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| Discovered date | Reporting criterion | Title | Event description | |
|---|---|---|---|---|
| ENS 56810 | 23 October 2023 00:48:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Average Power Range Monitors Declared Inoperable | The following information was provided by the licensee via phone and email: On October 21, 2023, at 2048 EDT, reactor recirculation pump (RRP) 12 tripped. The cause for the trip is under investigation. Following the RRP trip, the average power range monitors (APRMs) flow bias trips were inoperable due to reverse flow through RRP 12. The APRMs were restored to operable on October 21, 2023, at 2058 EDT, when the RRP 12 discharge blocking valve was closed. This 8-hour non-emergency report is being made based upon requirements of 10CFR50.72(b)(3)(v)(A) which states: "Licensee shall notify the NRC of any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down the reactor and maintain it in a safe shutdown condition." The NRC Resident Inspector has been notified. |
| ENS 55514 | 11 October 2021 17:21:00 | 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 | Automatic Reactor Scram After Main Turbine Trip | At 1321 EDT on October 11, 2021, Susquehanna Steam Electric Station Unit 2 reactor automatically scrammed due to a trip of the Main Turbine. Unit 2 reactor was being operated at approximately 95 percent RTP (rated thermal power) with no evolutions in progress. The Control Room received indication of a Main Turbine trip with both divisions of RPS (Reactor Protection System) actuated and all control rods inserted. Turbine bypass valves opened automatically to control reactor pressure and subsequently failed open causing the reactor to depressurize. When reactor pressure reached approximately 560 psig, the operations crew manually closed the Main Steam Isolation Valves (MISVs) to stop the depressurization. Reactor water level lowered to -31 inches causing Level 3 (+13 inches) isolations. No (automatic) ECCS (Emergency Core Cooling System) actuations occurred. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) were manually initiated to control reactor water level. The Operations crew subsequently maintained reactor water level at the normal operating band using RCIC and reactor pressure was controlled with HPCI in pressure control mode and main steam line drains. The Reactor Recirculation Pumps tripped as designed on EOC-RPT (end of cycle recirculation pump trip). The reactor is currently stable in Mode 3. An investigation into the cause of the turbine trip is underway. The NRC Resident Inspector was notified. A voluntary notification to PEMA will be made. This event requires a 4 hour ENS notification in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour ENS notification in accordance with 10 CFR 50.72(b)(3)(iv)(A). |
| ENS 55370 | 21 July 2021 22:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor SCRAM | At 1826 EDT on July 21, 2021, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed due to a trip of the Main Turbine. Unit 1 reactor was operating at 100 percent reactor power with no evolutions in progress. The Control Room received indication of a Main Turbine trip with both divisions of RPS (Reactor Protection System) actuated and all control rods inserted. The Reactor Recirculation Pumps tripped on EOC-RPT (end of cycle recirculation pump trip). Reactor water level lowered to +8 inches causing Level 3 (+13 inches) isolations. No ECCS (Emergency Core Cooling Systems) or RCIC (Reactor Core Isolation Cooling system) actuations occurred. The Operations crew subsequently maintained reactor water level at the normal operating band using Reactor Feed Water. The reactor is currently stable in Mode 3 with main condenser available. Investigation into the trip of the Main Turbine is in progress. The NRC Resident Inspector was notified. A voluntary notification to PEMA will be made. This event requires a 4 hour ENS notification in accordance with 10CFR50.72(b)(2)(iv)(B) and an 8 hour ENS notification in accordance with 10CFR50.72(b)(3)(iv)(A) and 10CFR50.72(b)(3)(iv)(B). |
| ENS 55137 | 13 March 2021 02:02:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Reactor Recirculation Pump Trip | On March 12, 2021, at 2102 (EST), Reactor Recirculation Pump (RRP) 13 tripped. The cause for the trip is under investigation. Following the RRP trip, the Average Power Ranger Monitors (APRMs) flow bias trips are inoperable due to reverse flow through RRP 13. The APRMs were restored to operable on March 12, 2021, at 2110 (EST) when the RRP 13 Discharge Blocking Valve was closed. This 8-hour non-emergency report is being made based upon requirements of 10 CFR 50.72(b)(3)(v)(A) which states: 'Licensee shall notify the NRC of any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down the reactor and maintain it in a safe shutdown condition.' The licensee has notified the NRC Resident Inspector. |
