Semantic search
| Entered date | Site | Region | Reactor type | Event description | |
|---|---|---|---|---|---|
| ENS 56875 | 29 November 2023 17:02:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: This notification is being made per 10 CFR 50.72(b)(2)(xi), as a result of notifications made to State and local government agencies for the discovery of an oil sheen in the discharge canal outside Unit 3. The New York State Department of Environmental Conservation and Westchester County Department of Health were notified. No sheen was observed in the river or at the southern end of the discharge canal near the outfall gates. Clean up efforts are underway. The licensee will notify the NRC Project Manager. |
| ENS 56503 | 4 May 2023 12:25:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | The following summary was provided by the licensee via phone: On May 4, 2023 at 1145 EDT, the licensee found contact dose rates of 215 and 555 millirem-per-hour at 2 separate spots on the top of an exclusive use package during receipt survey. These dose rates are above the 200 millirem-per-hour allowable. No loose surface contamination was identified. The package contains tools from Holtec and was intact on delivery. The package has subsequently been secured in a locked radiation storage building. No overexposure or unauthorized exposure resulted to plant personnel. The licensee suspects shielding, internal to the package, may have shifted and the licensee will investigate further. Dose rates at one foot from the package were recorded at 65 millirem per hour. |
| ENS 54037 | 30 April 2019 07:37:00 | Indian Point | NRC Region 1 | A non-licensed employee supervisor had a confirmed positive test for a prohibited substance during a follow-up fitness-for-duty test. The individual's unescorted access to the plant has been terminated. The NRC Senior Resident Inspector was notified by the licensee. | |
| ENS 53954 | 24 March 2019 17:40:00 | Indian Point | NRC Region 1 | On March 24, 2019, at 1445 EDT, Indian Point Unit 2 automatically tripped on a turbine trip due to a loss of excitation. All control rods fully inserted and plant equipment responded normally to the unit trip. This RPS (reactor protection system) actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B). The auxiliary feedwater system actuated following the automatic trip as expected. This specified system actuation is reportable under 10 CFR 50.72(b)(3)(iv)(A). During the event offsite power remained available and stable. No primary or secondary reliefs lifted. Unit 2 is in Mode 3 at normal operating temperature and pressure. Decay heat removal is via the steam generators to the atmospheric steam dumps. No radiation was released. Indian Point Unit 3 was unaffected by this event and remains defueled in a scheduled refueling outage. A post trip investigation is in progress. The licensee has notified the NRC Resident Inspector The New York State Public Service Commission, Consolidated Edison System Operator, and New York State Independent System Operator were also notified. | |
| ENS 53937 | 15 March 2019 13:39:00 | Indian Point | NRC Region 1 | On March 15, 2019 at 1300 EDT, Indian Point Unit 2 automatically tripped offline from mode 1 - 100% power operations. Reactor Operators verified the reactor trip and the plant is currently stable in mode 3. All automatic systems functioned as required. The auxiliary feedwater system actuated following the trip, as expected. All control rods fully inserted upon the trip, as expected. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(2)(iv)(B). The unit remains on offsite power in hot standby at normal operating temperature and pressure. Decay heat is being removed from the steam generators via the auxiliary feedwater system and the condensate steam dump valves. Unit 3 remains in mode 6 for a scheduled refueling outage. The licensee notified the NRC Resident Inspector, the local transmission company, and New York State Independent System Operator. The Indian Point Unit 2 automatic trip was caused by the trip of the main generator. The cause of the generator trip is unknown at this time. | |
| ENS 53929 | 12 March 2019 11:21:00 | Indian Point | NRC Region 1 | A contract employee failed to report for a random fitness for duty test. The contractor's access to the plant has been terminated. The licensee notified the NRC Resident Inspector and the NY Public Service Commission. | |
| ENS 53863 | 6 February 2019 12:46:00 | Indian Point | NRC Region 1 | On February 05, 2019 at approximately 1800 EST, candy that contained alcohol was discovered in the plant protected area. The candy was removed from the protected area by station security management. The licensee notified the NRC Resident Inspector and the State of New York Public Service Commission. | |
| ENS 53699 | 28 October 2018 15:10:00 | Indian Point | NRC Region 1 | During the performance of Service Water Essential header swap, SWN-6 (Supply to Turbine Building Oil Coolers) valve stem became disconnected from its gear box at 85% open and could not be operated. Therefore, the non-essential service water system was inoperable. LCO 3.0.3 was entered at 0930 (EDT) with required actions to be in Mode 3 in 7 hours, Mode 4 in 13 hours and Mode 5 in 37 hours. Repair efforts were successful at shutting SWN-6, and LCO 3.0.3 was exited at 1305 (EDT) before adding any negative reactivity in support of shutdown. ('TS Required S/D' box not checked.) This condition constituted a loss of safety function which requires an 8 hour report (in accordance with) IAW 50.72(b)(3)(v)(B): Without the ability to close SWN-6, the non-seismic portion of the conventional Service Water System could not be isolated as required in the event of either a seismic event or as required in the EOPs. The nonessential service water system is required to support the recirculation phase post (Design Basis Accident) DBA for accident mitigation. The licensee notified the NRC Resident Inspector and the State of New York. | |
| ENS 53611 | 18 September 2018 09:06:00 | Indian Point | NRC Region 1 | Due to a steam leak on the reheater line to 36C Feedwater Heater, operators initiated a manual trip of the reactor, verified the reactor trip, and closed all Main Steam Isolation Valves. The plant is currently stable in Mode 3 with the steam leak isolated. The Auxiliary Feedwater System actuated following the trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(2)(iv)(B). The unit remains on offsite power in Hot Standby at normal operating temperature and pressure. Decay heat is being removed from the steam generators via the Auxiliary Feedwater System and atmospheric steam dumps. Unit 2 was unaffected and remains at 100 percent power. The licensee notified the NRC Resident Inspector, the State of New York, and the local transmission company. | |
