwork. The plant will be taken to cold shutdown conditions. At 0545 EST the plant entered the emergency plan at the NUE level due to inability to successfully extinguish the fire. All control rods fully inserted following the reactor scram. MSIVs remain open with decay heat being removed via steam to the main condenser using the bypass valves. All electrical buses are powered from their normal offsite reserve source. The licensee notified the NRC Resident and appropriate State and local government agencies. Notified DHS SWO, FEMA, DHS NICC and NuclearSSA via email.

  • * * UPDATE FROM JOHN WALKOWIAK TO DONALD NORWOOD AT 0642 EST ON 11/11/2012 * * *

As of 0639 EST the fire in the T-1A bus ductwork has been extinguished.

  • * * UPDATE FROM MARK HAWES TO DONALD NORWOOD AT 0747 EST ON 11/11/2012 * * *

Local fire department is on-site. No radiological release and no protective actions required. Plant cooldown in progress.

  • * * UPDATE FROM MARK HAWES TO DONALD NORWOOD AT 0810 EST ON 11/11/2012 * * *

The Unusual Event (HU 6.1) has been terminated at 0801 EST. Cooldown in progress to cold condition. Reactor level at 206 inches and pressure is at 530 pounds. The licensee notified the NRC Resident and appropriate State and local government agencies. Notified R1DO (Dentel), NRR EO (McGinty), IRD (Gott). Notified DHS SWO, FEMA, DHS NICC and NuclearSSA via email.

ENS 484969 November 2012 03:03:00SusquehannaManual ScramNRC Region 1GE-4At approximately 0118 hours (EST) on November 9, 2012, Susquehanna Steam Electric Station Unit Two reactor was scrammed by plant operators due to a loss of ICS (Integrated Control System; which controls the reactor feed and reactor recirculation systems). The reactor operator placed the mode switch in shutdown when reactor water level reached +25 inches and lowering. All control rods inserted and both reactor recirculation pumps tripped at -38 inches. Reactor water level lowered to -52 inches causing Level 3 (+13 inches) and level 2 (-38 inches) isolations. HPCI and RCIC both automatically initiated. HPCI was overridden prior to injection and RCIC was utilized to restore reactor water level to the normal band. All isolations and initiations at this level occurred as expected. No steam relief valves opened. Pressure was controlled via turbine bypass valve operation. All safety systems operated as expected. The (Unit 2) reactor is currently stable in Mode 3. An investigation into the cause of the loss of ICS is underway. Unit One continued power operation (at 78% power). The NRC Resident Inspectors were notified. A press release will occur. The licensee will inform the State of Pennsylvania. Decay heat removal is being maintained through the main condenser. On-site electrical power is in the normal configuration.
ENS 484877 November 2012 12:39:00FermiManual ScramNRC Region 3GE-4At 09:21 EST 11/7/12, the reactor mode switch was taken to shutdown and the main turbine generator was manually tripped in response to hydrogen gas in-leakage into the stator water cooling system from the main turbine generator. The scram was uncomplicated, and all control rods, except one, fully inserted into the core. One control rod stopped at position 02 and was manually inserted. The lowest reactor vessel water level reached was 125 inches, and as expected, HPCI & RCIC did not actuate. No safety relief valves (SRV) actuated. Reactor water level is being controlled in the normal band using the control rod drive and reactor feedwater systems. All isolations and actuations for reactor water level 3 occurred as expected. The cause of the increased hydrogen gas in-leakage into stator water cooling is under investigation. At the time of the manual scram, all Emergency Diesel Generators were operable. All Emergency Core Cooling Systems were available and no significant safety related equipment was out of service. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B), as an event that results in actuation of the reactor protection system (RPS) when the reactor is critical. The MSIVs are open with decay heat being removed via steam to the main condenser using the bypass valves. The plant is in a normal shutdown electrical lineup. The licensee has notified the NRC Resident Inspector.
ENS 484795 November 2012 00:40:00FitzPatrickAutomatic ScramNRC Region 1GE-4The reactor was scrammed on a valid reactor protection system activation caused by a main turbine trip. The cause of the main turbine trip is under investigation. All control rods fully inserted. All isolations and initiations occurred as designed. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiated as expected. RCIC injected into the reactor coolant system, HPCI did not, as expected. This scram was characterized as uncomplicated and the reactor is stable in Mode 3. The plant is in a normal post shutdown electrical lineup. All systems functioned as required. The NRC Resident Inspector has been notified.
ENS 4830914 September 2012 19:27:00FermiAutomatic ScramNRC Region 3GE-4At 1603 EDT, Fermi 2 automatically scrammed due to onsite loss of 120 kV switchyard. All control rods fully inserted. The lowest Reactor Water Level (RWL) reached was 98 inches. Division I diesels, EDG-11 and EDG-12, automatically started and loaded. HPCI and RCIC automatically started and restored RWL. RWL is currently being maintained in the normal level band with Condensate/Feed and Control Rod Drive (CRD) systems. No Safety Relief Valves (SRV) actuated. All isolations and actuations for RWL 3 and 2 occurred as expected. Investigation into loss of 120 kV switchyard continues. At the time of the scram, all Emergency Core Cooling (ECCS) and Emergency Diesel Generators (EDG) were operable with the exception of EDG-11 which was available vice operable due to ventilation work, and no other safety related equipment was out of service. This report is being made in accordance with 10CFR50.72(b)(2)(iv)(A), any event that results in ECCS discharge into the reactor coolant system as a result of a valid signal and 10CFR50.72(b)(2)(iv)(B), any event that results in actuation of the reactor protection system (RPS) when the reactor is critical. EDG-11 and EDG-12 are performing all of their functions and providing power to the Division I AC buses. Temperatures are being monitored in the room containing EDG-11 and the room is not approaching any temperature limits. The MSIVs are open with decay heat being removed via steam to the main condenser using the bypass valves. The licensee has notified the NRC Resident Inspector.
ENS 4811718 July 2012 09:33:00LimerickManual ScramNRC Region 1GE-4

