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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56286 | 24 December 2022 07:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | 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. | Steam Generator Reactor Protection System Auxiliary Feedwater | |
| ENS 54607 | 25 March 2020 04:56:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Trip Due to Failure of a Rod Control Motor Generator | 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. | Steam Generator Reactor Protection System Auxiliary Feedwater | |
| ENS 54211 | 11 August 2019 12:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Feedwater Regulating Valve Malfunction | 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. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Steam | |
| ENS 53852 | 31 January 2019 08:01:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip - Circulating Water Icing Conditions | 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. | Steam Generator Reactor Protection System Auxiliary Feedwater Main Steam | |
| ENS 53606 | 14 September 2018 04:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to a Failure of the Steam Generator Feed Regulating Valve | 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. | Steam Generator Reactor Protection System Auxiliary Feedwater Main Steam | |
| ENS 53386 | 7 May 2018 07:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip | 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. | Reactor Coolant System Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | |
| ENS 52213 | 31 August 2016 19:11:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Low Reactor Coolant Flow | 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. | Reactor Coolant System Auxiliary Feedwater Emergency Core Cooling System Decay Heat Removal Control Rod Main Steam | |
| ENS 52048 | 28 June 2016 08:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Main Generator Protection Signal | 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. | Reactor Coolant System Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | 05000311/LER-2016-005 |
| ENS 51734 | 15 February 2016 01:58:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Main Generator Protection Trip | 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). | Steam Generator Reactor Coolant System Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | 05000311/LER-2016-003 |
| ENS 51708 | 4 February 2016 16:21:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Main Turbine Trip | 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. | Reactor Coolant System Service water Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | 05000311/LER-2016-002 |
| ENS 51290 | 5 August 2015 19:38:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Low Reactor Coolant Flow | 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. | Reactor Coolant System Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | 05000311/LER-2015-002 |
| ENS 50092 | 7 May 2014 08:56:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Trip Due to Generator Protection Relay | This 4 and 8 hour notification is being made to report that Salem Unit 1 experienced an unplanned automatic reactor trip and subsequent Auxiliary Feedwater System actuation. The automatic trip was initiated from a generator protection relay. A troubleshooting team is being assembled to determine the exact cause for the generator protection actuation. 11, 12 and 13 Aux Feedwater Pumps automatically started as expected on low Steam Generator water level following the reactor trip. All control rods inserted on the reactor trip. All ECCS and ESF systems functioned as expected. Salem Unit 1 is currently in Mode 3. Reactor Coolant system pressure is at 2235 PSIG and temperature is 547 F with decay heat removal via the Main Steam Dumps and Auxiliary Feedwater Systems. Condenser vacuum was available for the duration of the event. The 12 Chiller was tagged for maintenance prior to this event and had no adverse impact on the post trip plant response or stabilization. No personnel were injured during this event. The licensee notified the NRC Resident Inspector | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | |
| ENS 50032 | 14 April 2014 01:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Actuation of Main Generator Protection Relay | This 4 and 8 hour notification is being made to report that Salem Unit 1 experienced an unplanned automatic reactor trip and subsequent Auxiliary Feedwater system actuation. The automatic trip was initiated from a generator protection relay. A walkdown of the main generator and all protection circuitry has been completed with no visible problems identified. A troubleshooting team is being assembled to determine the exact cause for the generator protection actuation. 11, 12, and 13 Aux Feedwater Pumps automatically started as expected on low Steam Generator water level following the reactor trip. All control rods inserted on the reactor trip. All ECCS and ESF systems functioned as expected. Salem Unit 1 is currently in Mode 3. Reactor Coolant System pressure is at 2235 psig and temperature is 547 degrees F with decay heat removal via the main steam dumps and Auxiliary Feedwater Systems. There was no major primary or secondary equipment tagged for maintenance prior to this event. Condenser vacuum and Steam Generator Feed Pumps were available for the duration of the event. No personnel were injured during this event. The licensee notified the NRC Resident Inspector and the Lower Alloways Creek Township. Unit 1 is in a normal shutdown electrical lineup. Unit 2 is in a refueling outage and was unaffected by the Unit 1 trip. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | |
| ENS 50012 | 9 April 2014 01:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 Manually Tripped Following the Loss of an Operating Steam Generator Feed Pump (Sgfp) | This 4 and 8 hour notification is being made to report that Salem Unit 1 has performed an unplanned manual reactor trip and subsequent Auxiliary Feedwater system actuation. The trip was initiated due to loss of Steam Generator water level (SGWL) following the loss of one of the running SGFPs (#11 Steam Generator Feed Pump). The reactor was manually tripped prior to reaching the automatic SGWL low setpoints. #11/12/13 Aux Feedwater Pumps automatically started following the reactor trip on low Steam Generator water level. Salem Unit 1 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 1 has no active shutdown technical specification action statements in effect. All control rods inserted on the reactor trip. All ECCS (Emergency Core Cooling System) and ESF (Engineered Safety Feature) systems functioned as expected. There was no major primary or secondary equipment tagged for maintenance prior to this event. No personnel were injured during this event. The licensee will inform the Lower Alloways Creek Township and the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | |
