Semantic search
| 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 52347 | 5 November 2016 08:04:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Actuation of Rps While Reactor Shutdown | On November 5, 2016 an RPS (Reactor Protection System) actuation occurred from an actual high scram discharge volume level reaching the RPS actuation setpoint. This actuation was the result of a Redundant Reactivity Control System (RRCS) signal inadvertently generated during excess flow check valve testing with the reactor in cold shutdown. At the time of the actuation, all control rods were inserted. RCS pressure was approximately 830 psig to support excess flow check valve testing and shutdown cooling was removed from service. When RRCS initiated, the B Reactor Recirculation Pump tripped as expected and the scram air header depressurized as expected, which caused the high level in the scram discharge volume. The cause of the RRCS signal is being investigated. The A loop of RHR was placed back into the Shutdown Cooling mode of operation with reactor temperature being maintained at approximately 150 degrees F. There were no injuries as a result of this event. The licensee has notified Lower Alloways Creek Township and the NRC Resident Inspector. | Shutdown Cooling Reactor Recirculation Pump 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 45016 | 23 April 2009 13:10:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Actuation of Auto Load Sequencer During Fast Bus Transfer Testing | At 0910 during planned testing of the 10A401 vital bus, an unplanned momentary loss of power was experienced when testing the fast transfer function. A delay of approximately 1 second was experienced when forcing a fast transfer (simulating a transformer lockout via the 86 device) from the alternate to the normal 4KV bus infeeds. This condition resulted in an initiation of the 'A' channel LOP sequencer (ELS) as a direct result of bus under voltage and is considered valid ESF actuation. All other equipment operated as expected that was available for service. The bus de-energized for <1 second and restored power automatically from the bus normal infeed. No major loads were lost. The channel is currently inoperable for planned maintenance and is not required per tech specs. Prior to the transient, the reactor was shut down, in operational condition 5, with all control rods fully inserted, and the reactor cavity flooded and communicating with the spent fuel pool. All other 'A' and 'C' channel equipment, including the 'A' and 'C' emergency diesel generators are inoperable for maintenance. Decay heat is being removed by the 'B' loop of residual heat removal operating in the shutdown cooling mode. No personnel were injured. The licensee is conducting an investigation to understand the response and cause. The licensee has notified the NRC Resident Inspector and will notify the township.
During a subsequent review of the event, it has been determined that this was not a valid actuation of one of the systems listed in 10CFR50.72(b)(3)(iv)(B). The original report was made based on the event being an actuation of (8) Emergency AC electrical power systems, including: emergency diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs'. The Hope Creek (HC) Updated Final Safety Analysis Report (UFSAR) �8.3.1.1.2 states that there are three potential power sources for the Class 1E AC power system: (a) normal power source, (b) alternate power source and (c) dedicated standby diesel generator (SDG). The HC UFSAR �8.3.1.1.3 lists the components of the 'Standby Power System' (Emergency AC Power System). The ELS is not one of the components listed. Additionally, the sequence of events did not complete the required logic to actuate a start of the EDG. Since no EDG actuation signal was generated, this was neither a valid nor invalid actuation of a system listed in 10CFR50.72(b)(3)(iv)(B) and is not reportable. The licensee notified the NRC Resident Inspector. Notified the R1DO (Caruso). | ||
| ENS 43753 | 29 October 2007 02:03:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Refuel Floor Exhaust Hi-Hi Radiation Alarms During Reactor Reassembly | On 10/28/07 at 2203 in Operational Condition 5, two of three Refuel Floor Exhaust Radiation Monitors actuated on HI-HI Radiation during the Moisture Separator lift for Reactor Pressure Vessel Reassembly. Reactor Building Ventilation was procedurally secured at the time of the actuation to minimize the potential for airborne contamination. The cause of the elevated radiation levels at the Refuel Floor Exhaust Radiation Monitors is currently attributed to the lifting of the Moisture Separator closer to the surface of the water to obtain an unobstructed underwater path to the Reactor Pressure Vessel. The resulting ESF signal yielded an automatic start of the 'A' and 'D' Station Service Water Pumps, the Filtration Recirculation and Ventilation System, and load-shed of non-1E breaker loads. All systems responded as expected. Systems and components actuated by the condition have been restored to a normal stand-by alignment as required. No other systems were affected and the plant is stable in Operational Condition 5. The licensee will notify the NRC Resident Inspector. The licensee notified the Lower Alloways Creek Township.
