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Start date | Reporting criterion | Title | Event description | System | LER | |
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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 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 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 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 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.
In addition to the initiation of Division 1 and 3 Emergency Diesel Generators, Service Water Radiation Monitor 2SWP*RE146A 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 2SWP*RE146A 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 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.
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.
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).
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 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 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
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 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.
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.
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 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 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 Safety Relief Valve 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.
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 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.
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 40719 | 2 May 2004 06:17:00 | 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)(i), Tech Spec Required Shutdown | Manual Scram Due to Rising Torus Temperature Due to Stuck Open Erv | Operators inserted a manual scram at 0217 (EDT) in anticipation of torus bulk temperature approaching 110 degrees F following an Electromatic Relief Valve (ERV) failing to close during ERV testing. Technical Specification 3.3.2.e requires the insertion of a manual scram prior to torus bulk temperature reaching 110 degrees F. The maximum average torus temperature was 104 degrees F. Operators opened ERV 123 at 0209 for post maintenance testing from approximately 19% power. The ERV subsequently stuck open. Operators performed the off-normal procedure for a stuck open ERV, but this failed to shut the valve. This ported steam from the reactor vessel to the torus resulting in an uncontrolled torus temperature rise. Operators placed torus cooling in service but this did not stop the rising torus temperature. Following the manual scram, cooldown, and depressurization, shutdown cooling is being placed in service. Immediately following the manual scram, a turbine trip signal actuated the logic for high pressure coolant injection. High pressure coolant injection actuated and operators controlled level above 53 inches. With the ERV stuck open, the cooldown rate could not be controlled. The cooldown rate was approximately 190 degrees F during the first hour following the scram. All control rods fully inserted into the core. Decay heat is being removed by shutdown cooling (in service at the time of the report). The operators intend to cooldown to cold shutdown. The electrical buses are stable. Nine Mile Point Unit 2 was not affected. The licensee notified the NRC Resident Inspector. | High Pressure Coolant Injection Shutdown Cooling Control Rod | |
ENS 40684 | 18 April 2004 22:30:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Residual Heat Removal System Isolation Due to a Rise in Reactor Pressure | At 1830 hours on 18 April 2004, Nine Mile Point Unit 2 was restoring from reactor pressure vessel leakage test. Reactor pressure was ~82 psi, temperature ~178 degrees, reactor level was solid with both reactor recirculation pumps running in slow speed per the leak test procedure. 'B' residual heat removal system was being warmed up in preparation for going in to service. Reactor water cleanup reject from the vessel was secured to maintain RPV (Reactor Pressure Vessel) pressure stable in order to provide driving head for flow through the residual heat removal discharge line to radwaste. The line is warmed up from the reactor, back through the Shutdown cooling isolation valve, to radwaste prior to placing shutdown cooling in service. When warm-up criteria are met, the operating procedure directs securing flow. When flow was secured, this effectively isolated the solid reactor vessel, resulting in a rise in reactor pressure. Pressure peaked at ~146 psi before operators established reactor water cleanup reject flow. When RPV pressure reached 128 psi, the residual heat removal system isolation was automatically initiated as designed. The shutdown cooling injection valve, which was open to support piping warm-up, closed as designed. All other shutdown cooling valves were closed prior to the event per the warm-up lineup. RPV is currently depressurized with shutdown cooling in service. All systems are functioning as expected. The licensee notified the NRC Resident Inspector.
The following is a retraction of ENS Notification #40684: On April 18, 2004, Nine Mile Point Unit 2 reported Residual Heat Removal Shutdown Cooling System isolated due to a pressure spike while warming up the piping. The shutdown cooling injection valve, which was open to support piping warm-up, closed as designed. All other shutdown cooling valves were closed prior to the event per the warm-up lineup. Accordingly, Control Room personnel conservatively initiated ENS reporting under 10CFR50.72(b)(3)(iv)(A) in response to the valid containment isolation signal. Subsequent analysis has concluded that, although the isolation signal was valid, it was not a general containment isolation signal (i.e., the isolation signal only involved the shutdown cooling system) and the shutdown cooling function of the Residual Heat Removal System is not listed under 10CFR50.72(b)(3)(iv)(B) as a system that is required to be reportable. This system isolation does not meet the reporting requirements of 10CFR50.72(b)(3)(iv)(A). Therefore, this notification made on April 18, 2004, is being retracted. The NRC Resident Inspector was notified. NRC R1DO (Ron Bellamy) notified. | Shutdown Cooling Reactor Recirculation Pump Reactor Pressure Vessel Residual Heat Removal Reactor Water Cleanup | |
ENS 40322 | 14 November 2003 02:49:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of 115 Kv Line Results in Emergency Diesel Generator Auto-Start | Auto initiation of Emergency Diesel Generator (EDG) 103 and re-energizing of PB 103 from EDG due to loss of off-site 115 KV line #4. Line #4 was momentarily lost, then re-energized. The loss of power also resulted in a trip of Reactor Recirculating Pump 13. Power reduced to approximately 80% rated. Recovery operations are in progress. All systems functioned as required and there was not anything unusual or not understood. The NRC Resident Inspector was notified. | Emergency Diesel Generator |