| ENS 53601 | 13 September 2018 08:24:00 | Indian Point | NRC Region 1 | A hurricane is within 500 nautical miles from Indian Point Energy Center with wind speed in excess of 87 knots. The National Weather Service has issued a hurricane warning for a hurricane with wind in excess of 87 knots (approximately 100 mph) within 500 nautical miles of the facility. Per the Technical Requirement Manual a prompt report shall be made to the NRC Incident Response Center within 1 hour of receipt of that hurricane warning. The licensee has notified the NRC Resident Inspector. | |
| ENS 53589 | 7 September 2018 20:56:00 | Indian Point | NRC Region 1 | (Indian Point Unit 3) entered Technical Specification 3.0.3 for Safety Injection Boron Injection Tank Header Inoperable due to a thru weld leak at the thermal well rendering two Safety Injection Pumps Inoperable. Plant shutdown was started 1803 hours EDT. The licensee plans to be in Mode 4 at 0600 EDT on 9/8/2018. There is no impact to Unit 2. The licensee notified the NRC Resident Inspector and the State of New York. | |
| ENS 53578 | 1 September 2018 13:52:00 | Indian Point | NRC Region 1 | Twelve (12) cans of an alcoholic beverage were discovered inside the protected area. The 12 cans were removed from the protected area and taken to and secured by site station security. The licensee notified the NRC Resident Inspector. | |
| ENS 53348 | 19 April 2018 23:41:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | While performing a turbine startup, a turbine control anomaly caused a steam generator level transient. The rise in steam generator level above the setpoint caused the turbine to automatically trip. The high steam generator level of 73 percent caused a feedwater isolation signal at 2107 EDT, which also tripped both Main Boiler Feed Pumps. The tripping of the Main Boiler Feed Pumps auto started the motor driven Aux Boiler Feed Pumps 21 and 23. The reactor was manually tripped at 2108 EDT in accordance with AOP-FW-1 Loss of Main Feedwater. All control rods inserted. Electrical power is being provided from offsite via the Station Aux Transformer. Decay heat removal is being provided via the Atmospheric Dump Valves. An investigation into the cause of the turbine control anomaly is underway. The NRC Resident Inspector has been notified. The event did not have an affect on Unit 3 and there is no primary to secondary leakage. |
| ENS 53305 | 31 March 2018 19:33:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | On March 31, 2018, during the Indian Point Unit 2 refueling outage, with the reactor defueled and the head removed and located on the head stand, and all fuel from the reactor vessel removed and located in the spent fuel pool, while performing planned examinations on the 97 reactor vessel head penetrations, it was determined that one penetration could not be dispositioned as acceptable per the requirements of 10CFR50.55a for the reactor coolant system pressure boundary. The examinations are being performed to the meet the requirements of 10CFR50.55a(g)(6)(ii)(D), and ASME Code Case N-729-4, to find potential flaws/indications well before they increase to a degree that could potentially challenge the reactor vessel head pressure boundary. All other reactor vessel head penetrations have had a bare metal visual inspection completed with no other indications identified. The station is currently performing the remaining non-destructive examinations required by Code Case N-729-4. Repairs are currently being planned, and will be completed prior to entering Mode 5 from the current refueling outage. This is reportable, pursuant to 10CFR50.72(b)(3)(ii)(A) since the as found indications did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-4 to remain in-service without repair. The NRC Resident Inspector has been informed. The licensee has also notified the NY Public Service Commission. |
| ENS 53216 | 16 February 2018 04:43:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | On February 16, 2018 at 02:01 EST Indian Point Unit 3 automatically tripped on a turbine trip due to a loss of main generator excitation. All control rods fully inserted and all plant equipment responded normally to the unit trip. This is reportable under 10 CFR 50.72(b)(2)(iv)(B). The auxiliary feedwater system actuated following the automatic trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). During the event offsite power remained available and stable. No primary or secondary reliefs lifted. The plant is stable, in Mode 3, at no load operating temperature and pressure. Decay heat removal is via the steam generators to the main condenser via the condenser steam dumps. No radiation was released. Indian Point Unit 2 was unaffected by this event and remains at 100 percent power. A post trip investigation is in progress. The licensee has notified the NRC Resident Inspector. |
| ENS 53052 | 3 November 2017 21:08:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | On November 3rd, 2017 at 2022 EDT, the Indian Point Unit 3 Reactor Protection system automatically actuated at 100 percent power. Annunciator first out indication was from 33 SG (Steam Generator) Low Level. This automatic reactor trip is reportable to the NRC under 10 CFR 50.72(b)(2)(iv)(B). All control rods fully inserted on the reactor trip. All safety systems responded as expected. The Auxiliary Feedwater System actuated as expected. Offsite power and plant electrical lineups are normal. All plant equipment responded normally to the unit trip. No primary or secondary code safeties lifted during the trip. 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. Unit 3 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 the high pressure steam dumps. Unit 2 was unaffected and remains at 100 percent power. A post trip investigation is in progress. The licensee indicated that Radiation Monitor number 14 spiked twice during the transient, however, is currently not indicating any signs of radiation. The licensee will notify the NRC Resident Inspector and the NY Public Service Commission.