| ENS 49780 | 31 January 2014 15:01:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip | This 4-hour notification is being made to report that Salem Unit 2 has performed an unplanned manual reactor trip. The trip was initiated due to reactor coolant temperature approaching the minimum temperature for criticality, 543 degree F, due to boration to achieve shutdown margin requirements following identification of a misaligned control rod. All control rods inserted on the reactor trip. All ECCS (Emergency Core Cooling System) and ESF (Engineered Safety Features) systems functioned as expected with no equipment actuated. The 21 safety injection pump was out of service for scheduled maintenance during the event as was the 2R41 plant vent radiation monitor. Salem Unit 2 is currently in Mode 3. Reactor coolant system pressure is at 2235 psig and temperature is 547 degrees 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. There was no major secondary equipment tagged for maintenance prior to the event. Prior to the event, the licensee was conducting their monthly control rod surveillance. No primary or secondary relief valves lifted during the transient. The electrical grid is stable and the plant is in its normal shutdown electrical lineup. There was no effect on Unit 1. The licensee notified the NRC Resident Inspector, the State of New Jersey, the State of Delaware and the Lower Alloways Creek Township. | Reactor Coolant System Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | |
| ENS 48617 | 21 December 2012 10:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 Automatic Reactor Trip on Turbine Trip | Salem Unit 1 has experienced an automatic reactor trip at 0528 (EST) on 12/21/12. The unit tripped due to turbine trip above P-9 (Greater than 49% power). All shutdown and control rods fully inserted on the reactor trip. Prior to the trip, the unit was operating at 100% when the crew received the OHA (Overhead Alarm) for main power transformer over excitation which actuated generator protection, which initiated the turbine trip. The auxiliary feed water (AFW) system auto started on low steam generator levels as expected on a reactor trip. Numbers 11, 12 & 13 AFW pumps all automatically started to provide feed to the steam generators. The (Operating Crew) entered EOP-Trip-1, then transitioned to EOP-Trip-2 and stabilized the plant. The unit is currently in mode 3. The OCC (Outage Control Center) is manned and the cause of the main power transformer over excitation is under investigation at this time. RCS temperature is 547 degrees F, RCS pressure is 2235 psig. The 11-14 RCP's are in service. There is one shut down technical specification action statement in effect. Unit 1 containment APD (Containment Radiation Monitor) is inoperable for DCP (Design Change Package) work. This is a 30 day shutdown LCO that expires on 1/16/2013 at 0830. All ECCS and ESF systems are available. Decay heat removal is being provided by 11 and 12 AFW pumps and the main steam dump system. The 13 AFW pump operation is not required and has been removed from service. The plant is aligned with a normal electrical line-up from offsite power sources. There were no personnel injuries associated with this event This event is reportable per 10CFR50.72(b)(2)(iv)(b) due to the automatic reactor trip. This event is reportable per 10CFR50.72(b)(3)(iv)(a) due the AFW actuation on low steam generator levels. There was no lifting of PORVs or primary to secondary leakage. There was no impact on Unit 2. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | |
| ENS 48534 | 25 November 2012 16:30:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Steam Generator Feedwater Regulator Valve Not Responding to Demand Signal | This notification is being made to report that Salem Unit 2 has experienced an automatic reactor trip at 1130 EST. The reactor trip was caused by 24BF19, 24 steam generator feedwater regulation valve, not responding to demand signal that resulted in a reactor trip due to 24 steam generator lo-lo level. The cause of the 24BF19 failure to respond to demand is under investigation. 21, 22 & 23 AFW pumps all auto started as expected on the reactor trip feeding the steam generators. 23 AFW pump has been shutdown by procedure. 21 & 22 AFW pumps are maintaining steam generator levels with decay heat removal via the main steam dump system. All control & shutdown rods fully inserted on the reactor trip. Salem Unit 2 is currently in mode 3. Reactor coolant system temperature is 547 degrees F with pressure at 2235 psig. There are no shut down technical specifications action statements in effect. All ECCS and ESF systems are available. All safety related equipment responded as expected. There were no personnel injuries associated with this event. The licensee will inform the Lower Alloways Creek Township, the State of New Jersey, and the State of Delaware. The licensee has informed the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Decay Heat Removal Main Steam | |
| ENS 48457 | 30 October 2012 05:09:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip from 100% Power | This report if being made under the requirements of 10 CFR 50.72(b)(2)(iv)(B), Actuation of the Reactor Protection System While Critical, except preplanned, and under the requirements of 10 CFR 50.72(b)(3)(iv)(A), Valid Actuation of Listed System, except preplanned. Salem Unit 1 was operating at 100% reactor power when a loss of 4 condenser circulators required a manual reactor trip in accordance with station procedures. The cause of the 4 circulators being removed from service was due to a combination of high river level and detritus from Hurricane Sandy's transit. All control rods inserted. A subsequent loss of the 2 remaining circulators required transition of decay heat removal from condenser steam dumps to the 11-14 MS10s (atmospheric steam dump). Decay heat removal is from the 11/12 Aux Feed Pumps to all 4 steam generators via the 11-14 MS10s. 11/12/13 AFW pumps started due to low level on all steam generators due to shrink from full power operation (this is a normal response). All safety related equipment functioned as expected. No one has been injured. As an additional note, Hurricane Sandy had recently moved past artificial island. Salem Unit 1 is currently in Mode 3. Salem Unit 2 reactor is currently in its 2R19 refueling outage and is shutdown and defueled with no fuel movement in progress. The licensee has notified the NRC Resident Inspector.
