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Start date | Site | Region | Reactor type | Event description | |
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ENS 57302 | 10 July 2024 13:02:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: A 10 CFR 50.73(a)(1) invalid specified system actuation reported under 10 CFR 50.73(a)(2)(iv)(a) invalid actuation of residual heat removal (RHR). This 60-day telephone notification is being made per 10 CFR 50.73 (a)(2)(iv)(a) under the provision of 10 CFR 50.73 (a)(1) as an invalid actuation of the RHR. On July 10, 2024, while at 100 percent power, a partial train actuation of RHR was initiated by an invalid actuation signal while performing RHR valve logic testing. The cause for the RHR system logic actuation was due to improper configuration of an emergency core cooling system (ECCS) logic tester. The RHR system started and functioned as designed for the actuation signals it received from the ECCS logic tester. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector was notified. |
ENS 57103 | 3 May 2024 08:11:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 0411 EDT on 5/03/2024, it was determined that primary containment did not meet TS (Technical Specification) 4.6.1.2 (surveillance) requirement due to a primary containment leak rate test exceeding `La (allowable leakage rate). This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The final observed leak rate is still being calculated as the test is still within the stabilization period. Testing is allowed within the stabilization period for an unspecified amount of time. Short term corrective actions are to identify and repair any leak paths. No mode changes are required due to this event. |
ENS 56952 | 7 February 2024 21:00:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: A programmatic vulnerability, failure, or degradation was discovered within the fitness for duty (FFD) program that may permit undetected drug or alcohol use or abuse by individuals within the protected area, or by individuals who are assigned to perform duties that require them to be subject to the FFD program. Public and plant safety have not been affected. The NRC Resident Inspector was notified. |
ENS 56889 | 15 December 2023 00:39:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone call and email: On December 14, 2023, at 1939 EST, Hope Creek reactor scrammed following closure of turbine control valve number 4. All control rods fully inserted into the core. All safety systems responded as designed and expected. There was no radiological release. The unit is stable in mode 3 with decay heat being removed via the turbine bypass valves rejecting steam to the main condenser. Normal feedwater level control is providing makeup to the reactor vessel. No personnel injuries resulted from the event. The outage control center has been staffed to determine the cause of the reactor scram. The Hope Creek NRC Resident Inspector has been notified. |
ENS 56495 | 30 April 2023 06:00:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 0200 EDT on 04/30/23, a Technical Specification required shutdown was initiated at Hope Creek Unit 1. Technical Specification Action 3.6.1.1 Primary Containment Integrity was entered on 04/30/23 at 0100 with a required action to restore primary containment integrity within 1 hour. This required action was not completed within the allowed outage time; therefore, a Technical Specification required shutdown was initiated, and this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
ENS 56494 | 30 April 2023 05:00:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 0100 EDT on 04/30/23, it was determined that the primary containment integrity did not meet (Technical Specification) TS 4.6.1.1.d requirement, suppression chamber in compliance with TS 3.6.2.1 due to the inability to establish test conditions for the bypass leakage test in accordance with TS 4.6.2.1.f. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D) & 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
ENS 56286 | 24 December 2022 07:22:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 0222 (EST) on 12/24/22, with Unit 2 at 100 percent power during steady state operation, the reactor automatically tripped and a safety injection actuated due to steam generator differential pressure. The trip and safety injection were not complex, with all systems responding normally post-trip. An actuation of the auxiliary feedwater system occurred following the reactor trip as expected due to low level in the steam generators. The unit is stable in Mode 3. The turbine bypass steam dumps and auxiliary feedwater system are removing decay heat. Salem Unit 1 was not affected. Due to the actuation of the reactor protection system while critical, this event is being reported as a four hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported as an eight hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the auxiliary feedwater system. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
ENS 54716 | 15 May 2020 13:47:00 | Salem | NRC Region 1 | At 0947 (EDT) on 5/15/20, Salem reported to the New Jersey Department of Environmental Protection a sheen on the Delaware River. This discovery did not violate any NRC (Nuclear Regulatory Commission) regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The licensee will be notifying the National Response Center and Lower Alloways Creek Township. The substance spilled was less than one pint of hydraulic oil. | |
ENS 54607 | 25 March 2020 04:56:00 | Salem | NRC Region 1 | At 0056 EDTon March 25, 2020, with Unit 1 at 17 percent power during a unit power ascension, the reactor was manually tripped due to the failure of the 11 Rod Control Motor Generator caused by a malfunction of its associated Voltage Regulator. The trip was not complex, with all systems responding normally post-trip. An actuation of the Auxiliary Feedwater system occurred following the manual reactor trip as expected due to low level in the steam generators. The unit is stable in Mode 3. Decay heat is being removed by the Atmospheric Steam Dumps and Auxiliary Feedwater System. Salem Unit 2 was not affected. Due to the actuation of the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B), This event is also being reported as an eight hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Auxiliary Feedwater System. There was no impact to the health and safety of the public or plant personnel. The NRC Resident lnspector has been notified. | |
ENS 54211 | 11 August 2019 12:14:00 | Salem | NRC Region 1 | At 0814 EDT on 8/11/19, with Unit 2 at 83 percent power during a planned load reduction, the reactor was manually tripped due to degraded feedwater flow control to the 23 Steam Generator caused by a malfunction of the associated Feedwater Regulating Valve, 23BF19. The trip was not complex, with all systems responding normally post trip. An actuation of the Auxiliary Feedwater system occurred following the manual reactor trip as expected due to low level in the steam generators. The unit is stable in Mode 3. Decay heat is being removed by the Main Steam Dumps and Auxiliary Feedwater System. Due to the actuation of the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported as an eight-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Auxiliary Feedwater System. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The licensee notified the State of New Jersey. Unit 1 remains at 100 percent power. | |
ENS 54198 | 3 August 2019 23:47:00 | Hope Creek | NRC Region 1 | At 1947 (EDT) on 8/3/19, with Hope Creek in Mode 1 at 37 percent power, the reactor was manually scrammed due to loss of condenser vacuum. All control rods fully inserted into the core. All safety systems responded as designed and expected. Reactor level was stabilized using Reactor Core Isolation Cooling (RCIC) and Reactor Feedwater Pumps. Currently reactor water level is being maintained by the feedwater system and decay heat is being removed by the main condenser using the main turbine bypass valves. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Due to the manual actuation of RCIC, this event is also being reported as an eight-hour, non-emergency notification per 10 CFR 50. 72(b )(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The plant is in its normal shutdown electrical lineup with all safe shutdown equipment available. The licensee will be notifying the state of Delaware, state of New Jersey and the Lower Alloway Creek township. | |
