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 Entered dateSiteScramRegionReactor typeEvent description
ENS 5421111 August 2019 11:40:00SalemManual ScramNRC Region 1At 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 541983 August 2019 23:33:00Hope CreekManual ScramNRC Region 1At 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 5385231 January 2019 04:23:00SalemManual ScramNRC Region 1At 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 5360614 September 2018 16:28:00SalemAutomatic ScramNRC Region 1At 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 533867 May 2018 05:23:00SalemManual ScramNRC Region 1Westinghouse PWR 4-LoopThis 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 5221331 August 2016 18:12:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThis 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 5204828 June 2016 06:58:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThis 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 5173414 February 2016 23:13:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThis 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 517084 February 2016 13:33:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThis 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 5143028 September 2015 22:49:00Hope CreekAutomatic ScramNRC Region 1GE-4On 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 512905 August 2015 18:51:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThis 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 5055020 October 2014 01:36:00SalemManual ScramNRC Region 1Westinghouse PWR 4-Loop

This is an 8-hour notification being made to report that a valid ESF (emergency safety features) auxiliary feed water system actuation occurred. Salem Unit 1 was in mode 1 at 19% reactor power and executed a planned manual reactor trip to begin a scheduled refueling outage (1R23). The 11, 12, and 13 AFW (auxiliary feedwater) pumps were not in service prior to the reactor trip and as a result, narrow range levels in 12, 13, and 14 steam generators reached 14% before recovering. This resulted in a valid ESF actuation for low steam generator water level. Steam generator water levels were restored to normal post trip values as part of the procedurally directed response to a reactor trip. All safety related equipment was operational before the reactor trip and AFW actuation and responded as required. There were no equipment failures that contributed to this event. There were no personnel injuries as a result of' this event. The NRC Resident Inspector has been informed. A transformer alarm caused the operators to trip the reactor prior to reaching the step in the procedure to start the AFW pumps.

  • * * UPDATE FROM JOHN OSBORNE TO DONALD NORWOOD AT 1555 EST ON 11/24/2014 * * *

On October 20, 2014 at 0136 EDT, Salem Unit 1 reported an 8 hour notification of a valid ESF Auxiliary Feed Water System actuation following a manual reactor trip at the start of the S1R23 refueling outage. After additional review it was determined that the manual reactor trip met the criteria for a four hour report in accordance with 10CFR50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical...'. The licensee notified the NRC Resident Inspector. Notified R1DO (Noggle).

ENS 500927 May 2014 07:45:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThis 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
ENS 5003213 April 2014 23:49:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThis 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.
ENS 500128 April 2014 23:13:00SalemManual ScramNRC Region 1Westinghouse PWR 4-LoopThis 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.
ENS 4978031 January 2014 11:44:00SalemManual ScramNRC Region 1Westinghouse PWR 4-LoopThis 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.
ENS 496085 December 2013 05:40:00Hope CreekAutomatic ScramNRC Region 1GE-4

While operating at 76% power on 12/5/13 at 0325 EST, the main turbine tripped on moisture separator hi level. The reactor scrammed along with the main turbine trip. All safety systems responded as designed and expected. There was no radiological release. There were no injuries. During the scram, all rods inserted into the core. Plant is stable in Mode 3 in its normal S/D (shutdown) electrical line up. Decay heat is being removed via the turbine bypass valves dumping steam to the main condenser. At 0505 EST while securing from cooldown in an attempt to start a recirc pump, BPVs (Bypass Valve) opened causing reactor level swell and subsequent shrink. During this time, RPV (Reactor Pressure Vessel) level lowered to below RPV level 3 and caused a RPS (Reactor Protection System) actuation. RPV level was recovered and is now stable in normal band. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE ON 12/5/13 AT 1000 EST FROM LINDSAY KOBERLEIN TO DONG PARK * * *

This update to ENS #49608 adds reporting criterion 10CFR50.72(b)(3)(iv)(A) for the RPS actuation at 0505 EST during post-scram recovery.

The licensee notified the NRC Resident Inspector and the Lower Alloways Creek township. The licensee will be making a press release. Notified R1DO (Cook).

