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{{#ask: [[Category:ENS Notification]] [[Site::Nine Mile Point]] [[Reporting criterion::10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown||10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation||10 CFR 50.72(b)(3)(iv)(A), System Actuation]]
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	Start date	Reporting criterion	Title	Event description	System	LER
ENS 57004	4 March 2024 00:42:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram Due to Main Turbine Trip on LOW Condenser Vacuum	The following information was provided by the licensee via email: On 3/3/24 at 1942 EST, while performing a plant shutdown in preparation for a refuel outage, Nine Mile Point Unit 2 experienced a reactor scram due to a main turbine trip on low condenser vacuum. The plant was at approximately 55 percent power at the time of the reactor scram. Additionally, following the scram a low RPV (reactor pressure vessel) level scram and containment isolation signal on level 3 was received, as expected. The containment isolation signal impacted RHR (residual heat removal) shutdown cooling, RHR letdown to radwaste, and RHR sampling. All impacted valves were closed at the time the isolation occurred. All control rods were fully inserted. Plant response was as expected. Post scram, the main turbine bypass valves are being used to control decay heat, and normal post scram level control is via the feed / condensate system. This is being report under 10 CFR 50.72(b)(2)(iv)(B), 'RPS Actuation', and 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'. Unit 1 is not affected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the low condenser vacuum was a momentary loss of sealing steam. The condenser remained viable for decay heat removal. All safety equipment is available. The grid is stable with the plant in its normal shutdown electrical configuration.	Shutdown Cooling Decay Heat Removal Control Rod
ENS 56710	2 September 2023 10:32:00	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)(2)(iv)(A), System Actuation - ECCS Discharge	Automatic Reactor Scram Due to Feedwater Transient	The following information was provided by the licensee via email: On 9/2/2023 at 0632 EDT, a feedwater transient occurred resulting in an reactor protection system (RPS) automatic reactor scram on low level (Level 3, 159.3 inches). Following the scram, reactor water level dropped below Level 2 (108.8 inches) resulting in a Group 2 recirculation sample system isolation, Group 3 traveling in-core probe (TIP) isolation valve isolation, Group 6 and 7 reactor water cleanup isolation, and Group 9 containment purge isolations. All control rods inserted as expected. High pressure core spray and reactor core isolation cooling initiated and injected as expected. ECCS systems have been secured and normal reactor pressure and level control has been established for hot shutdown. Nine Mile Point Unit 2 is stable and in Mode 3. These 4 hour and 8 hour non-emergency reports are being made in accordance with 10 CFR 50.72(b)(2) (iv)(A), 10 CFR 50.72(b)(2)(iv)(B), and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident was informed. There was no impact on Unit 1.	Feedwater Reactor Protection System Reactor Core Isolation Cooling High Pressure Core Spray Reactor Water Cleanup Control Rod
ENS 56089	4 September 2022 20:39:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Specified System Activation	The following information was provided by the licensee via fax: On September 4, 2022 with Unit 2 in Mode 3, an (Reactor Protection System) RPS actuation and Containment Isolation occurred on (Reactor Pressure Vessel) RPV Low Level (Level 3) of 159.3 inches due to issues with the normal feedwater level control system during plant cooldown. The RPS actuation occurred with control rods already inserted and a containment isolation on Level 3. The containment isolation signal impacted (Residual Heat Removal) RHR Shutdown Cooling, RHR letdown to radwaste, and RHR sampling. All impacted valves were closed at the time the isolation occurred. Operators took manual control of RPV level and restored level to the normal operating band shortly after the low level was received. This is being reported under 10CFR50.72(b)(3)(iv)(A) and 10CFR50.72(b)(3)(iv)(B). The NRC Resident Inspector was notified.	Feedwater Shutdown Cooling Control Rod
ENS 55821	5 April 2022 06:23:00	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)(2)(iv)(A), System Actuation - ECCS Discharge	Scram on LOW Level	The following information was provided by the licensee via telephone and email: On 4/5/2022, at time 0223, during maintenance on Feedwater Level Control Valve 2FWS-LV10B, a Feedwater transient occurred resulting in an RPS Automatic Reactor Scram on Low Level (Level 3, 159.3 inches). Following the scram, reactor water level dropped below Level 2 (108.8 inches) resulting in a Group 2 Recirculation Sample System Isolation, Group 3 TIP ((Traversing Incore Probe)) Isolation Valve Isolation, Group 6 and 7 Reactor Water Cleanup Isolation and Group 9 Containment Purge Isolations. All control rods inserted as expected. High Pressure Core Spray and Reactor Core Isolation Cooling initiated and injected as expected. ECCS Systems have been secured and normal reactor pressure and level control has been established for hot shutdown. Nine Mile Point Unit 2 is stable in Mode 3. These 4 hour and 8-hour non-emergency ENS ((Emergency Notification System)) reports are being made in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B), and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident was informed. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: There was no impact on Unit 1.	Feedwater Reactor Core Isolation Cooling High Pressure Core Spray Reactor Water Cleanup Control Rod
ENS 54562	4 March 2020 17:05:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Manual Scram in the Turbine Control System	At 1205 EST, on March 4, 2020, Nine Mile Point Unit 2 initiated a manual reactor scram due to lowering Electrohydraulic Control System (EHC) level in the turbine control system. The cause of the lowering level was a leak in the EHC system piping. All control rods inserted. There were no safety system actuations. The cause of the EHC leak is being investigated. The NRC Resident has been notified. Additionally, the licensee notified the New York State Public Service Commission.	Control Rod
ENS 54035	29 April 2019 20:33:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Reactor Scram Due to Power Oscillations and High Pressure Coolant Injection System Initiation	During power ascension on April 29, 2019, at 1630 (EDT), Nine Mile Point Unit 1 power and pressure oscillations were observed with reactor power at approximately 82 (percent). At time 1633 (EDT), the reactor was manually scrammed when the scram criteria of greater than 4 (percent) APRM power oscillations were observed in accordance with special operating procedures. All control rods fully inserted and all plant systems responded per design following the scram. Following the manual scram, the High Pressure Coolant Injection (HPCI) System automatically initiated as expected. At Nine Mile Point Unit 1, a HPCI system actuation signal on low Reactor Pressure Vessel (RPV) level is normally received following a reactor scram, due to level shrink. HPCI is a flow control mode of the normal feedwater systems, and is not an Emergency Core Cooling System. At 1633 (50 seconds after the reactor scram), RPV level was restored above the HPCI System low level actuation setpoint and the HPCI System initiation signal was reset. Pressure control was established on the Turbine Bypass Valves, the preferred system. No Electromatic Relief Valves actuated due to this scram. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor water level and pressure maintained within normal bands. The offsite grid is stable with no grid restrictions or warnings in effect. The cause of the power oscillations is currently under investigation. The NRC Resident Inspector was notified. The New York State public service commission was notified.	Feedwater High Pressure Coolant Injection Reactor Pressure Vessel Emergency Core Cooling System Control Rod
ENS 53998	14 April 2019 04:03:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram and Specified System Actuation	On April 14, 2019 at 0003 (EDT), Nine Mile Point Unit 1 experienced an automatic reactor scram during reactor startup. The cause of the automatic scram was due to high (Reactor Pressure Vessel) pressure following closure of the turbine stop valves. All control rods fully inserted and all plant systems responded per design following the scram. Following the automatic scram, the High Pressure Coolant Injection (HPCI) System automatically initiated as expected. At Nine Mile Point Unit 1, a HPCI System actuation signal on low Reactor Pressure Vessel (RPV) level is normally received following a reactor scram, due to level shrink. HPCI is a flow control mode of the normal feedwater systems, and is not an Emergency Core Cooling System. At 0004, RPV level was restored above the HPCI System low level actuation set point and the HPCI System initiation signal was reset. Pressure control was established on the Turbine Bypass Valves, the preferred system. No Electromatic Relief Valves actuated due to this scram. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor water level and pressure maintained within normal bands. The offsite grid is stable with no grid restrictions or warnings in effect. The unit is currently implementing post scram recovery procedures. The NRC Resident Inspector has been notified. The Licensee will notify the State of New York.	Feedwater High Pressure Coolant Injection Reactor Pressure Vessel Emergency Core Cooling System Control Rod