An electrical transformer fault occurred resulting in a loss of both Recirc Pumps. The reactor was manually scrammed from 100% power as required by Plant Procedure OT-112. The electrical transformer was walked down by Operations supervisor. Licensee's assessment was that a flashover occurred, and was confined to the load center transformer cabinet. Based on observed damage, EAL declaration of HU3 was made. HU3 is identified as an explosion within the Protected Area. The licensee has notified the NRC Resident Inspector. Licensee also notified state, local and other government agencies. Notified other agencies (DHS SWO, FEMA, DHS NICC)

  • * * UPDATE FROM BRIAN DEVINE TO JOHN KNOKE AT 1022 EDT ON 07/18/12 * * *

Limerick, Unit 1 is terminating from their Unusual Event (HU3) due to the initiating event and conditions no longer being present. The 124A Fault was isolated by the trip of the designed protection features (feeder breaker trip). A walkdown of the area/equipment was completed with no adverse conditions noted. Normal plant shutdown activities are in progress. The area/equipment is quarantined for investigation. The licensee will be issuing a press release. Notified other agencies (DHS SWO, FEMA, DHS NICC) The licensee has notified the NRC Resident Inspector. Notified R1DO (Dimitriadis) and NRR EO (Davis)

ENS 4804725 June 2012 16:38:00FermiManual ScramNRC Region 3GE-4At 1330 EDT on June 25, 2012, while restoring the Main Turbine Generator (MTG) to service after repairs to Main Unit Transformer 2B (MUT2B), Main Control Room (MCR) staff manually initiated a reactor scram in response to trip of both Reactor Feed Pumps (RFP). All control rods fully inserted. The lowest Reactor Water Level (RWL) reached was 154 inches and, as expected, HPCI and RCIC did not actuate. RWL was restored to normal using the Standby Feedwater (SBFW) system. RWL is currently being maintained in the normal level band with SBFW and Control Rod Drive (CRD) systems. No Safety Relief Valves (SRV) actuated. All isolations and actuations for RWL 3 occurred as expected. Investigation into the trip of RFPs continues. At the time of the scram, all Emergency Core Cooling Systems (ECCS) and Emergency Diesel Generators (EDG) were operable and no safety related equipment was out of service. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an event that results in actuation of the reactor protection system (RPS) when the reactor is critical. The plant is in a normal shutdown electrical lineup with decay heat being removed via steam to the main condenser using the bypass valves. The licensee notified the NRC Resident Inspector.
ENS 4797229 May 2012 07:22:00Browns FerryAutomatic ScramNRC Region 2GE-4On 5/29/2012 at 0331 (CDT) the Unit 3 reactor scrammed due to turbine control valve fast closure initiated by a load reject signal on the Main Generator. The cause of the load reject signal is Main Transformer differential relay 387T. Reactor power at the time of the SCRAM was approximately 75%. All systems responded as expected to the load reject signal. Main Steam Isolation Valves remained open and reactor pressure is being controlled on the Main Turbine Bypass Valves. No Main Steam Relief Valves lifted during the transient. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation setpoints were reached. Primary Containment Isolation Signals (PCIS) Groups 2, 3, 6 and 8 were received. The lowest reactor water level observed was -41 inches. Reactor water level was restored to and is being controlled by the Feedwater system in the normal band. This event is reportable within 4 hours per 10CFR50.72(b)(2)(iv)(B), 'any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR50.73(a)(2)(iv)(A). This event is documented in the station corrective action program on SR# 557947. The NRC Resident Inspector has been notified. All control rods inserted into the reactor core. Electrical power is being back fed from offsite power through the 161 KV feeder line. The reactor is being cooled down within the Technical Specification rates.
ENS 4795524 May 2012 11:10:00Browns FerryManual ScramNRC Region 2GE-4At 0639 CDT on 5/24/2012. Unit 3 initiated a manual scram due to multiple rods inserting. At 0637 CDT during Unit 3 start-up Intermediate Range Monitor (IRM) 'H' was ranged down instead of up resulting in half scram on Reactor Protection System (RPS) 'B' trip system. The half scram was being reset after IRM 'H' was properly ranged. The operator placed the scram reset switch in Group 2/3 position. As the operator reset groups 2 and 3, a spike on IRM 'A' was received on the RPS 'A' trip system, resulting in rod insertion for groups 1 and 4. When the operator identified multiple rods inserting, the actions of procedure 3-AOI-l00-1 were followed and a manual scram was inserted. Investigation is ongoing. All safety systems remained in standby readiness configuration. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. Primary Containment lsolations Systems did not received actuation signals and performed as designed. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the RPS when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of systems listed in paragraph (b)(3)(iv)(B) 'Reactor Protection System (RPS) Including reactor scram and reactor trip.' This event requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.
ENS 4794222 May 2012 07:38:00Browns FerryAutomatic ScramNRC Region 2GE-4

