Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 40067 | 14 August 2003 20:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 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 Greater than 15 Minutes. | Automatic reactor scram due to a loss of offsite power. All rods fully inserted. Supplying power to vital buses via emergency diesel generators. All systems operating properly. NRC Resident Inspector was notified of the event by the licensee.
The following information was provided as an update: RPS Actuation (loss of flow) due to loss of offsite power. Auto actuation of AFW in response to the unit trips. Auto start & load of Emergency Diesel Generators in response to the loss of offsite power. The licensee informed the state/local authorities and the NRC resident inspector. Notified R1IRC.
The licensee exited the Unusual Event at 0210 upon confirmation of a stable off-site power. Notified R1 IRC and FEMA(Heyman). | Emergency Diesel Generator | |
| ENS 40069 | 14 August 2003 20:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Unusual Event Declared Due to Loss of Offsite Power Greater than 15 Minutes. | Automatic reactor scram due to a loss of offsite power. All rods fully inserted. Supplying power to vital buses via emergency diesel generators. All systems operating properly. NRC Resident Inspector was notified of the event by the licensee.
The following information was provided as an update: RPS Actuation (loss of flow) due to loss of offsite power. Auto actuation of AFW in response to the unit trips. Auto start & load of Emergency Diesel Generators in response to the loss of offsite power. Unit 3 also experienced a second AFW auto start during the event. Unit 3 also entered Tech Spec LCO 3.0.3 for loss of two offsite circuits and one EDG inoperable. #31 EDG was declared inoperable when its associated Fuel oil storage tank inventory decreased below required. The licensee informed the state/local authorities and the NRC resident inspector. Notified R1IRC.
The licensee exited the Unusual Event at 0210 upon confirmation of a stable off-site power. Notified R1 IRC and FEMA(Heyman). | Emergency Diesel Generator | |
| ENS 41003 | 1 September 2004 04:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Oscillating Steam Generator Water Level | On September 1, 2004, at 00:01 hours, operations noticed the 22 Main Feedwater (FW) regulating valve operating erratically and FW and 22 Steam Generator (SG) level oscillating. Operators placed the 22 Main FW regulating valve in manual and attempted to control FW flow and SG level but were unsuccessful. A Nuclear Plant Operator (NPO) in the plant reported water hammer in the Auxiliary Building that contains the FW regulating valves. At 00:05 hours operations initiated a manual trip of the reactor based on the degrading condition with FW flow and SG level. All control rods fully inserted and safety equipment operated as expected. Auxiliary Feedwater started as expected, the Emergency Diesel Generators did not start and offsite power remained available. The plant is in Hot (Standby) at Normal Temperature and Pressure and there was no radiation release. The 22 Main FW regulating valve was discovered to be partially stuck open. At 00:19 hours operations isolated the 22 FW flow path by closing the 22 FW line motor operated valve (MOV) BFD 5-1. The condition is under investigation and a post trip review is being conducted." State of NY was notified of this event by the licensee. The NRC Resident Inspector was notified of this event by the licensee.
After further review of the reporting guidelines and discussion with the NRC R1, Licensing concluded the event notification should be reported under 10CFR50.72 (b)(1)(iv)(B) for a 4-hour report. A Condition Report was initiated for the late notification. Notified NRR (Reis) and R1DO (Henderson). | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater Control Rod | |
| ENS 41066 | 24 September 2004 15:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip with No Complications Due to Lowering Steam Generator Water Level | On September 24, 2004, at 11:53 hours, operations noticed the 23 Main Feedwater (FW) regulating valve operating erratically and FW flow and 23 Steam Generator (SG) level decreasing. At 11:55 hours operations initiated a manual trip of the reactor based on the degrading condition with FW flow and SG level. All control rods fully inserted and safety equipment operated as expected. Auxiliary Feedwater started as expected, the Emergency Diesel Generators did not start as offsite power remained available. The plant is in Hot (Standby) at Normal Temperature and Pressure and there was no radiation releases. The condition is under investigation and a post trip review is being conducted." State was notified of this event by the licensee. The NRC Resident Inspector was notified of this event by the licensee. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater Control Rod | |
| ENS 41187 | 10 November 2004 04:39:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Actuation of the Emergency Ac Electric Power System | On November 9, 2004 at 2339 hours, an actuation of the emergency AC electrical power system occurred. While in a refueling outage a planned evolution was in progress to tie 480V Bus 3A to 480V Bus 6A using the tie breaker. A Bus under-voltage condition occurred when the Normal supply breaker was opened to bus 6A. When the normal supply breaker was opened, the tie breaker also opened causing a loss of power to bus 6A. With bus 6A deenergized an under-voltage signal was generated. 21 and 22 Emergency Diesel Generators automatically started and supplied bus sections 5A, 2A, and 3A. 480V Bus 6A remained deenergized because 23 Emergency Diesel Generator was out of service for planned maintenance. As a result, Residual Heat Removal cooling was lost for 5 minutes until power was restored and 21 Residual Heat Removal Pump was started. In addition, normal Spent Fuel Pool Cooling was lost for 39 minutes until 21 Spent Fuel Pool Pump was started. An activity is in progress to determine why the tie breaker opened when the normal supply breaker to Bus 6A was opened. On November 10, 2004 at 0058 hours unit 2 was returned to its normal 480V lineup. At the time of the event, refueling outage 16 was in progress on Unit 2 with all fuel assemblies installed in the core after refueling with the Reactor Vessel Head and Upper Internals removed. This results in a condition that resulted in a valid actuation of the emergency AC electrical power systems which is reportable under 10 CFR 50.72(b)(3)(iv)(A). The loss of the 21 spent fuel pool cooling pump resulted in a 3 degree rise in spent fuel pit temperature. The loss of the 21 RHR pump resulted in no appreciable increase in reactor coolant temperature. The licensee will be notifying the New York Public Service Commission of this incident and has notified the NRC Resident Inspector. | Emergency Diesel Generator Residual Heat Removal | |
| ENS 41227 | 26 November 2004 18:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rps Actuation Due to Turbine Trip Caused by Loss of Stator Cooling Water | At 1322 EST on 11/26/04, the Indian Point Unit 2 Reactor automatically shutdown due to a turbine generator trip resulting from a loss of stator cooling water. All systems responded as expected, no PORV's opened. The reactor is currently shutdown with all control rods inserted. Temperature is being maintained in Mode 3 with the steam dumps. AFW was initiated and is being used to maintain SG levels. Investigation into the loss of stator cooling water is ongoing. The licensee has notified the NRC Resident Inspector. | Stator Water Control Rod | |
| ENS 41673 | 6 May 2005 14:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Steam Flow/ Feed Flow Mismatch Signal | On May 6, 2005, at 10:32 hours, an automatic reactor trip occurred due to a Main Steam Flow/Feedwater Flow mismatch on the 32 Steam Generator (SG). All control rods fully inserted and safety equipment operated as expected. Auxiliary Feedwater started as expected. The Emergency Diesel Generators did not start as offsite power remained available. The plant is in Hot Shutdown at Normal Temperature and Pressure and there was no radiation release. The condition is under investigation and a post trip review is being conducted. No primary or secondary relief valves lifted during the reactor trip. Decay heat is being discharged to the main condenser via the high pressure steam dumps. Auxiliary feedwater is being used to maintain SG level. The electrical grid is stable. There was no affect on Unit 2. The licensee notified the NRC Resident Inspector and the State of New York. | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Main Condenser Control Rod Main Steam | |