ENS 53852 | 31 January 2019 08:01:00 | Salem | NRC Region 1 | At 0301 (EST) on 1/31/19, with Unit 2 in Mode 1 at 100% power, the reactor was manually tripped due to icing conditions requiring the removal of 4 Circulating Water Pumps from service. The trip was not complex, with all systems responding normally post-trip. 21 CFCU (Containment Fan Cooler Unit) was inoperable prior to the event for a planned maintenance window and did not contribute to the cause of the event and did not adversely impact the plant response to the trip. An actuation of the Auxiliary Feedwater System occurred following the manual reactor trip. The reason for the Auxiliary Feed Water System auto-start was due to low level in a steam generator. Operations responded and stabilized the plant. Decay heat is being removed by the Main Steam Dumps and Auxiliary Feedwater System. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported as an eight hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Auxiliary Feed Water System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The icing condition was described as frazil ice. Unit-1 reduced power to 88% because one circulating water pump was shutdown. | |
ENS 53691 | 23 October 2018 04:00:00 | Salem | NRC Region 1 | At 1616 EDT on 10/23/18, Salem reported to the New Jersey Department of Environmental Protection a sheen on ground water discovered during excavation in the Salem Switchyard. This discovery did not violate any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The licensee notified the NRC Resident Inspector and will notify Lower Alloway Creek Township. | |
ENS 53786 | 15 October 2018 04:00:00 | Salem | NRC Region 1 | This 60-day telephone notification is being made in accordance with the reporting requirements of 10 CFR 50.73(a)(2)(iv)(A). The successful, complete train actuation of the 22 Auxiliary Feedwater Pump was initiated by an invalid signal during testing. The Auxiliary Feedwater System was not impacted in its ability to perform its function. There were no safety consequences or impacts to the health and safety of the public as a result of this event. The NRC Resident Inspector has been notified. | |
ENS 53625 | 26 September 2018 04:00:00 | Hope Creek | NRC Region 1 | On 9/26/2018 at 1530 EDT, it was discovered that the HPCI system was inoperable due to a blown fuse in the 10C617 Panel, E21-F15A. Therefore, this condition Is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). The blown fuse also impacts 'A' channel Residual Heat Removal (RHR) subsystem and 'A' Core Spray (CS) subsystem. These Emergency Core Cooling subsystems have been declared inoperable. Remaining Emergency Core Cooling subsystems and the Reactor Core Isolation Cooling (RCIC) system remain OPERABLE. There was no impact on the health and safety of the public or plant personnel." The licensee notified the NRC Resident Inspector and will notify the local authorities. | |
ENS 53606 | 14 September 2018 04:00:00 | Salem | NRC Region 1 | At 1323 (EDT) on 9/14/18, with Unit 2 in Mode 1 at 90% power, the reactor automatically tripped due to a failure of 23BF19, 23 Steam Generator (SG) Feed Regulating Valve. The trip was not complex, with all systems responding normally post-trip. No equipment was inoperable prior to the event. An actuation of the auxiliary feedwater system occurred following the automatic reactor trip. The reason for the auxiliary feed water system auto-start was due to low level in the steam generator. Operations responded and stabilized the plant. Decay heat is being removed by the main steam dumps and auxiliary feedwater system. Unit 1 is not affected. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non- emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the auxiliary feed water system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The Lower Alloways Creek Township will be notified. | |
ENS 53438 | 1 June 2018 04:00:00 | Salem | NRC Region 1 | During the period of evaluation of tornado missile vulnerabilities and the potential impacts to technical specification (TS) plant equipment, it was determined that the power cables to a safety related motor control center (MCC) in the service water (SW) intake structure are not adequately protected from tornado generated missiles. During walk downs, it was identified that the installed SW pipe tunnel barrier is not adequate. A tornado could generate missiles capable of striking the power cables and rendering a SW MCC inoperable. These conditions are reportable per 10 CFR 50.72(b)(3)(ii)(B) and per 10 CFR 50.72(b)(3)(v)(D). This condition is being addressed in accordance with NRC enforcement guidance provided in Enforcement Guidance Memorandum (EGM) 15-002, 'Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' Revision 1, and DSS-ISG-2016-01, 'Clarification of Licensee Actions in Receipt of Enforcement Discretion per Enforcement Guidance Memorandum EGM 15-002, Enforcement Discretion for Tornado- Generated Missile Protection Noncompliance,' Revision 1. Compensatory measures have been implemented in accordance with these documents. The NRC Resident Inspector has been notified. The licensee will be notifying the Lower Alloways Creek Township.
During subsequent walk downs, PSEG (Public Service Enterprise Group) identified that both the Unit 1 and Unit 2 turbine driven auxiliary feedwater pumps are also not adequately protected from tornado generated missiles. The steam exhaust pipe could be potentially impacted and cause crimping that could reduce steam exhaust flow and pump capacity. EN 53438 is updated to include both Salem units and these additional components. This condition is being addressed in accordance with NRC enforcement guidance provided in enforcement Guidance Memorandum (EGM) 15-002, 'Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' Revision 1, and DSS-ISG-2016-01, 'Clarification of Licensee Actions in Receipt of Enforcement Discretion per Enforcement Guidance Memorandum EGM 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' Revision 1. Compensatory measures have been implemented in accordance with these documents." The NRC Resident Inspector has been notified. The licensee will be notifying the Lower Alloways Creek Township. Notified R1DO (Burritt). | |
ENS 53386 | 7 May 2018 07:25:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | This 4 and 8 hour notification is being made to report that Salem Unit 2 initiated a manual reactor trip and subsequent automatic Auxiliary Feedwater system actuation. The trip was initiated due to a 21 Reactor Coolant Pump reaching its procedural limit for motor winding temperature of 302F. Salem Unit 2 is currently stable in Mode 3. Reactor Coolant system pressure is 2235 PSIG and Reactor Coolant System temperature is 547 F with decay heat removal via the Main Steam Dump and Auxiliary Feedwater Systems. Unit 2 has no active shutdown technical specification action statements in effect. All control rods inserted on the reactor trip. All ECCS (emergency core cooling systems) and ESF (emergency safety function) systems functioned as expected. No safety related equipment or major secondary equipment was tagged for maintenance prior to this event. No personnel were injured during this event. The NRC Resident Inspector was notified. The Lower Alloways Creek Township will be notified. |
ENS 53101 | 30 November 2017 21:57:00 | Hope Creek | NRC Region 1 | GE-4 | An Unusual Event was declared at 1657 EST due to an earthquake detected onsite. The Unusual Event was declared under EAL HU1.1. There is no release in progress due to this event. There are no protective actions recommended at this time. The Licensee will notify the NRC Resident Inspector. Note: See also EN #53099 for Salem Unusual Event.