At 0249 CDT on 5/22/2012, Unit 3 reactor automatically scrammed due to de-energization of Reactor Protection System (RPS) from actuation of 3A Unit Station Service Transformer (USST) differential relay 387SA, which resulted in a loss of 500KV power to Unit 3. This relay was picked up during a transfer of 4KV Unit Board 3C from alternate power (161KV) to normal power (3A USST). Investigation is in progress as to the cause of relay actuation. 500KV power was restored through the alternate feeder breakers from 161KV to all Unit 3 4KV Unit Boards successfully. 161KV remained available during the entire event. Loss of 500KV power lasted less than 30 seconds and power has been restored to all safety related boards. All Unit 3 diesel generators successfully started and tied to their respective 4KV Shutdown Boards.

All safety systems responded as expected to the loss of 500KV power. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. RCIC was manually started to control reactor water level. Primary Containment Isolation System (PCIS) and PCIS initiation signals for groups 1, 2, 3, 6 & 8 were received as designed. At the time of the scram, High Pressure Coolant Injection (HPCI) system was tagged out for removal of temporary instrumentation following planned maintenance. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation'. It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.

ENS 4785319 April 2012 20:58:00Browns FerryManual ScramNRC Region 2GE-4On 04/19/12 at 1430 while performing 1-SR-3.5.1.7, HPCI (High Pressure Coolant Injection) Main & Booster Pump Set developed head & flow rate at rated reactor pressure. The HPCI turbine failed to trip using the manual trip pushbutton. This manual trip pushbutton should have caused the 1-FCV-73-18, HPCI TURBINE STOP VALVE, to go closed. HPCI was secured by taking the 1-FCV-73-16, HPCI TURBINE STEAM SUPPLY VALVE, to close. The 1-FCV-73-18, HPCI TURBINE STOP VALVE, also failed to go closed locally using the 1-XCV-73-18, HPCI TURBINE MECHANICAL TRIP, nor did it go closed when the auxiliary oil pump was secured. With the 1-FCV-73-18, HPCI TURBINE STOP VALVE, open, the HPCI ramp generator is no longer in the circuit therefore, should an initiation occur and cause the 1-FCV-73-16, HPCI TURBINE STEAM SUPPLY VALVE, to open there is the potential for the HPCI turbine to over speed. Therefore, HPCI was isolated using 1-FCV-73-3, HPCI STEAM LINE OUTBD ISOL VALVE. This incident is reportable as an 8-hour ENS notification under 10CFR 50,72 (b)(3)(v) as '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.' It also requires a 60 day written report in accordance with 10CFR 50.73(a)(2)(vii). The NRC Resident Inspector has been notified.
ENS 4785019 April 2012 11:10:00LimerickManual ScramNRC Region 1GE-4