ENS 53565	27 August 2018 04:00:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	En Revision Imported Date 8/29/2018	EN Revision Text: AUTOMATIC SCRAM DUE TO A GENERATOR TRIP At 0033 EDT Nine Mile Point Unit 2 experienced an automatic scram on high reactor pressure due to a turbine trip. The cause of the turbine trip was due to a generator trip. All control rods inserted. There were no safety system actuations. The cause of the generator trip is being investigated. This is a 4-Hour report for 10CFR50.72(b)(2)(iv)(B) RPS Actuation. The NRC Resident Inspector has been notified. Decay heat is being removed via the main condenser. Reactor vessel water level is being maintained by the condensate and feedwater systems. The licensee will be notifying the state of New York. * * UPDATE FROM DANIEL CIFONELLI TO HOWIE CROUCH AT 1653 EDT ON 8/28/18 * * * After further review, the licensee has determined that the cause of automatic scram was due to turbine control valve fast closure as a result of the turbine trip, not high reactor pressure, as originally reported. The licensee has notified the NRC Resident Inspector. Notified R1DO (Lilliendahl).	Feedwater Main Condenser Control Rod
ENS 53395	10 May 2018 06:48:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Valid Emergency Diesel Start on Partial Loss of Offsite Power	At 0248 (EDT), with the plant shutdown in Mode 4, Nine Mile Point Unit 2 experienced a partial loss of off-site power during relay testing that resulted in an automatic start of the Division 2 Emergency Diesel Generator. All systems responded as expected for the event. The cause is being investigated. The station responded in accordance with appropriate Special Operating Procedures and restored impacted systems. This event is being reported in accordance with 10CFR50.72(b)(3)(iv)(A) At the time of the report, the emergency diesel generators are loaded and supplying plant safety equipment. The licensee has notified the state of New York Emergency Management Agency and the NRC Resident Inspector.	Emergency Diesel Generator
ENS 52950	6 September 2017 15:57:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram on Low Reactor Vessel Water Level	On September 6, 2017 at 1157 (EDT), Nine Mile Point Unit 1 experienced an automatic reactor scram with a Main Steam Isolation Valve (MSIV) and Containment isolation. The scram was due to reactor vessel low water level. The cause of the reactor vessel low water level is under investigation. All control rods fully inserted. Following the scram, pressure was momentarily controlled through the use of the Emergency Condenser (EC) system. At 1205, pressure control was established through the main steam lines to the condenser through the turbine bypass valves. All plant systems responded per design following the scram. The reactor scram is a 4-hour report per 10 CFR 50.72(b)(2)(iv)(B). The following systems automatically actuated after the scram as expected. These system actuations are an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A): 1. The High Pressure Coolant Injection (HPCI) system. HPCI initiated at 1157 and was reset at 1158 when RPV level was restored above the HPCI system low level actuation set point. HPCI is a flow control mode of the normal feedwater systems, and is not an Emergency Core Cooling System. 2. The Core Spray system actuated, but did not discharge to the Reactor Coolant system. The Core Spray system was secured at 1216. 3. Containment and MSIV isolation on reactor vessel low-low water level signal. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor vessel water level and pressure maintained within normal bands. Decay heat is being removed via steam to the main condenser using the turbine bypass valves. The offsite grid is stable with no grid restrictions or warnings in effect. The licensee has notified the NRC Resident Inspector. No safety relief valves lifted during the transient. The main steam isolation valves were opened after the isolation signal cleared to facilitate decay heat removal. Offsite power is supplying all plant loads. There was no effect on Unit 2. The licensee notified New York State Department of Environmental Protection and will be issuing a press release.	Reactor Coolant System Feedwater High Pressure Coolant Injection Main Steam Isolation Valve Core Spray Emergency Core Cooling System Decay Heat Removal Main Steam Line Safety Relief Valve Main Condenser Control Rod	05000220/LER-2017-003
ENS 52889	6 August 2017 02:35:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Automatic Reactor Scram During Testing	At 2235 (EDT) Nine Mile Point Unit 2 experienced an automatic scram on high reactor pressure. Turbine stop valve testing was in progress at the time of the scram. All control rods inserted. Pressure control is via the turbine bypass valves. The cause of the scram is being investigated. This is a 4-Hour report for 10CFR50.72(b)(2)(iv)(B) RPS (Reactor Protection System) Actuation. The NRC Resident Inspector has been notified. Reactor water level is being maintained with normal feedwater flow. No safety or relief valves lifted. The plant is in its normal shutdown electrical lineup.	Feedwater Control Rod
ENS 52624	20 March 2017 06:27:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Manual Scram Due to Pressure Oscillations	On March 20, 2017 at 0227, Nine Mile Point Unit1 was manually scrammed due to pressure oscillations. The Unit was offline and reactor shutdown was in-progress at the time of the scram. The scram was inserted at approximately 4% reactor power when pressure oscillations occurred exceeding the procedurally required limit for pressure oscillations. The cause of the scram was due to Operators manually inserting the scram. The cause of the pressure oscillations is being investigated. The licensee notified the NRC Resident Inspector. Notified the R1DO (Gray).		05000220/LER-2017-002
ENS 52425	10 December 2016 13:48:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Reactor Scram Due to High Main Turbine Vibrations	On December 10, 2016 at 0848 EST, (operators at) Nine Mile Point Unit 1 manually scrammed the reactor due to high vibrations on the Main Turbine. Cause of the high vibrations is being investigated. Following the scram, the High Pressure Coolant Injection (HPCI) system automatically initiated as expected. At Nine Mile Point Unit 1, a HPCI system actuation signal on low reactor pressure vessel (RPV) level is normally received following a reactor scram, due to level shrink. HPCI is a flow control mode of the normal feedwater system and is not an emergency core cooling system. At 0849, RPV level was restored above the HPCI system low level actuation set point and the HPCI system initiation signal was reset. Pressure control was established on the turbine bypass valves, the preferred system. No Electromatic relief valves actuated due to this scram. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor water level and pressure maintained within normal bands. Decay heat is being removed via steam to the main condenser using the bypass valves. The offsite grid is stable with no grid restrictions or warnings in effect. The unit is currently implementing post scram recovery procedures. The licensee has notified the state of New York Public Service Commission and the NRC Resident Inspector.	Feedwater High Pressure Coolant Injection Main Turbine Reactor Pressure Vessel Emergency Core Cooling System Main Condenser	05000220/LER-2017-001
ENS 51369	4 September 2015 13:16:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Unplanned Automatic Scram and Specified System Actuations Due to Msiv Closure	On September 4, 2015, at 0916 (EDT), Nine Mile Point Unit 1 experienced an automatic reactor scram following Main Steam Isolation Valve (MSIV) closure and isolation of both main steam lines. The cause of the MSIV closure is not known at this time. All control rods fully inserted. Following the scram, pressure was momentarily controlled through the use of the Emergency Condenser system. At 0950, pressure control was established through the main steam lines to the condenser through Main Steam Isolation Valves (MSIVs) 01-02 and 01-04. MSIV 01-03 would not reopen. All other plant systems responded per design following the scram. The reactor scram is a 4-hour report per 10 CFR 50.72(b)(2)(iv)(B)." The following systems automatically actuated after the scram as expected. These system actuations are an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A). 1. The High Pressure Coolant Injection (HPCI) system. HPCI initiated at 0916 and reset at 0917 when RPV level was restored above the HPCI system low level actuation set point. HPCI initiated and was reset a second time at 0922. HPCI is a flow control mode of the normal feedwater systems, and is not an Emergency Core Cooling System. 2. The Core Spray system actuated, but did not discharge to the Reactor Coolant system. The Core Spray system was secured at 1033. 3. The Emergency Condenser (EC) system actuated to control pressure. EC-11 was secured at 0917. EC-12 was secured at 0921 . The maximum shell temperature of EC system was 193 degrees Fahrenheit. 4. Containment Isolation actuation. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor water level and pressure maintained within normal bands. Since the scram, there have been no anomalies observed with feedwater system operation. Decay heat is being removed via steam to the main condenser using the turbine bypass valves. The offsite grid is stable with no grid restrictions or warnings in effect. The Reactor is being supplied by the normal feedwater system and there was indication of a partial lift/reset on one Electrometric Relief Valve (ERV). The licensee notified the NRC Resident Inspector and the State of New York Public Service Commission.	Reactor Coolant System Feedwater High Pressure Coolant Injection Main Steam Isolation Valve Core Spray Emergency Core Cooling System Main Steam Line Main Condenser Control Rod	05000220/LER-2015-004