| ENS 41701 | 16 May 2005 14:39:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Afw Actuation During Surveillance Testing | On May 16, 2005, at approximately 1039 hours, both motor driven Auxiliary Boiler Feedwater Pumps automatically started after exceeding their 28 second time delay during performance of the low-low Steam Generator Water Level section of procedure 3PT-M13B1, 'Reactor Protection Logic Channel Functional Test.' The operators secured the pumps and saw no indication of reactivity addition. The operators anticipated the possibility of these results because the test states in its precaution and limitations section, that the pumps would start if the action required by the test is not completed within the 28 second time delay. The cause of the event is under investigation. The licensee notified the NRC Resident Inspector and the NY State Public Service Commission. | Steam Generator Feedwater | |
| ENS 41762 | 10 June 2005 13:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Service Water Leak in Exciter Cabinet | At 0924 on 06/10/05, Indian Point 3 was manually tripped due to a service water leak in the main generator exciter. All control rods fully inserted. Plant response was as designed. Unit 3 is stable in Mode 3. Investigation is ongoing. Unit 2 (was not affected and) remains at 100% power. The steam generators are discharging steam to the main condenser to remove decay heat. The licensee notified the NRC resident inspector. | Steam Generator Service water Main Condenser Control Rod | |
| ENS 42217 | 22 December 2005 07:08:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Afw Start During Main Feedwater Pump Trip | While performing a plant shutdown due to a packing leak on #24 Feedwater Regulating Valve on 12/22/05 at 0208 at about 67% power, #22 MBFP tripped while swapping lube oil coolers. #21 and #23 motor driven auxiliary feedwater pumps auto started. All systems responded properly. Entered appropriate abnormal operating procedure (2-AOP-FW-1, Loss of Main Feedwater). Reduced power to within the capacity of one MBFP and established conditions to shutdown and realign #21 and #23 ABFPs to Auto. On 12/22/05 at 0218, #21 and #23 ABFPs were shutdown and aligned for automatic operation. The licensee notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater | |
| ENS 42220 | 22 December 2005 10:50:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Actuation Signal Due to High Steam Generator Water Level | While performing a plant shutdown due to a packing leak on #24 Feedwater Regulating Valve (see event #42217) on December 22, 2005 at 0550 hours, Indian Point Unit 2 received a 22 Steam Generator Water High-High Level signal at 73% narrow range level. This resulted in tripping the standby 21 Main Boiler Feed Pump which then resulted in a start signal being sent to both motor driven Auxiliary Feedwater Pumps. Both of the motor driven Auxiliary Feedwater Pumps were already operating and feeding the steam generators when this start signal was received. Operators restored 22 Steam Generator Level to normal. All systems responded properly. The licensee will notify the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 42378 | 1 March 2006 19:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip as a Result of Dropped Control Rods | On March 1, 2006, at 1435 hours, the Reactor Protection System was manually actuated while at 100 percent power upon indication of 12 dropped rods. All safety systems actuated as required and safety equipment operated as expected. There were eight control rods that did not indicate full insertion. One control rod had a bad rod bottom light bulb and the other seven control rods require calibration to indicate full insertion. Auxiliary Feedwater started as expected. The Emergency Diesel Generators did not start as offsite power remained available and stable. No primary or secondary relief valves lifted. The plant is in Hot Standby at normal temperature and pressure with residual heat removal using auxiliary feedwater and normal heat removal through the condenser via steam dumps. There was no radiation released. An investigation revealed that a manual disconnect switch for power control cabinet 1AC was open and scaffold work was being performed in the area. Cabinet 1AC is the power supply for the 12 dropped rods. The event is still under investigation and a post trip review is being conducted. The licensee notified the State and the NRC Resident Inspector. | Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Residual Heat Removal Control Rod | |
| ENS 42687 | 6 July 2006 07:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Generator Differential | Indian Point Unit 3 tripped on Generator Differential (Direct Generator Trip). Current conditions are stable, mode 3 operation with all automatic actions occurring properly. Steam Generators levels are stable on aux. feed water. Unit 2 was unaffected by the trip. All control rods inserted fully and decay heat is being removed via steam dump to the condenser. All required safety related systems are working properly and the electrical buses and diesel generators are in their normal alignment. The licensee will notify the NRC Resident Inspector.
The Auxiliary Feedwater System actuated and the Auxiliary Feedwater Pump started as designed due to the reactor trip. The actuation of the Auxiliary Feedwater System was the result of Steam Generator shrink/swell effect due to the reactor trip at full power. The actuation of the Auxiliary Feedwater System is an eight-hour non-emergency event reportable under 10 CFR 50.72(b)(3)(iv). The licensee will notify the NRC Resident Inspector. Notified R1DO (P. Henderson) | Steam Generator Auxiliary Feedwater Control Rod | |
| ENS 42720 | 21 July 2006 14:30:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Sparks from Under Main Generator | At 1031 hours on July 21, 2006, the reactor operator manually tripped the reactor from 100% power resulting in turbine trip and actuation of auxiliary feedwater. This is reportable under 10 CFR 50.72(b)(2)(iv)(B). 31 Reactor Coolant Pump tripped after the reactor trip. All other systems operated as expected. Steam is being condensed by the condenser. The reactor was tripped due to an abnormal condition under the main generator; electrical arcing/sparking was observed. The reactor will remain shutdown until the cause of the arcing/sparking is identified and corrected. There was no indication of fire but the fire brigade was called out. The observed arcing and sparking from underneath the main generator secured after the turbine trip. All control rods fully inserted on the reactor trip. The electric plant is in a normal shutdown lineup and the EDGs are operable. The unit is stable and Unit 2 was not affected. The licensee notified the NRC Resident Inspector and the New York Public Service Commission.
Licensee reported the AFW actuation as a Specified System Actuation per 10 CFR 50.72(b)(3)(iv)(A). The licensee notified the NRC Resident Inspector. Notified the R1DO (Kinneman) | Auxiliary Feedwater Control Rod | |
| ENS 42797 | 23 August 2006 14:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Scram Due to Mismatch of Primary Power and Secondary Load Indications | On August 23, 2006, at 1035 hours, the Reactor Protection System was manually actuated while at reduced load of approximately 68 percent reactor power upon indication of an apparent disparity in net electric load. It was determined that the cyclic operation of the condenser steam dumps varied steam to the turbine resulting in the indicated drop in electrical output power from about 500 Mwe to 150 Mwe. The manual trip is reportable under 10 CFR 50.72(b)(2)(iv). The event started when heater drain tank pumps 21 and 22 tripped, apparently due to a failure of the automatic heater drain level control valves that could not be manually corrected from the single controller. Reactor operators reduced load in accordance with station procedures to respond to the trip of the heater drain tank pumps and stabilized reactor power at 77%. As a result of the required load reduction delta flux was outside of the operational limit envelope specified at 77% reactor power. This required a further load reduction to less than 50 percent as per Technical Specification 3.2.3. At approximately 68% percent power the reactor was manually tripped due to the mismatch of primary power and secondary load indications. The Auxiliary Feedwater System actuated following the manual trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). A high-high water level signal was received on the 22 Steam Generator. This was due to a feedwater regulating valve that did not fully close. The feedwater isolation valves closed. All other safety systems actuated as required and safety equipment operated as expected. The Emergency Diesel Generators did not start as offsite power remained available and stable. No primary or secondary relief valves lifted. The plant is in Hot Standby at normal temperature and pressure with residual heat removal using auxiliary feedwater and normal heat removal through the condenser via steam dumps. There was no radiation released. A post trip investigation will be performed prior to restart. The licensee notified the NRC Resident Inspector.