The initial notification stated that Indian Point Unit 3 Reactor Tripped on 33 SG (Steam Generator) Low Level, this is incorrect. Indian Point Unit 3 Reactor Tripped on a Turbine Trip. The Turbine Trip was caused by a Generator Back-up Lockout Relay. The Turbine Trip was the 'first' annunciator first-out but was acknowledged instead of silenced during initial operator actions. The Turbine Trip first-out being acknowledged allowed a Low Steam Generator first-out to later annunciate. A Low Steam Generator Level is an expected condition post trip. This update does not change any actions taken by the operating team or required notifications under 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). A post trip investigation remains in progress. The licensee will notify the NRC Resident Inspector and the NY Public Commission. Notified the R1DO(Cook) |
| ENS 52836 | 30 June 2017 09:04:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | On June 30, 2017 at time 0130 (EDT), the Condensate Storage Tank (CST) was declared inoperable per Technical Specification 3.7.6. A pin hole sized through wall leak was discovered on the downstream side of CD-123, 32 Auxiliary Boiler Feed Pump Bearing Cooling Relief Valve, which is unisolable to the CST. No type of Non-Destructive Examination (NDE) can be used to effectively characterize the defect. Based on the inability to characterize the defect and the fact that the degradation mechanism is not readily apparent, the valve is considered inoperable because the valve has lost its ability to maintain a pressure boundary. As a result, the CST is inoperable. Currently, the CST has 580,000 gallons of water contained in the tank, which is well above the minimum required amount of 360,000 gallons. City water is the backup means to supply water to the Auxiliary Feedwater System, and this has been verified to be operable in accordance with the Actions of the Technical Specification. The CST must be restored to operable status within 7 days. The CST provides cooling water to remove decay heat and the minimum amount of water in the Condensate Storage Tank is the amount needed to maintain the plant for 24 hours at hot shutdown following a trip from full power. The CST satisfies Criteria 2 and 3 of 10 CFR 50.36. This event was determined to be reportable as a Loss of Safety Function pursuant to 10 CFR 50.72(b)(3)(v)(B). The licensee will notify the NRC Resident Inspector and the State of New York. |
| ENS 52829 | 26 June 2017 18:39:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | On June 26, 2017, at 1531 (EDT), Indian Point Unit 2 inserted a manual reactor trip prior to Steam Generator levels reaching the automatic reactor trip setpoint. Steam Generator water level perturbation resulted from a loss of 22 Main Boiler Feed Pump. All Control Rods verified inserted. The Auxiliary Feedwater System started as designed and supplied feedwater to the Steam Generators. Heat removal is via the Main Condenser through the High Pressure Steam Dumps. Offsite power is being supplied through the normal 138kV feeder 95332. The cause of the 22 Main Boiler Feed Pump loss is currently under investigation. Entergy is issuing a press release/news release on this issue. Unit 2 is stable and in Mode 3. There was no impact on Unit 3. The licensee notified the State of New York and the NRC Resident Inspector. |
| ENS 52801 | 11 June 2017 14:35:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | At Indian Point Energy Center (IPEC), Unit 3 normal letdown was isolated due to a significant body/bonnet leak on the inlet valve to Reactor Coolant Filter. Shift team took action to isolate letdown and stop the leak. The Abnormal Operating Procedure (AOP) was entered, normal letdown was isolated per procedure, and excess letdown was placed in service to balance inventory at 61 percent Pressurizer Level. This was above the Technical Specification (3.4.9) level of 54.3 percent which was exceeded at 0911 (EDT), putting the unit in a 6-hour shutdown action statement. The valve body/bonnet was torqued, successfully eliminating the leakage. Pressurizer Level was restored to the normal control band and the AOP was exited at 1136. The SM (Shift Manager) estimated the leakage at 18 gallons per minute when leak was active. No EAL (Emergency Action Level) thresholds were exceeded. This occurrence is considered a safety system functional failure per 10 CFR 50.72(b)(3)(v)(A), requiring an 8-hour NRC report. Both IPEC units are stable at full power." The licensee has notified the NRC Resident Inspector and the New York State Public Service Commission. |
| ENS 52751 | 14 May 2017 15:32:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | On May 14, 2017, with the Unit in Mode 4, it was identified that a single flow barrier access point (Gate C) to the 46 ft. elevation of Containment was unbolted for access to the inner crane wall to perform surveillance testing. This single flow barrier access point is required to be bolted closed to ensure the Emergency Core Cooling System (ECCS) operability basis which requires the sump barrier system to be operable in Modes 1- 4. The sump barrier system is required to prevent the transport of debris to the recirculation and containment sumps. This event is a safety system functional failure as the condition could have prevented adequate post-accident core cooling due to Design Basis Accident debris blockage of the recirculation and/or the containment sump and is reportable under 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector and the New York Public Service Commission. |
