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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56889 | 15 December 2023 00:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram | The following information was provided by the licensee via phone call and email: On December 14, 2023, at 1939 EST, Hope Creek reactor scrammed following closure of turbine control valve number 4. All control rods fully inserted into the core. All safety systems responded as designed and expected. There was no radiological release. The unit is stable in mode 3 with decay heat being removed via the turbine bypass valves rejecting steam to the main condenser. Normal feedwater level control is providing makeup to the reactor vessel. No personnel injuries resulted from the event. The outage control center has been staffed to determine the cause of the reactor scram. The Hope Creek NRC Resident Inspector has been notified. | Feedwater Main Condenser Control Rod | |
| ENS 54198 | 3 August 2019 23:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip and Manual Actuation of Reactor Core Isolation Cooling | 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. | Feedwater Reactor Protection System Reactor Core Isolation Cooling Main Condenser Control Rod | |
| ENS 51430 | 29 September 2015 00:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Following Trip of Both Reactor Recirculation Pumps | 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. | Feedwater Reactor Recirculation Pump Main Condenser Control Rod | |
| ENS 49608 | 5 December 2013 08:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Following Turbine Trip on High Moisture Separator Level | 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). | Main Turbine Main Condenser | 05000354/LER-2013-009 |
| ENS 49592 | 1 December 2013 11:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Main Turbine Trip | 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. | Main Turbine Main Condenser | 05000354/LER-2013-008 |
| ENS 49108 | 12 June 2013 17:33:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Circulating Water Pump Trip Leads to Reactor Scram | 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. | Secondary containment Service water Reactor Recirculation Pump Control Rod | 05000354/LER-2013-002 |
| ENS 45074 | 17 May 2009 07:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Loss of Scram Air Header | 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). | Reactor Pressure Vessel Safety Relief Valve Main Condenser Control Rod | 05000354/LER-2009-004 |
| ENS 43132 | 30 January 2007 04:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram on Low Reactor Pressure Vessel (Rpv) Water Level | 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. | Reactor Pressure Vessel Main Steam Line Control Rod | |
| ENS 41109 | 10 October 2004 22:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Manual Reactor Scram Due to a Steam Leak in the Turbine Building | 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.
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 | High Pressure Coolant Injection Reactor Protection System Main Steam Isolation Valve Reactor Core Isolation Cooling Primary containment Decay Heat Removal Main Condenser Control Rod | |
| ENS 40437 | 12 January 2004 15:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Scram Following Invalid Containment Isolation Signal | 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. | Feedwater Main Condenser | |
| ENS 40224 | 4 October 2003 21:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Rps Actuation While Critical | 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. | Reactor Protection System Emergency Diesel Generator Primary Containment Isolation System Main Condenser Control Rod |