An earthquake was felt onsite at time 1645 EST. Multiple phone calls were made to the Control Room confirming the earthquake. It was verified there was an earthquake felt in Delaware with a magnitude of 4.4. Neither seismic monitor at Salem Unit 1, Salem Unit 2, and Hope Creek actuated. There is no indication of any damage to any systems or plant structures. Plant walk-downs have been initiated in accordance with plant operating procedures for a seismic event. No injuries have been reported to the Control Room. The licensee will notify the NRC Resident Inspector and State and local government agencies. Notified R1DO (Gray), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), and DHS Nuclear SSA (email).
The licensee terminated the Unusual Event at 2125 EST on 11/30/2017 following plant walkdowns that revealed no damage to plant structures, systems, or components. The NRC Resident Inspector has been notified. Notified R1DO (Gray), IRD (Grant), and NRR EO (Miller), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), and DHS Nuclear SSA (email). |
ENS 53099 | 30 November 2017 21:57:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | An Unusual Event was declared at 1657 EST due to an earthquake detected onsite. The Unusual Event was declared under EAL HU1.1. There is no release in progress due to this event. There are no protective actions recommended at this time. The Licensee will notify the NRC Resident Inspector. Note: See also EN #53101 for Hope Creek Unusual Event.
An earthquake was felt onsite at time 1645 EST. Multiple phone calls were made to the Control Room confirming the earthquake. It was verified there was an earthquake felt in Delaware with a magnitude of 4.4. Neither seismic monitor at Salem Unit 1, Salem Unit 2, and Hope Creek actuated. There is no indication of any damage to any systems or plant structures. Plant walk-downs have been initiated in accordance with plant operating procedures for a seismic event. No injuries have been reported to the Control Room. The licensee will notify the NRC Resident Inspector and State and local government agencies. Notified R1DO (Gray), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), and DHS Nuclear SSA (email).
The licensee terminated the Unusual Event at 2125 EST on 11/30/2017 following plant walkdowns that revealed no damage to plant structures, systems, or components. The NRC Resident Inspector has been notified. Notified R1DO (Gray), IRD (Grant), and NRR EO (Miller), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), and DHS Nuclear SSA (email). |
ENS 52699 | 21 April 2017 01:10:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | At 2110 EDT, Salem control room received data that supported unacceptable levels of hydrazine concentration in the U2 Containment atmosphere based on Site Protection atmospheric sampling. The high hydrazine levels were caused due to Steam Generator (S/G) venting into the Containment atmosphere in support of maintenance for the current Salem Unit 2 Refueling Outage (2R22). The NIOSH habitability limit for hydrazine is 0.03 ppm (2 hour limit). Area samples indicated concentrations as high as 0.25 ppm. Salem Unit 2 Containment has been evacuated while a mitigation plan is being developed. There were no personnel injuries as a result of this occurrence. Salem Unit 2 defueling activities were in progress during this event. All fuel assemblies have been placed in a safe condition. All Salem Unit 2 Containment activities are currently on hold. There has been no impact to the equipment in the Unit 2 Containment, no adverse impact to any equipment located in the vicinity of the high hydrazine concentration, and no operational impact to the plant including Shutdown Cooling which is currently on RHR. The Unusual Event was declared under EAL HU3.1, Toxic/Flammable Gas Release Affecting Plant Operations. The licensee plans to issue a press release. The licensee notified the NRC Resident Inspector, Lower Alloways Creek Township, State of New Jersey and State of Delaware. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
Salem Unit 2 terminated the Unusual Event at 1258 EDT on 4/21/17. The basis for termination was no longer restricting access to the containment after getting two sets of satisfactory air sample results. With the access restored, normal plant operations can resume and EAL HU3.1 is no longer applicable. The details of the sample results are: Fire Protection performed satisfactory results of no detectable Hydrazine (0.01 ppm with a NIOSH limit of 0.03 ppm) completed both at 1001 EDT and 1247 EDT at the following locations: - (3) at 130 ft. elevation - at 78 ft. in the bioshield - at 78 ft. outside the bioshield. Additional mitigating actions taken following U2 Containment evacuation were as follows: - FME screen installed on open manways for 21/23 S/G with additional plastic covering and tape to prevent further gas release into containment. - Modified Containment Purge in service to maximize ventilation in Containment. - 21/24 S/G draining to support filling and draining evolutions to reduce Hydrazine concentrations in the S/G's. - Releasing tags on the AFWST to commence filling and further support filling and draining evolutions on the U2 S/G's. The licensee notified the NRC Resident Inspector. Notified R1DO (Arner), NRR EO (King), and IRD (Stapleton). Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email). |
ENS 52698 | 20 April 2017 17:06:00 | Hope Creek | NRC Region 1 | GE-4 | At 1306 (EDT) on April 20, 2017, Control Room Overhead Alarm C6-C4, Seismic Monitor Panel, was received. This alarm normally indicates an actuation of the Hope Creek OBE (Operational Basis Earthquake) Seismic Switch at 0.1g, however, this alarm was accompanied by no other indications of seismic activity. The indicator for a 0.01g earthquake was not actuated, no earthquake was felt on site, and the National Earthquake Information Center recorded no seismic activity in the area. The Seismic Monitor Panel was considered non-functional. With the Seismic Monitor Panel non-functional, the ability to classify EAL HA1.1, Operating Basis Earthquake Detected Onsite, was lost. Testing of the Seismic Monitor Panel is in progress to determine system functionality. The licensee will notify the NRC Resident Inspector. |
ENS 52805 | 16 April 2017 14:53:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | This is to report the Salem Unit 2, 2C Emergency Diesel Generator (EDG) actuation due to an invalid signal. This report is being made per paragraphs 10CFR50.73(a)(1) and 10CFR50.73(a)(2)(iv)(A) to address the invalid actuation of the 2C EDG on April 16, 2017, while performing the 2C Safeguards Equipment Controller (SEC) Mode OPS Surveillance test. Plant conditions: Salem Unit 2 was in mode 5 at the time of the invalid actuation. On April 16, 2017, at approximately 1053 (EDT) while performing Solid State Protection System (SSPS) testing of the 2C SEC, the 2C Emergency Diesel Generator (EDG) output breaker was manually opened per the associated procedure step. The EDG output breaker unexpectedly reclosed and the 2C 4kV vital bus loaded onto the EDG in SEC Mode 2. The cause of the 2C EDG output breaker reclosure and 2C 4kV Vital bus loading during testing was determined to be two faulty input block switches in the 2C SEC. When Step 5.2.27 of the test procedure was performed, the 2C SEC 'input block' switches failed to block a 'blackout' actuation signal. This resulted in the breaker reclosure and loading of the 2C Vital Bus onto the EDG. Trouble shooting identified that the two failed switches exhibited high resistance across the switch contacts which is indicative of being in a 'fail to block' (the input signal) condition. The licensee has notified the NRC Resident Inspector. The licensee will also notify the States of New Jersey and Delaware. |