| ENS 52688 | 18 April 2017 14:57:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | A non-licensed employee supervisor had a confirmed positive for a prohibited substance during a random fitness-for-duty test. The individual's unescorted access to the plant has been suspended. The licensee has notified the NRC Resident (Inspector). |
| ENS 52388 | 21 November 2016 21:22:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | At 1738 (EST) on November 21, 2016, a leak was identified inside the Vapor Containment building on a Service Water line associated with 24 Fan Cooler Unit. The leak was isolated at 1743 by shutting the Service Water Isolation valves to 24 Fan Cooler Unit. This isolation meets the Technical Specifications of 3.6.1 Condition A Required Action. The leaking defect could have resulted in post-LOCA air leakage out of containment in excess of that allowed by Technical Specification 3.6.1 (Containment) which requires leakage rates to comply with 10 CFR 50, Appendix J. This event had no effect on the health and safety of the public. This event is being reported under 10 CFR 50.72(b)(3)(v) and the guidance of NUREG 1022, section 3.2.7 as a loss of safety function. The service water was quantified at approximately 15 gpm. The licensee notified the NRC Resident Inspector. |
| ENS 52353 | 8 November 2016 08:59:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | On November 8, 2016, at approximately 0840 EST, plant personnel reported an explosion within the Protected Area resulting in a fire with potential damage to plant structures or equipment. An Unusual Event was declared at 0851 EST. The onsite Fire Brigade was mobilized. The fire was extinguished at 0848 EST. The explosion was to the 138 kV power cross connect cable between the Unit 2 and 3 Station Auxiliary Transformers. Both Units are operating normally. There was no release of radioactive material. An investigation of the event is in progress. Offsite assistance was not required. Unit 2 was in a normal electrical line-up and Unit 3 was not affected. The licensee has notified the NRC Resident Inspector, State, Local, and Other Government Agencies. The licensee will issue a press release. Notified DHS SWO, FEMA OPS Center, FEMA National Watch (email only), DHS NICC, Nuclear SSA (email only).
The Unusual Event was terminated at 0946 EST based on the fire being out. The licensee has notified the NRC Resident Inspector. Notified R1DO (McKinley), IRD (Gott), and NRR EO (Miller). Notified DHS SWO, FEMA OPS Center, FEMA National Watch (email only), DHS NICC, Nuclear SSA (email only). |
| ENS 52344 | 3 November 2016 08:54:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | At 0300 EDT on November 3, 2016, it was identified that a Service Water leak existed in the Vapor Containment Building of Indian Point Unit 3. The leak was determined to be from 31 Fan Cooler Unit and was subsequently isolated at 0344 EDT by shutting the Service Water isolation valves to 31 Fan Cooler Unit. This isolation meets the Technical Specification 3.6.1 Condition A Required Action. The leaking defect could have resulted in post-LOCA air leakage out of containment in excess of that allowed by Technical Specification 3.6.1 (Containment) which requires leakage rates to comply with 10 CFR 50, Appendix J. This event had no effect on the health and safety of the public. This event is being reported under 10 CFR 50.72(b)(3) and the guidance of NUREG 1022, section 3.2.7 as a loss of safety function. The licensee notified the NRC Resident Inspector. |
| ENS 52254 | 21 September 2016 09:20:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | At 0221 (EDT) on 9/21/16, Operators at Unit 2 Secured the 21 Component Cooling Water (CCW) Pump for planned maintenance while 22 and 23 CCW pumps were in operation. When the 21 pump was secured, the discharge check valve failed to seat. This resulted in a low system pressure and reverse rotation of the 21 CCW Pump due to the discharge of the 22 and 23 CCW pumps to a common header. When system pressure dropped below 107 psig the 21 CCW pump received an auto start signal. Due to the reverse rotation, the 21 CCW pump tripped on overcurrent. Reactor Operators directed Field Operators to manually shut the 21 CCW Pump discharge valve. The 21 CCW pump Discharge Valve was closed at 0223 (EDT). This action was successful in stopping the reverse flow and restoring system parameters. During this two minute period the CCW system was declared inoperable and LCO 3.0.3 was entered. Unit 2 exited LCO 3.0.3 at 0223 (EDT) after observing system pressure and flow return to normal. The declaration of inoperability on the CCW system is considered a Loss of Safety Function for purposes of reporting under 50.72(b)(3)(v)(D). There was no reduction in power while in LCO 3.0.3 and no other issues arose. The Licensee notified the NRC Resident Inspector. The Licensee notified the Public Service Commission.