ENS 495921 December 2013 10:02:00Hope CreekAutomatic ScramNRC Region 1GE-4While operating at 100% power on 12/01/2013, at 0613 EST, the main turbine tripped on moisture separator hi level. The reactor scrammed along with the main turbine trip. All safety systems responded as designed and expected. There was no radiological release. There were no injuries. During the scram, all rods inserted into the core. The plant is stable in mode 3 in its normal shutdown electrical line up. Decay heat is being removed via the turbine bypass valves dumping steam to the main condenser. The licensee notified the NRC Resident Inspector and will be notifying Lower Alloways Creek township.
ENS 4910812 June 2013 16:59:00Hope CreekManual ScramNRC Region 1GE-4This is a report of a manual RPS actuation and manual RCIC actuation per 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). At 1332 (EDT), on 6/12/13, the 'B' Circulating Water Pump tripped with a stuck open discharge valve resulting in a vacuum transient. Operators lowered reactor power from 100% in an effort to stabilize condenser vacuum. When vacuum reached 6.5 inches, the operators inserted a manual reactor scram at 1333 (EDT). All control rods inserted as required. No automatic ECCS or RCIC initiations occurred. No primary or secondary containment isolations occurred. The plant is stable in OP CON 3 HOT SHUTDOWN with the condensate pumps in service. The Reactor Recirculation Pumps are in service. At the time of the event, a RCIC surveillance was in progress, but did not contribute to the event. The RCIC pump was secured and subsequently placed in service for inventory control. The only safety-related equipment out of service at the time of the scram was the C Service Water Pump, which was tagged for scheduled maintenance. No personnel injuries occurred. No radiation releases occurred. The NRC Resident Inspector has been informed.
ENS 4861721 December 2012 07:06:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopSalem 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.
ENS 4853425 November 2012 15:13:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThis 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.
ENS 4845730 October 2012 04:10:00SalemManual ScramNRC Region 1Westinghouse PWR 4-Loop

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.

  • * * UPDATE ON 10/30/12 AT 0835 EDT FROM JOHN OSBORNE TO DAN LIVERMORE * * *

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).

ENS 4776623 March 2012 17:00:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-Loop

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.

  • * * UPDATE FROM WILLIAM MUFFLEY TO PETE SNYDER AT 2127 EDT ON 3/23/12 * * *

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).

ENS 4699226 June 2011 22:08:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThis 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.
ENS 4677421 April 2011 19:16:00SalemManual ScramNRC Region 1Westinghouse PWR 4-Loop

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.

  • * * UPDATE FROM MATTHEW MOG TO VINCE KLCO ON 4/21/2011 AT 2241 EDT * * *

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).

ENS 4633717 October 2010 06:38:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-Loop

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.

  • * * UPDATE FROM BILL MUFFLEY TO JOHN SHOEMAKER 1138 EDT ON 10/18/10 * * *

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.

ENS 4633616 October 2010 01:52:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopAn 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.
ENS 460757 July 2010 13:22:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-Loop

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.

  • * * UPDATE FROM ERIC POWELL TO CHARLES TEAL ON 7/8/10 AT 1535 * * *

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).

ENS 4564721 January 2010 21:36:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopOn 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.
ENS 456043 January 2010 11:52:00SalemManual ScramNRC Region 1Westinghouse PWR 4-LoopA 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.
ENS 4507417 May 2009 05:24:00Hope CreekManual ScramNRC Region 1GE-4

At 0335, Hope Creek was manually scrammed due to indications of multiple control rods drifting. All rods indicate fully inserted. Reactor level is being controlled in the normal band with Start-up level control in automatic. Reactor pressure is being controlled by bypass valves to the main condenser. Recirc pumps are in service and no ECCS system actuations were reached. The failure is a solder joint on the air supply to HCU 22-15. A manual scram was reinserted at 0445 to mitigate the air leak. The licensee reset the scram to re-pressurize the scram air header. Once the leak was located, a second manual scram signal was initiated to secure the leak. No safety relief valves lifted during the transient. The electrical grid is stable and the plant is in a normal shutdown electrical lineup. The licensee will be notifying the Lower Alloways Creek Township and has notified the NRC Resident Inspector.