At 0513, following (a) Unit 1 manual (reactor) trip due to loss of condenser cooling, a manual steam line isolation was initiated due to a high condenser back pressure. All main steamline isolation valves responded as expected. The high condenser back pressure resulted in the #11 low pressure turbine rupture disc relieving. Unit 1 remains in mode 3 with Reactor Coolant System temperature at 549 (degrees) and stable. Reactor Coolant System pressure is 2235 psig and in automatic control. Pressurizer level is on program at 26 percent level and in automatic control. Core cooling is via aux feed water and the steam generator levels are on program. There were no (personnel) injuries. The licensee has notified the NRC Resident Inspector. Notified R1DO (Caruso). | Steam Generator Reactor Coolant System Reactor Protection System Decay Heat Removal Control Rod | |
| ENS 47874 | 30 April 2012 14:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Unusual Event Due to a Potential Fire in Containment | Performing an I&C functional (test) caused an inadvertent Safety Injection signal resulting in a reactor trip/safety injection. All safety systems responded as designed for a safety injection. Electrical systems are aligned to normal offsite power sources. All fire alarms have been validated by the Fire Protection Department as invalid alarms and confirmed that no fire event in the protected area. The reactor trip was successful and all rods (fully) inserted. Decay heat removal is via auxiliary feedwater through the atmospheric (steam) dumps. Unknown at this time is the cause of the inadvertent safety injection signal. No injuries occurred as a result of this event. The licensee believes that the trip/safety injection may have caused piping to shake resulting in dust near the fire detection equipment resulting in the invalid fire indication. The instrument being tested was the high steam flow channel-1 bistable for PT505. The maximum pressurizer level during this event was 95%. The licensee notified the NRC Resident Inspector.
At 1003 hours on April 30, 2012, Salem Unit 1 experienced a reactor trip and safety injection (SI) signal due to a high steam flow coincident with a low steam pressure signal. At the time of the safety injection signal, function testing of the 1PT505 turbine inlet pressure channel was in progress. This testing required the tripping of the high steam flow bistables. As a result of the reactor trip and safety injection signal, the Emergency Diesel Generators started but did not load, the ECCS system (high head safety injection pumps actuated and injected into the reactor vessel, intermediate head safety injection pumps and low head (RHR) safety injection pumps) actuated. All 4 main steam isolation valves closed along with feedwater isolation and start of the auxiliary feedwater pumps. All control rods fully inserted following the reactor trip. Following the main steam line isolation, the atmospheric relief valves opened along with the lifting of several main steam safety valves. The unit is currently in Mode 3 and will be cooling down to Mode 4. Train A SSPS (Solid State Protection System) is currently out of service and suspected of causing the safety injection signal. Train B SSPS has not been reset due to the standing safety injection signal. With Train A SSPS inoperable and Train B SSPS not reset, TS 3.0.3 was entered and a shutdown required by TS 3.0.3 was commenced at 1345 hours. This report is being made in accordance with 10CFR50.72(b )(2)(iv)(B), 50.72(b)(3)(iv)(A), 50.72(b)(2)(i) and 50.72(b)(2)(iv)(A). The licensee exited the Unusual Event at 1249 EDT. The licensee notified the NRC Resident Inspector. Notified the R1DO (Conte). The NRC Operations Center notified other Federal Agencies (DHS SWO, FEMA Ops, DHS NICC, and NuclearSSA via e-mail). | Feedwater Emergency Diesel Generator Main Steam Isolation Valve Auxiliary Feedwater Main Steam Safety Valve Decay Heat Removal Main Steam Line Control Rod | |
| ENS 47766 | 23 March 2012 18:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Turbine Trip Followed by Automatic Reactor Trip | Salem Unit 2 has experienced an automatic reactor trip at 1428 hours on 3/23/12. Salem Unit 2 tripped due to a turbine trip above (permissive) P-9. All shutdown and control rods fully inserted on the reactor trip. At the time of the trip 22 station power transformer (SPT), the 2F and 2G 4KV group buses were not energized causing a loss of 23 and 24 reactor coolant pumps. The auxiliary feedwater (AFW) system auto-started to provide feed to the steam generators. The crew entered EOP-TRIP-1, then transitioned to the EOP-TRIP-2 and stabilized the plant. Salem Unit 2 is currently in Mode 3 at normal operating temperature and pressure with 21 and 22 reactor coolant pumps in service. There are no shutdown technical specifications in effect. All ECCS and ESF systems are available. Decay heat removal is being provided by 21 and 22 AFW pumps and the main steam dump system. The plant is aligned with a normal electrical line-up from offsite power sources with the exception of 22 SPT. Restoration of the 22 SPT and 2F and 2G 4KV group buses is in progress. There were no personnel injuries associated with this event. This event is reportable per 10 CFR 50.72 (b)(3)(iv)(A) due to the auto-start of the auxiliary feed water pumps. The licensee notified the NRC Resident Inspector. The licensee will notify Lower Alloways Creek Township and the States of New Jersey and Delaware. The licensee anticipates making a press release.