Indian Point Unit 2 is retracting the 8-hour non-emergency notification made on September 21, 2016, at 0920 EDT (EN#52254). The notification on September 21, 2016, reported a safety system functional failure (SSFF) as a result of declaring the Component Cooling Water System (CCW) inoperable due to failure of the 21 CCW pump discharge check valve (761C) to close. This condition was discovered during planned maintenance after securing the 21 CCW pump while the 22 and 23 CCW pumps were in operation. When the 21 CCW pump was secured, the discharge check valve failed to seat. This resulted in a low system pressure and reverse rotation of the 21 CCW pump due to the discharge of the 22 and 23 CCW pumps to a common header. Condition was reported as a safety system functional failure (SSFF) under 10 CFR 50.72(b)(3)(v)(D). After further investigation of the condition, a revised calculation was prepared for the CCW hydraulic model which is used to analyze CCW system performance for normal and DBA (design basis accident) modes of operation and documented in a calculation. The new calculation included the as-found condition of the 21 CCW pump discharge check valve failure to seat. Based on the results of the new calculation, the CCW system is capable of performing its design basis heat removal function during a design basis accident. Calculated flow rates with CCW aligned for Post-LOCA recirculation demonstrates that with failed open check valve 761C, the 22 CCW pump and 23 CCW pump have adequate NPSH margin, are operating below analyzed pump run out and deliver flow to the CCW system that is significantly greater than the flow required for post-LOCA recirculation. Therefore the CCW system was operable and a safety system functional failure (SSFF) did not occur as a result of failed open 21 CCW pump discharge check valve 761C. The licensee has notified the NRC Resident Inspector and will be notifying the New York Public Service Commission. Notified R1DO (Bickett). |
| ENS 52224 | 7 September 2016 16:47:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | At 1347 EDT on September 6, 2016, the non-essential service water header was declared inoperable due to what appeared to be a leak on a filet weld on a slip-on flange upstream of the 23 service water pump discharge valve SWN-2-2. (Operators) entered LCO 3.0.3 based on Engineering input with no code case to support an immediate determination of operability. The 23 service water pump was secured and the discharge valve SWN-2-2 was closed isolating the leak and restoring the non-essential SW header to operable status. With the discharge valve closed, LCO 3.0.3 was exited at 1417 EDT on September 6, 2016. The declaration of the inoperability of the non-essential service water header is considered a loss of safety function for purposes of reporting under 10 CFR 50.72(b)(3)(v). Further investigation of the leak revealed (that the leak had occurred at) a pipe flaw rather than at the weld location. An operability evaluation is ongoing. There was no reduction in power while in LCO 3.0.3 and no other issues arose. The licensee notified the NRC Resident Inspector.
Indian Point Unit 2 is retracting the 8-hour non-emergency notification made on September 7, 2016, at 1647 hours EDT (EN #52224). The notification on September 7, 2016, reported a safety system functional failure (SSFF) as a result of declaring the non-essential service water (SW) header inoperable due to discovery of a through wall leak on the discharge line from the 23 SW Pump. The leak appeared to be on the downstream fillet weld on the north side of the pipe spool piece. At the time the exact location of the leak could not be established due to the geometry of the area and its carbon fiber coating. Because there is no qualified NDE inspection technique for fillet welds and the analytical methods provided in ASME code Case N-513-3 are not applicable to fillet welds, operability of the effected section of the SW System could not be demonstrated and the SW piping was conservatively considered inoperable. Further investigation of the leak determined the leak occurred mid-pipe 1/2 inch upstream from a slip-on flange socket weld and not on the weld itself. The weld location is ASME Section XI Class 3, nuclear safety-related, Seismic Class 1. An engineering evaluation was performed of the pipe defect and concluded the identified through-wall defect on the pipe spool piece did not result in any structural, flooding, spraying condition or SW system capacity issues which would adversely impact any normal or accident design functions of the SW system or any other nuclear safety-related equipment or component to perform their design safety functions from the time of the leak occurrence to the isolation of the pipe. The structural portion of the evaluation was performed using the ASME CC N-513-3 methodology but the code case was not invoked because the flaw was isolated and a code repair was made. The 23 SW Pump discharge line spool piece is a 14 inch NPS schedule STD cement-lined carbon steel pipe having a nominal wall thickness of 0.375 inches with slip-on flanges welded on both sides with fillet welds on the exterior. The function of this spool piece is to connect the 23 Zurn Strainer outlet to the isolation valve SWN-2-2. The degradation mechanism was likely due to corrosion of the carbon steel on the underside of the cement lining due to a lining defect. The licensee will notify the NRC Resident Inspector. Notified R2DO (Krohn). |