ENS 50830	18 February 2015 19:06:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Manual Scram Due to Rapidly Rising Reactor Water Level	At 1406 (EST) Nine Mile Point Unit 2 inserted a manual scram due to rapidly rising reactor water level. The cause of the rapidly rising water prior to the manual scram is unknown and under investigation. Reactor water level is currently being maintained in normal control band post scram. The reactor scram is reportable in accordance with 10 CFR 50.72(b)(2)(iv)(B) as, 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical.' The NRC Resident has been notified. All rods fully inserted. Nine Mile Point Unit 2 is in a normal electrical shutdown configuration. The licensee reported no impact on Unit 1. The New York State Public Service Commission was also notified.	Reactor Protection System
ENS 49888	10 March 2014 20:28:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Reactor Scram Due to Actuation of the Alternate Rod Insertion System	At 1628 EDT Nine Mile Point (NMP) Unit 2 experienced an actuation of the Alternate Rod Insertion (ARI) system which resulted in a reactor scram. Coincident with the scram, the Reactor Core Isolation Cooling (RCIC) system initiated. Prior to the event, maintenance personnel were working in the vicinity of a reactor vessel level instrumentation rack and may have agitated the common drain line of the transmitters. A prompt investigation is underway to investigate the incident. The actuation signal for the RCIC system was invalid because reactor vessel level did not reach level two and the actuation was not in response to actual plant conditions or parameters. The reactor scram is reportable in accordance with 10 CFR 50.72(b)(2)(iv)(B) as, 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical.' The event has been entered into the NMP corrective action program as CR-2014-001963. The NRC Resident Inspector has been notified. The licensee has notified the State of New York. The reactor is shutdown with all rods inserted. Decay heat is being rejected to the condenser and reactor water level is being maintained by condensate, feedwater, reactor water clean up, and control rod drive systems.	Feedwater Reactor Protection System Reactor Core Isolation Cooling Control Rod
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.	Reactor Recirculation Pump Reactor Building Ventilation Standby Gas Treatment System Control Rod
ENS 49832	16 February 2014 17:16:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness	Unit 2 Experienced a Partial Loss of Offsite Power with Autostart of Division I and III Edgs	An automatic initiation of Emergency Diesel Generators was received due to a loss of one of two off-site power sources. On February 16, 2014 at 1216 (EST), Nine Mile Point Unit 2 experienced an automatic initiation of the Division 1 and 3 Emergency Diesel Generators (EDGs) due to a loss of Line 5. Line 5 is one of the two 115KV offsite power sources. Line 5 was lost due to an offsite event that is currently being evaluated by the grid operator. During the electrical transient, Nine Mile Point Unit 2 also experienced a feedwater level control lockup, requiring manual control. No Emergency Core Cooling Systems actuated, and feedwater level control was returned to automatic. Nine Mile Point Unit 2 remained at 100% power during the loss of Line 5, one of the offsite power sources. The NRC resident has been notified. The licensee also notified the State of New York. Unit 2 is currently in a 24-hour Technical Specification Limiting Condition of Operation (TS LCO) Action Statement 3.8.1.A3. Once Division III is realigned to the "B" side Line 6 offsite power, Unit 2 will be in a 72-hour LCO. Both EDGs are currently running. There was no effect on Unit 1. * * UPDATE PROVIDED BY MARK GREER TO JEFF ROTTON AT 1727 EST ON 2/17/2014 * * * In addition to the initiation of Division 1 and 3 Emergency Diesel Generators, Service Water Radiation Monitor 2SWPRE146A and Radwaste/Reactor Building Vent Gaseous Effluent Monitoring Systems were also lost. Compensatory actions were established as required by station procedures and the Offsite Dose Calculation Manual. The loss of these instruments is reportable as a major loss of Emergency Assessment Capability per 10CFR50.72(b)(3)(xiii). The Service Water Radiation Monitor 2SWPRE146A was restored to service on February 16, 2014 at 1450 EST. The Radwaste/Reactor Building Vent Gaseous Effluent Monitoring Systems were restored to service on February 17, 2014 at 0240. It is recognized that this notification is not within eight hours of the event. Line 5 is restored to OPERABLE February 17, 2014 at 1628. The NRC Resident Inspector has been informed of this additional reportable condition (and the update on the lost off-site power source). The condition has been entered into the corrective action program. The licensee also notified the State of New York Public Service Commission. The licensee exited the 72-hour LCO for the lost off-site power at 1628 on February 17, 2014. Notified R1DO(Krohn), NRREO (Monninger) and ILTAB (Whitney).	Feedwater Service water Emergency Diesel Generator Emergency Core Cooling System
ENS 49677	26 December 2013 18:01:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Technical Specification Required Shutdown Due to Blown Fuse	On December 26, 2013 at 1301 EST, Nine Mile Point Unit 1 (NMP1) commenced a reactor shutdown in accordance with Technical Specification (TS) 3.2.7.c as a result of Main Steam Isolation Valve (MSIV) 111 being declared inoperable due to a surveillance test failure when power was lost to the control circuit. The cause of the power loss was a blown fuse due to a short in the test circuitry. TS 3.2.7.c requires that a normal orderly shutdown be initiated within one hour and have the reactor in cold shutdown condition within ten hours. The blown fuse was replaced, test circuit isolated, power to the control circuit restored and MSIV 111 was declared operable at 1549 EST. Power ascension is in progress. Post maintenance operability test have been completed and the surveillance will be completed at a later time. The licensee has notified the NRC Resident Inspector and the State of New York.	Main Steam Isolation Valve
ENS 49593	2 December 2013 14:04:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Manual Scram Following Trip of Both Reactor Recirculation Pumps	At 0904 (EST) on Monday, December 2, 2013, Nine Mile Point Unit 2 was manually scrammed from approximately 40% thermal power due to the loss of both reactor recirculation pumps during a planned downpower evolution. Manual scram of the unit is procedurally required upon loss of both recirculation pumps to avoid potential power/flow oscillations. The reactor recirculation pumps failed to transfer to the low frequency motor generators when downshifted from fast speed. The cause of the loss of both reactor recirculation pumps is not known at this time. (Nine Mile Point Unit 2) NMP2 has commenced cooldown in preparation for the forced outage to investigate and commence repairs. 10 CFR 50.72(b)(2)(iv)(B) requires reporting within 4 hours of any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical. All control rods fully inserted. No safety systems actuated. Decay heat is being removed via the main condenser. The "A" recirculation pump was restarted in low speed at 1045 EST. Unit 2 is in a normal shutdown electrical lineup. The licensee informed the NRC Resident Inspector and will inform the New York State Public Services Commission.	Reactor Protection System Reactor Recirculation Pump Main Condenser Control Rod	05000410/LER-2013-004
ENS 48590	13 December 2012 21:30:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Technical Specification Required Shutdown Due to Primary Containment System Declared Inoperable	An increased usage of Nitrogen to maintain Primary Containment pressure within specification was noticed during steady state operation. Investigation into the extra Nitrogen usage revealed that Primary Containment Leakage was in excess of that allowed per Technical Specification 3.3.3.a. No action statement is provided for leakage in excess of Technical Specification 3.3.3.a; therefore in accordance with Technical Specification 3.0.1, the reactor shall be placed in an operational condition in which the specification is not applicable. This requires the plant to be shutdown and cooled down to less than 216 degrees F. Additionally, this is reportable as an event or condition that could have prevented fulfillment of a safety function of a system needed to control the release of radioactive material. The primary containment was declared inoperable at 1630 EST and a normal orderly plant shutdown was commenced at 1645 EST and will be less than 215 degrees F within 10 hours. Investigation of containment leakage is in progress. An update will be provided when the plant is in an operational condition in which Technical Specification 3.3.3.a is not applicable. The licensee has notified the NRC Resident Inspector. Licensee has notified the State of New York. * * UPDATE AT 0011 EST ON 12/14/12 FROM CHRISTOPHER GRAPES TO BILL HUFFMAN * * * As of 2333 EST on 12/13/2012, the reactor is below 215 degrees F, and containment is no longer required to be operable by Technical Specification 3.3.3. As part of the shutdown, a manual reactor scram was initiated as part of the pre-planned shutdown sequence and the High Pressure Coolant Injection (HPCI) System automatically initiated as expected. At Nine Mile Point Unit 1, a HPCI System actuation signal on low Reactor Pressure Vessel (RPV) level is normally received following a reactor scram, due to level shrink. HPCI is a flow control mode of the normal feedwater systems and is not an Emergency Core Cooling System. At 1913 EST, RPV level was restored above the HPCI system low level actuation set point and the HPCI system initiation signal was reset. Pressure control was established on the turbine bypass valves, the preferred system. No Electromatic Relief Valves actuated due to the scram. Nine Mile Point Unit 1 is currently in Cold Shutdown, with reactor water level and pressure maintained within normal bands. Decay heat is being removed via shutdown cooling (SDC). The offsite grid is stable with no grid restrictions or warnings in effect." The licensee has notified the NRC Resident Inspector and State authorities. Notified R1DO (Holody) and NRR EO (Lund).	Feedwater High Pressure Coolant Injection Primary containment Shutdown Cooling Reactor Pressure Vessel Emergency Core Cooling System	05000220/LER-2012-006