During the post trip review ... two additional AFW pump start signals were identified. At 1429 hrs. EDT with RCPs 23 and 24 tripped the 21 and 22 steam generators (SG) dipped below the lo-lo level setpoints and caused an AFW actuation to occur. At 1457 hrs. EDT 23 turbine driven AFW pump was stopped in accordance with operating procedures. At 1509 hrs. EDT the 21 SG cleared the low level setpoint and at 1513 hrs. EDT the 22 SG cleared the low level setpoint. At 1523 hrs. EDT the 22 SG dipped below the lo-lo level setpoint and caused an AFW actuation to occur. The two motor drive AFW pumps continued to run throughout this event. The licensee notified the NRC Resident Inspector and will notify Lower Alloways Creek Township and the States of New Jersey and Delaware. Notified R1DO (Schmidt). | Steam Generator Auxiliary Feedwater Decay Heat Removal Control Rod Auxiliary Feed Water Main Steam | |
| ENS 46992 | 26 June 2011 04:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Loss of Reactor Coolant Pump | This notification is being made to report that Salem Unit 2 has experienced an automatic reactor trip. The trip was caused by 23 Reactor Coolant (RC) loop low flow and P-8 caused by 23 reactor coolant pump trip. The cause of the 23 Reactor Coolant Pump trip is under post reactor trip investigation. Main Feedwater to the Steam Generators was isolated due to the 'Feed Water Interlock' which is expected on a reactor trip. The Auxiliary Feed Water (AFW) system auto started on low Steam Generator levels. This is also expected on a reactor trip. 21, 22 and 23 AFW pumps all automatically started and fed the Steam Generators. 23 AFW pump has been shutdown by procedure and is available. The crew entered EOP-TRIP-1, and then transitioned to EOP-TRIP-2 and stabilized the plant at no load conditions. Salem Unit 2 is currently in Mode 3. Reactor Coolant System temperature is 547 degrees F with pressure at 2235 psig NOP/NOT. 21, 22, and 24 Reactor Coolant Pumps are in service. Decay heat removal is being provided by 21 and 22 AFW pumps via the Main Steam Dump System. All shutdown and control rods fully inserted on the reactor trip. The plant is aligned with normal electrical lineup from offsite power sources. There were no personnel injuries associated with this event. The NRC Resident Inspector has been informed. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | |
| ENS 46774 | 21 April 2011 20:01:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Circulating Water Pumps Out of Service | A manual trip of Salem Unit 1 was initiated due to a loss of circulating water pumps from heavy grassing at the circulating water intake. All rods fully inserted on the trip and all systems responded as designed with decay heat being removed via the steam dump system. All Auxiliary Feed Water (AFW) pumps auto started as expected on low steam generator levels from the trip. During a period of heavy grassing with one circulating water pump out of service for maintenance and another out of service for water box cleaning, a third circulating water pump's screen stopped and this circulating water pump was emergency tripped. When this circulating water pump was emergency tripped, the adjacent circulating water pump received a large intake of grass and automatically tripped. The abnormal operating procedure provides guidance to trip the plant if less than three circulating water pumps are in service and power is above P-10. At this point the manual reactor trip was initiated. The crew entered EOP-TRIP-1, appropriately transitioned to EOP-TRlP-2 and stabilized the plant at no load conditions. Salem Unit 1 is currently in mode 3. Reactor Coolant System temperature is 547F with pressure at 2235 psig. All ECCS and ESF Systems are available. Salem Unit 2 is currently defueled in a planned refueling outage and there is no impact to Salem Unit 2. No personnel injuries have occurred as a result of the trip. This event is reportable per 10CFR50.72(b)(3)(iv)(A) due to the auto start of the AFW pumps. This event is reportable per 10CFR50.72(b)(2)(iv)(B) due to the manual reactor trip. The plant is aligned with a normal electrical line-up from offsite power sources. The Licensee has notified the NRC Resident Inspector. The Licensee will notify Lower Alloways Creek Township and the States of New Jersey and Delaware.
During the post-trip data review for the manual reactor trip that occurred on April 21, 2011 at 1601 (EDT), 2 additional automatic Auxiliary Feedwater (AFW) pump start signals were identified. Operators were controlling AFW level to maintain steam generator levels within the band required by the emergency operating procedures. At approximately 1615 (EDT) the 13 steam generator level had cleared the low level setpoint but subsequently dipped below the low level setpoint at 1623 (EDT). At approximately 1640 (EDT), the 14 steam generator had recovered above the low level setpoint and subsequently dipped below the low level setpoint. The Unit 1 reactor had been previously tripped at 1601 (EDT) and all AFW pumps had automatically started. The 13 turbine driven AFW pump was stopped at approximately 1625 (EDT). The two motor driven AFW pumps continued to run throughout this event. Low level in one steam generator generates an automatic start of the motor driven AFW pumps and is reportable per 10CFR50.72(b)(3)(iv)(A). The licensee notified the NRC Resident Inspector and will notify Lower Alloways Creek Township and the States of New Jersey and Delaware. Notified the R1DO (Cook). | Steam Generator Reactor Coolant System Auxiliary Feedwater Auxiliary Feed Water | |
| ENS 46337 | 17 October 2010 09:12:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to 4 Kv Group Bus Undervoltage Condition | On 10/17/10 at 0512 (EDT), Unit 2 experienced an (automatic) reactor trip due to 4 kV group bus under voltage. While restoring the voltage regulator to automatic, following a swap to manual at 0123 (EDT) on 10/17/10 while 500 kV line 5014 was being restored to service, the reactor tripped after placing the voltage regulator in automatic. The crew entered EOP (Emergency Operating Procedure) Trip 1, and then EOP Trip 2 as required. The plant was stabilized at no load conditions. All rods fully inserted on the trip, and all systems responded as designed with RCS (Reactor Coolant System) temperature being controlled via 21 RCP (Reactor Coolant Pump) and the steam dump system. 22, 23, and 24 RCP's tripped on group bus under voltage. The AFW (Auxiliary Feed Water) Pumps started in response to low steam generator levels. 23 AFW was subsequently tripped per procedure since 21 and 22 AFW pumps were in service feeding 21-24 steam generators. Salem Unit 2 is currently in Mode 3. RCS temperature is 547 degrees. RCS pressure is 2235 pounds. All Emergency Core Cooling Systems and Engineered Safety Function Systems are available. No personnel injuries occurred as a result of the trip. No radiological release is in progress due to this event. At 0123 EDT, following the restoration of a 500 kV line, perturbation in the line caused an over current condition which shifted the voltage regulator to manual. After engineering's recommendation, the voltage regulator was placed in automatic shortly prior to the reactor trip. The plant is in a normal shutdown electrical lineup. There was no impact on Salem Unit 1 as a result of this event. The licensee has notified the NRC Resident Inspector.