Limerick Unit 1 was manually scrammed from 100% power at 0753 hours on 4/19/12 in accordance with plant procedure OT-112 'Recirculation Pump Trip' when both 1A and 1B Recirculation Pump Adjustable Speed Drives (ASDs) tripped due to an electrical fault affecting the 144D and 114A non-safety related 480V Load Centers. The shutdown was normal and the plant is stable in Hot Shutdown with normal pressure control via the Main Steam Bypass valves to the main condenser and normal level control using feedwater. The manual RPS actuation is reportable under 10 CFR 50.72(b)(2). The Technical Support Center (TSC) Normal Air conditioning systems shut down due to loss of power from the 144D Load Center. The loss of power also affects the flow indication for the Emergency Ventilation system. This is considered a Loss of Emergency Assessment Capability, and reportable under 10 CFR 50.72(b)(3)(xiii). The Emergency TSC Ventilation system is available but flow cannot be verified. During a required activation the TSC, responders would report to the TSC. If conditions required use of the Emergency Ventilation system, the Station Emergency Director would assess habitability in accordance with Station procedures. TSC relocation of personnel would be directed as required until such time that the TSC ventilation system is returned to service The licensee notified the NRC Resident Inspector.

  • * * UPDATE AT 1726 ON 4/20/2012 FROM BRANDON SHULTZ TO MARK ABRAMOVITZ * * *

The Technical Support Center (TSC) 144D load center has been re-energized, restoring the emergency ventilation flow indication and emergency assessment capability to its normal stand-by condition." The switchgear was inspected for any potential grounds and then reenergized at approximately 0800 EDT on 4/20/2023. The licensee notified the NRC Resident Inspector. Notified the R1DO (Joustra).

ENS 4769023 February 2012 02:55:00BrunswickManual ScramNRC Region 2GE-4

At 2319 hours EST, a manual Reactor Protection System (RPS) actuation was inserted on Unit 1 in anticipation of a loss of condenser vacuum. Shortly before the manual RPS actuation, Circulating Water Intake Pump (CWIP) 1B tripped due to high delta-pressure across the intake traveling screen. This caused the trip of the remaining pumps. Previously, at 1859 hours, balance of plant (BOP) bus Common C unexpectedly de-energized. This caused loss of power to the CWIP traveling screen motors which, in turn, lead to the high delta-pressure across the traveling screen(s). All control rods inserted properly. As a result of the scram, reactor water level reached the Low Level 1 actuation set point and Primary Containment (i.e., Group 6) isolation occurred. All systems functioned as designed. The High Pressure Coolant Injection (HPCI) system is being used, as needed, for pressure control. The Reactor Core Isolation Cooling (RCIC) system is being used, as needed, for level control. No Safety/Relief Valves (SRVs) actuated as a result of the manual RPS actuation. The manual RPS actuation is reportable in accordance with 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). The actuation of the HPCI and RCIC systems and the Group 6 isolation are reportable in accordance with 10CFR50.72(b)(3)(iv)(A). The unit is currently in Mode 3 with a cooldown in progress. The licensee notified the NRC Resident Inspector. Notified R2DO (Ernstes).

  • * * UPDATE FROM STEWART BYRD TO CHARLES TEAL AT 0741 EST ON 2/23/12 * * *

At 2319 hours EST, a loss of all Circulating Water Intake Pumps caused a lowering vacuum on Unit 1. As previously reported (i.e. Event Notification 47690), a manual Reactor Protection System (RPS) actuation was inserted on Unit 1 at this time. In addition, a valid actuation of the RPS, High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC), and a Group 6 isolation was reported in accordance with 10CFR50.72(b)(3)(iv)(A). At 2342, Main Condenser vacuum was 15 in. Hg and lowering. All Main Steam Isolation Valves were slow closed in anticipation of Group 1 isolation at this time. This follow-up notification is being made to report the manual actuation of the Group 1 isolation valves in accordance with 10 CFR 50.72(b)(3)(iv)(A). The Group 1 isolation was discussed with the NRC during initial notification of EN 47690, and this follow-up is providing written notification of the MSIV closure. The NRC Resident Inspector has been notified. Notified R2DO (Ernstes).