| ENS 52067 | 6 July 2016 13:16:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | On July 6, 2016 at 0938 EDT Indian Point Unit 2 experienced a trip from 100% steady state power during the performance of reactor protection testing. The cause of the trip is under investigation. All control rods fully inserted and all systems responded as expected. The auxiliary feedwater system actuated as expected on a low level in the steam generators which occurs as a result of a trip from full power. Auxiliary feedwater is maintaining steam generator levels and decay heat removal is via the steam generators to the main condenser. Offsite power and plant electrical line ups are normal with the exception of 13.8kV feeder 33332 which remains out of service during the replacement of breaker BT4-5. No primary or secondary code safety relief valves lifted. The reactor is in Mode 3 and stable. Indian Point Unit 3 was unaffected and remains at 100% steady state power. The NRC Resident Inspector has been notified. The licensee notified the New York Public Service Commission and the New York Independent System Operator. Indian Point indicated they have issued a press release regarding the event. |
| ENS 52039 | 24 June 2016 04:05:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | At 0400 (EDT) on June 24, 2016, Indian Point Unit 2 initiated actions to commence reactor shutdown to comply with Technical Specification (TS) LCO 3.7.7, Condition B. TS LCO 3.7.7, Condition A had been entered at 0230 on June 21, 2016 in order to repair a leaking weld on the 20 inch service water pipe to nozzle weld on the 21 Component Cooling Water Heat Exchanger (CCW HX). Condition A allows 72 hours to restore the inoperable CCW train to service or Condition B is entered which requires the plant to be in Mode 3 in 6 hours and Mode 4 in 12 hours. The initiation of a nuclear plant shutdown required by TS requires a 4-hour report in accordance with 50.72(b)(2)(i) which is being made by this notification. The licensee notified the New York Independent System Operator and the New York Public Service Commission. The licensee notified the NRC Resident Inspector. |
| ENS 51829 | 29 March 2016 17:04:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | This is a non-emergency report. Indian Point Unit 2 is shut down for a scheduled refueling outage and is performing an industry required MRP-227-A inspection of the reactor vessel internals. During this inspection, Entergy identified baffle/former bolts with either visual anomalies or ultrasonic indications. All vessel internal examinations have been successfully completed with no anomalies other than the baffle/former bolts. Entergy will be developing a repair plan that will be implemented prior to startup. The unit is in a safe and stable condition. No leakage was detected as a result of this condition. On March 29, 2016, this event was determined to be reportable pursuant to 10 CFR 50.72(b)(3)(ii)(B), since the as-found conditions were not previously analyzed. Entergy is planning a press release to communicate this condition to our stakeholders. Therefore this is also being reported under 10 CFR 50.72(b)(2)(xi) since a news release is planned. The NRC Resident Inspector has been informed. The licensee will notify State and local government agencies as appropriate. |
| ENS 51775 | 7 March 2016 17:12:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | During Emergency Diesel Generator (EDG) surveillance testing, the normal 480V Bus 3A supply breaker tripped open. All EDGs auto-started as per plant design. During restoration of normal power to the 480V Buses, 23 EDG output breaker opened on overcurrent. All plant systems responded as per design. There was no loss of core cooling since steam generators were coupled with an operational RCP. No other ESF (Engineered Safety Feature) equipment automatically started. The 480V Bus 3A normal supply breaker was replaced and the cause of both issues is under investigation. Indian Point Unit 2 remains in a stable MODE 5 condition with all 480V Buses energized via normal power. The NRC Resident has been informed. The licensee will inform the New York Public Service Commission. |
| ENS 51724 | 10 February 2016 12:20:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | This report is being made pursuant to 10 CFR 50.72(b)(2)(xi) for any event or situation related to the health and safety of the public or on-site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. On February 10, 2016 at approximately 0916 EST, Entergy provided a news release of updated findings from follow-up groundwater tests at Indian Point that confirm anticipated fluctuations in tritium levels. The most recent samples from on-site groundwater monitoring wells show elevated levels of tritium. The levels of tritium identified including the new readings pose no threat to public health and safety. The tritium levels remain less than one-tenth of one percent of Federal reporting guidelines. Entergy continues to provide voluntary notifications to state agencies and stakeholders. Sampling will continue to be taken regularly from the monitoring wells in accordance with plant procedures. The licensee notified the NRC Resident Inspector. |
| ENS 51606 | 14 December 2015 19:49:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | At 1906 (EST) on 12/14/2015, Indian Point Unit 3 received a Main Generator Lockout trip signal, and the reactor automatically tripped. Site personnel reported seeing arcing on a 345kV output transmission line tower. At the time of the trip, there was moderate rain and fog in the area. The site fire brigade leader investigated the reports of arcing and found no evidence of fire; fire brigade response was not required. All automatic systems functioned as designed and all control rods inserted automatically. Auxiliary Feedwater Pumps started automatically due to expected low steam generator levels following a reactor trip from 100% power. Unit 3 is being maintained in Mode 3 with decay heat removal via steam dumps to the condenser. Offsite power remains available and in service from 138kV to the 480V safeguards buses. The cause of the Main Generator Lockout signal is being investigated. Unit 2 was not impacted and continues to operate at 100% power. The NRC Resident Inspector has been notified. After the trip, operators observed high vibrations on the 33 reactor coolant pump which eventually returned to normal range. The licensee will be notifying the New York Public Service Commission and their local Independent System Operator. A press release will be issued by the Communications Department. |