Notified by the licensee that the AFW Pumps started on loss of Steam Generator Feed Pumps which had tripped on low suction pressure. The licensee has notified the NRC Resident Inspector. | Steam Generator Emergency Core Cooling System | |
| ENS 46336 | 16 October 2010 03:21:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to a Turbine Trip | An automatic reactor trip of Salem Unit 1 (occurred) due to a turbine trip above P-9. The crew entered EOP (Emergency Operating Procedure) Trip 1, appropriately transitioned to EOP Trip 2, and stabilized the plant at no load conditions. All rods fully inserted on the trip, and all systems responded as designed with decay heat removal being removed via the steam dump system with condenser vacuum being maintained. The Auxiliary Feed Water Pumps automatically started due to low steam generator levels. Salem Unit 1 is currently in Mode 3. Reactor Coolant System temperature is 547 degrees, and pressure is 2235 pounds. All Emergency Core Cooling Systems and Engineered Safety Function Systems are available. No personnel injuries occurred as a result of the trip. No radiological release (occurred) due to this event. Approximately 5 minutes prior to the turbine trip, Salem Unit 1 lowered reactive loading on the voltage regulator as requested by the load dispatcher while Hope Creek Unit 1 was coming off line due to their outage. The cause of the trip is still under investigation. There was no impact on Salem Unit 2 as a direct result of this event. The licensee will notify the NRC Resident Inspector. | Steam Generator Reactor Coolant System Emergency Core Cooling System Decay Heat Removal Auxiliary Feed Water | |
| ENS 46075 | 7 July 2010 15:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Caused by Main Generator and Turbine Trip | Salem Unit 1 experienced an automatic reactor trip at 1117 hours due to a trip of the main generator and turbine. The main generator tripped as a result of a fault on the B Main Power Transformer (MPT). The B MPT fire protection deluge system automatically actuated and extinguished a small fire that occurred on the B MPT. The fire existed for approximately 5 minutes prior to being extinguished. All control rods fully inserted on the trip and all systems responded as designed with decay heat being removed via the Steam Dump system with condenser vacuum maintained. All three AFW (Auxiliary Feed Water) Pumps auto started as expected due to low Steam Generator level. Following the trip, the Reactor Coolant Pump (RCP) thermal barrier supply valve 1CC131 automatically closed and was subsequently reopened with no issue. Salem Unit 1 is currently in mode 3. The Reactor Coolant System temperature is 547 degrees F with pressure at 2235 psig (NOT and NOP). All ECCS and ESF Systems are available and no ECCS (Emergency Core Cooling System) systems actuated during the event. The 13 AFW pump has not yet been reset following the trip, per the Emergency Operating Procedures and is unavailable. No personnel injuries occurred as a result of the trip. There is no primary to secondary leakage. There was no impact on Unit 2. The licensee has notified the NRC Resident Inspector.