This is an update of non-emergency Event Notification 42797 that was previously made on August 23, 2006. Clarification is provided to include reporting of the Event Classification for actuation of a second system (automatic feedwater isolation) as an 8-Hr. Non-Emergency per 10 CFR 50.72(b)(3)(iv)(A). This update revises the wording as follows: The feedwater isolation valve was manually closed. After receipt of the high-high level trip signal, 21, 23, and 24 feedwater isolation valves automatically closed. This is also reportable under 10 CFR 50.72(b)(3)(iv)(A). The licensee notified the NRC Resident Inspector. Notified R1DO (Conte). | Steam Generator Feedwater Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Residual Heat Removal | |
| ENS 42800 | 24 August 2006 15:35:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Main Feedwater Pump Trip During Startup | On August 24, 2006, at approximately 1135 hours, during startup from a reactor trip on August 23, 2006, a trip of Main Boiler Feedwater Pump (MBFP) 21 sent an actuation signal to automatically start both motor driven Auxiliary Feedwater Pumps (AFPs) as well as isolate steam generator blowdown, and close MBFP discharge valve BFD-2-21. At the time, the plant was at about 1% power and both motor driven AFPs were in operation. All plant responses were as expected. The cause of the MBFP trip is under evaluation. Offsite power is normal and emergency diesel generators are operable." The plant is currently at 8% power. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 42993 | 15 November 2006 19:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip During Electrical Main Generator Troubleshooting | On November 15, 2006, at 14:00 hours, an automatic reactor trip occurred during troubleshooting of the Main Generator Exciter Generrex power supply, when an electrical spike resulted in a turbine trip which caused a reactor trip. All control rods fully inserted and safety equipment operated as expected. The Auxiliary Feedwater (AFW) System actuated as expected as a result of a low steam generator level from the reactor trip and started the motor driven AFW pumps. The Emergency Diesel Generators did not start as offsite power remained available. The plant was stabilized in hot standby with decay heat being removed by the main condenser. There were no radiation releases. The condition is under investigation and a post trip review is being conducted. No PORVs or SG safety valves lifted during the event. The electrical plant is in a normal shutdown lineup and the EDGs are operable in standby. There was no affect on Unit 3 operation due to this event. The licensee notified the NRC Resident Inspector, and stated that IPEC communications has transmitted an IPEC Status Report which provides a status to local, state, and federal agencies. | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 43016 | 30 November 2006 19:45:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Auxiliary Feedwater Pump Automatic Start | On November 30, 2006, at approximately 14:45 hours, an Auxiliary Feed Water (AFW) System actuation signal was generated as a result of maintenance activities on the 21 Main Boiler Feed Pump (MBFP). Both motor driven AFW pumps were running at the time and were not automatically started due to previously being placed in service to provide feed water flow to the steam generators. The actuation signal isolated the Steam Generator blowdown Containment Isolation Valves. Both MBFPs were isolated and not feeding the steam generators at the time the actuation signal was generated. The plant was in Mode 3 (Hot Standby) for a forced outage and 21 MBFP was in a maintenance / test condition. The automatic actuation occurred due to the reset oil pressure lowering below the trip set point and initiating an automatic actuation signal. The licensee notified the NRC Resident Inspector.
At 2100 on 11/30/2006, as a result of continuing work activities on the 21 MBFP an AFW System actuation signal was generated. Both motor driven AFW pumps were running at the time and were not automatically started due to previously being placed in service to provide feed water flow to the steam generators. 21 MBFP has been successfully placed in service." The NRC Resident Inspector was notified of this update by the licensee. Notified R1DO (James Dwyer). | Steam Generator Auxiliary Feedwater | 05000425/LER-2015-002 |
| ENS 43199 | 28 February 2007 11:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Following Malfunction of Pressure Transmitter | At approximately 0633 AM on February 28, 2007, control room operators manually initiated a reactor trip after observing decreasing steam generator levels as a result of low feed water flow. Control room alarms indicated low main feed water pump suction pressure which was subsequently, attributed to a malfunction of the main feed water pump low suction pressure cutback pressure transmitter (PT-408B) on the common main feed pump water supply header. The malfunction of the main feed pump water low suction cutback pressure transmitter resulted in a cutback of both main feed water pumps reducing main feed water flow to the steam generators and causing decreasing steam generator levels. All control rods fully inserted and all safety systems responded as expected. The auxiliary feedwater system actuated as expected from low steam generator levels which occurs as a result of a full power reactor trip. The emergency diesel generators (EDGs) did not start as offsite power remained available. The plant was stabilized in hot shutdown with the (Auxiliary Feedwater System) providing decay heat removal via the main condenser. The main feed water pumps are shutoff and secured. The event is under investigation and a post trip review is being conducted. A courtesy call to stakeholders will be made. All EDGs remain available in standby. No steam generator Power Operated Relief Valves or Main Steam Safety Valves lifted. The licensee notified state and will notify local authorities. The licensee plans to issue a press release. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater Main Steam Safety Valve Decay Heat Removal Main Condenser | |
| ENS 43272 | 3 April 2007 08:16:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip While Performing Boiler Feed Pump Maintenance | The Indian Point Unit 3 reactor was manually tripped at 04:16 on 4/3/2007 due to low steam generator levels that occurred while the 31 MBFP (main boiler feed pump) speed control was being isolated for maintenance. Indian Point Unit 3 is currently in mode 3 with all automatic actions for a manual reactor trip occurring as required. Unit 2 was unaffected and remains in mode 1 at 100% power. The reactor was manually tripped when steam generator levels reduced to 18% in all four steam generators. All rods inserted on the trip. No safety or relief valves lifted due to the trip. The motor driven aux feedwater pumps automatically started on low steam generator level and are being used to maintain steam generator level. Atmospheric steam dumps are maintaining reactor temperature. The Unit 3 electrical lineup is the normal shutdown electrical lineup. The licensee has notified the state Public Service Commission. The licensee will notify the NRC Resident Inspector. | Steam Generator Feedwater | |