ENS 48481	6 November 2012 05:06:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	High Pressure Coolant Injection Actuation Signal	On Tuesday, November 06, 2012, at 00:06 EST, during the application of a tag-out associated with feedwater level control, the 12 feedwater flow control valve (FCV-29-137) unexpectedly partially opened. As a result, reactor vessel water level rose to the high level turbine trip set point causing the main turbine to trip. The turbine trip signal then resulted in the initiation of High Pressure Coolant Injection (HPCI) channels 11 and 12 logic. No actual system component starts or actuations occurred as a result of the logic initiation and no actual HPCI injection occurred due to the system configuration, nor was injection required. Actions were taken to manually isolate the 12 feedwater flow control valve and reactor vessel water level was restored to normal. This meets NRC 8-Hour reporting criteria per 10 CFR 50.72(b)(3)(iv)(A) due to a valid actuation of the High Pressure Coolant Injection System. The licensee has notified the NRC Resident Inspector. * * RETRACTION FROM JERRY HELKER TO CHARLES TEAL ON 12/17/12 AT 1543 EST * * * This notification is being made to retract Event Notification (EN) #48481, which reported an automatic actuation of the High Pressure Coolant Injection (HPCI) system initiation logic. The HPCI system is automatically initiated based on conditions representing a small break loss of coolant accident (LOCA). The initiation signals are: - Low reactor water level - This is a direct indication of a potential loss of adequate core cooling. - Turbine trip - During a LOCA within the drywell, high drywell pressure due to the line break will cause a reactor scram, which causes a turbine trip, which then by design initiates the HPCI system. The event occurred with the reactor in the cold shutdown condition, with the main turbine and main turbine shaft-driven feedwater pump (#13) out of service. In the cold shutdown condition, the probability of a LOCA is low and the HPCI system is not required by the Technical Specifications to be operable. Neither of the conditions requiring actuation of the safety function of the HPCI system (high drywell pressure or low reactor water level) was present. Although the turbine trip signal was in response to an actual sensed high reactor water level condition, high reactor water level is not a plant condition satisfying the requirement for actuation of the safety function of the HPCI system. With reactor vessel water level high, the safety function of the HPCI system (i.e. to provide adequate core cooling) was already completed. Thus, the HPCI initiation signal was invalid, and the event is not reportable under 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been informed. Notified the R1DO (Hunegs). * * UPDATE FROM JOHN APRIL TO VINCE KLCO ON 4/24/13 AT 0158 EDT * * * Upon further review, it has been determined the event did constitute a valid actuation of the HPCI system and is reportable per 10CFR50.72(b)(3)(4)(A). The licensee will notify the NRC Resident Inspector. Notified the R1DO (Joustra).	Feedwater High Pressure Coolant Injection Main Turbine	05000220/LER-2012-007
ENS 48477	3 November 2012 12:23:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Scram on Low Reactor Water Level	On November 3, 2012 at 0823 EDT, Nine Mile Point Unit 1 experienced an automatic reactor scram on low reactor water level. All control rods fully inserted and all plant systems responded per design following the scram. Prior to the automatic scram, an unexpected high Reactor Pressure Vessel (RPV) water level was experienced, followed by a turbine trip and subsequent lowering of RPV water level to the RPV low level scram set point. The cause of the water level transient is unknown. Following the automatic scram, the High Pressure Coolant Injection (HPCI) System automatically initiated as expected. HPCI is a flow control mode of the normal feedwater systems, and is not an Emergency Core Cooling System. At 0824 EDT, RPV level was restored above the HPCI System low level actuation set point and the HPCI System initiation signal was reset. Pressure control was established on the Turbine Bypass Valves, the preferred system. No Electromatic Relief Valves actuated due to this scram. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor water level and pressure maintained within normal bands. Since the scram, there have been no anomalies observed with feedwater system operation. Decay heat is being removed via steam to the main condenser using the bypass valves. The offsite grid is stable with no grid restrictions or warnings in effect. The unit is currently implementing post scram recovery procedures. The licensee has notified the NRC Resident Inspector. Unit 2 was not affected during this event.	Feedwater High Pressure Coolant Injection Reactor Pressure Vessel Emergency Core Cooling System Main Condenser Control Rod	05000220/LER-2012-005
ENS 48456	30 October 2012 01:00:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Initiation of Emergency Diesel Generator Due to Loss of One Off-Site Power Source	On October 29, 2012 at 2100 EDT, Nine Mile Point Unit 2 experienced an automatic initiation of the Division 1 Emergency Diesel Generator due to a loss of line 5. Line 5 is one of the 115KV offsite power sources. Line 5 was lost due to a lightening arrestor falling onto electrical components in the Scriba switchyard during high winds. During the electrical transient, Nine Mile Point Unit 2 also experienced a feedwater level control lockup, requiring manual control. No Emergency Core Cooling Systems actuated and feedwater level control was returned to automatic. Nine Mile Point Unit 2 remained at 100% power during the loss of line 5. The off-site power source, line 5, was restored on October 30, 2012 at 0326 EDT. The licensee notified the NRC Resident Inspector.	Feedwater Emergency Diesel Generator Emergency Core Cooling System	05000410/LER-2012-005
ENS 48453	29 October 2012 04:00:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Rps Actuation While Critical Due to Generator Load Reject	On October 29, 2012 at 2100 EDT, Nine Mile Point Unit 1 experienced an automatic reactor scram due to a generator load reject. The cause of the load reject is currently under investigation. All control rods fully inserted and all plant systems responded per design following the scram. Following the automatic scram, the High Pressure Coolant Injection (HPCI) System automatically initiated as expected. At Nine Mile Point Unit 1, a HPCI System actuation signal on low Reactor Pressure Vessel (RPV) level is normally received following a reactor scram, due to level shrink. HPCI is a flow control mode of the normal feedwater systems, and is not an Emergency Core Cooling System. At 2101 EDT, RPV level was restored above the HPCI System low level actuation set point and the HPCI System initiation signal was reset. Pressure control was established on the Turbine Bypass Valves, the preferred system. Three Electromatic Relief Valves actuated due to this scram and re-closed automatically. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor water level and pressure maintained within normal bands. Decay heat is being removed via steam to the main condenser using bypass valves. Both Reserve Station Transformers are in service and being supplied by their normal power sources. Both Emergency Diesel Generators are operable and in standby. The unit is currently implementing post scram recovery procedures. The licensee has notified the NRC Resident Inspector. Notified R1DO (Caruso).	Feedwater High Pressure Coolant Injection Emergency Diesel Generator Reactor Pressure Vessel Emergency Core Cooling System Main Condenser Control Rod
ENS 48323	20 September 2012 13:23:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Scram and High Pressure Coolant Injection System Initiation	On September 20, 2012 at 0923 EDT, Nine Mile Point Unit 1 experienced an automatic reactor scram due to a turbine trip at power. The cause of the turbine trip is currently under investigation. All control rods fully inserted and all plant systems responded per design following the scram. Following the automatic scram, the High Pressure Coolant Injection (HPCI) System automatically initiated as expected. At Nine Mile Point Unit 1, a HPCI System actuation signal on low Reactor Pressure Vessel (RPV) level is normally received following a reactor scram, due to level shrink. HPCI is a flow control mode of the normal feedwater systems, and is not an Emergency Core Cooling System. At 0924 EDT, RPV level was restored above the HPCI System low level actuation set point and the HPCI System initiation signal was reset. Pressure control was established on the Turbine Bypass Valves, the preferred system. No Electromatic Relief Valves actuated due to this scram. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor water level and pressure maintained within normal bands. Decay heat is being removed via steam to the main condenser using the bypass valves. The offsite grid is stable with no grid restrictions or warnings in effect. One 115kv off site power source (Line 4) is unavailable for planned maintenance at the James A Fitzpatrick Nuclear Power Plant. Both Reserve Station Transformers are in service and being supplied by the other 115kv offsite power source (Line 1). Both Emergency Diesel Generators are operable and in standby. The unit is currently implementing post scram recovery procedures. The licensee has notified the NRC Resident Inspector. Licensee has notified the state.	Feedwater High Pressure Coolant Injection Emergency Diesel Generator Reactor Pressure Vessel Emergency Core Cooling System Main Condenser Control Rod	05000220/LER-2012-002