Based on additional reviews, the reactor trip occurred at 1118 hours following actuation of the B Main Power Transformer (MPT) fire protection deluge system at 1116 hours. Personnel in the area stated that an arc flash occurred on the high voltage bushing of the B MPT following the deluge system actuation and there was no fire present when the deluge system actuated. Notified the R1DO (Dimitriadis). | Steam Generator Reactor Coolant System Control Rod | |
| ENS 45647 | 21 January 2010 23:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Low Steam Generator Water Level | On January 21, 2010, at 1820 hours, Salem Unit 2 had an Automatic Reactor Trip due to low steam generator (SG) water level in the 22 SG. Prior to the reactor trip, the 21 SG feedwater pump tripped due to low oil pressure. A turbine runback was in progress and SG levels failed to recover. (Also at the time of the reactor trip), the 22 SG feedwater pump tripped on overspeed. All rods fully inserted on the trip and all systems responded as designed with decay heat being removed via the Steam Dump system, with condenser vacuum maintained. All three AFW Pumps auto started due to low Steam Generator level. Following the trip, the 22 Reactor Coolant Pump (RCP) flange vibration level spiked to approximately 9 mils and returned to a normal value of 2.6 mils. 22 RCP remained in service following the trip. The 21RS12 low pressure turbine intercept valve failed to indicate closed following the turbine trip but was locally verified closed. Salem Unit 2 is currently in mode 3. The Reactor Coolant System temperature is 547 degrees F with pressure at 2235 psig (NOT and NOP). All ECCS and ESF Systems are available and no ECCS systems actuated during the event. No personnel injuries occurred as a result of the trip. There is no primary to secondary leakage. The electrical shutdown configuration is normal. There was no impact on Unit-1. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater | |
| ENS 45604 | 3 January 2010 13:04:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Loss of Circulators After Uptake of River Ice | A manual trip of Salem Unit 2 was initiated due to a loss of 4 circulators due to an excessive uptake of river ice. Strong northwest winds, freezing temperatures and abnormally low tide levels contributed to the ice formation on the river. The operating crew entered EOP-TRIP-A, appropriately transitioned to EOP-TRIP-2 and stabilized the plant at no load conditions (MODE 3 Hot Stand-by). All rods fully inserted on the trip and all systems responded as designed with decay heat being removed via the Steam Dump system, with condenser vacuum maintained. All three AFW Pumps auto started due to low Steam Generator level due to closure of 21-23CN27, low pressure feed water heater inlet isolation valves. There were no other significant equipment challenges associated with the reactor trip. Salem Unit-2 is currently in mode 3. The Reactor Coolant System temperature is 547 (degrees) F with pressure at 2235 psig (NOT and NOP). All ECCS and ESF Systems are available and no ECCS systems actuated during the event. No personnel injuries occurred as a result of the trip. (The Unit 2 reactor trip recovery is uncomplicated.). Unit 1 power was also reduced to 80%. Although the combination of strong northwest winds, freezing temperatures and an abnormally low tide contributed to the ice formation on the river, operators foresee no further complication due to the ice conditions. The licensee has notified the NRC Resident Inspector and will notify the States of Delaware and New Jersey. | Steam Generator Reactor Coolant System | |
| ENS 43873 | 28 December 2007 22:54:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Trip on Failure of Station Power Transformer | This 4 hour notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) to report that Salem Unit 1 has experienced an automatic reactor trip. The trip was initiated by failure of 12 Station Power Transformer (SPT). The cause of the trip was the loss of two Reactor Coolant Pumps (RCP) (13 and 14). These RCPs were powered from 'F' and 'G' group busses (non safeguards), which were powered by 12 SPT. Post trip the following occurred: 11, 12 & 13 Auxiliary Feed Water (AFW) Pumps (automatically) started after the trip due to valid steam generator low levels (14%) as expected. The automatic start of the Auxiliary Feedwater System due to a valid steam generator low level is an ESF actuation and is an 8 hour report in accordance with 10 CFR 50.72 (b)(3)(iv)(B). 13 AFW pump was tripped (in accordance with) the Emergency Operating Procedure and will be returned to an operable condition following recovery of steam generator levels. All control rods fully inserted following the trip. No ECCS actuation occurred. There was no major equipment out of service prior to the event. No injuries resulted from this event. Salem Unit 1 is currently in Mode 3. Reactor Coolant System pressure is at 2235 PSIG and temperature is at 547 degrees with decay heat removal via the Main Steam Dump System. Unit 1 has no active shutdown (Technical Specification) action statements in effect. The NRC Resident Inspector has been notified. | Steam Generator Reactor Coolant System Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | |
| ENS 43550 | 6 August 2007 17:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Spurious Interlock Signal to Feedwater Regulating Valves | This 4-hour notification is being made to report that Salem Unit 2 has experienced an automatic reactor trip. The trip was initiated by a low steam generator level on 22 SG (14 percent narrow range). The low steam generator level was apparently caused due to the feedwater regulating valve closing as a result of a spurious 'Feedwater interlock' signal. Post trip the following occurred: 21, 22, & 23 auxiliary feedwater pumps auto started after the trip due to valid steam generator low levels (14 percent) as expected on a unit trip. 23 Auxiliary feedwater pump (turbine driven) was tripped in accordance with the emergency operating procedure and will be returned to an operable condition following recovery of steam generator levels. Salem Unit 2 is currently in mode 3. Reactor coolant system pressure at 2235 psig and temperature is 547 degrees F with decay heat removal via the main steam dump system. Unit 2 has no active shutdown Tech Spec Action statements in effect. There was no major secondary equipment tagged for maintenance prior to the event. On the reactor trip, all control rods fully inserted. No primary or secondary relief valves or safety valves lifted. The electric plant is in a normal shutdown lineup. This event had no effect on Unit 1. The licensee notified the NRC Resident Inspector. The licensee will notify the Lower Alloways Creek township, the State of NJ Bureau of Nuclear Engineering, and the State of Delaware. The licensee also plans on issuing a press release. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | |
| ENS 43382 | 24 May 2007 06:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip on Low Steam Generator Water Level | This 4-hour notification is being made to report that Salem Unit 2 automatically tripped on low steam generator water level. The automatic trip was preceded by lowering steam generator feed pump suction pressure, followed by the trip of both steam generator feed pumps on low suction pressure. The cause of the low suction pressure was due to a condensate system breach that occurred in the Condensate Polisher building. The condensate polishing system has been isolated and condensate pressure has stabilized. Salem Unit 2 is currently in mode 3 with reactor coolant system temperature at 547 degrees with pressure at 2235 psig. Auxiliary feedwater pumps 21, 22, and 23 started automatically, as expected, due to the low steam generator level. Decay heat removal is via the steam dumps to the main condenser. All equipment functioned as designed. There were no personnel injuries. A 15 minute notification was made to the State of New Jersey due to hydrazine laden water exiting the Condensate Polisher building and entering the storm drainage system. All control rods fully inserted following the reactor trip. Emergency Diesel Generator are available, if needed. The licensee informed the NRC Resident Inspector.