  • * * UPDATE ON 5/17/2009 AT 0552 FROM MICHAEL REED TO MARK ABRAMOVITZ * * *

The failure was on HCU 22-11 not 22-15. The licensee notified the NRC Resident Inspector. Notified the R1DO (Holody) via e-mail.

  • * * UPDATE ON 5/29/2009 AT 1133 FROM JIM PRIEST TO VINCE KLCO * * *
On 5/17/09, at 0335, Hope Creek automatically scrammed due to low Reactor Pressure Vessel water level approximately two seconds prior to locking the Reactor Mode Switch in Shutdown due to indications of multiple control rods drifting. All rods indicate fully inserted. Reactor level is being controlled in the normal band with Start-up level control in automatic. Reactor pressure is being controlled by bypass valves to the main condenser. Recirc. Pumps are in service and no ECCS system actuations were reached. The failure is a solder joint on the air supply to HCU 22-11. A manual scram was reinserted at 0445 to mitigate the air leak.

The licensee notified the NRC Resident Inspector. Notified the R1DO(Dentel).

ENS 4387328 December 2007 19:58:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThis 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.
ENS 435506 August 2007 15:21:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-Loop

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.

ENS 4339529 May 2007 11:22:00Hope CreekManual ScramNRC Region 1GE-4On 5/28/07 with Hope Creek in Operating Condition 1 at 100% Reactor power, an electrical transient and loss of 'A' and 'B' reactor feed pumps resulted in lowering reactor water level. Operators inserted a manual reactor scram at 0835 in response to the lowering reactor water level. Reactor water level lowered to (-) 38 inches subsequent to the manual scram, resulting in initiation of High Pressure Coolant Injection (HPCl) and injection to the reactor vessel. The Reactor Core Isolation Cooling (RCIC) system also initiated but tripped. Investigation of the cause of the electrical transient, loss of the 'A' and 'B' RFP's, and trip of the RCIC system are currently in progress. Initial review of the event indicates that all other systems operated as expected. Current plant conditions as of 1100 are: Hope Creek is in mode 3 at 715 psig with heat removal to the main condenser via the Main Turbine Bypass valves. All control nods fully inserted on the scram. This report also documents a 4 hour report under 10CFR50.72(b)(2)(iv)(A) for valid ECCS initiation and injection to the reactor vessel (RAL 11.3.1). The reactor is stable with the water level currently at 17 inches and feedwater being supplied by the 'C' feed pump. No Safeties lifted during the transient. All systems functioned as required except for the trip of the RCIC. The licensee was not in any major technical specification LCO at the time of the trip. The licensee notified the NRC Resident Inspector. The licensee will also notify the States of NJ and Delaware, and Lower Alloways Creek Township.
ENS 4338224 May 2007 05:11:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-Loop

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.

  • * * UPDATE AT 17:18 ON 5/25/2007 FROM ERIC POWELL TO MARK ABRAMOVITZ * * *

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.

ENS 4332930 April 2007 16:49:00SalemManual ScramNRC Region 1Westinghouse PWR 4-Loop

A manual trip of Salem Unit 1 was initiated due to a loss of the circulators from heavy grassing at the circulating water intake structure. All rods fully inserted on the trip and all systems responded as designed, with decay heat being removed via the atmospheric relief valves and subsequently the steam dump system. All Auxiliary Feedwater pumps auto started as expected on low steam generator levels from the trip. Salem Unit 1 had entered the abnormal operating procedure for circulating water due to two (2) circulators being out of service; one circulator previously emergency tripped due to excessive traveling water screen (differential pressure) and a second circulator emergency tripped soon thereafter. The procedure provides guidance to trip the plant if less than three (3) circulators are in service and power is above ten (10) percent. Two (2) of the remaining four (4) in-service circulators emergency tripped due to excessive traveling water screen differential pressure. At this point the manual reactor trip was initiated. The crew entered emergency operating procedures and stabilized the plant at no load conditions. No personnel injuries have occurred as a result of the trip. Salem Unit 2 was not affected and is operating at 100% power.

All ECCS and ESF Systems are available. Actions are being taken to return circulators to service. The licensee notified the NRC Resident Inspector.