| ENS 43285 | 6 April 2007 15:11:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Main Generator Transformer Explosion - No Impact on Plant Safety Related Systems | At 1143 on 4/6/07 the licensee declared an Unusual Event due to a main transformer explosion in the protected area. The event was declared in accordance with EAL 8.2.2. Unit 3 automatically tripped due to a load reject at the main generator. The Unit is currently stable at normal operating temperature and pressure. Decay heat is being removed via the steam dumps to the main condenser. All emergency diesel generators are available if needed but safety buses are currently supplied with offsite power. At the transformer site the deluge system actuated and the fire brigade responded. The fire was out at the time of the report The licensee informed the NRC Resident Inspector. * * * UPDATE AT 1232 ON 4/6/07 FROM LAUGHLIN TO HUFFMAN * * * The offsite local fire department was called by the licensee but their support was not needed and they were released. * * * UPDATE AT 1259 ON 4/6/07 FROM LAUGHLIN TO HUFFMAN * * * The licensee exited Unusual Event at 1247 based on the fact that the fire was confirmed to be extinguished, there was no damage to safety equipment and the plant was in a safe and stable condition. * * * UPDATE AT 1440 ON 4/6/07 FROM PRUSSMAN TO KNOKE * * * On April 6 at 1143 hours, an unusual event was declared based on Emergency Action Level 8.2.2, an explosion in the 'B' phase of the 31 main transformer. This is a one hour reportable event to the NRC made at 1159 hours. This is an update of EN 43285. The NUE was terminated at 1247 hours. The State and County were also notified of the unusual event. A press release was made. Although no one was injured as a result of this event, a four hour report is being made since this is related to on-site personnel safety. The main generator tripped and this resulted in a consequential reactor trip. The Reactor Protection System shut down the reactor at 1109 hours, a four hour report. The plant operated as designed and the Auxiliary Feedwater System actuated, an eight hour report. The fire was reported to the Control Room at 1111 hours. The fire was put out in less than 15 minutes. The event is currently under investigation. The header information in this event was revised as a result of this update to indicate 'Offsite Notification.' | Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Main Transformer Main Condenser | |
| ENS 44089 | 24 March 2008 02:16:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip - Loss of Speed on Mbfp | The Indian Point Unit 2 reactor was manually tripped from 94% power at 2216 on 3/23/08 due to a loss of speed on 22 main boiler feed pump (MBFP). Indian Point Unit 2 is currently in mode 3 with all automatic actions for a manual reactor trip occurring as required. Indian Point Unit 2 was in a coast down in advance of a scheduled refueling outage. The reactor was manually tripped as required by abnormal operating procedure 2-AOP-FW-001. All control rods inserted on the trip. No safety or relief valves lifted due to the trip. The motor driven aux feedwater pumps automatically started on low steam generator level and are being used to maintain steam generator level. Condenser steam dumps are maintaining reactor temperature. The Unit 2 electrical lineup is the normal shutdown electrical lineup. The licensee has notified the state Public Service Commission. Unit 3 was unaffected and remains in mode 1 at 100% power. The licensee has notified the NRC Resident Inspector. The licensee expects to issue a media release. | Steam Generator Feedwater Control Rod | |
| ENS 44091 | 25 March 2008 15:18:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Inadvertent Actuation of Emergency Diesel Generator During Surveillance Testing | On March 25, 2008, during scheduled monthly testing per surveillance procedure M628, '480 Volt Undervoltage Degraded Grid Protection System Bus 5A Functional', an Instrumentation & Control technician conducting the test inadvertently caused an undervoltage condition on 480 volt AC safeguards Bus 5A, resulting in the automatic actuation of the associated Emergency Diesel Generator (EDG)-33. In accordance with design EDG-33 started and loads on Bus 5A were stripped and required loads were sequenced back onto Bus 5A. All equipment performed as designed. Normal power for Bus 5A has been restored and all equipment returned to pre-event condition. Investigation into the cause of the event is in progress. Unit 2 is currently in a refueling outage and was unaffected. Licensee has notified the NRC Resident Inspector and the State Public Service Commission. | Emergency Diesel Generator | |
| ENS 44153 | 21 April 2008 15:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Lowering Steam Generator Water Level | On April 21 2008, at approximately 1123 hours, during power ascension from a scheduled refueling outage, Unit 2 experienced an inadvertent main turbine runback. The runback initiated for unknown reasons from an initial reactor power of approximately 37%. As a result of the runback, operators observed decreasing steam generator levels and initiated a manual reactor trip at 1125 hours. As a result of the reactor trip, the auxiliary feedwater pumps actuated on a low steam generator level as designed. Prior to the runback, one of two main feedwater pumps was operating. The unit is stable and in Mode 3. Operators closed the main steam isolation valves as a result of indications of lowering RCS temperature. A 4-hour non-emergency notification is provided for a reactor trip while critical and 8 hour non-emergency notification is provided for a valid actuation of the auxiliary feedwater system. All rods inserted as expected. Decay heat is being removed by AFW through the atmospheric steam dumps. Offsite power is in a normal configuration. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Main Steam Isolation Valve Auxiliary Feedwater Main Turbine | |
| ENS 44383 | 4 August 2008 16:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation - Both Auxiliary Boiler Feedwater Pumps Started Unexpectedly | On August 4, 2008 at approximately 1200 hours, both motor driven auxiliary boiler feed pumps automatically started during the performance of procedure 3PT-M13B1, 'Reactor Protection Logic Channel Functional Test Train B'. Operations secured the pumps and returned them to auto at approximately 1207 hours. The event did result in water addition to the steam generators but Operations did not observe any reactivity changes on the nuclear instrumentation. The cause of the event is under investigation. Operators did observe a 2% increase in steam generator water level but no limits or alarms were approached. The licensee has notified the NRC Resident Inspector and New York State Public Service Commission. | Steam Generator Feedwater | |
| ENS 44554 | 9 October 2008 16:54:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator and Auxillary Feedwater Actuation Due to Bus Undervoltage | On October 9, 2008, during scheduled monthly testing per surveillance procedure 3-PT-M62C, '480 volt Undervoltage Degraded Grid Protection System Bus 6A Functional,' the normal supply power to 480 volt safeguards Bus 6A de-energized resulting in the start and load of Bus 6A by the 32 Emergency Diesel Generator (EDG), and the actuation of the '33' motor driven Auxiliary Feedwater Pump (AFWP), and the '32' steam driven AFWP. All equipment performed as designed. Auxiliary feedwater (AFW) was injected into the Steam Generators (SG). Core reactivity changes as a result of AFW injection resulted in an 0.1% increase in reactor power. SG level change from normal of approximately 1-2% with no actuation of SG alarms, no challenge to automatic control, and no rod movement. No significant change to the Nuclear Instrumentation was observed. Normal supply power was restored to bus 6A and the EDG-32 returned to Auto. Investigation into the cause of the event is in progress. Unit 2 was unaffected and is at 100% power. The licensee notified the NRC Resident inspector. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 44749 | 2 January 2009 15:09:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unexpected Autostart of Emergency Diesel Generator During Surveillance Testing | During performance of 3-PT-M62C (Rev 7) 480V Undervoltage/Degraded Grid Protection System Bus 6A Functional Test, Bus 6A was inadvertently de-energized. (Number) '32' Emergency Diesel Generator (EDG) autostarted and is supplying Bus 6A. Additionally, '32' and '33' Auxiliary Feed Pumps started due to ESF actuation based on loss of Bus 6A. All equipment operated as required. Investigation is in progress as to the cause of loss of power to Bus 6A. Bus 6A remains energized by '32' EDG. The licensee informed the NRC Resident Inspector. | Emergency Diesel Generator | |