In addition to hydrazine that was contained in the condenser water that exited the Condensate Polisher Building, supplemental samples determined that the water also contained tritium at approximately 50,000 picoCuries per liter. The state of New Jersey was updated with this information. Notified the R1DO (Trapp). The licensee will notify the NRC Resident Inspector. | Steam Generator Reactor Coolant System Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal Main Condenser Control Rod | |
| ENS 43329 | 30 April 2007 19:02:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Trip Due to Grass Intrusion Into Circulating Water Intake Structure | |||
| ENS 43317 | 25 April 2007 02:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Trip Due to Loss of Circulators | A manual trip of Salem Unit 1 was initiated due to a loss of circulators from heavy grassing at the circulating water intake. All rods fully inserted on the trip and all systems responded as designed with decay heat being removed via the steam dump system. All Auxiliary Feedwater pumps automatically started as expected on low steam generator levels from the trip. The unit had previously entered the abnormal operating procedure for circulating water due to two circulators being out of service; one had previously emergency tripped due to high traveling water screen differential pressure and a second had been out of service for condenser waterbox cleaning. Alarms were received indicating rapid rising differential pressure on the remaining four in-service circulators' traveling water screens. In accordance with the abnormal procedure, the Control Room Supervisor directed the tripping of the circulators and the reactor. There were no complications encountered as a result of the trip and all equipment operated as expected. The operating crew stabilized the plant at no load conditions. Salem Unit 1 is currently in Mode 3 with RCS at normal operating temperature and pressure. Actions are being taken to return the circulators to service. Unit 2 was not affected at this time and is operating at 100% power. The licensee notified the NRC Resident Inspector and the local township Licensee will be issuing a press release.
The power level was at 40% when the manual trip occurred and not 100% as previously reported. Notified R1DO (Trapp) | Steam Generator Auxiliary Feedwater | |
| ENS 42864 | 26 September 2006 19:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Reactor Coolant Pump Seal Leakage. | Salem Unit 2 was manually tripped from 94.1 percent reactor power due to Reactor Coolant Pump "21" number 1 seal leakoff exceeding 6 gpm in accordance with abnormal operating procedure. Reactor Coolant Pump "21" was subsequently stopped upon reactor trip verification in accordance with the Abnormal Operating Procedures. Forced reactor coolant system circulation is via the remaining 3 reactor coolant pumps. All safety systems functioned as designed. Auxiliary feedwater pumps started as expected, offsite power is available, Emergency Diesel Generators are available but not required at this time. Decay heat removal is via steam dumps to the main condenser. Salem Unit 2 is currently stable in Mode 3 (Hot Standby). There is no equipment out of service that contributed to this event. There were no personnel injuries or radiological occurrences associated with this event. The licensee is investigating to determine the cause of the abnormal reactor coolant pump seal leakoff flow rate. The licensee notified local government officials. The licensee notified the NRC Resident Inspector. | Reactor Coolant System Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal Main Condenser | |
| ENS 42395 | 8 March 2006 16:09:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip as a Result of a Turbine Generator Trip | This 4-hour notification is being made in accordance with 10CFR50.72(b)(2)(iv)(b). Salem Unit-1 reactor automatically tripped at 1109 (RPS actuation). The trip was initiated due to turbine trip. The cause of the turbine trip is being investigated. All safety systems functioned as designed with the exception of control rod 1SC1, which did not fully insert. 1SC1 indicates 16 steps. It should indicate zero steps when fully inserted. Auxiliary Feedwater pumps started as expected. Off-site power is available. Emergency diesel generators are available but not required at this time. During the implementation of the EOP's, a steam leak was reported in the Turbine Building. The Main Steam Isolation Valves (MSIV's) were closed as a conservative measure. This is an 8-hour reportable occurrence in accordance with 10CFR50.72(b)(3)(iv)(a). The leak was subsequently identified as a feedwater leak on the 11CN32 (11 Steam Generator Feed Pump suction valve). The Condensate System was placed in a normal shutdown line-up and the leak is not an impact to personnel safety or plant stability. Decay heat removal is via the atmospheric steam dumps at this time. The MSIV's are being bypassed to restore the main condenser as a heat sink. Salem Unit-1 is currently in Mode 3 with reactor coolant system temperature at approximately 549 deg F with pressure at 2235 psig. There was no equipment out of service that contributed to this event and there were no personnel injuries or radiological occurrences associated with this event. The licensee has notified the NRC Resident Inspector and will be making State and local notifications. A press release is expected. | Steam Generator Reactor Coolant System Feedwater Emergency Diesel Generator Main Steam Isolation Valve Auxiliary Feedwater Decay Heat Removal Main Condenser Control Rod | |