ENS 48110	17 July 2012 15:18:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram Due to High Neutron Flux	On July 17 2012, at 1118 (EDT), Nine Mile Point Unit 1 experienced an automatic reactor scram due to high neutron flux as measured by the Average Power Range Monitoring system. The cause is currently under investigation. All control rods fully inserted and all plant systems responded per design following the scram. The High Pressure Coolant Injection (HPCI) system, which is an operating mode of the feedwater system, initiated as expected. There were no planned activities in progress at the time which could have contributed to the event. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor water level and pressure maintained within normal bands. Decay heat is being removed via steam to the main condenser using the bypass valves. The offsite grid is stable with no grid restrictions or warnings in effect. All emergency power sources are available in standby. The unit is currently implementing post scram recovery procedures. This was an uncomplicated Reactor Scram and the plant is stable with normal levels, temperature, and pressure. Offsite power is in a normal alignment. HPCI has been secured and returned to normal alignment. The cause to the high neutron flux is still under investigation. This event had no impact on Unit 2. The licensee has informed the NRC Resident Inspector and State Authorities.	Feedwater High Pressure Coolant Injection Main Condenser Control Rod	05000220/LER-2012-001
ENS 48097	12 July 2012 06:20:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Manual Reactor Scram Due to Lowering Condenser Vacuum	On July 12, 2012, at 0200 EDT, clean steam reboiler 'B' failed, causing a loss of sealing steam. This resulted in degrading condenser vacuum and rising off gas system pressure. The main steam backup supply for sealing steam also failed, and condenser vacuum and off gas system pressure continued to degrade. In response to rising off gas system pressure and lowering condenser vacuum, reactor power was lowered to 85% in accordance with Special Operating Procedures. With off gas system pressure approaching the procedural limit and condenser vacuum degrading rapidly, a manual reactor scram was inserted at 0220 EDT. All control rods fully inserted and all systems functioned as expected on the scram. Plant is currently shutdown and parameters are stable. The unit is currently implementing post scram recovery procedures and plant cooldown is in progress. The cause of the loss of both the primary and backup sources of sealing steam is under investigation. The shutdown electrical lineup is normal and decay heat is being removed via steam bypass valves to the main condenser. The licensee has notified the NRC Resident Inspector and the Public Service Commission.	Main Condenser Control Rod Main Steam	05000410/LER-2012-004
ENS 47514	9 December 2011 15:46:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Unidentified Rcs Leakage Greater than 2 Gpm Increase within 24 Hours	At 0908 (EST) Nine Mile Point Unit 2 received alert alarms for containment monitoring particulate channels. These alarms were accompanied by a rise in drywell floor drain leakage and drywell pressure. At 0915 (EST) Nine Mile Point Unit 2 entered technical specification action statement (TS 3.4.5 B) due to unidentified leakage rise of greater than 2 gpm within 24 hours. At 1046 (EST) the control room commenced power reduction for shutdown of the plant. Peak drywell pressure was approximately 0.16 psig and is lowering since the commencement of plant shutdown. Peak drywell floor drain leakage was approximately 3.7 gpm and is also lowering since the commencement of the plant shutdown. The cause of the rise in unidentified drywell leakage is unknown at this time. The licensee has notified the NRC Resident Inspector and the state.
ENS 47141	11 August 2011 04:16:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Manual Reactor Scram Inserted Due to a Feedwater Leak on a Min Flow Line	NMP2 inserted a manual reactor scram due to a through wall leak on feedwater pump 2FWS-P1A minimum flow line. 2FWS-P1A has been removed from service to minimize the leak. All control rods inserted and all systems functioned as designed. The unit will remain in hot shutdown until plant startup. All systems functioned as required and the scram was uncomplicated. The licensee plans to issue a press release and has informed the NRC Resident Inspector and the New York State Public Service Commission. * * UPDATE AT 1035 EDT ON 8/11/11 FROM MANLEY TO HUFFMAN * * * The licensee will not be issuing a press release regarding this manual scram. R1DO (Powell) notified.	Feedwater Control Rod
ENS 47129	6 August 2011 07:22:00	10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Unusual Event and Plant Shutdown Due to Drywell Unidentified Leakage Greater than 10 Gpm	At 0152 (EDT), Nine Mile Point Unit 2 received red alarms for containment monitoring cabinets 10A and 10B along with a rise in drywell floor drain leakage and drywell pressure. At 0205 (EDT), Nine Mile Point Unit 2 entered technical specification action statement (TS 3.4.5 B) due to unidentified leakage rise of greater than 2 gpm within 24 hours. At 0217 EDT, the control room commenced power reduction to mitigate the rise in drywell leakage and drywell pressure. At 0322 EDT, Nine Mile Point Unit 2 declared an Unusual Event (EAL 2.1.1) due to unidentified drywell leakage greater than 10 gpm. Actual drywell floor drain leakage reached 11.3 gpm. The cause of the rise in unidentified leakage is unknown at this time. At 0345 EDT, Nine Mile Point Unit 2 has commenced a plant shutdown. The licensee has notified State and local authorities. The NRC Resident Inspector has been notified. * * UPDATE ON 08/06/2011 AT 1140 EDT FROM JULIAN THOMPSON TO DAN LIVERMORE * * * The licensee terminated Notice of Unusual Event at 1127 EDT because leakage rates, drywell level, drywell pressure, rad levels, and all other parameters on the licensee checklist were normal and stable. The licensee is continuing to power down in order to perform inspections in the drywell area. The licensee notified State and local authorities and the NRC Resident Inspector. Notified: IRD MOC (Gott), R1RA (Dean), R1DO (Bellamy), NRR (Leeds), NRR EO (Cheok), DHS (Inzer) and FEMA (O'Connel) .
ENS 46808	3 May 2011 00:51:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram	Nine Mile Point Unit 1 experienced an automatic RPS actuation (scram) on Monday, May 02, 2011 at 2051. Cause of the RPS actuation is not understood at this time. All control rods inserted to position 00. N1-EOP-2 entered due to low reactor water level following trip (expected plant response). Reactor pressure stabilized 800 to 1000 psig, reactor water level restored and maintained 53 to 95 inches. Automatic initiation of the High Pressure Coolant Injection system occurred following plant trip (normal actuation). The reactor is stable in Hot Shutdown. The licensee characterized the scram as uncomplicated. All rods fully inserted, normal feed water is supplying the reactor with decay heat being removed to the condenser. The unit is in a normal shutdown electrical lineup and the scram had no impact on Unit 2. All systems functioned as required. The unit was operating at 47% power at the time of the event due to feed pump maintenance. The licensee does not have a first-out indication for the reactor scram and is still investigating the cause of the scram. The NRC Resident Inspector has been notified.	High Pressure Coolant Injection Control Rod
ENS 46409	10 November 2010 15:56:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Unplanned Automatic Scram and High Pressure Coolant Injection System Initiation During Testing	At 10:56 (EST) on Wednesday, November 10, 2010, Nine Mile Point Unit (1) One automatically scrammed from rated power. The cause of the scram was Main Steam Isolation Valve (MSIV) closure. The MSIV closure occurred during Instrument and Control Lo-Lo Level Surveillance Testing. The Lo-Lo Surveillance Test has been secured and all Reactor Protection System (RPS) Level Signals returned to normal. Two of four MSIVs went closed; troubleshooting to follow to determine the cause of the equipment malfunction (unexpected MSIV closure). Following the automatic scram, the High Pressure Coolant Injection (HPCI) System automatically initiated. At Nine Mile Point Unit One, a HPCI System actuation signal on low Reactor Pressure Vessel (RPV) level is normally received following a reactor scram, due to level shrink. At 10:58, RPV level was restored above the HPCI System low level actuation set point and the HPCI System initiation signal was reset. Pressure control was initially established on the Emergency Condensers (ECS). The MSIVs have been re-opened and the normal heat removal capability has been re-established (to the Main Condensers). All off-site power sources remain available (with a normal electrical alignment). 10 CFR 50.72(b)(2)(iv)(B) requires reporting within 4 hours when a valid actuation of the Reactor Protection System occurs. 10 CFR 50.72(b)(3)(iv)(A) requires reporting within 8 hours when a valid actuation of the High Pressure Coolant Injection System occurs. There are no other adverse impacts to the station based on this event. All control rods inserted and the unit is stable in Mode 3 with reactor pressure and temperature approximately 600 psig and 480 degrees. All other safety systems operated as expected. The licensee notified the NRC Resident Inspector and the New York State Public Service Commission.	High Pressure Coolant Injection Reactor Protection System Main Steam Isolation Valve Reactor Pressure Vessel Main Condenser Control Rod