ENS 476518 February 2012 19:31:00Browns FerryManual ScramNRC Region 2GE-4During BFNP NFPA 805 transition review, it was determined in the vent of an Appendix-R fire, the Reactor Protection System (RPS) function could be rendered not functional. The current Appendix R Safe Shutdown Analysis states: "The safe shutdown function of the Reactor Protection System (RPS) is to initiate reactor scram through actuation of the control rod drives. The RPS includes the RPS motor-generator power supplies and associated control and indicating devices, sensors, relays, bypass circuitry, and switches that initiate rapid insertion of control rods (scram) to shutdown the reactor. The RPS utilizes a fail-safe design so that device failures or a loss of power will result in control rod insertion. The scram function will remain available despite any fire-induced spurious signals that may be generated due to the effects of a postulated fire in any fire area. This system is expected to perform its function automatically, however credit is taken only for manual scram. No additional analysis is needed to ensure the availability of reactor scram in the even of a fire. Due to lack of physical separation with 120 volt AC lighting circuitry, the RPS system potentially could remain energized due to a postulated hot short circuit during a fire which could potentially prevent the control rods from inserting. Therefore, the fail-safe design of the RPS system would not be maintained. Compensatory actions in the form of fire watches to mitigate this condition are in place in accordance with the BFNP Fire Protection Report. This event is reportable as an 8-hour notification to the NRC in accordance with 10CFR50.73(a)(2)(ii)(B). The NRC Resident Inspector has been notified of this event. This event was entered into the licensee's Corrective Action Program as PER 503304.
ENS 4744416 November 2011 03:33:00BrunswickManual ScramNRC Region 2GE-4

On 11/16/11 at 0208 EST, Brunswick Nuclear Plant, Unit 2 calculated a drywall floor drain 42 minute leak rate of 5.88 gpm, following several hours of gradually rising floor drain leakage during a plant startup. Tech Spec 3.4.4 A was entered, requiring floor drain leakage to be restored below 5 gpm within 8 hours. At 0253 EST, a 45 minute leak rate of 10.11 gpm was calculated. At 0301 EST, Unusual Event SU 6.1 was declared for unidentified leakage exceeding 10 gpm, and at 0309 EST, a manual reactor scram was inserted from approximately 7% power (10 CFR 50.72(b)(2)(iv)(B)). Following the scram, the reactor was depressurized at a maximum cooldown rate of 92.5 deg F/hr, and the unidentified leak rate fell less than 10 gpm within 1 hour and less than 5 gpm within 2 hours. Leak rate at 0614 EST on 11/16/11 is 3.82 gpm with reactor pressure at 228 psig. The exact nature of the leak is unknown at this time. The current plan is to continue to depressurize and cool down the reactor to Mode 4, such that a full drywall inspection can commence. At present, Brunswick has not terminated the Unusual Event. Level control is currently being maintained with control rod drives (CRD). The MSIVs were manually closed to control cooldown. The maximum cooldown was observed to be 92.5 F/hour. The plant plans to reopen MISIVs and depressurized to condensate booster pump injection pressure of 350 psig. The plan is to achieve Mode 4 for a leak inspection. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM DAVID FASHCHER TO CHARLES TEAL AT 0550 EST ON 11/16/11 * * *

The leakage rate is currently 3.73 gpm. The decrease is due to lower pressure which is currently at 258 psig. There are no additional changes. The leakage source is not identified at this time. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).

  • * * UPDATE FROM DUSTIN LUPTON TO JOHN SHOEMAKER AT 0648 EST ON 11/16/11 * * *

The leakage rate is currently below the T.S. limit due to lower pressure which is currently at 210 psig. There are no additional changes. The plant will remain in an Unusual Event (UE) until further notice. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).

  • * * UPDATE FROM DUSTIN LUPTON TO CHARLES TEAL AT 0749 EST ON 11/16/11 * * *

The leakage rate is stable. The leak rate is calculated at 3.04 gpm at 183 psig at 0708 EST. The current reactor pressure is 162 psig. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).

  • * * UPDATE FROM DUSTIN LUPTON TO JOHN KNOKE AT 0832 EST ON 11/16/11 * * *

The licensee terminated from their Unusual Event at 0815 EST. The leakage is still unidentified. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).

ENS 4736924 October 2011 17:01:00HatchManual ScramNRC Region 2GE-4While performing a startup of HNP-2, after reaching criticality, the crew observed erratic indications on two Intermediate Range Monitors (IRMs), 2C51K601A and 2C51K601C. IRM 2C51K601A had been spiking and was subsequently bypassed. The 2C51K601C was spiking downscale and could not be bypassed due to the 2C51K601A being bypassed already. Both IRMs are in the 'A' RPS trip system. At the time when the second IRM was acting erratic, the crew identified the condition as both IRMs in the same quadrant and did not continue withdrawal of control rods. As a result of not withdrawing control rods, reactor power began to decrease and the crew conservatively inserted a manual reactor scram to shutdown the reactor. All rods did fully insert into the core. No PCIS (Primary Containment Isolation System) actuations occurred and none were expected to occur based on plant conditions following the scram. At this time, investigation is in progress, but the investigation and corrective actions have not yet been completed. The crew is maintaining HNP-2 in Hot Standby (Mode 3) at this time. The licensee informed the NRC Resident Inspector.