Follow up investigation has revealed that the initiation of the signals was the result of radiation levels outside of the Refuel Floor Exhaust duct, and not from the in-duct activity. This would not be considered the parameter satisfying the requirements for initiation of the safety function of the system. The actual plant condition that satisfies the requirements for initiation of the safety function of the system is to detect conditions and initiate protective actions due to a release of steam or radioactive material internal to the duct that could result in the release of fission products to the environment. As a result, this is not considered a 'valid actuation' and is being retracted. Also, upon review of the refuel floor camera system, the reason for the HI-HI actuation was that during the Moisture Separator lift, the Moisture Separator broke the surface of the water by approximately 24 inches. This caused the radiation monitors to alarm on radiation shine from the moisture separator rather than airborne activity in the exhaust duct. This notification also provides telephone notification of an invalid actuation other than actuation of the reactor protection system instead of submitting a written LER 1. This is not considered an LER 2. This report is being made under 10 CFR 50.73(a)(2)(iv)(A). 3. The resulting ESF signal yielded an automatic start of the 'A' and 'D' Station Service Water Pumps, the Filtration, Recirculation and Ventilation System (FRVS), and load-shed of non-1E breaker loads: all systems responded as expected. This was a complete actuation. All required systems started and functioned successfully. The licensee notified the NRC Resident Inspector. Notified the R1DO (Holody). | Service water Reactor Protection System Reactor Pressure Vessel Reactor Building Ventilation | |
| ENS 43395 | 29 May 2007 12:35:00 | 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 Loss of Feed Pumps After Electrical Transient | 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. | Feedwater High Pressure Coolant Injection Reactor Core Isolation Cooling Main Condenser | |
| 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 41110 | 11 October 2004 01:53:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | a Rps Actuation Signal Occurred Due to Low Reactor Water Level While in Mode 3 | 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 | High Pressure Coolant Injection Main Steam Isolation Valve Reactor Core Isolation Cooling Reactor Water Cleanup Safety Relief Valve | |
| 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 41094 | 5 October 2004 17:34:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Inadvertent Undervoltage Condition Initated Diesel Load Sequencer, Etc. Diesel Did Not Start. | At approximately 1334, during realignment from monthly surveillance testing of the normal and alternate power supply breakers to the 10 A404 vital 4 Kv bus, an inadvertent undervoltage condition appears to have occurred. This condition resulted in initiation of the diesel load sequencer and tripping of the normal loads supplied by this bus. The undervoltage condition was momentary in nature, the load sequencer stopped upon restoration of voltage prior to starting the emergency diesel generator and operators successfully restarted the equipment that had tripped and restored the load sequencer. The D Emergency Diesel Generator is considered to be operable at this time. The cause of the occurrence is under investigation. The plant is stable in Operational Condition 1 at 100% power. This event is being reported in accordance with 10CFR50.72(b)(3)(iv)(A) 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) subsection 8', specifically in this case, Emergency ac electrical power systems." LAC Township notified by licensee. The NRC Resident Inspector was notified of this event by the licensee.
Upon further review this event was determined to not meet the reportability requirements of 10CFR50.72. The event that caused the momentary interruption to the 10A404 vital bus (i.e., less then a second) did not result in an actuation of a listed system (Emergency ac electrical power systems, including: Emergency diesel generators (EDGs)). Because the event did not meet the NUREG 1022 reporting requirement specified in 10CFR50.72 (b)(3)(iv)(A) it has been determine to not be reportable." The NRC Resident Inspector was notified of this event by the licensee. LAC Township notified by licensee. Notified R1DO (Dimitriadis) | Emergency Diesel Generator | |
| 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 40378 | 6 December 2003 04:49:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | a Level 3 Low Reactor Water Level Was Reached During a Reactor Water Level Transient | 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) | Reactor Protection System Reactor Water Cleanup | |
| 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 |