| ENS 44967 | 3 April 2009 15:36:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Loss of 21 Mbfp | At 1136 hours on April 3, 2009 with the plant in mode 1, Indian Point Energy Center Unit 2 experienced a 21 Main Boiler Feed Pump (MBFP) trip. Following the MBFP trip the turbine runback circuit did not actuate as expected. The plant operators attempted to perform a manual main turbine runback and inserted a manual reactor trip based on the impending trip signal of Steam Generator low level. Auxiliary feedwater automatically started as expected. Troubleshooting of the cause of the transient is in progress. Unit 2 is stable in mode 3 with AFW in service. Heat removal is to the main condenser via the steam dumps. Offsite power is available and supplying all safeguards busses. Unit 3 is not affected and remains in mode 6 for 3R15. The (NRC) Resident Inspector has been notified. | Steam Generator Auxiliary Feedwater Main Turbine Main Condenser | |
| ENS 45069 | 15 May 2009 05:53:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Failure of the Main Feedwater Regulating Valve | On May 15, 2009 Indian Point Unit 3 initiated a manual Reactor Trip at 0153 in response to a failure of 33 Main Feed Regulating Valve. 33 Main Feed Regulating Valve failed open causing 33 Steam Generator Level to rise. A manual Reactor Trip signal was inserted when it was determined that the level rise in 33 Steam Generator could not be corrected. As a result of the reactor trip, the Auxiliary Feedwater System automatically actuated per plant design. Unit 3 is currently in Mode 3. This results in a condition that resulted in an actuation of the Reactor Protection System which is reportable under 10 CFR 50.72(b)(2)(iv)(B). The valid actuation of the Auxiliary Feedwater System is reportable under 10 CFR 50.72(b)(3)(iv)(A). All rods inserted during the trip. There were no relief or safety valves that lifted during the transient. The electrical grid is stable and is in the normal shutdown electrical lineup. The plant is being maintained at normal operating temperature and pressure using steam dumps to condenser to remove decay heat. Unit 2 was not affected by the trip. The licensee notified the New York Public Service Commission and the NRC Resident Inspector. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | |
| ENS 45098 | 28 May 2009 10:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Resulting from a High Steam Generator Level Turbine Trip | At 0530 EDT a power reduction commenced due to elevated vibrations on 32 MBFP (Main Boiler Feedwater Pump). When reactor power reached 60% the team was attempting to stabilize reactor power when a level excursion occurred in 32 steam generator. When level reached its high level turbine trip set point an automatic reactor trip occurred and all systems responded as expected. Auxiliary feedwater actuated as expected. The Unit is currently in mode 3 and stable with auxiliary feed water in service and reactor temperature maintained with the steam dumps to the main condenser. Investigation into the cause of the steam generator level excursion is in progress. Offsite power is available and supplying safeguard busses. Unit 2 was unaffected and remains at 100% power. The (NRC) Resident Inspector has been notified. No safety or relief valves lifted during this event. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | |
| ENS 45255 | 11 August 2009 00:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Indian Point Unit 3 Automatically Tripped During Thunderstorm | Indian Point Unit 3 automatically tripped from 100% power at 2032 (EDT) on 8/10/2009 due to (a) generator trip. Indian Point Unit 3 is currently in mode 3. All control rods inserted on the trip. No safety relief valves lifted due to the trip. The motor driven aux feedwater pumps automatically started on low steam generator level and are being used to maintain steam generator level. Condenser steam dumps are maintaining reactor temperature. After the trip , (the) 6.9kv bus 2 did not transfer to offsite power and is de-energized resulting in a loss of (the) '34' RCP (Reactor Coolant Pump). (The) 480v bus 2A was energized by the diesel generator. Plant operators are investigating the cause of the generator trip and loss of bus 2. Unit 2 was unaffected and remains in mode 1 at 100% power. The licensee has notified the NRC Resident Inspector. The licensee informed the State of New York and plans on issuing a press release. | Steam Generator Feedwater Safety Relief Valve Control Rod | |
| ENS 45306 | 27 August 2009 23:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Turbine Trip | At 1945 on 8/27/09 Indian Point Unit #3 automatically tripped due to a turbine trip signal. The cause for this automatic trip is under investigation. All auxiliary feedwater pumps started as expected. All control rods inserted as expected. Currently, Indian Point #3 is in Mode 3. The reactor coolant system is at Normal Operating Temperature, Pressure and Level. Offsite power is available and supplying all safeguards busses. Heat removal is to the main condenser via the steam dumps. Indian Point Unit 2 is unaffected and remains in Mode 1 at 100% power. The (NRC) Resident Inspector has been notified. | Reactor Coolant System Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 45474 | 3 November 2009 02:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Turbine Trip Caused by a Generator Lock-Out Relay Actuation | At 2242 on November 2, 2009 Indian Point Unit 2 reactor tripped from 100% due to a turbine trip. The turbine tripped due to a generator lockout relay actuation. The investigation into the relay actuation is ongoing. All systems responded as expected. Auxiliary feed water is providing steam generator feed and levels are being maintained. No steam safeties opened and no coolant system relief or safety valves actuated. The reactor plant is in mode 3 and stable. Unit 3 was unaffected and remains at 100%. The trip was characterized as uncomplicated. All rods fully inserted and stabilized at the normal mode 3 no-load Tave and pressure. The shutdown electrical lineup is normal and the unit was not in any significant LCOs at the time of the trip. The licensee stated that they were not performing any maintenance or surveillances at the time of the trip. Initial investigation has not revealed any obvious generator electrical system issues but a detailed investigation is ongoing. The licensee has notified State and local authorities and the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater | |
| ENS 45624 | 11 January 2010 20:59:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Trip Due to Main Generator Electrical Trip | At 1559 on January 11, 2010, Indian Point Unit 2 tripped from 100% power due to a main generator electrical trip. The investigation into the generator trip is ongoing. All systems responded as expected. The auxiliary feedwater system responded as expected and is maintaining steam generator water levels. Decay heat removal is via the steam generators to the main condenser. Offsite power and plant electrical lineups are normal. No primary or secondary side relief valves lifted. The reactor plant is in mode 3 and stable. Indian Point Unit 3 was not affected and remains at 100% power. The licensee notified the NRC Resident Inspector and the N.Y. State Public Service Commission. The licensee intends to notify the Mayor's Office and issue a press release. | Steam Generator Auxiliary Feedwater Decay Heat Removal Main Condenser | |
| ENS 46229 | 3 September 2010 14:58:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 2 Experienced an Automatic Reactor Trip Due to High Steam Generator Level | On September 3, 2010, at approximately 1058 hrs., Indian Point Unit 2 automatically tripped from 41% power due to High Steam Generator Level. At the time of the trip, Unit 2 was performing a power reduction to take the unit offline for a planned maintenance outage. Unit 2 is currently stable in Mode 3. All automatic actions occurred as required. All control rods fully inserted with the exception of rod H-8 - which indicates 26 inches withdrawn. No primary or steam generator safety or relief valves lifted. The motor driven auxiliary feedwater pumps automatically started on low steam generator level as designed. Decay heat removal is via the steam generators to the main condenser. Offsite power is available and supplying all safeguards busses. The cause of the trip and rod H-8 position indication are under investigation. Unit 3 is unaffected and remains in Mode 1 at 100% power. The NRC Resident Inspector has been notified. Offsite notification was made to the State of New York. No press releases are planned. | Steam Generator Auxiliary Feedwater Decay Heat Removal Main Condenser Control Rod | |