| ENS 54849 | 21 August 2020 14:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram | On August 21, 2020 at 0908 CDT, River Bend Station was operating at 100% reactor power when reactor recirculation pump 'B' tripped. At 0918 CDT, a manual reactor scram was inserted at 67% reactor power after receiving indications of thermal hydraulic instability as indicated by flux oscillations on the period based detection system (PBDS) and average power range monitors (APRMs). All control rods fully inserted and there were no complications. All systems responded as designed. Currently River Bend Station Unit 1 is stable and pressure is being maintained using turbine bypass valves. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical and 10 CFR 50.72 (b)(3)(iv)(A) Specified System Actuation as result of Group 3 isolations. NRC Resident Inspector has been briefed on this event. No radiological releases have occurred due to this event from the unit. |
| ENS 54691 | 3 May 2020 12:21:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Main Turbine Trip | At 0821 EDT on May 3, 2020, the Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed due to a trip of the Main Turbine. The Unit 1 reactor was operating at 76 percent reactor power following a ramp schedule to full power subsequent to a maintenance outage. The Control Room received indication of a Main Turbine trip with both divisions of the Reactor Protection System actuated and all control rods inserted. The Reactor Recirculation Pumps tripped on End of Cycle - Recirculation Pump Trip. Reactor water level lowered to -1 inch causing Level 3 (+13 inches) isolations. No Emergency Core Cooling System or Reactor Core Isolation Cooling actuations occurred. The operations crew subsequently maintained reactor water level at the normal operating band using Reactor Feed Water. No Steam Relief Valves opened. The reactor is currently stable in Mode 3. Investigation into the trip of the Main Turbine is in progress. The NRC Resident Inspector was notified. A voluntary notification to the Pennsylvania Emergency Management Agency and press release will occur. This event requires a 4-hour Emergency Notification System (ENS) notification in accordance with 10 CFR 50.72(b)(2)(iv)(B) and an 8-hour ENS notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(3)(iv)(B). |
| ENS 54525 | 14 February 2020 05:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Rising Condenser Backpressure | At 0025 EST on February 14, 2020, Susquehanna Steam Electric Station Unit 2 reactor was manually scrammed due to rising Main Condenser backpressure caused by a loss of the Unit 2 Offgas Recombiner. Unit 2 reactor was being operated at maximum facility output, approximately 98% RTP, when at 0012 EST, Unit 2 Recombiner 0C145 Panel Trouble and 2C198 HWC Panel Trouble alarms were received along with rising Main Condenser backpressure. Initial Main Condenser backpressure was 2.6 in HgA and was rising at approximately 0.3 HgA/min. A Recirc Lim 2 was inserted to lower reactor power and condenser backpressure continued to rise following the reduction in reactor power. A manual scram was inserted at 0025 EST by placing the Mode Switch to Shutdown when condenser backpressure rose to 6 in HgA. All control rods inserted. Reactor water level lowered to -30 inches causing Level 3 (+13 inches) isolation and partial (Division 2) Level 2 (-38 inches) isolation. No ECCS actuations occurred and RCIC initiated. The Operations crew subsequently maintained reactor water level at the normal operating band using Reactor Feed Water. No steam relief valves opened. The Reactor Recirculation Pumps remained in service. The reactor is currently stable in Mode 3. Investigation into the cause of the loss of Unit 2 recombiner is underway. The NRC Senior Resident Inspector was notified. A voluntary notification to PEMA and press release will occur. |
| ENS 54199 | 4 August 2019 21:45:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Average Power Range Monitors Declared Inoperable | On August 4, 2019 at 1745 (EDT), Reactor Recirculation Pump (RRP) 11 tripped. The cause for the trip is under investigation. Following the RRP trip, the Average Power Range Monitors (APRMs) flow bias trips are inoperable due to reverse flow through RRP 11. The APRMs were restored to operable on August 4, 2019 at 1807, when the RRP 11 Discharge Blocking Valve was closed. This 8-hour non-emergency report is being made based upon requirements of 10CFR50.72(b)(3)(v)(A) which states: 'Licensee shall notify the NRC of any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down the reactor and maintain it in a safe shutdown condition.' The licensee has notified the NRC Resident Inspector. |