ENS 4331725 April 2007 01:33:00SalemManual ScramNRC Region 1Westinghouse PWR 4-Loop

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.

      • UPDATE FROM LICENSEE(MEEKINS) TO KNOKE AT 0853 EDT ON 04/25/07 ***

The power level was at 40% when the manual trip occurred and not 100% as previously reported. Notified R1DO (Trapp)

ENS 4325928 March 2007 01:51:00SalemManual ScramNRC Region 1Westinghouse PWR 4-LoopThis 8-hour notification is being made to report that an Auxiliary Feedwater (AFW) system actuation signal was generated and a Reactor Protective Signal (RPS) generated with the reactor subcritical. On March 27, 2007 at approximately 2050, Salem Unit 1 was in Mode 3 following a planned manual reactor trip to begin a scheduled refueling outage. Operators had established initial Reactor Coolant System cooldown using the steam dumps for heat removal and the 11, 12, and 13 AFW pumps for steam generator make-up. All three pumps had been previously placed in-service in accordance with plant procedures. As plant cooldown proceeded, narrow range levels in 11, 12, and 13 Steam Generators reached the low steam generator setpoint trip, resulting in a valid ESF actuation (i.e., start signal to the AFW pumps); however all AFW pumps were already in-service. The actuation signal also generated a reactor trip signal to be initiated by the reactor protection system; however, the plant was already in a shutdown condition with the reactor trip breakers open. The lowest level during this transient occurred in 13 Steam Generator and was 11.3% narrow range level. Steam generator water level was restored to a normal value and the RCS cooldown recommenced. There were no equipment failures that contributed to this event. The NRC Resident has been informed.
ENS 4313230 January 2007 00:49:00Hope CreekAutomatic ScramNRC Region 1GE-4

On 1/29/07 with Hope Creek reactor startup in progress, in mode 1 at 22% Reactor power, Secondary Condensate Pump Minimum Flow Control Valves began to cycle. Reactor water level reached 39" (RPV level 7) and then lowered to 30" (RPV level 4). Manual control of the Reactor Feed pumps was taken, however, RPV water level continued to lower to 15" (Reactor scram on RPV water level is 12.5") at which time the reactor mode switch was locked in shutdown. There were no ECCS injections and all ECCS systems are operable. Initial review of the event indicates that all systems operated as expected with the exception of the Secondary Condensate pump minimum flow valves and the 'A' IRM failed to insert. Current plant conditions as of 1/30/07 at 0010 are: Hope Creek is in mode 3 at 565 psig. The Main Steam Line Isolation valves are open. 'B' and 'C' Primary Condensate Pumps, 'B' and 'C' Secondary Condensate Pumps, and 'A' Rx Feed Pump are feeding the vessel. All control rods have fully inserted on the scram and Main Turbine Bypass valves are removing decay heat. The licensee will inform the LAC (Lower Alloways Creek Township) and has informed the NRC Resident Inspector.

  • * * UPDATE FROM REED TO HUFFMAN AT 1719 EST ON 2/01/07 * * *

Based on the post-trip review performed by the licensee, it was determined that the in-service Reactor Feed Pump Minimum Flow Recirculation Valve opened in response to the feed flow adjustments. Reactor vessel level reached the low-level trip set point and an automatic reactor scram occurred. During the event analysis it was determined that the operator initiated the manual scram two seconds after the automatic low level scram occurred. A post event equipment performance review concluded that the Secondary Condensate Pump Minimum Flow Recirculation Valves operated as expected. The licensee notified the NRC Resident Inspector. The R1DO (Cahill) notified.