| ENS 46241 | 10 September 2010 01:29:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Service Water Leak in the Main Generator Exciter Housing | On September 9, 2010, at approximately 2129 hours (EDT), Indian Point 3 manually tripped the reactor from the control room following indications of a service water leak in the Main Generator Exciter housing. Unit 3 is currently stable in mode 3. All automatic actions occurred as required. All control rods fully inserted. No primary or steam generator safety relief valves lifted. The motor driven auxiliary feedwater pumps automatically started on low steam generator level as designed. Decay heat removal is via the steam generators to the main condenser. Offsite power is available and supplying all safeguards busses. Unit 2 is unaffected and remains in Mode 5 in a forced outage for 21 RCP (Reactor Coolant Pump) repair. During the Fast Bus Transfer (the) 34 Reactor Coolant Pump tripped. The cause is being investigated. There was an inadvertent release of CO2 to 31 Main Boiler Feed Pump. The cause was due to the suction relief valve lifting which has subsequently reseated. The NRC Sr. Resident Inspector and New York State Public Service Commission have been notified. The reactor is currently being maintained at normal operating temperature and pressure. | Steam Generator Service water Auxiliary Feedwater Decay Heat Removal Safety Relief Valve Main Condenser Control Rod | |
| ENS 46400 | 7 November 2010 23:39:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Alert Declared Due to Explosion in the 21 Main Transformer | At 1849 EST, the licensee declared an Alert due to an explosion in the 21 Main Transformer. As a result of the loss of the transformer, Unit 2 experienced a reactor trip. Plant response to the trip was normal and the plant is stable in Mode 3. At 1839 hrs., the licensee activated their Emergency Response Organization. Additionally, there was a report of smoke issuing from a metering station at the site boundary. Offsite assistance has been requested to investigate the metering station. No injuries were reported and the licensee is investigating whether any other equipment damage occurred as a result of the transformer failure.
On November 7, 2010, at 1839 hours, the Reactor Protection System automatically actuated at 100% Reactor Power when the 21 Main Transformer failed. The failure resulted in a turbine / reactor trip. All plant equipment responded normally to the unit trip. This is reportable under 10 CFR 50.72(b)(2)(iv). The transformer failure was characterized as an explosion that meets the IPEC (Indian Point Energy Center) EAL criteria 8.2.3. An Alert was declared at 1849 hours. The plant is stable in Mode 3 at this time. The Auxiliary Feedwater System actuated following the manual trip, as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). The Emergency Diesel Generators did not start as offsite power remained available and stable. The Station Auxiliary Transformer tap changer has remained at the maximum tap changer position and is inoperable, however the unit remains on offsite power and all electrical loads are stable. No primary or secondary relief valves lifted. The plant is in Hot Standby at normal temperature and pressure with residual heat removal using auxiliary feedwater and normal heat removal through the condenser via steam dumps. There was no radiation released. Indian Point Unit Three was not affected by this event and remains at 100% power. A post trip investigation is in progress. A one hour notification of the Alert Declaration was made to the NRC Emergency Operations Duty Center officer Howie Crouch at 1907 (hrs. EST). The licensee has notified the NRC Resident Inspector, the New York State Public Service Commission, New York State EMA and local county officials. Notified R1DO (White).
At 2218 EST, the licensee exited the Alert emergency declaration. The exit criteria was that: 1) The plant was in or was proceeding to cold shutdown, 2) No radiation release was in progress or anticipated 3) All radiation levels in the plant were stable or decreasing, and 4) No limitation on access to all plant areas. The licensee has notified the NRC Resident Inspector. Notified IRD (Gott), R1DO (White), NRR EO (Bahadur), FEMA (Heyman), DHS (Doyle), HHS (Gruenspecht), USDA (John) and DOE (Turner).
Approximately 50-100 gallons of oil from the transformer reached the Hudson River. The licensee notified the National Response Center and the New York State Department of Environmental Conservation. The licensee notified the NRC Resident Inspector. Notified the R1DO (White). | Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Main Transformer Residual Heat Removal | 05000247/LER-2010-009 |
| ENS 46649 | 1 March 2011 16:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generators Start on a Loss of Power from 138 Kv Circuit | At 1100 EST, Indian Point Unit 2 experienced a loss of offsite power from the 138 Kv circuit. All three Emergency Diesel Generators automatically started as required. All other plant systems functioned as required. Restoration of offsite power from the 13.8 Kv offsite circuit is in progress. Investigation into the loss of the 138 Kv circuit is ongoing. Indian Point, Unit 2 continues in Mode 1 at 100 % power. Indian Point, Unit 2 is in a 72 hour LCO due to a loss of 1 of 2 offsite circuits. Unit 3 was not affected. The licensee has notified the NRC Resident Inspector and will be notifying the Public Service Commission of the State of New York. | Emergency Diesel Generator | |
| ENS 47177 | 19 August 2011 22:05:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Partial Loss of Offsite Power Resulting in Edg Autostart | At 1805 Indian Point Unit 3 experienced a loss of normal 138 KV Offsite Power (LOOP) during a thunderstorm in the area. The Station Auxiliary Transformer (SAT) protection circuitry actuated and de-energized 6.9 KV buses 5 and 6. Thirty two and 33 Emergency Diesel Generators (EDGs) auto-started and loaded onto Safeguards buses 5A and 6A per plant design. The corresponding blackout logic resulted in an auto start of 32 and 33 Auxiliary Feed Water pumps (AFW). The plant remained in Mode 1; however, due to the loss of two 6.9 KV buses, three Circulating Water pumps (CW) tripped as designed, which caused main condenser backpressure to rise (i.e., vacuum degraded). As a result of rising condenser backpressure, a manual turbine load reduction to 75% power was performed. The reactor remains stable at 75% power and buses 5 and 6 have been manually re-energized from the alternate 13.8 KV offsite power source and the Emergency Diesel Generators have been unloaded, shutdown, and realigned for auto start capability. The only unusual system response was 32 Component Cooling Water (CCW) pump did not auto-start on a low pressure signal when the remaining pumps were stripped and re-Ioaded by the EDG sequencing logic due to a problem with the closing spring on the supply breaker. CCW system pressure was observed to be at normal levels when 31 and 33 CCW pumps were loaded on the EDGs. 32 CCW pump remains inoperable due to the breaker issue. There was no impact to the operation of Unit 2 during this event. The licensee has notified the NRC Resident Inspector and the New York State Public Service Commission. | Emergency Diesel Generator Auxiliary Feedwater Main Condenser | |
| ENS 47582 | 10 January 2012 09:36:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Excessive Leakage from Rcp Seal Return | Plant shutdown from 100% due to increased leakage from 21 RCP (Reactor Coolant Pump) seal return was commenced at 0100 EST. A manual reactor trip was actuated in accordance with plant procedures at 25% power. Auxiliary feedwater pump was placed in service prior to the reactor trip, per plant procedures. Decay heat is being removed via the condenser steam dumps. All offsite power sources are available. All equipment operated as expected. The licensee has notified the NRC Resident Inspector and other government agencies.