| ENS 51586 | 5 December 2015 18:48:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | At 1731 (EST) on December 5, 2015, Indian Point Unit 2 Control Room operators initiated a Manual Reactor Trip due to indications of multiple dropped Control Rods. The initiating event was a smoldering Motor Control Center (MCC) cubicle in the Turbine Building that supplies power to the Rod Control System. The unit is stable in Mode 3 with heat sink provided by Auxiliary Feedwater and decay heat removal is via the steam dumps to the condenser. Offsite Power remains in service. The smoldering MCC cubicle had power removed from it when 24 MCC breaker tripped on overcurrent. The affected cubicle has ceased smoldering and is being monitored by on-site Fire Brigade trained personnel. The trip of 24 MCC removed power to 22 Battery Charger, 22 DC Bus remained powered from the 22 Battery without interruption, and 22 Battery Charger was subsequently repowered. The cause of the smoldering MCC is being investigated and a post reactor trip evaluation is being conducted by the licensee. There was no impact on Unit 3, which continues to operate at 100% power. The licensee has notified the NRC Resident Inspector and appropriate State and Local authorities. |
| ENS 51329 | 18 August 2015 19:38:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | At 1331 hours (EDT) on August 18, 2015 with the plant in mode 1, IP2 entered TS (Technical Specification) 3.0.3 upon determination by the Shift Manager that MOV-746 and 747 at the outlet of the Residual Heat Removal (RHR) Heat Exchangers may not (fully) open on an Sl (Safety Injection) signal if there was a degraded grid voltage (where the voltage at the 480V Safeguards Buses is below the minimum drop-out value of 415V and above the loss-of-voltage value of 206.6V). The MOV 746 and 747 valves are normally closed so the Sl signal with degraded voltage present could cause the fuses to fail. The time when the fuses would fail and the extent to which the MOVs open has not yet been analyzed. Immediate corrective actions was taken to replace the fuses with fuses that would not fail. The RHR trains were restored to an operable condition at 1419 and 1431 hours. This event is potentially reportable under 10 CFR 50. 72(b)(3)(ii) and 10 CFR 50.72(b)(3)(v) since the condition has not yet been fully analyzed but has been corrected. The plant remained at 100% power during the time of this event. All technical specification requirements were followed. The licensee will notify the NRC Resident Inspector and the State. |
| ENS 51211 | 8 July 2015 15:37:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | Indian Point Unit 3 was manually tripped at 1427 EDT due to lowering steam generator water levels. At 1425 EDT, #31 condensate pump tripped, causing the lowering water levels. There were no immediate complications on the trip and the unit is stable in Mode 3. Auxiliary feedwater actuated as expected and is in service. All rods inserted and decay heat is being rejected to the condensers. Offsite electrical power is in service. Unit 2 is stable at 100% power. The licensee plans on issuing a press release. The licensee notified the NRC Resident Inspector and New York Public Service Commission. |
| ENS 51156 | 15 June 2015 20:15:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | 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). |
| ENS 51060 | 9 May 2015 18:19:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | 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. |
| ENS 51046 | 7 May 2015 07:11:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | At 0700 on 5/7/15, Indian Point Unit 3 commenced a shutdown due to the inability to isolate a steam leak on a feedwater instrument line. Offsite power is available. Indian Point Unit 2 is unaffected and remains in Mode 1 at 100% power. The NRC Resident Inspector, the NYISO (New York Independent System Operator), and NY Public State Commission have been notified. |
| ENS 50939 | 30 March 2015 12:23:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | The licensee reported that there was an unscheduled removal of service of the Unit 3 Main Generator automatic voltage regulator. This is required to be reported per IP-SNM-LI-108, "Event Notification and Reporting". This does not significantly affect the unit operation. The investigation into the loss of automatic voltage regulation is under investigation. The licensee notified the New York Public Service Commission and the New York Independent System Operator. The licensee will be notifying the NRC Resident Inspector. |
| ENS 50879 | 11 March 2015 11:54:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | A non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector was notified and the NY Public Service Commission will be notified. |