ENS 46144	3 August 2010 18:52:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Unit 1 Entered a Tech Spec Required Shutdown After Declaring Edg 102 Inoperable	Nine Mile Point Unit 1 entered a Technical Specification (TS) Required Shutdown due to unplanned inoperability of Emergency Diesel Generator (EDG) 102 during planned maintenance on opposite division Liquid Poison pump 12. In accordance with TS 3.0.1, Limiting Condition for Operation (LCO) Applicability and TS 3.1.2, Liquid Poison System, normal orderly shutdown commenced at 1551 (EDT) which was within one hour of declaring Emergency Diesel Generator 102 inoperable. The unplanned inoperability of EDG 102 was caused by the failure of its lube oil circulating pump due to thermal overload. The licensee is pursuing restoration of EDG 102 and the Liquid Poison pump 12 in order to exit the LCO Action Statement. The licensee informed the NRC Resident Inspector. * * UPDATE FROM ROY GREEN TO VINCE KLCO ON 8/3/10 @ 2118 EDT * * * The Nine Mile Point Unit 1 TS required shutdown was suspended at 1929 (EDT) upon restoration of Emergency Diesel Generator (EDG) 102 to operable. With operability of EDG 102 restored, Liquid Poison Pump 11 was also made operable. Accordingly, the TS shutdown action for an inoperable liquid poison system was exited. The lowest power level reached was 92.5% of the licensed thermal power limit. Power ascension is in progress, and (the licensee expects) to be at rated power by 2200 (EDT). The licensee notified the NRC Resident Inspector. Notified the R1DO (DeFrancisco)	Emergency Diesel Generator
ENS 45612	7 January 2010 06:00:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Automatic Reactor Scram During Transmitter Venting	At 0100, Nine Mile Point Unit 2 reactor scrammed (on low reactor water level). The cause is currently under investigation. Maintenance activities were in progress on Residual Heat Removal (RHR) B & C systems. It is suspected that transmitter venting is the most probable cause of the trip. Reactor core isolation system (RCIC) injected on invalid level 2 signal. No safety relief valves (SRV's) actuated. B and C Residual Heat Removal systems (RHR) remain inoperable. All other ECCS systems and Emergency Diesel Generator are currently operable. Reactor water level is normal. No emergency core cooling system injection was required. The plant is currently in the hot shutdown condition. No emergency plan activation is required. Activation of forced outage plan is in progress. Lowest reactor water level reached during the transient was 140 inches. This is less than the reactor water level scram set point of 159.3 inches. Reactor pressure before scram was 1020 psig, current reactor pressure is 530 psig and reactor coolant temperature is 470 degrees F. Plant stack monitor is not available. Last grab sample reading at time 0257 is 857 micro-curies/sec. All systems functioned as required. All rods fully inserted on trip. Decay heat is being removed by steaming to the main condenser. The plant is in its normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector.	Emergency Diesel Generator Residual Heat Removal Main Condenser
ENS 45412	5 October 2009 15:58:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Scram and High Pressure Coolant Injection Following a Loss of Feedwater Level Control	At 1158 on Monday, October 5, 2009, Nine Mile Point Unit One was manually scrammed from approximately 100% rated power due to failure of the Feedwater Level Control System, in anticipation of a reactor scram. Following the manual scram insertion at 11:58, High Pressure Coolant Injection (HPCI) System automatically initiated on low Reactor Vessel (RPV) level. At 11:59, RPV level was restored above the HPCI System low level actuation setpoint, and the HPCI System initiation signal was reset. At Nine Mile Point Unit One, a HPCI System actuation signal on low RPV level is normally received following a reactor scram, due to level shrink. Unit 1 has commenced cooldown, in preparation for the forced outage to commence repairs. Reactor water level is being controlled in the normal operating band; reactor temperature is 450?F and reactor pressure is approximately 500 psig. 10 CFR 50.72(b)(2)(iv)(B) requires reporting within 4 hours when any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical. 10 CFR 50.72(b)(3)(iv)(A) requires reporting within 8 hours when a valid actuation of the High Pressure Coolant Injection System occurs. All control rods fully inserted. All systems functioned as required following the reactor scram. The reactor is currently stable in Mode 3. HPCI has been secured. Makeup water is being provided by the Reactor Feedwater System and decay heat removal is through the bypass valves to the condenser. There was no impact on Unit 2 and the plant is in a normal post-scram electrical lineup. The licensee notified the NRC Resident Inspector.	Feedwater High Pressure Coolant Injection Reactor Protection System Decay Heat Removal Control Rod
ENS 44982	10 April 2009 09:38:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	High Pressure Coolant Injection Automatic Initiation Due to Turbine Trip	At 0538 on Friday, April 10, 2009, Nine Mile Point Unit One manually tripped the turbine at approximately 28% rated power in response to rising turbine bearing vibrations. Following the manual turbine trip, the High Pressure Coolant Injection (HPCI) system automatically initiated on the turbine trip signal. The HPCI system initiation signal was reset immediately following the turbine trip. Reactor water level was maintained in the normal operating range throughout the transient. At Nine Mile Point Unit One the HPCI system is a mode of operation of the feedwater and condensate system. It is not an Emergency Core Cooling System (ECCS). A HPCI system actuation signal on a turbine trip is an expected response of the HPCI system. Nine Mile Point Unit 1 remains at power since the turbine was tripped at a power level that is within the capacity of the turbine bypass valves. 10 CFR 50.72(b)(3)(iv)(A) requires reporting within 8 hours when a valid actuation of the feedwaler coolant injection system occurs. The event has been entered into the corrective action program. There are no other adverse impacts to the station based on this event. The NRC Resident Inspector was notified.	Feedwater High Pressure Coolant Injection Emergency Core Cooling System
ENS 44923	21 March 2009 05:21:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	High Pressure Coolant Injection System Initiation Following Planned Manual Scram	At 0121 on Saturday, March 21, 2009, Nine Mile Point Unit One was manually scrammed from approximately 18% rated power in preparation for Refueling Outage 20. The plant scram was a planned evolution. Following the manual scram insertion at 0121, the High Pressure Coolant Injection (HPCI) System automatically initiated on low Reactor Pressure Vessel (RPV) level. At 0123, RPV level was restored above the HPCI System low level actuation setpoint, and the HPCI System initiation signal was reset. At Nine Mile Point Unit One the HPCI system is a mode of operation of the feedwater and condensate system. It is not an Emergency Core Cooling System (ECCS). A HPCI System actuation signal on low RPV level is normally received following a reactor scram, due to level shrink. 10 CFR 50.72(b)(3)(iv)(A) requires reporting within 8 hours when a valid actuation of the feedwater coolant injection system occurs. The event has been entered into the corrective action program. There are no other adverse impacts to the station based on this event. All control rods fully inserted. Electrical lineup is normal and there are no safety equipment out of service. The licensee has notified the NRC Resident Inspector * * UPDATE FROM BRIAN FINCH TO JOE O'HARA AT 1611 ON 5/15/09 * * * Event notification 44923 is being retracted after further evaluation because it has been determined that the High Pressure Coolant Injection (HPCI) system initiation was part of a pre-planned sequence. 10 CFR 50.72(b)(3) requires that any event or condition that results in valid actuation of any of the systems listed in this section, including a feedwater injection system, must be reported within 8 hours. Except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. The HPCI actuation that occurred after the manual scram from 18% Rated Thermal Power (RTP) was part of a pre-planned sequence in accordance with an operating procedure on plant shutdown. The HPCI initiation was also discussed during Just-in-Time training for the operating crew prior to performance of the soft shutdown. At Nine Mile Point Unit 1, the HPCI system is a mode of operation of the feedwater and condensate systems. The HPCI system is not an Emergency Core Cooling System (ECCS). A HPCI system actuation signal on low reactor water level is normally received following a reactor scram, due to reactor water level shrink. The licensee will notify the NRC Resident Inspector. Notified R1DO (Holody).	Feedwater High Pressure Coolant Injection Reactor Pressure Vessel Emergency Core Cooling System Control Rod
ENS 44635	7 November 2008 01:27:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Tech Spec Required Shutdown Required Due to Failed Surveillance Test	Nine Mile Point (NMP) Unit 1 initiated a Technical Specification (TS) shutdown at 2027. During a normal Instrumentation Surveillance Test, a relay for Emergency Condenser (ECCS) Initiation failed to meet its acceptance criteria (at 1245). This placed the plant in an action statement that required placing a channel trip in for the affected relay channel in 24 hours. To perform the relay replacement NMP Unit 1 entered 1 hour TS shutdown Limiting Condition of Operation at 1928. At 2027 Shutdown of the plant was initiated by lowering reactor recirculation flow and thus reactor power. At 2031, relay replacement was completed, (at) 2038 plant returned to full power operations. The plant remained in 24 hour action statement until post maintenance testing was completed at 2124. The licensee notified the NRC Resident Inspector.