Subsequent review concluded the condition did not meet reporting criteria. A normal plant shutdown was initiated in accordance with plant procedures for a known problem with the 21 RCP # 1 seal whose seal leakoff rate had reached an administrative alarm limit. Shutdown was not required by the Technical Specifications. On January 10, 2012, at 0100 hours (EST) a plant shutdown commenced per plant operating procedure 2-POP-2.1. At 0436 hours, the reactor was manually tripped in accordance with procedure 2-POP-3.1. This event is not reportable as the reactor trip was part of a normal shutdown for corrective maintenance. The licensee has notified the NRC Resident Inspector. Notified R1DO (Powell). | Auxiliary Feedwater | |
| ENS 47999 | 6 June 2012 10:12:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 2 Experienced an Automatic Reactor Trip on Turbine Main Generator Trip | On June 6, 2012, at approximately 0612 hours, Indian Point Unit 2 reactor automatically tripped from 100% steady state power due to a turbine-main generator electrical trip. All rods fully inserted. Investigation into the generator trip is in progress. All systems responded as expected. The Auxiliary Feedwater system actuated and responded as expected and is maintaining steam generator levels. Decay heat removal is via the steam generators to the main condenser. Offsite power and plant electrical lineups are normal. No primary or secondary code safety relief valves lifted. The reactor is in Mode 3 and stable. Unit 3 was unaffected and remains at 100% power. The NRC Resident Inspector has been notified, and the licensee issued a press release. | Steam Generator Auxiliary Feedwater Decay Heat Removal Safety Relief Valve Main Condenser | |
| ENS 48454 | 30 October 2012 02:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Rps Actuation on Generator Trip | On October 29, 2012, at 2241EDT, the Reactor Protection System automatically actuated at 100% reactor power due to a direct electrical trip to the Unit 3 Main Turbine Generator. The generator trip resulted in a turbine/reactor trip. All control rods fully inserted on the reactor trip. All plant equipment responded normally to the unit trip. This is reportable under 10 CFR 50.72(b)(2)(iv)(B). The plant is stable in Mode 3 at this time. The Auxiliary Feedwater System actuated following the automatic trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). The Emergency Diesel Generators did not start as offsite power remained available and stable. The unit remains on offsite power and all electrical loads are stable. No primary or secondary relief valves lifted. The plant is in Hot Standby at normal operating temperature and pressure with decay heat removal using auxiliary feedwater to the steam generators and normal heat removal through the condenser via condenser steam dumps. There was no radiation released. Indian Point Unit 2 was not affected by this event and remains at 100% power. A post trip investigation is in progress. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Main Turbine Decay Heat Removal Control Rod | |
| ENS 48750 | 13 February 2013 18:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip on Low Steam Generator Water Level | At 1352 (hrs. EST), the Unit 2 CCR (Central Control Room) noted a trip of both heater drain tank pumps and entered Abnormal Operating Procedure 2-AOP-FW-1, 'Loss of Feedwater'. Prior to the event, Instrumentation and Controls personnel were performing testing on the heater drain tank level control system. Turbine load was reduced per plant procedures, however a manual reactor trip was initiated at 1355 due to an inability to maintain steam generator water levels. The team subsequently entered E-0, 'Reactor Trip or Safety Injection'. All control rods fully inserted. All safety systems responded as expected with the exception of source range detector N-31 and intermediate range detector N-35. N-31 and N-35 were declared inoperable. The auxiliary feedwater system actuated as expected and provided feedwater to maintain steam generator water level. Decay heat removal is via the steam generators to the main condensers. Offsite power and plant electrical lineups are normal. No primary or secondary code safety relief valves lifted. The reactor is in Mode 3 and stable. Unit 3 was unaffected and remains at 100% power. An investigation is in progress. Unit 2 is currently at normal operating pressure and temperature. The licensee plans to issue a press release on this event. The licensee notified the State of New York Public Service Commission and the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal Safety Relief Valve Main Condenser Control Rod | |
| ENS 48854 | 27 March 2013 10:01:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Safety Injection Actuation During Testing | On March 27, 2013 at 0601 EDT, the Safety Injection System automatically actuated while in Mode 3 during performance of I&C (Instrument and Control) testing due to faulty test equipment. All plant equipment responded normally to the safety injection. This is reportable under 10 CFR 50.72(b)(2)(iv)(A). The plant is stable in Mode 3 at this time. The Auxiliary Feedwater System actuated following the safety injection signal as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). The unit remains on offsite power and all electrical loads are stable. No primary or secondary relief valves lifted. The plant is in Hot Standby at 1140 psig and 395 degrees F with decay heat removal using auxiliary feedwater to the steam generators and normal heat removal through the atmospheric steam dumps. There was no radiation released. Indian Point Unit Two was not affected by this event and remains at 100% power. Notified NRC Resident Inspector. Notified NRC Emergency Operations Center Duty Officer. The Safety Injection was reset and all plant equipment was restored to normal alignment. Pressurizer level remained in the indication range during the Safety Injection. The cause of the Safety Injection is still under investigation, but appears to be related to a faulty jumper necessary for the test.
