|Entered date||Site||Region||Scram||Reactor type||Event description|
|ENS 54198||3 August 2019 23:33:00||Hope Creek||NRC Region 1||Manual Scram||At 1947 (EDT) on 8/3/19, with Hope Creek in Mode 1 at 37 percent power, the reactor was manually scrammed due to loss of condenser vacuum. All control rods fully inserted into the core. All safety systems responded as designed and expected. Reactor level was stabilized using Reactor Core Isolation Cooling (RCIC) and Reactor Feedwater Pumps. Currently reactor water level is being maintained by the feedwater system and decay heat is being removed by the main condenser using the main turbine bypass valves. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Due to the manual actuation of RCIC, this event is also being reported as an eight-hour, non-emergency notification per 10 CFR 50. 72(b )(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The plant is in its normal shutdown electrical lineup with all safe shutdown equipment available. The licensee will be notifying the state of Delaware, state of New Jersey and the Lower Alloway Creek township.|
|ENS 51430||28 September 2015 22:49:00||Hope Creek||NRC Region 1||Automatic Scram||GE-4||On September 28, 2015 at 2046 EDT, the Hope Creek reactor scrammed following a trip of both reactor recirculation pumps. All control rods fully inserted into the core. All safety systems responded as designed and expected. There was no radiological release. The unit is stable in Mode 3 with decay heat being removed via the turbine bypass valves rejecting steam to the main condenser. Normal feedwater level control is providing makeup to the reactor vessel. No personnel injuries resulted from the event. The Outage Control Center has been staffed to determine the cause of the reactor scram. The Hope Creek NRC Resident Inspector has been notified. The licensee notified Lower Alloways Creek township of the event.|
|ENS 49608||5 December 2013 05:40:00||Hope Creek||NRC Region 1||Automatic Scram||GE-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.
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 49592||1 December 2013 10:02:00||Hope Creek||NRC Region 1||Automatic Scram||GE-4||While 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 49108||12 June 2013 16:59:00||Hope Creek||NRC Region 1||Manual Scram||GE-4||This 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 45074||17 May 2009 05:24:00||Hope Creek||NRC Region 1||Manual Scram||GE-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.
The failure was on HCU 22-11 not 22-15. The licensee notified the NRC Resident Inspector. Notified the R1DO (Holody) via e-mail.
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 43395||29 May 2007 11:22:00||Hope Creek||NRC Region 1||Manual Scram||GE-4||On 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 43132||30 January 2007 00:49:00||Hope Creek||NRC Region 1||Automatic Scram||GE-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.
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 41753||7 June 2005 15:30:00||Hope Creek||NRC Region 1||Manual Scram||GE-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.
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.
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.
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.
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 41117||12 October 2004 19:45:00||Hope Creek||NRC Region 1||Manual Scram||GE-4||During 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 22.214.171.124 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 41110||11 October 2004 00:49:00||Hope Creek||NRC Region 1||Automatic Scram||GE-4||At 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 41109||10 October 2004 21:48:00||Hope Creek||NRC Region 1||Manual Scram||GE-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 126.96.36.199 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.
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 40437||12 January 2004 12:30:00||Hope Creek||NRC Region 1||Manual Scram||GE-4||On 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 40378||6 December 2003 04:44:00||Hope Creek||NRC Region 1||Manual Scram||GE-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
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 40224||4 October 2003 20:04:00||Hope Creek||NRC Region 1||Manual Scram||GE-4|
Hope Creek Generating Unit was manually scrammed at 1713 hours (EDT) on 10/04/03 due to an Electro Hydraulic Control (EHC) System oil leak. Prior to the event the unit was at 100% power. The plant responded as designed for the scram, with lowest reactor level reaching -8 inches. Reactor level is currently being maintained between +12.5 inches and +54 inches with secondary condensate pumps. The unit is currently in Mode 3 - Hot Shutdown with reactor pressure being maintained between 500-600 psig with the main turbine bypass valves utilizing the main condenser as a heat sink. The EHC leak was validated to be associated with the #4 Combined Intermediate Control Valve (CIV) and has since been isolated. This report is being generated Law Event Classification Guide section 11.3.2 - Actuation of Reactor Protection System (RPS) when critical except preplanned. Current safety system status is normal with the exception that the `B' Emergency Diesel Generator (EDG) is inoperable as the result of a relay failure and the 'B' Control Room chiller is inoperable as the result of a failed economizer float. The 'B' EDG has been retested and validation of test results are currently underway to determine operability. Common mode failure testing is in progress for the remaining three EDG's. All control rods inserted fully during the reactor scram. No relief valves lifted during the transient and there were no ECCS actuations or Primary Containment Isolation System actuations. The electrical grid remained stable during the event. NRC Resident was notified by Licensee.
During reactor level recovery following the scram, a second level 3 (+12.5 inches) RPS scram signal was received. The RPS system was still actuated upon receipt of this second level 3 signal. R1DO (Cobey ) notified of update. NRC Resident will be notified by licensee.