ENS 52681 | 14 April 2017 17:57:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | At 1357 (EDT) on April 14, 2017, an unplanned automatic start signal was generated for the 2C Emergency Diesel Generator (EDG). The station was in the process of transferring the 2C 4160 volt vital bus from the 24 Station Power Transformer (SPT) to the 23 SPT, which are the offsite power in-feeds for the 2C 4160 volt vital bus. The 24 SPT infeed breaker opened as expected; however, the 23 SPT infeed breaker failed to close. The failure to swap from the 24 SPT to 23 SPT resulted in a momentary loss of power to the 2C 4160 volt vital bus generating the automatic start signal for the 2C EDG. The 2C 4160 volt vital bus was automatically re-energized by the 2C EDG as expected. Abnormal operating procedures were entered for loss of the 2C 4160 volt vital bus. Salem Unit 2 was in Mode 1 operating at 100% power. All equipment operated as expected. At 1555 (EDT), 2C 4160 volt vital bus was reenergized from 24 SPT, and the 2C EDG was secured in accordance with station implementing procedures. There was no impact to the health and safety of the public. The Resident Inspector has been notified. The Lower Alloways Creek Township, State of New Jersey, and State of Delaware will be notified. |
ENS 52581 | 28 February 2017 15:00:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | On February 28, 2017 at 0930 (EST), a containment visual inspection was performed to identify the source of elevated RCS (Reactor Coolant System) leakage. A leak was identified between 13RC6 and 13SS661, 13 RCS hot leg sample isolation valves at 1000 (EST). These valves are manual isolation valves in the reactor coolant hot leg sample line. Leak isolation could not be initially verified and is considered RCS pressure boundary leakage. Salem Unit 1 entered Technical Specification 3.4.6.2a, RCS operational leakage, for the existence of pressure boundary leakage. This event is being reported under the requirements of 10 CFR 50.72(b)(2)(i) for 'The initiation of a plant shutdown required by Technical Specifications' and 10 CFR 50.72(b)(3)(ii)(A) or 'Any event of condition that results in the condition of the nuclear power plant, including its principal safety barriers being seriously degraded.' The unit was placed in mode 3 at 1554 (EST) on 02/28/2017. This condition has no impact on public health and safety. Per Technical Specifications, the unit is proceeding to mode 5. The leak rate at the time of shutdown was 0.33 gpm. This event has no effect on Unit 2. The licensee has notified the NRC Resident Inspector. The licensee will be notifying the Lower Alloways Creek Township, the State of New Jersey and the State of Delaware.
The purpose of this notification is to retract event report number 52581 made on 2/28/2017 at 1624 (EST). Previously, PSEG notified the NRC that Salem Unit 1 initiated a shutdown required by Technical Specifications (TS) for Reactor Coolant System (RCS) Pressure Boundary Leakage. Subsequent to the initial report, PSEG has determined that the leak occurred in tubing downstream of the design specification break between Safety Related, Nuclear Class 1, Seismic Class1 and Non-Safety-Related, Nuclear Class 2, Seismic Class 2. Therefore, the observed leakage is not RCS pressure boundary leakage as defined in the Salem Unit 1 Technical Specifications and in the tubing design classification specification. At the time of the event, during initial entry into the containment, the volume of steam present and the height of the break above the floor made it difficult to ascertain the location of the steam source with certainty. The initial judgment of RCS Pressure Boundary Leakage was conservative under these circumstances. The plant was taken offline to minimize radiation exposure when personnel operated the isolation valves. Following the shutdown, the leak was isolated. Based on an observed reduction in RCS leak rate and visual verification of leakage isolation, the TS Limiting Condition for Operation (LCO) was exited and the unit remained in Mode 3, Hot Standby, to affect repairs. The condition did not meet the Technical Specification Pressure Boundary Leakage definition of leakage through a non-isolable fault in a RCS component body, pipe wall or vessel wall. The leakage did not impact the ability to shut down the unit and no TS limits were exceeded during this event. Therefore, the plant shutdown to investigate and correct leakage from flawed sample system tubing does not meet the reporting requirements of 10 CFR 50.72 and PSEG is retracting the notifications made under 10 CFR 50.72(b)(2)(i) and 10 CFR 50.72(b)(3)(ii)(A). The NRC Resident Inspector was notified of this retraction by the licensee. Notified R1DO (Jackson). |
ENS 52471 | 2 January 2017 00:55:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | A non-licensed employee had a confirmed positive for alcohol during a "for cause" fitness-for-duty test. The employee's access to the plant has been suspended. The licensee notified the NRC Resident Inspector. |
ENS 52347 | 5 November 2016 08:04:00 | Hope Creek | NRC Region 1 | GE-4 | On November 5, 2016 an RPS (Reactor Protection System) actuation occurred from an actual high scram discharge volume level reaching the RPS actuation setpoint. This actuation was the result of a Redundant Reactivity Control System (RRCS) signal inadvertently generated during excess flow check valve testing with the reactor in cold shutdown. At the time of the actuation, all control rods were inserted. RCS pressure was approximately 830 psig to support excess flow check valve testing and shutdown cooling was removed from service. When RRCS initiated, the B Reactor Recirculation Pump tripped as expected and the scram air header depressurized as expected, which caused the high level in the scram discharge volume. The cause of the RRCS signal is being investigated. The A loop of RHR was placed back into the Shutdown Cooling mode of operation with reactor temperature being maintained at approximately 150 degrees F. There were no injuries as a result of this event. The licensee has notified Lower Alloways Creek Township and the NRC Resident Inspector. |
ENS 52294 | 11 October 2016 13:38:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | At 0938 (EDT), the watertight door in the Unit 1 Inboard Service Water Penetration Area was unable to be closed. The watertight door serves as a High Energy Line Break (HELB) and Medium Energy Line Break (MELB) barrier between the mechanical penetration room and the service water penetration room. A HELB/MELB event occurring in a room with its barrier door open could adversely affect equipment in an adjacent room. Consequently, a HELB/MELB event could have rendered equipment in the adjacent room inoperable. At 1005, station maintenance was able to successfully close and latch the door restoring the barrier. This event is being reported under the requirements of 10CFR50.72(b)(3)(ii)(B) as, 'the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' An ENS notification is required if an unanalyzed condition occurred within 3 years of the date of discovery even if the event is not on-going at the time of discovery. The licensee notified Lower Alloways Creek Township and the NRC Resident Inspector.