| ENS 41028 | 9 September 2004 05:06:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Turbine Trip/Reactor Trip | Salem Unit 2 experienced a Turbine Trip > P-9 which initiated a Reactor Trip. All rods fully inserted. Aux Feedwater auto start initiated on steam generator low-low level. The cause of the turbine trip is Generator Diff (Differential Current) or Loss of Field. Decay heat removal is via the steam dumps to the main condenser. All safety related systems are available and functioned as designed. The 500 KV breaker 2-10 was cleared and tagged prior to the trip. Upon the turbine/reactor trip, the 500KV breakers 1-9 and 9-10 opened as designed. Due to the 2-10, 1-9, 9-10 breakers being opened, the 500 KV transmission line 5037 is de-energized. Two offsite power sources are available and all 3 Emergency Diesel Generators are operable. The 5037 line will be energized following restoration from scheduled maintenance. There were no personnel injuries. The licensee stated that the NRC resident inspector and the local township were notified. | Steam Generator Feedwater Emergency Diesel Generator Decay Heat Removal Main Condenser | |
| ENS 40875 | 15 July 2004 23:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Trip from 8% Power Due to Lowering Water Level on 23 Steam Generator | During plant start up, a manual Reactor trip was initiated in response to lowering water level on 23 Steam Generator. All systems responded as required. All control rods fully inserted and no ECCS actuated or relief valves lifted. Both motor driven auxiliary feedwater pumps (21 and 22) started as expected on the low level signals from 23 Steam Generator. The steam driven auxiliary feed (23) pump was not required to start and remained in standby. Decay heat is being removed via the main steam dump system to the main condenser. The reactor is currently at normal operating temperature and pressure. No major equipment was unavailable at the time of the trip. No personnel injuries occurred as a result of this event. The cause of the low steam generator level is being investigated. The NRC Resident Inspector was notified along with Lower Alloway Creek Township. The States of New Jersey and Delaware will be notified. | Steam Generator Auxiliary Feedwater Main Condenser Control Rod Main Steam | |
| ENS 40865 | 13 July 2004 17:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Low Low Steam Generator Water Level. | An automatic Reactor Trip occurred on 21 Steam Generator low low level. The cause is currently under investigation. All systems responded as required. All control rods fully inserted. All auxiliary Feedwater pumps started as expected, on the low level signals and the steam driven auxiliary feed (23) pump was secured by procedure. Currently the steam dump system is rejecting heat to the main condenser for reactor coolant system temperature control, currently at normal operating temperature and pressure. No major equipment was unavailable at the time of the trip. All Emergency Core Cooling Systems and the Emergency Diesel Generators are fully operable if needed and the electrical grid is stable. The Licensee is investigating the cause of Steam Generator "21" low low water level. The NRC Resident Inspector was notified of this event by the licensee. | Steam Generator Reactor Coolant System Emergency Diesel Generator Auxiliary Feedwater Emergency Core Cooling System Main Condenser Control Rod | |
| ENS 40350 | 23 November 2003 10:19:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Was Manually Tripped After a Control Rod Dropped During Low Power Physics | During performance of low power physics testing for dynamic rod worth measurements, control rod bank D was being withdrawn. At control bank D position of 209 steps, control rod 1D4 dropped into the reactor core. The control room crew entered the abnormal operating procedure for a dropped (control) rod at 0507. Based upon the dropped control rod causing the reactor to go subcritical, the abnormal operating procedure directs that all (control) rods to be inserted. Based upon the control bank D not being fully withdrawn and not in the proper bank overlap due to low power physics testing, the Control Room Supervisor directed the reactor to be (manually) tripped. The crew entered the emergency operating procedure at 0519. All equipment functioned as designed and all major equipment is available. (The crew) exited the emergency operating procedures at 0538. The plant is currently stable in mode 3 at normal operating temperature and pressure. The cause of the drop rod is reported to be a blown fuse on the stationary coil. All reactor coolant pumps are in service and decay heat removal is through the steam dumps to the condenser. All control rods fully inserted when the reactor was manually tripped. Feedwater to the steam generators is being supplied by the auxiliary feedwater system. The licensee has notified the NRC Resident Inspector and will be notifying the LAC (Lower Alloways Creek) Township. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 39713 | 29 March 2003 09:42:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | MANUAL REACTOR TRIP DUE TO LOSS OF SECONDARY CIRCULATING WATER PUMP
"Salem Unit 2 was manually tripped in response to loss of circulating water pumps. Loss of circulators was caused by heavy grass intrusion from the Delaware River. At time of trip, a load reduction was in progress and power level was approximately 80%. All safety systems operated as designed during and after the reactor trip. Unit 2 is currently stable in Mode 3 at normal operating temperature and pressure. All reactor cooling pumps are in service and core cooling is being supported by auxiliary feedwater pumps and main steam dumps to the condenser. No safety systems or equipment was out of service at the time of the trip. Salem Unit 1 was also affected by the heavy river grass conditions and is currently in Mode 1 and stable at 55% reactor power. The licensee has notified the NRC Resident Inspector. | Auxiliary Feedwater Main Steam |