| ENS 50722 | 8 January 2015 10:09:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | At 0400 (EST), Indian Point Unit 3 entered LCO 3.5.4 Condition C due to both Refueling Water Storage Tank (RWST) Low Low Level Alarm channels failing. The failure was a result of freezing in the instrument lines. LCO 3.5.4 Condition C requires at least one channel of RWST Low Low Level Alarm be restored to Operable within 1 hour. At 0500, Unit 3 entered Condition D due to the required action and completion time of Condition C not being met. This requires the unit be placed in Mode 3 in 6 hours and Mode 4 in 12 hours. This failure also has resulted in a loss of safety function. Operators rely upon the RWST Low Low Level Alarms during an accident to alert them of the need to transfer injection from the RWST as a source of water to the containment sump. At 0700, Indian Point Unit 3 commenced a shutdown to be in compliance with the requirements of LCO 3.5.4 Condition D. Investigation and repair efforts were immediately put in place to correct the failure and return the function to operable. At 1000, one Channel of RWST Low Low Level Alarm was returned to operable. LCO 3.5.4 Conditions C & D were exited. The shutdown was stopped and safety function was restored. Unit 3 remains in a 7-day shutdown (LCO action statement) for one RWST Low Low Level Alarm Channel inoperable as per LCO 3.5.4 Condition B. Unit 2 is unaffected and remains at 100% power. The State of New York and the NRC Resident Inspector were notified. |
| ENS 50467 | 18 September 2014 13:01:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | During the performance of PT-Q89, Control Rod Exercise Test, Rod G-3 in Shutdown Bank 'B' misaligned (stepped in) to approximately 195-200 steps indicated with group demand at 217 steps. The movable gripper fuse was suspect, found not blown (so further) troubleshooting required. At 1150 (EDT), the decision was made to reduce power to approximately 70% within two hours per I.T.S. (Improved Technical Specification) 3.1.4.b.2.2. The plant is currently stable at 68% power. I.T.S. 3.1.5.b.1 was also entered for Shutdown Bank 'B' less than allowable insertion limits at time 1228 (EDT) to be in Mode 3 in 6 hrs. (1828 (EDT)). The licensee notified the New York Independent System Operator, the New York Public Service Commission and the NRC Resident Inspector. |
| ENS 50361 | 13 August 2014 13:06:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | 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. |
| ENS 50258 | 3 July 2014 21:40:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | A hurricane is within 500 nautical miles from Indian Point Energy Center with wind speed in excess of 87 knots. The National Weather Service has issued a hurricane warning for a hurricane with wind in excess of 87 knots (approximately 100 mph) within 500 nautical miles of the facility. Per Technical Requirement Manual 5.4.A, a prompt report shall be made to the NRC Incident Response Center within 1 hour of receipt of that hurricane warning. The licensee notified the NRC Resident Inspector. |
| ENS 50250 | 2 July 2014 14:49:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | On July 1, 2014, at approximately 1500 EDT, as a result of an investigation into a report that the electronics associated with the air operated source controller of a J. L. Shepherd Model 149 Neutron Calibrator was not working properly, it was concluded that the electronics for the air operated source controller had failed. The failure would have allowed the source to be withdrawn without the electronic circuit being energized. The original source contained 8.5 Ci of Pu-Be. The calibrator is designed to expose a neutron source on a signal from a control circuit using air to withdraw the source. In this event, source withdrawal was initiated when air was aligned without the control circuit being energized; the air was immediately isolated preventing the source from being withdrawn. The calibrator has been removed from service due to the control circuit failure. The source is in the safe position, tagged and locked with the air line disabled. This failure did not result in any personnel exposure. This is reportable under 10 CFR 70.50(b)(2) since the circuit is required by design to be operable when the calibrator is used. The cause of the failure has not been determined at this time. The affected calibrator is not part of any installed plant equipment and has no impact on plant operation. The licensee has notified the NRC Resident Inspector. The licensee will notify the New York Public Service Commission. |
| ENS 49949 | 24 March 2014 13:38:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | A non-licensed supervisory employee had a confirmed positive for alcohol during a for-cause fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector and the New York Public Service Commission. |
| ENS 49802 | 6 February 2014 20:17:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | 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. |
| ENS 49698 | 6 January 2014 21:55:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | 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. |
| ENS 49573 | 22 November 2013 15:18:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | On November 21, 2013 during an Access Authorization Fitness for Duty (FFD) Baseline Inspection, an NRC Inspector identified laboratory testing of a Blind FFD sample failed to provide anticipated testing results. Laboratory testing results for sample (specimen number 422136066) should have reported the sample as 'Dilute' but the laboratory report results came back as 'Negative.' In addition, the specific gravity of the sample was also outside the expected range. It should have been reported at or near 1.0015 but was reported at 1.0224 instead. The sample was submitted on March 3, 2013. The results error was not identified when the results were received which caused a violation of 10CFR26. An investigation with the laboratory has been initiated through the Medical Review Officer for IPEC (Indian Point Energy Center). Additionally an investigation with the vendor supplier of the Blind sample will be initiated. The corrective action for this event is that an extent of condition is being performed to verify that no other Blind errors exist. The licensee notified the NRC Resident Inspector and will be notifying New York State Public Service Commission. |
| ENS 49171 | 3 July 2013 10:46:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | 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. |