The purpose of this notification is to retract event report number 52294 made on 10/11/2016 at 1648 (ET). Previously PSEG reported that Salem Unit 1 was determined to be in an unanalyzed condition due to being unable to close the Unit 1 Inboard Service Water Penetration Area Water Tight Door. The watertight door was reported to serve as a High Energy Line Break (HELB) and Medium Energy Line Break (MELB) barrier between the mechanical penetration room and the service water penetration room. Subsequent review identified that the condition did not meet the reporting criterion. Engineering evaluation determined that the service water penetration room had been previously evaluated for the impact due to a HELB event occurring in the adjacent mechanical penetration area with the watertight door open and that the event would not impact the operability of the service system. Engineering evaluation also determined that the Salem Unit 1 Design Basis does not require analysis of a MELB event occurring in the service water penetration room. This is due to the timing of the Unit 1 operating license issue date. At the time of the issuance of the Unit 1 operating license analysis of a MELB event was not required. Therefore an unanalyzed condition for a MELB event did not exist. Additionally while the door was difficult to close, it was able to be closed and dogged in a reasonable time interval (27 minutes), therefore, any potential internal flooding which would have been detected immediately by the attendant with required actions taken to close the door and isolate the leak rapidly. Therefore PSEG is retracting the notification made under 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector was notified of this retraction by the licensee. Notified R1DO (Schroeder). |
ENS 52245 | 16 September 2016 13:13:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | At 0858 (EDT) on September 16th, 2016, approximately 3 ounces of hydraulic fluid (fish oil) was spilled in front of the Unit Two Circulating Water System (CWS) Intake trash racks at the Salem Generating Station. The spill of hydraulic fluid (fish oil) was caused by a leak from the crane used to rake debris from the Unit Two trash racks. The crane was stopped and the leak terminated at the time of discovery. Nuclear Environmental Affairs Department determined a 4 hr report to the NRC, under RAL 11.8.2.a, was warranted due to the 15 minute notification to the New Jersey Department of Environmental Protection at 0913 (EDT). Nuclear Environmental Affairs Department intends to retract the report to the New Jersey Department of Environmental Protection based on the fluid remained within the Circulating Water System (CWS) Intake Structure. The licensee has notified the New Jersey Department of Environmental Protection, the NRC Resident Inspector, and will notify the Lower Alloways Creek Township. |
ENS 52222 | 3 September 2016 13:02:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | At 0902 EDT, on September 3, 2016, a leak developed from 13 Charging Pump seals that resulted in an unidentified leak rate of 1.3 gpm. Salem Unit 1 entered Technical Specification 3.4.6.2.b for unidentified leakage greater than 1 gpm. The Technical Specification requires the leak rate to be restored to below 1 gpm in a 4 hour period or place the Unit in Hot Standby in the next 6 hours and Cold Shutdown within the following 30 hours. In addition there is an allowable total ECCS leakage outside of Containment value of 0.45 gpm associated with Control Room habitability to comply with GDC-19 (General Design Criteria) limits. 12 Charging Pump was placed in service and 13 Charging Pump was removed from service and isolated. The unidentified leak rate lowered to below the 0.45gpm requirement for the total ECCS leakage outside of containment and the Technical Specification value of 1 gpm. This event is being reported under the requirements of 10CFR50.72(b)(3)(ii)(B) as 'the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety' for exceeding the 0.45gpm total ECCS leakage allowable in accordance with GDC-19 limitations. The Licensee has notified the NRC Resident and the Lower Alloways Creek Township.
The purpose of this notification is to retract event report number 52222 made on 09/03/2016 at 1247 (EDT). Previously PSEG reported that Salem Unit 1 was in an unanalyzed condition due to a leak from the 13 Charging Pump seals exceeding the 0.45 gpm limit for total ECCS leakage outside containment associated with the Control Room Habitability analysis. A subsequent engineering evaluation determined that the post-accident ECCS leak rate from the 13 Charging Pump seals that would have existed is 0.33 gpm. This is within the limits of the existing Control Room Habitability analysis and no unanalyzed condition existed. Therefore, PSEG is retracting the notification made under 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector was notified of this retraction by the licensee. Notified R1DO (Schroeder). |
ENS 52213 | 31 August 2016 19:11:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | This 4-hour and 8-hour notification is being made to report that Salem Unit 2 had an unplanned automatic reactor trip and automatic actuation of the auxiliary feedwater system. The trip occurred due to the loss of the 21 reactor coolant pump (RCP) resulting in a reactor trip on low reactor coolant flow. The 21 RCP remains unavailable. The cause of the loss of the 21 reactor coolant pump is unknown at this time. All control rods inserted on the reactor trip. All emergency core cooling systems and engineered safety feature systems functioned as expected. The auxiliary feed pumps started as expected. Salem Unit 2 is currently in Mode 3. Reactor coolant system pressure is at 2235 psig and temperature is 547 degrees Fahrenheit with decay heat removal via the main steam dumps and auxiliary feedwater systems. Unit 2 has no active technical specification action statements in effect requiring a lower mode of operation due to the transient. The 21 and 22 containment fan coil units (CFCU) were out of service for surveillance testing prior to the event. There was no major secondary equipment tagged for maintenance prior to the event. There were no personnel injuries as a result of this event. Normal offsite power is available to the site. There is no effect on Unit 1. The licensee notified the NRC Resident Inspector. |
ENS 52159 | 7 August 2016 05:01:00 | Hope Creek | NRC Region 1 | GE-4 | HPCI (high pressure coolant injection) governor valve did not respond as expected. During performance of a planned HPCI valve functional test the HPCI governor valve (FD-FV-4879) did not reposition as expected. The HCPI system has been declared inoperable based on the response per Technical Specification 3.5.1 action C.1. All other emergency core cooling systems and the reactor core isolation cooling (RCIC) system remain operable. The unit remains at 100% power. The station has initiated an event response team to identify and correct the cause of the failure. No personnel injuries resulted from the event. The licensee notified the NRC Resident Inspector and the Lower Alloways Creek Dispatch. |
ENS 52129 | 28 July 2016 09:41:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | This four and eight hour notification is being made to report that at 0541 (EDT) on 7/28/16, Salem Unit 1 initiated a shutdown to comply with Technical Specifications due to the inoperability of both source range nuclear instruments. During a reactor startup, with Unit 1 in Mode 2, both source range instruments were reading approximately one decade lower than expected compared to intermediate range and Gamma-Metric instruments and due to the proximity to the estimated critical condition. The condition could also have prevented the fulfilment of the source range instruments safety function to trip the reactor when required. Salem Unit 1 is currently stable in Mode 3. Reactor Coolant system pressure is 2235 psig and reactor coolant system temperature is 547 F with decay heat removal via the main steam dump and auxiliary feedwater systems. Unit 1 has one active shutdown tech spec action statement in effect due to the inoperability of the containment radiation monitor 1R11A. The inoperability of this radiation monitor had no effect on the event. All control rods were manually inserted to place Unit 1 in Hot Standby (Mode 3). No ECCS (emergency core cooling system) or ESF (emergency safety features) systems were required to function during this event. No major secondary equipment was tagged for maintenance prior to this event. No personnel were injured during this event. The reactor was manually shut down and a shutdown margin calculation verified sufficient margin. The licensee notified the NRC Resident Inspector and the local township.