ENS 4286426 September 2006 17:18:00SalemManual ScramNRC Region 1Westinghouse PWR 4-LoopSalem 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.
ENS 423958 March 2006 14:08:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThis 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.
ENS 417537 June 2005 15:30:00Hope CreekManual ScramNRC Region 1GE-4

Hope Creek manually scrammed the reactor from 100% power at 1413 and declared an Unusual Event due to unidentified drywell leakage exceeding 10 gpm (EAL 2.1.1.b) at 1437. The drywell unidentified leak rate peaked at approximately 15 gpm and is currently 12 gpm and slowly lowering. All safety systems were operable prior to the transient and responded as expected. Drywell pressure peaked at approximately 0.5 psig and is steady using normal drywell cooling (the normal pressure band is 0.1 to 0.7 psig). Drywell and suppression pools sprays were not required to mitigate the drywell pressure transient. Reactor vessel level lowered to approximately -30 inches following the scram and was returned to the normal level band using the feedwater and condensate systems. The expected vessel level 3 (setpoint +12.5 inches) ESF actuations occurred. The plant is proceeding to cold shutdown to investigate the drywell leak. The licensee notified the NRC Resident Inspector.

  • * * UPDATE BY NRC (HOLIAN) TO HUFFMAN AT 0330 EDT ON 6/08/05 * * *

As of 0330, Region I IRC in consultation with NRC/IRD (McGinty) secured from Monitoring Mode based on the plant being stable at about 55 psig (about 300 degrees F) and preparing to initiate shutdown cooling. The leak rate remains at about 8 gpm, and an initial drywell entry determined the source of the leakage to be from the A-loop of shutdown cooling testable check valve (50A ). The valve was found with the position indication failed/separated and an approximate 20 foot plume of steam/liquid coming out. Plans are to continue to cool down the plant, go onto the B-loop of shutdown cooling , and isolate valve 50A. The licensee has conservatively remained in the UE and plans to exit when leak rate is assured to remain below 10 gpm (EAL entry condition) or cold shutdown is achieved and the EAL is no longer applicable. The licensee's outage center remains manned, and an NRC inspector remains on site around the clock. DHS (Hoisington), FEMA (Sweetser), DOE (Turner) EPA (Crews), USDA (Pimmons), HHS (Williams) were notified.

  • * * UPDATE BY NRC (MANGAN - RG 1) TO HUFFMAN AT 0500 EDT ON 6/08/05 * * *

The licensee has placed RHR loop B into service and reached Mode 4 (cold shutdown) at 0455. Preparations are in progress to isolate the leak by closing manual valve 183.

  • * * UPDATE FROM LICENSEE (WILSON) TO HUFFMAN AT 0530 EDT ON 6/08/05 * * *

The licensee terminated the Unusual Event at 0515 EDT based on reaching cold shutdown with the leak rate less than 10 gpm. The licensee notified the NRC Resident Inspector. DHS (Hoisington) and FEMA (Sweetser), R1D0 (Jackson), NRR EO (Hannon), and IRD (Leach) notified.

  • * * UPDATE FROM LICENSEE (BAUER) TO HUFFMAN AT 1101 EDT ON 6/08/05 * * *

On the morning; of 06/08/05, investigation into a previously reported increase in unidentified drywell leakage (Event #41753) identified the leak location as the F050A residual heat removal (RHR) check valve. The F050A check valve is on the return line to the 'A' recirculation loop,. The F050A check valve was isolated at approximately 0545 hours. The position indication magnatrol assembly for the F050A check valve appears to be the source of the leakage. Additional investigation is proceeding to identify the exact location of the leak as well as address the structural integrity of the valve. This updated report is being made in accordance with 10CFR50.72(b)(3)(ii). At the time of this notification Hope Creek Generating station is in OPCON 4 (Cold Shutdown) at 122 degrees reactor coolant temperature. The licensee notified the NRC Resident Inspector. R1DO (Jackson) and NRR EO (Jung) notified.