ENS 44598	24 October 2008 01:26:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Reactor Scram Due to Loss of the Electronic Pressure Regulator	Control Room Operators observed slight (reactor) pressure rise during panel walk down. Investigation of pressure indication led Control room staff to determine that (the) EPR (Electronic Pressure Regulator) was not functioning properly (noise in the output signal). Control Room Staff entered Special Operating Procedure for failed pressure regulator. EPR could not be moved and this was confirmed by operators in the field. Control Room Staff (then) inserted a manual scram. Immediately after the scram reactor water level reached a low of 36", Emergency Operating Procedures for Level (EOP-2) were entered. HPCI initiated on the turbine trip to control water level. After the turbine tripped, all turbine bypass valves failed open; MSIVs (main steam isolation valve) were manually shut to control pressure. (The) EPR eventually disengaged from control, allowing the operator control of the turbine bypass valves. MSIVs were then reopened. (The) Scram has been reset. (The) turbine driven shaft pump did not initially disengage, pump (was) manually tripped after turbine speed reduced to 1500 rpm. All other systems responded correctly. (The) plant is not currently in any SOPs or EOPs and is proceeding to cold shutdown using normal operating procedures. All control rods fully inserted as expected. The plant is in a normal shutdown electrical lineup. At the time of the event, containment spray loop 1-12 was out of service for routine surveillance. The plant is currently cooling down and is at 365 psi. The licensee notified the NRC Resident Inspector.	Main Steam Isolation Valve Control Rod Containment Spray
ENS 44206	13 May 2008 12:25:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Unusual Event Declared Due to Loss of Offsite Power	At 0806 EDT, a loss of offsite power occurred when the 115 KV line #4 was lost combined with the unavailability of 115 KV line #1 which was out of service at the time for planned maintenance. Both EDGs started and loaded on the safety busses as expected. The licensee declared an Unusual Event at 0825 EDT per EAL 6.1.1 based on loss of offsite power for greater than 15 minutes. The Unit 1 recirc pump #13 tripped as a result of the event and spent fuel pool cooling was also lost. The licensee remained at power and there was no impact on Nine Mile Unit 2. Power for 115 KV line #1 was restored at 0845 EDT. Cause of the loss of Line #4 is still under investigation. The reactor is stable and the safety busses continue to be power by the EDGs. Restoration of a normal electrical power lineup and termination of the UE is still under review. The licensee notified the NRC Resident Inspector, the State, and local authorities. * * UPDATE AT 10:29 EDT ON 5/13/08 FROM HELKER TO HUFFMAN * * * The licensee terminated the Unusual Event at 10:22 EDT based on restoration of offsite power to safety bus 103 from line #1. Bus 102 still remains energized by the EDG at this time. Investigation into the loss of line #4 continues. The loss of line #4 has been attributed to a Fitzpatrick protective relay 87B actuation for the B phase differential current. Investigation continues. The NRC Resident has been notified by the licensee. NRR EO (Thorp), Rg 1 (Dentel), IRD (McDermott), R1DO (Krohn), DHS (Dwight) and FEMA (McKentry) notified. * * UPDATE AT 12:53 EDT ON 5/13/08 FROM HELKER TO SNYDER * * * This condition is also reportable under 10 CFR 50.72(b)(3)(v)(D) 'Any event or condition that at the time of discovery could have prevented fulfillment of a safety function of systems that are needed to mitigate the consequences of an accident.' Specifically, both offsite power 115 KV lines were not available. The second paragraph which applies is 10 CFR 50.72(b)(3)(iv)(B) 'Any event or condition that results in the valid actuation of the systems listed in paragraph (b)(3)(iv)(B).' Specifically, the auto start of the Emergency Diesel Generators on a valid loss of offsite power signal. The cause of the loss of line 4 is still being investigated. Offsite power has been restored to onsite buses including PB 102, 103 and 101. The licensee will notify the NRC Resident Inspector of this update. Notified R1DO (Krohn).	Emergency Diesel Generator
ENS 43223	8 March 2007 19:50:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Technical Specification Shutdown Due to a Degraded Seal in the "a" Reactor Recirculation Pump	Reactor Recirculation Pump 'A' removed from service due to a degraded seal. TS 3.4.1 entered to meet requirement for single loop operation at 1050. When the pump was secured, the 3D monicore system failed preventing verification of thermal limits. TS 3.4.1 action D entered at 1450 required to be in mode 3 (Hot Shutdown) within 12 hours. The NRC Resident Inspector was notified of this event by the licensee. Reactor Recirculation Pump 'A' seal is degraded, no increase in leakage. Licensee is in contact with General Electric in trying to restore the 3D monicore system. All Emergency Core Cooling Systems and the Emergency Diesel Generators (all 3 divisions) are fully operable if needed and the electrical grid is stable. Nine Mile Point Unit 1 is in coastdown for a scheduled refueling outage. Nine Mile Point Unit 2 should be in Mode 3 (Hot Shutdown) by 0100 hours on 03/09/07. * * UPDATE AT 1241 EST ON 03/09/07 FROM CHRIS SKINNER TO MACKINNON * * * On March 8, 2007 at 1709, surveillance for thermal limit verification was completed satisfactorily and TS 3.4.1 was exited. The reactor was no longer required to be shutdown per Technical Specifications; however, the plant shutdown continued for performance of maintenance activities. R1DO (Neil Perry) notified. The NRC Resident Inspector was notified of this update by the licensee.	Emergency Diesel Generator Reactor Recirculation Pump Emergency Core Cooling System
ENS 42633	12 June 2006 17:02:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Hpci Inititation Signal Due to Turbine Trip Signal During Startup	Feedwater was being place into long path recirculation mode in accordance with N1-OP-16 to support a chemistry sample. During this evolution, reactor water level rose high enough such that the High Level Annunciator (92.2") was received, the turbine tripped (off of the turning gear) and a HPCI (High Pressure Coolant Injection) initiation signal was received as per design. HPCI system did not initiate flow and no pumps started because the Feedwater Booster Pumps were not in service (pull-to-lock). HPCI was reset and Reactor Water level (was) restored to the operating band 65" - 83". A copy of this Notification Worksheet will be provided to the Resident Inspector. This event has been captured on Condition Report 2006-2703. A prompt investigation has been performed. The licensee notified the NRC Resident Inspector.	Feedwater
ENS 42630	11 June 2006 05:26:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Technical Specification Shutdown Due to High Rcs Drywell Leakage	At 01:26 on 6/11/2006 Nine Mile Point Unit 1 experienced a rise in drywell leakage of 2 gallons per minute increase of unidentified leakage in a 24 hour period (placing the Unit in an LCO) per Technical Specification 3.2.5.a.2. Present Drywell leakage is 2.33 gpm. A normal orderly shutdown commenced at 01:26, per Technical Specification 3.2.5.b, (to) place the reactor in cold shutdown within 24 hours. Drywell leakage has trended up very slowly over the past several weeks. On 6/10/2006 the plant reduced power, deinerted the primary containment, (and) attempted to repair packing leak on VLV-32-194. (This valve is a 1 inch drain valve on recirculation pump #12). Following attempts to repair (the) valve, (the) plant inerted the primary containment (and) raised power to approximately 45% while main condenser tube cleaning continued. Unidentified drywell leakage began rising at a faster rate beginning on 6/10/2006 at 17:55. Drywell leakage has stabilized at a little over 2 gpm. The TS shutdown is required because of the 2 gpm increase in leak rate over the last 24 hour period. There are no other equipment operability concerns or TS LCOs that present a challenge to a normal safe plant shutdown. The licensee notified the NRC Resident Inspector.	Primary containment Main Condenser
ENS 42403	10 March 2006 03:14:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Automatic Reactor Scram Due to Loss of Condenser Vacuum Turbine Trip	This notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) RPS Actuation (scram) for Nine Mile Point Unit 2 which states '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.' On 03/09/2006 at 2214 while operating at approximately 85 % reactor power (coast down to refueling) a condenser low vacuum condition resulted in a turbine trip and a subsequent reactor scram. A loss of sealing steam most probably caused the loss of condenser vacuum and an investigation is in progress. Special Operating procedure N2-SOP-101 C was entered. All control rods inserted (fully) as expected. Condenser vacuum has been stabilized and the main steam isolation valves are open. The plant is stable and recovery actions are in progress. Decay heat is being removed via the turbine bypass valves to the main condenser. The condensate and feedwater system is in operation maintaining reactor vessel level. The electric plant is in a normal shutdown lineup and there was no effect from this transient on Unit 1. The licensee notified the NRC Resident Inspector. * * UPDATE AT 0035 EST ON 3/14/06 FROM P. WALSH TO W. GOTT * * * This update is being made to provide additional information to EN# 42403 which was communicated via ENS on 3/10/06 at 0058 hours. During the scram that occurred at Nine Mile Point Unit 2 on 3/9/06 at 2214 hours, a primary containment isolation signal to RHR Shutdown Cooling, RHR Head Spray and RHR sample valves was received as designed. No components repositioned as the valves are normally closed during plant operations. The licensee will notify the NRC Resident Inspector. Notified R1DO (P. Krohn).	Feedwater Reactor Protection System Main Steam Isolation Valve Primary containment Shutdown Cooling Main Condenser Control Rod