The purpose of this notification is to retract event report number 52129 made on 7/28/2016 at 0925 (EDT). Previously PSEG reported that Salem Unit 1 initiated a shutdown to comply with Technical Specifications (TS) due to the inoperability of both source range nuclear instruments. Additionally PSEG reported that the condition could have prevented the fulfillment of the safety function needed to, 'Shut down the reactor and maintain it in a safe shutdown condition.' Subsequent review identified that the condition did not meet either reporting criteria. Maintenance and Engineering evaluation of the source range nuclear instruments determined that the instruments were fully operable at the time of the event. TS 3.3.1.1, Reactor Trip Instrumentation remained met, no TS shutdown was required and the instruments were capable of performing their required function. Therefore PSEG is retracting the notifications made under 10 CFR 50.72(b)(2)(i) and 10 CFR 50.72(b)(3)(v)(A). The NRC Resident Inspector was notified of this retraction by the licensee. Notified the R1DO (Cook). |
ENS 52048 | 28 June 2016 08:23:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | This 4 and 8 hour notification is being made to report that Salem Unit 2 suffered an unplanned automatic reactor trip and subsequent automatic auxiliary feedwater system actuation. The trip was initiated due to a Main Turbine Trip above P-9 (49% power). The Main Turbine trip was caused by a Main Generator Protection signal. Salem unit 2 is currently stable in Mode 3. Reactor coolant system pressure is 2235 psig and Reactor Coolant System temperature is 547 F with decay heat removal via the main steam dump and auxiliary feedwater systems. Unit 2 has no active shutdown tech spec action statements in effect. All control rods (fully) inserted on the reactor trip. All ECCS (Emergency Core Cooling System) and ESF (Emergency Safety Features) systems functioned as expected. No safety related equipment or major secondary equipment was tagged for maintenance prior to this event. No personnel were injured during this event. The main generator protection signal was either a ground fault or a differential current trip. The plant is in its normal shutdown electrical lineup. No safeties or relief valves lifted during this event. Unit 1 is defueled and was not affected by this event. The licensee notified the NRC Resident Inspector and will notify the Lower Alloways Creek Township. |
ENS 51902 | 3 May 2016 17:01:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | With Salem Unit One in a defueled condition during a planned refueling outage, anomalies were identified on baffle-former bolts while conducting a scheduled visual inspection of reactor vessel internals. Due to the visual anomalies, PSEG commenced ultrasonic inspection of the baffle-former bolts to determine the extent of condition and determine a repair plan. Based on initial results of ultrasonic inspections received on May 03, 2016, this condition was determined to be reportable pursuant to 10 CFR 50.72(b)(3)(ii)(B), since the as-found conditions were not previously analyzed. Additional details regarding the extent of condition will be supplied in the 60 day report. The NRC Resident Inspector has been informed. The licensee will notify State and local government agencies as appropriate. |
ENS 51901 | 3 May 2016 15:30:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | A non-licensed employee supervisor violated the fitness-for-duty policy during a random fitness-for-duty test. The employee's access to the plant has been denied. The licensee has notified the NRC Resident Inspector. |
ENS 52007 | 18 April 2016 01:04:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | This is to report the Salem Unit 1, 1B Emergency Diesel Generator (EDG) actuation due to an invalid signal. This report is being made per paragraphs 10CFR50.73 (a)(1) and 10CFR50.73(a)(2)(iv)(A) to address the invalid actuation of the 1B EDG on April 17, 2016, during replacement of an indicating bulb. Plant conditions; Salem Unit 1 was in mode 6 at the time of the invalid actuation. On April 17, 2016, at approximately 2055 (EDT) while performing Solid State Protection System (SSPS) testing of the 1B Safeguards Equipment Cabinet (SEC), operators identified that an input test light was not lit as expected. At approximately 2104 (EDT) while attempting to replace the light bulb, the 1B EDG unexpectedly automatically started. The 1B EDG responded properly to the auto start signal and started in SEC Mode 1, accident only, and did not load. The cause of the inadvertent start was determined to be a loss of the block circuit which allowed an output to the logic module which then caused the EDG to auto start. Subsequent testing of the input block switches demonstrated that, due to switch degradation, slight pressure applied to the switch was enough to allow the block signal to be momentarily interrupted, even without repositioning of the switch. It was determined that the loss of the block was most likely due to the operators finger coming in contact with the switch during the bulb replacement. The licensee has notified the NRC Resident Inspector. The licensee will notify the State of New Jersey and Delaware. |
ENS 51814 | 22 March 2016 13:14:00 | Hope Creek | NRC Region 1 | GE-4 | A non-licensed supervisor tested positive for alcohol during a random fitness for duty test. The individual's access to the facility has been denied. The licensee has notified the NRC Resident Inspector. |
ENS 51738 | 16 February 2016 13:27:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | At 0827 (EST), the 21 and 22 Aux Feedwater pumps auto started due to a trip of the running Steam Generator Feed pump (22). The 22 Steam Generator Feed pump tripped while swapping from heating steam to main steam in accordance with normal operating procedures. The cause of the 22 Steam Generator Feed pump trip is being investigated at this time. The auto start of the Unit 2 Auxiliary Feedwater pumps is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any system listed in Technical Basis 11.3.3 except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' The licensee has notified the NRC Resident Inspector and will notify the Lower Alloways Creek Township. |