ENS 4111712 October 2004 19:45:00Hope CreekManual ScramNRC Region 1GE-4During a review of post trip activities associated with MANUAL REACTOR SCRAM DUE TO A STEAM LEAK IN THE TURBINE BUILDING (Event 41110) on 10/10/04, it was determined that Technical Specifications actions requirements were inappropriately applied. With both loops of RHR in suppression pool cooling (necessary with SRV's controlling reactor pressure), procedural guidance requires that the affected loop of RHR be declared inoperable when in a secondary mode of operation. With both loops of RHR thus inoperable, the applicable Technical Specification Action TS 3.6.2.3 Action b requires that the plant be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. In accordance with the Technical Specification, this action was entered on 10/10/04 at 1831. The required time to cold shutdown was incorrectly noted as 0631 on 10/12/04. The required time was based on the combination of the 12 hours to hot shutdown and 24 to cold shutdown (or 36 hours). Because the plant was already in hot shutdown, the action should have been to place the plant in cold shutdown within 24 hours or by 1831 on 10/11/04. As a result of this error, planning activities and cooldown to cold shutdown condition was predicated on a target time of 0631 on 10/12/04 resulting in the plant exceeding the 24 hour AOT. This constitutes a condition prohibited by Technical Specifications. The plant achieved cold shutdown on 10/12/04 at 0509 hours. In addition, Emergency Classification Guide (ECG) Initiating Condition 8.5 states that the inability to reach required operational condition within Technical Specification Limits and requires the declaration of an Unusual Event if the plant is not brought to the required Operational Condition within the Technical Specification required time limit. There are no safety consequences associated with this error. There were no issues associated with the transition to cold shutdown that would have constituted an emergency condition requiring initiation of the Emergency Plan. The missed LCO and subsequent classification was based on an erroneous TS Action time and, as such, exceeding the specification occurred as a result of scheduling not plant conditions. The licensee will inform the NRC resident inspector.
ENS 4111011 October 2004 00:49:00Hope CreekAutomatic ScramNRC Region 1GE-4At 2153 (hrs. EDT) on October 10, 2004, the Hope Creek Generating Station experienced an automatic reactor scram signal on low reactor level +12.5 inches (Level 3) while cooling down following a manual scram. As previously reported under Event Notification 41109, the Main Steam Isolation Valves (MSIV's) were closed as the result of a steam leak in the Turbine Building. The +12.5 inch (Level 3) scram occurred from the manual closure of a Safety Relief Valve (SRV) while it was being manually operated to reduce reactor pressure. The SRV was closed when reactor level was +24 inches, resulting in a reactor level shrink. Reactor level lowered to +8 inches, and stabilized. The secondary condensate pumps immediately restored reactor level to its normal band following the scram signal. SRV's were being utilized to assist the plant cool down because the High Pressure Coolant Injection (HPCI) system had been manually taken out of service. The HPCI vacuum tank vacuum pump tripped on an overload/power failure condition, and use was not desired. The Reactor Core Isolation Cooling (RCIC) system was out of service because of a high reactor level condition, due to plant cool down. Also, the Reactor Water Cleanup (RWCU) system was out of service due to the initial manual scram that occurred at 1814 hours which prevented normal reactor level blow down. The NRC Resident Inspector was notified and Lower Alloway Creek Township will be notified. HOO note: See Event # 41109
ENS 4110910 October 2004 21:48:00Hope CreekManual ScramNRC Region 1GE-4

At 1814 (hrs. EDT) on October 10, 2004, Hope Creek Generating Station was manually scrammed due to a steam leak in the Turbine Building. All Control Rods inserted fully. Subsequent to the manual actuation of the Reactor Protection System, reactor pressure was reduced to minimize the effects of the steam leak. Degrading Main Condenser Vacuum following the scram resulted in trips of all operating Reactor Feed Pump Turbines at 10 (inches) HgA. The High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems were manually initiated for reactor level control and the Main Steam Isolation Valves (MSIV's) were closed to isolate the leak - MSIV closure was completed prior to reaching the Main Condenser Vacuum isolation setpoint of 21.5 (inches) HgA. During plant stabilization, Reactor Water Level lowered below the RPS actuation setpoint of 12.5 inches four separate times. First, following the initial scram. Second, immediately following initiation of the HPCI and RCIC systems, when the 'A' and 'B' Reactor Water Level channels lowered to -38 inches (Level 2). Level 2 is the HPCI and RCIC actuation setpoint and Primary Containment Isolation actuation setpoint for Groups 2, 7, 8, 9, 12, 13, 14, 17, 18, 19, and 20 valves. Because only two of the four Level 2 instrument channels actuated, the isolation of these systems was channel dependent and occurred as required by the respective isolation logic. Third, following manual closure of the MSIVs. Finally, Reactor Water Level lowered below 12.5 inches following reset of the original manual scram signal which resulted in an automatic scram signal. RCIC was re-initiated manually to restore Reactor Water Level. No personnel were injured during this event. The plant is currently stable in OPCON 3 with reactor pressure at 615 psig. Pressure control (decay heat removal) was transitioned to HPCI in pressure control mode during plant stabilization. Reactor Water Level is being maintained with the Secondary Condensate Pumps. Two loops of RHR in Suppression Pool Cooling mode are in service with Suppression Pool Temperature at 110 degrees F in compliance with Technical Specification 3.6.2.1 Action b.2. Actions to determine the cause of the steam leak and effect repairs are in progress. The licensee will inform Lower Alloway Creek Township and has informed the NRC resident inspector.