ENS 41927	18 August 2005 13:49:00	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 Loss of a Power Board	Unit 1 scrammed from 100% power due to a loss of power board 11 coincident with 1/2 scram present already on RPS channel 12 due to (instrumentation and control) (I&C) testing. A loss of power board 11 causes a loss of 11 RPS trip bus which in turn produces a 1/2 scram. Loss of power board 11 is currently under investigation. In addition during the scram, HPCI injected into the reactor vessel on a turbine trip signal to maintain reactor water level. Currently, the reactor is in hot shutdown with reactor water levels being maintained in the normal level band at 74 inches with feedwater in automatic. Reactor pressure is currently 920 psig and being maintained in automatic with turbine bypass valves. Plan is to stay in hot shutdown and complete scram recovery procedures. All control rods fully inserted. No safety relief valves actuated. Electrical busses were being supplied by normal offsite power. Emergency diesel generators are available. The decay heat removal path is currently normal feedwater to the reactor vessel through the turbine bypass valves to the condenser. There was no impact on Unit 2. The licensee is going to suspend any high risk maintenance activities on Unit 2. The licensee notified the NRC Resident Inspector.	Feedwater Emergency Diesel Generator Decay Heat Removal Control Rod
ENS 41584	10 April 2005 10:00:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Scram Signal Resulting from Failure of a Supply Valve to the Scram Air Header	Nine Mile Point Unit 1 received a valid RPS SCRAM signal from high water level in the SCRAM Dump Volume (SDV). While restoring the Hydraulic Control Rod Unit (HCU) for control rod 02-35 to service, the internals to the plug valve for the Instrument Air Supply (116 Valve) to the SCRAM Inlet and Outlet valves failed. This failure caused an approximate 1/2" hole in the SCRAM Air Header, which resulted in the SCRAM Air Header pressure lowering rapidly due to the leak. Operators at the HCU recommended isolating the SCRAM Air Header. The Shift Manager was contacted by the job supervisor and received permission to isolate the SCRAM Air Header. SCRAM inlet and outlet valves opened, SDV vents and drains closed due to the loss of SCRAM Air Header pressure. Approximately 3 minutes after SCRAM Air Header depressurization, a full SCRAM signal occurred as expected due to the water level in the SDV. There was no fuel in the Reactor Vessel (RPV). No Control Rod motion occurred due to all Control Rods being inserted or isolated for maintenance. Immediate (8 Hour Non-Emergency) notification of this event being made as a result of the requirements of 10CRF50.72(b)3(iv)(A). The licensee stated that more information on the event can be found in Nine Mile Point Internal document DER - NM-2005-1565. The license will be notifying the NRC Resident Inspector. * * RETRACTION PROVIDED BY LICENSEE (SHEEHAN) TO NRC (HELD) AT 1739 EDT ON 5/19/05 * * * The scram event that occurred on April 10, 2005 was not initiated from a "valid" scram initiation signal (i.e., none of the instrumentation signals identified in Technical Specification Table 3.6.2a triggered the scram). To the contrary, a valve on a CRD hydraulic control unit (HCU) failed and Operations took action to isolate instrument air from the scram air header. This operator action had the identical effect that a scram signal would have had - the scram air header completely vented through the broken valve and caused the scram inlet and outlet valves on the HCUs to open and the scram discharge volume vents and drains to close. At the time, the reactor was defueled and all control rods were either already inserted or properly removed from service for maintenance, thus, the event did not result in any control rod movement (i.e., the system had been properly removed from service and the safety function had already been performed). Subsequent to the initiating event, as per the design of the CRD and RPS systems, the scram discharge volume filled and a full RPS scram signal was generated. Conclusion: The scram event that occurred on April 10, 2005, resulted from an invalid scram initiation signal. At the time, the reactor was defueled, the CRD system had been properly removed from service and the safety function had been properly performed (no control rods moved). Thus, pursuant to the guidance in NUREG-1022, it is appropriate to conclude that the event is not reportable under 10CFR50.72(b)(2)(iv) or 10CFR50.73(a)(2)(iv). As such, the 8-hour ENS notification that was made at 11:46 on 4/10/05 (reference Event Number #41584) is being retracted The licensee notified the NRC Resident Inspector. R1DO (Bellamy) was contacted.	Control Rod
ENS 41464	7 March 2005 09:37:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	Automatic Turbine Trip/ Reactor Scram	The following information was provided by the licensee (text in quotes): At 0437 Unit 1 scrammed. The initiating event appears to be a turbine tripped initiated by a High level in 122 moisture separator drain tank. Investigating (the cause) of the scram will determine the actual cause. All control rods fully inserted into the core. HPCI initiated. Three relief valves lifted and reseated normally. The Reactor water clean up system is removing decay heat. The licensee will notify the NRC Resident Inspector.	Control Rod	05000220/LER-2005-001
ENS 40998	30 August 2004 12:35:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Unit 1 Manually Scrammed Due to Oscillating Reactor Vessel Water Level	Nine Mile Point, Unit One is initiating a 10 CFR 50.72 (b) (2) (iv) (B) 4-Hour Non-Emergency notification based upon insertion of a 'MANUAL' scram that occurred at 08:35 on Monday, August 30, 2004. At time of transient, plant was operating in Mode 1, Power Operating Condition, at 99.7% of rated power. At 08:25 on Monday, August 30, 2004, Operators noted oscillations on 13 Feedwater flow control valve (FCV) while in 'AUTOMATIC' mode of operation (normal mode of operation for this equipment). Operators took "MANUAL" control of 13 FCV per Plant Operating Procedures. 13 FCV oscillations continued while in the 'MANUAL' mode, and a decision was made to insert a 'MANUAL' scram at 08:35. All control rods fully inserted and the plant responded as designed to the scram. At 08:44, the scram signal was reset per procedure. Currently, plant is in Mode 2, Hot Shutdown Condition with cooldown in progress. Plant is transitioning to Mode 3, Cold Shutdown Condition, per Plant Operating Procedures. At the time the manual scram was inserted, Reactor Vessel Water Level (RVWL) was 67 inches and decreasing (automatic scram setpoint is 53 inches). The 13 FCV is on the discharge of the turbine-driven feedwater pump. Decay heat is currently being removed by the main condenser via the steam bypass valves. All ECCS and safety-related equipment is available, if needed. At the time of the transient there was no plant maintenance on-going which could have been a contributing factor. The licensee informed the NRC Resident Inspector. * * UPDATE ON 8/30/04 AT 2358 EDT FROM M. MINNICK TO J. ROTTON * * * The notification sent to the NRC on 8/30/04 at 11:42 was found to be incomplete. As a normal and expected response to a manual scram at high power, the High Pressure Coolant Injection System (feedwater) automatically initiated during the transient following the manual scram. This should have been reported as an 8 hour Non-Emergency 10CFR50. 72 (b) (3) (iv) (A) notification. The licensee notified the NRC Resident Inspector. Notified R1DO (Henderson).	Feedwater High Pressure Coolant Injection Main Condenser Control Rod	05000220/LER-2004-004
ENS 40919	3 August 2004 23:30:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Plant Shutdown Initiated Due to Inoperable Emergency Ups Inverter	Division 2 Emergency Uninterruptible Power Supply (UPS) Inverter is inoperable. The 24-hour completion time, for Technical Specification 3.8.7.A, has expired and Technical Specification 3.8.7.B was entered. Technical Specification 3.8.7.B requires the plant to be in Hot Shutdown by 0503 on August 4, 2004 and Cold Shutdown by 0503 on August 5, 2004. A plant shutdown was initiated at 1930. The licensee has notified the NRC Resident Inspector.

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