ENS 51734 | 15 February 2016 01:58:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | This 4-hour and 8-hour notification is being made to report that Salem Unit 2 had an unplanned automatic reactor trip. The trip occurred because the Unit 2 main generator tripped on generator protection with reactor power greater than P-9 (49%). The cause of the generator protection trip that resulted in the reactor trip is unknown at this time. A troubleshooting team is being assembled to determine the exact cause for the generator protection actuation. All control rods inserted on the reactor trip. All ECCS and ESF systems functioned as expected. The motor driven and steam driven auxiliary feed pumps started as expected on steam generator low level. Salem Unit 2 is currently in mode 3. Reactor Coolant System pressure is at 2235 psig and temperature is 547 degrees Fahrenheit with decay heat removal via the main steam dumps and auxiliary feedwater systems. Unit 2 has no active technical specification action statement in effect requiring a lower mode of operation. No safety related equipment or major secondary plant equipment was tagged for maintenance prior to the event. There were no personnel injuries as a result of this event. There was no impact on Salem Unit 1. The licensee informed the NRC Resident Inspector and will inform the Lower Alloways Creek Township (LAC). |
ENS 51708 | 4 February 2016 16:21:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | This 4 and 8 hour notification is being made to report that Salem Unit 2 suffered an unplanned automatic reactor trip and subsequent automatic Auxiliary Feedwater system actuation. The trip was initiated due to a Main Turbine trip above P-9 (49% power). The Main Turbine trip was caused by a Main Generator Protection signal. Salem Unit 2 is currently stable in Mode 3. Reactor Coolant system pressure is 2235 PSIG and Reactor Coolant system temperature is 547 F with decay heat removal via the Main Steam Dump and Auxiliary Feedwater Systems. Unit 2 has no active shutdown tech spec action statements in effect. All control rods inserted on the reactor trip. All ECCS and ESF systems functioned as expected. No major secondary equipment was tagged for maintenance prior to this event. The 24 Service Water pump is tagged for scheduled preventive maintenance and did not affect post trip plant response. No personnel were injured during this event. The licensee has notified the NRC Resident Inspector and will notify the Lower Alloway Creek Township. |
ENS 51663 | 19 January 2016 21:30:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | On January 19, 2016, while reviewing outage data, plant staff recognized that anomalous data collected in October, 2015, for the 21 Auxiliary Feed Pump time response loop resulted in an unanalyzed condition. Preliminary investigation has revealed that the condition most likely existed since April 20, 2015, when maintenance activities were performed on the auxiliary feedwater pump discharge pressure transmitter. Consequently, there were multiple instances when one of the other auxiliary feedwater pumps was removed from service, thus creating a condition which did not meet the accident analysis assumptions for auxiliary feedwater flow initial response. This event is being reported under the requirements of 10 CFR 50.72(b)(3)(ii)(B) as 'the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' This condition was corrected on November 20, 2015. The auxiliary feedwater pump discharge pressure transmitter instrument isolation valve was inadvertently left closed after the April 20, 2015 maintenance. The licensee has notified the NRC Resident Inspector and will notify the State of New Jersey, State of Delaware, and the local township.
This event is being retracted. An engineering review determined that while the auxiliary feedwater flow loop response time test results did not meet the procedural acceptance criteria, the accident analysis assumptions remained valid. The ATWS (Anticipated Transient Without Scram) was the limiting accident, and the loop response time results were still bounded by the existing analyses. The failed loop response time did result in a condition prohibited by TS which will be reported in a Licensee Event Report (LER). This condition has been documented in the licensee's Corrective Action Program. The licensee has notified the NRC Resident Inspector. Notified R1DO (Dimitriadis). |
ENS 51603 | 14 December 2015 14:04:00 | Hope Creek | NRC Region 1 | GE-4 | A non-licensed supervisor tested positive for alcohol during a for-cause fitness-for-duty test. The individual's access to the facility has been denied. The licensee has notified the NRC Resident Inspector. |
ENS 51563 | 24 November 2015 02:48:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | At 2148 EST on November 23, 2015, Salem Unit 2 declared an Unusual Event due to reactor coolant system leakage greater than 10 gallons per minute. While performing troubleshooting to determine the source of leakage from the Emergency Core Cooling System high head safety injection piping, a motor operated valve was opened and the high head piping relief valve lifted. Indications in the control room calculated the leak rate at 16 gallons per minute based on the change in Pressurizer level. The leak was terminated when the motor operated valve was closed and the relief valve reseated. The time (duration) of the leak was about one minute." The licensee notified the NRC Resident Inspector. Notified DHS SWO, FEMA Ops Center, and NICC Watch Officer. E-mailed FEMA NWC and Nuclear SSA.
The licensee terminated the Unusual Event at 0100 EST on 11/24/15. The licensee will be cooling down and depressurizing to mode 5 in order to replace the affected valve. The licensee will be notifying the NRC Resident Inspector as well as the New Jersey State police, Delaware State Police and local emergency dispatch. Notified the R1DO (Dwyer), NRR EO (Morris), IRD MOC (Stapleton), DHS SWO, FEMA Ops enter, and NICC Watch Officer. E-mailed FEMA NWC and Nuclear SSA.