The event reported above resulted in additional reportable actuations under 10 CFR 50.72(b)(3)(iv)(A). These were: 1) RPS actuation, 2) Phase A containment isolation, 3) Containment fan cooler unit actuation, and 4) Emergency Diesel Generator actuation (start but did not load). The licensee notified the NRC Resident Inspector of the 1335 EDT update. Notified R1DO (Krohn). | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal | 05000286/LER-2013-005 |
| ENS 49171 | 3 July 2013 11:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Following Loss of Instrument Air to Main Feedwater Regulating Valves | On July 3, 2013, at 0741 EST, the Indian Point Unit 2 CCR received a trip of both Main Boiler Feed Pumps (MBFP) and entered 2-AOP�FW-1, Loss of Feedwater. The unit was manually tripped at 0741 per 2-AOP-FW�1 due to the trip of both MBFPs. Operators entered E-O, Reactor Trip or Safety Injection. All control rods fully inserted. All safety systems responded as expected. The Auxiliary Feedwater (AFW) System actuated as expected. Offsite power and plant electrical lineups are normal. No primary or secondary code safety valves lifted. The 23 and 24 Main Steam Isolation Valves (MSIV) failed closed as a result of the loss of Instrument Air (IA). The 21 and 22 MSIVs remain open with the Main Condensers being used for heat sink. The reactor is in Mode 3 and stable. Unit 3 was unaffected and remains at 100% power. Preliminary investigations determined a two inch copper IA line in the switchyard which is normally buried had a failed coupling causing loss of IA to the main feedwater regulating valves. The IA line traversed an excavated area of the switchyard going to the Auxiliary Boiler Feed Pump (ABFP) Building. AFW operated using the Nitrogen backup supply to ABFP control valves until Instrument Air was restored to the ABFP building. An investigation is in progress. The licensee notified the NRC Resident Inspector. | Feedwater | |
| ENS 49698 | 7 January 2014 02:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Low Steam Generator Water Level Due to a Failed Main Feedwater Reg Valve | On January 6th, 2014 at 2115 EST Indian Point Unit 3 experienced an Automatic Reactor Trip due to '33 Steam Generator Steam flow/Feed flow Mismatch.' Operators entered emergency procedure E-0, Reactor Trip or Safety Injection. All control rods fully inserted, all safety systems responded as expected, and off-site power remained in-service. No primary or secondary safety valves actuated due to the trip. This is reportable under 10CFR50.72(b)(2)(iv)(B). The main condenser is being used for heat sink. Unit 2 remains stable at 100% power. The Auxiliary Feedwater System actuated following the automatic trip as expected. This is reportable under 10CFR50.72(b)(3)(iv)(A). Investigation is underway to determine the cause of the 33 Steam Generator Mismatch condition. The licensee will inform local and other government agencies and issue a press release. The licensee has informed the State of New York and the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 49802 | 6 February 2014 20:53:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unanticipated Auxiliary Feedwater Pump Start After Securing Emergency Diesel Generator | This report is being made in accordance with 10CFR50.72(b)(3)(iv)(A) for an Auxiliary Feedwater System Actuation. The monthly surveillance on 31 Emergency Diesel Generator (EDG) was conducted on 6 February 2014. The EDG was unloaded and its output breaker opened at 1553 (EST). At this time, the Non-SI Blackout Logic Defeated indication in the control room changed state from 'not illuminated' (logic defeated) to 'illuminated' (logic not defeated) without operator action. The steam-driven 32 Auxiliary Boiler Feed Pump (ABFP) auto started but did not inject water into the steam generators. The discharge valves are normally closed. Operators verified normal steam generator levels and level control and that all 480VAC Safeguards buses remained energized, then secured 32 ABFP and placed it back into AUTO. Indian Point 3 remains at full power in Mode 1. This event did not cause any change in power. The Senior NRC Resident and the NY State Public Service Commission have been informed. | Steam Generator Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 50361 | 13 August 2014 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 Trip | On August 13, 2014 at 1157 EDT, the Indian Point Unit 3 Reactor Protection System automatically actuated at 100% power due to Over Temperature Delta Temperature logic. At the time of the trip, pressurizer pressure Channel 1 was in test for maintenance, though testing was suspended at this time for lunch. All control rods fully inserted on the reactor trip. All plant equipment responded normally to the unit trip. This is reportable under 10 CFR 50.72(b)(2)(iv)(B). The plant is stable in Mode 3 at this time. The Auxiliary Feedwater System actuated following the automatic trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). The Emergency Diesel Generators did not start as offsite power remained available and stable. The unit remains on offsite power and all electrical loads are stable. No primary or secondary relief valves lifted. The plant is in Hot Standby at normal operating temperature and pressure with decay heat removal using auxiliary feedwater to the steam generators, and normal heat removal through the condenser via condenser steam dumps. There was no radiation released. Indian Point Unit 2 was not affected by this event and remains at 100% power. A post trip investigation is in progress. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 51060 | 9 May 2015 21:50:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 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 Main Transformer Fire | At 1750 EDT (05/09/15,) Indian Point Unit 3 experienced a fire on the 31 Main Transformer resulting in a unit trip. An Unusual Event was declared at 1801 EDT. The onsite fire brigade was mobilized. Offsite fire fighting assistance was requested. The fire was reported extinguished at 1815 EDT. The reactor was shutdown by an automatic trip. Plant response to the trip was as expected with no complications. The 31 and 33 Auxiliary Feed Pumps are operating and feeding the steam generators. Accountability is being performed. The plant is stable in mode 3, all control rods fully inserted, with normal offsite electrical power, and decay heat is being released to the main condenser. There was no impact on Unit 2 which continues to operate at 100% power. The licensee has notified the NRC Resident Inspector and state and local authorities. Notified DHS SWO, FEMA OPS Center, DHS NICC Watch Officer, and Nuclear SSA via email.
Oil from 31 Main Transformer has spilled into the discharge canal and has made its way into the river. Plant personnel are sandbagging drains and release paths. IPEC (Indian Point Energy Center) has contacted its environmental contractor, who is expected onsite at 2100 EDT to assist with cleanup. The National Response Center was notified at 1945 EDT and issued notification number 1116011. A message was left with the Westchester County Department of Health at 1953 EDT. The NY State DEC (Department of Environment Conservation) was contacted at 1955 EDT and issued notification number 1501459. The licensee has notified the NRC Resident Inspector. Indian Point Unit 3 remains in an Unusual Event at this time. Notified R1DO (Schroeder).
Indian Point Unit 3 exited the Unusual Event at 2103 EDT. The basis for exiting the Unusual Event is that the fire is out and field operators report they have been successful in cooling the transformer. The licensee has notified the NRC Resident Inspector and state and local authorities. Notified R1DO (Schroeder), R1RA (Lew), NRR (Dean), NRR EO (Morris), NRR EO (Howe), and IRD (Grant). Notified DHS SWO, FEMA OPS Center, DHS NICC Watch Officer, and Nuclear SSA via email. | Steam Generator Main Transformer Main Condenser Control Rod | |
| ENS 51156 | 15 June 2015 23:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Turbine Trip | On June 15, 2015 at 1920 EDT, Indian Point Unit 3 received a Turbine trip which directly led to a Reactor trip. Operators entered (plant procedure) E-0, Reactor Trip or Safety Injection. All control rods fully inserted. All safety systems responded as expected. The Auxiliary Feedwater (AFW) system actuated as expected. Offsite power and electrical lineups are normal. No primary or secondary code safety valves lifted. All MSIVs are open and the Main Condensers are being used as the heat sink. The Reactor is in Mode 3 and stable. Unit 2 was unaffected and remains at 100% power. Preliminary investigation determined that Breaker Number 1 in the Ring Bus was intentionally opened (by plant personnel on switching orders from the district operator) due to a problem on W93 (output feeder from Ring Bus). Subsequently Breaker #3 went open and caused a Turbine/Reactor trip of the Unit. The licensee notified the NRC Resident Inspector.
The licensee previously identified "a problem on W93" which was a mistake and should have been stated as "a problem on W97". All other information was stated correctly. The NRC Resident Inspector has been notified. Notified R1DO (Bickett). | Auxiliary Feedwater Main Condenser Control Rod |