  • * * UPDATE ON 10/11/04 @ 0049 HRS EDT BY BAUER TO GOULD * * *

On steam leak investigation, a walk down of the turbine building condenser bay determined the source of the leak to be a failure of an 8 inch moisture separator dump line. The line break is located approximately one foot from the condenser shell penetration. An additional investigation into the root cause of the failure has commenced. The NRC Resident Inspector was notified and Lower Alloway Creek Township will be notified. The Reg 1 RDO (Richard Barkley) and EO (Chris Grimes) were informed. HOO Note: See Event # 41110

ENS 4087515 July 2004 22:18:00SalemManual ScramNRC Region 1Westinghouse PWR 4-LoopDuring 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.
ENS 4086513 July 2004 14:53:00SalemAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopAn 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.
ENS 4043712 January 2004 12:30:00Hope CreekManual ScramNRC Region 1GE-4On 01/12/04 at 1048 hours, the Hope Creek Generating Station reactor was manually scrammed following an invalid containment isolation signal on Reactor Building High-High Radiation. The invalid signal was caused by the combination of a scheduled sensor calibration on channel 'C', coincident with an emergent failure on channel 'A.' This combination of trip signals made up the two out of three trip logic for the Reactor Building High-High Radiation containment isolation signal. While recovering from the spurious isolation signal, the operating crew observed two of the inboard MSIV's drifting closed from a loss of pneumatic pressure as a result of the isolation signal. In response to this condition, the operating crew manually scrammed the reactor. A low reactor water level scram signal was received at 12.5 inches as expected, and reactor level was subsequently returned to the normal band using the reactor feedpumps. At the time of this event, the 'A' Control Room Ventilation Train was inoperable but available pending emergent corrective maintenance. The 'C' channel Reactor Building Radiation monitor has been returned to service and is operable, and the 'A' channel remains failed in the tripped condition. All other systems functioned as expected, and a post-transient review team is being assembled to investigate the event. Decay heat is being removed via steam to the main condenser using the bypass valves. The condensate and feedwater system is in operation maintaining reactor vessel water level. No SRVs lifted during the transient and the electrical system is stable in a normal lineup. The licensee notified the NRC Resident Inspector and will be notifying the LAC Township.
ENS 403786 December 2003 04:44:00Hope CreekManual ScramNRC Region 1GE-4

The reactor was being shutdown as part of a planned evolution to allow repairs on the Reactor Water Cleanup flange leak. After the Reactor Protection System Mode Select Switch had been placed in shutdown, the resulting reactor level transient caused the Level 3 low reactor level set point to be reached. The Reactor Protection System had already been de-energized and the lowest level reached during the transient was +2 inches. This level transient is a normal occurrence on a reactor shutdown, and level was restored to the normal operating band. There was no effect on the plant due to reaching the low level set point. No other abnormal plant response was noted. The licensee will notify the NRC Resident Inspector

  • * * RETRACTION ON 1/9/04 AT 1310 FROM CLYDE BAUER TO E. THOMAS * * *

Upon further review of this event, the resulting Level 3 low reactor water level signal following the manual scram is considered part of the pre-planned sequence in accordance with the guidance of NUREG-1022. Therefore, this event is not reportable under 10 CFR50.72(b)(3)(iv)(A) and is being retracted. Notified R1DO (J. Noggle)

ENS 4035023 November 2003 06:48:00SalemManual ScramNRC Region 1Westinghouse PWR 4-LoopDuring 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.