This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D). 'Any event of condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident,' due to inoperability of both trains of the Emergency Core Cooling System high head safety injection systems. The unusual event was exited at 0100 (EST) on November 24th, 2015, when the Emergency Core Cooling System high head safety injection piping inlet valves were closed, ensuring isolation of the relief valve. The criteria for exit was leakage rate was below the 10 GPM rate. The plant is in mode 3 cooling down to mode 5. The licensee notified the states of New Jersey and Delaware, Lower Alloways Township, and the NRC Resident Inspector. Notified the R1DO (Dwyer). |
ENS 51624 | 29 October 2015 16:12:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | This is to report the Salem Unit 2, 2C Safeguards Equipment Cabinet (SEC) actuation due to an invalid signal. This report is being made per paragraphs 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to address the invalid actuation of the 2C SEC on October 29, 2015, during performance of maintenance on the Solid State Protection System (SSPS). Plant conditions: Salem Unit 2 was in Mode 6 at the time of the invalid actuation. On October 29, 2015, at approximately 1112 (EDT) during performance of maintenance to replace slave relays in the SSPS output test panel 131, the maintenance technician inadvertently made contact with the relay plunger for the 2C SEC. This caused a signal to be sent to the 2C SEC which actuated the equipment associated with it. The 2C SEC train responded to the signal as required. The equipment associated with the 2C SEC that was not previously removed from service for outage related activities or already in service responded as required. As a result of the actuation, the 2C Emergency Diesel Generator (EDG) started but did not load (no undervoltage condition existed on the associated vital bus requiring loading), the Unit 2 Emergency Control Air Compressor (ECAC) started. The 22 Charging Pump (CVC) started, and the 22 Safety Injection Pump (SI) started (the Sl pump did not inject as it was isolated and in the process of being tagged out of service at the time of the invalid signal). The cause of the invalid actuation signal was a human performance error. The licensee notified the NRC Resident Inspector. |
ENS 51504 | 28 October 2015 10:28:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | At 0628 (EDT) on 10/28/15, the Salem Unit 1 Control Room Emergency Air Conditioning System (CREACS) train was declared inoperable due to the results of charcoal filter testing not meeting the surveillance requirement. Prior to declaring the Unit 1 train inoperable, the system was operating in the single train filtration operation mode due to refueling outage activities being performed on Unit 2 with the Unit 2 CREACS train inoperable. This resulted in both units' CREACS trains being inoperable. Prior to the event on 10/28/15, Salem Unit 1 was operating in Mode 1 at 100% power and Unit 2 was in Mode 6 with fuel movement in progress. At 0628, Unit 1 entered Technical Specification 3.0.3, and Unit 2 suspended fuel movement to comply with Technical Specification 3.7.6 (Modes 5 and 6, or during movement of irradiated fuel assemblies), action c. The Unit 2 train was placed in service at 0755, allowing Unit 1 to exit Technical Specification 3.0.3. Operators continued to align the Unit 2 CREACS to single train filtration operation to comply with Unit 1 Technical Specification Action Statement 3.7.6.1, Action a. Actions to place the Unit 2 CREACS in single train filtration operation were completed at 0950. This 8 hour notification is being made pursuant to the requirements of 10 CFR 50.72(b)(3)(v) for 'any condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to : (D) Mitigate the consequences of an accident.' The licensee notified the NRC Resident Inspector. |
ENS 51430 | 29 September 2015 00:46:00 | Hope Creek | NRC Region 1 | GE-4 | On September 28, 2015 at 2046 EDT, the Hope Creek reactor scrammed following a trip of both reactor recirculation pumps. All control rods fully inserted into the core. All safety systems responded as designed and expected. There was no radiological release. The unit is stable in Mode 3 with decay heat being removed via the turbine bypass valves rejecting steam to the main condenser. Normal feedwater level control is providing makeup to the reactor vessel. No personnel injuries resulted from the event. The Outage Control Center has been staffed to determine the cause of the reactor scram. The Hope Creek NRC Resident Inspector has been notified. The licensee notified Lower Alloways Creek township of the event. |
ENS 51349 | 26 August 2015 16:15:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | On 8/26/2015 at 1215 (EDT), a review of the Salem Appendix R Safe Shutdown Analysis in response to the Byron Event Notification #51334 and Braidwood Event Notification #51335, identified a fire scenario that could cause spurious operation of the pressurizer power operated relief valves (PORVs) (PR1 and PR2) and also prevent the ability to close the PORV block valves (PR6 and PR7) until AC power is restored to close the block valves. This scenario would result in the loss of reactor coolant system (RCS) inventory and pressure control that is not accounted for in safe shutdown analysis. The above fire scenario is applicable to a fire in the Control Room and Relay Room fire areas. Hourly fire watches of the Relay Room have been implemented. In addition, the Control Room is continuously staffed by the Operating Shift. In addition, the Relay Room is equipped with automatic detection and suppression. This event is being reported under 10CFR50.72(b)(3)(ii)(B), for 'any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' The NRC Resident Inspector has been notified. The licensee will be notifying the Lower Alloways Township, the State of New Jersey and the State of Delaware. |
ENS 51290 | 5 August 2015 19:38:00 | Salem | NRC Region 1 | Westinghouse PWR 4-Loop | This 4-hour and 8-hour notification is being made to report that Salem Unit 2 had an unplanned automatic reactor trip. The trip occurred because the 2H 4kV infeed breaker tripped on over current protection which deenergized the 2H 4kV group bus resulting in a reactor trip on reactor coolant flow due to the loss of the 21 reactor coolant pump (RCP). The 21 RCP remains unavailable. The cause of the over current protection trip on the 2H 4kV infeed breaker is unknown at this time. All control rods inserted on the reactor trip. All ECCS (emergency core cooling systems) and ESF (engineered safety feature) systems functioned as expected. The auxiliary feed pumps started as expected. Salem Unit 2 is currently in Mode 3. Reactor coolant system pressure is at 2235 psig and temperature is 547 degrees Fahrenheit with decay heat removal via the main steam dumps and auxiliary feedwater systems. Unit 2 has no active technical specification action statements in effect requiring a lower mode of operation due to the transient. The 22 auxiliary building supply fan was tagged for maintenance prior to this event and has no adverse impact of the post trip plant response or stabilization. There was no major secondary equipment tagged for maintenance prior to the event. There were no personnel injuries as a result of this event. There is no physical evidence of damage to the 2H 4kV bus based on visual observations and thermography. Normal offsite power is available to the site. There is no effect on Unit 1. The licensee has notified the NRC Resident Inspector and will notify the Lower Alloways Creek Township Police Department. |
ENS 51274 | 28 July 2015 17:58:00 | Hope Creek | NRC Region 1 | GE-4 | At 1358 (EDT) on July 28, 2015, a 1 inch diameter hole was discovered in the secondary containment wall, between the Reactor Building and the Auxiliary Building, causing the Secondary Containment to become inoperable under Technical Specification 3.6.5.1. Reactor Building pressure was maintained at a negative pressure as required by Technical Specification 3.6.5.1 with the Reactor Building ventilation system in service before and after discovery of the hole. In addition, the Filtration, Recirculation and Ventilation system remained fully operable and remained in standby. The hole was sealed at 1600 and technical specification 3.6.5.1 was exited. Plant operation was not impacted by the event and was operating at 100% power. No personnel injuries resulted from this event. The hole was discovered by plant personnel that were walking past the wall. Due to the discovery of the hole, the plant is performing an extent of condition inspection. The licensee notified the NRC Resident Inspector and the Lower Alloways Creek Dispatch.
This event is being retracted. Hope Creek Generating Station Unit 1, is retracting the 8-hour non-emergency notification (EN# 51274) made on July 28, 2015, at 1855 EDT. The notification on July 28, 2015, reported that secondary containment was declared inoperable when a 1 inch hole was discovered in the secondary containment wall, between the Reactor Building and the Auxiliary Building. Secondary containment was declared inoperable based on the initial interpretation of the definition of secondary containment. The hole did not impact the ability to maintain the Tech Spec required negative pressure. Subsequent evaluation determined that secondary containment was always operable. Based on the engineering evaluation, the condition reported in EN# 51274 did not result in an inoperability of the secondary containment. Therefore, there is no reportable condition and this event report is being retracted. The NRC Resident Inspector has been briefed on the evaluation results and informed of this retraction. The licensee also notified the Lower Alloways Creek Dispatch. Notified R1DO (Powell). |