| ENS 53584 | 19 March 2018 04:00:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Specified System Actuation | Pursuant to 50.73(a)(1) the following information is provided as a sixty (60) day telephone notification to the NRC. This notification, reported under 50.73(a)(2)(iv), is being provided in lieu of the submittal of a written LER (Licensee Event Report) to report a condition that resulted in an invalid actuation of the high pressure coolant injection (HPCI). At Nine Mile Point Unit 1, HPCI is a flow control mode of the normal feedwater system and is not an emergency core cooling system. On March 19, 2018 Nine Mile Point Unit 1 (NMP1) was at 0 percent power and in cold shutdown in support of a planned maintenance outage. At approximately 0118 (EDT), a reactor water level transient initiated by the fill and vent of 12 Reactor Recirculation Pump (12 RRP) occurred. During the fill and vent, Reactor Pressure Vessel (RPV) level lowered quickly from the initial level of 68 inches and a low level alarm was received. Control Room Operators reduced Reactor Water Clean-Up (RWCU) reject flow to turn the level trend and clear the low level alarm generated off of the compensated, GEMAC, level instrumentation. RWCU reject flow was reduced by 50 percent which caused RPV level to start to rise. RPV level was raised to approximately 72 inches at which time the Reactor Operator began to raise reject flow to reestablish the normal level band. During the RPV level transient, with actual water level at 74 inches on the GEMAC, the Yarway level instrumentation, which is not density compensated and therefore invalid, reached 92 inches causing an invalid high RPV water level turbine trip signal and associated invalid HPCI initiation signal. At no point in time did actual RPV water level reach the high RPV water level turbine trip set point of 92 inches. The potential for a turbine trip signal to occur due to shutdown activities was understood and tags were hung to lockout the Feedwater Pumps to prevent the HPCI start signal. Therefore, no HPCI injection occurred. The Licensee has notified the NRC Resident Inspector. |
| ENS 53410 | 18 May 2018 13:51:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Caused by Main Transformer Trip | At 0651 (PDT) on May 18th, 2018, Columbia Generating station experienced a Main Transformer trip, that caused a Reactor Scram. Reactor Power, Pressure and Level were maintained as expected for this condition. MS-RV-1A (Safety Relief Valve) and MS-RV-1B (Safety Relief Valve) opened on reactor high pressure during the initial transient. MS-RV-1B appeared to remain open after pressure lowered below the reset point. The operating crew removed power supply fuses for MS-RV-1B and it currently indicates intermediate position. SRV (Safety Relief Valve) tail pipe temperatures indicate all valves are closed. Suppression pool level and temperature have remained steady within normal operating levels. All control rods inserted and reactor power is being maintained subcritical. RPV (Reactor Pressure Vessel) water level is being maintained with condensate and feed system with startup flow control valves in automatic. Reactor Pressure is being maintained with the Turbine Bypass valves controlling in automatic. The main condenser is the heat sink. No ECCS (Emergency Core Cooling Systems) systems actuated or injected; the EOC-RPT (End of Cycle-Recirculation Pump Trip) and RPS (Reactor Protection System) systems actuated causing a trip of the RRC pumps and a reactor scram. Core recirculation is being maintained with RRC-P-1A (Reactor Recirculation Pump) running. No release has occurred. At this time there will be no notifications to state, local or other public agencies. The NRC Senior Resident has been notified. The cause of the event is currently under investigation. Plant conditions are stable. The plant is in its normal electrical alignment and offsite power is available to the site. |
| ENS 53276 | 22 March 2018 08:00:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Technical Specification Required Shutdown Due to Pressure Boundary Leakage | This notification is being provided in accordance with 10CFR50.72(b)(2)(i), Plant Shutdown required by Technical Specifications, and 10 CFR 50.72(b)(3)(ii)A, Degraded or Unanalyzed Condition. At 0300 CDT on 3/22/18, on LaSalle Unit 1, a through-wall (welded joint) leak was identified on a 3/4 inch vent line off of the bonnet of the 1B33-F067B, 1B Reactor Recirculation Pump Discharge Valve. This condition qualifies as pressure boundary leakage, which requires entry into Technical Specification 3.4.5, Reactor Coolant System Operational Leakage, Required Action C, to be in Mode 3, Hot Shutdown, by 1500 on 3/22/18 and Mode 4, Cold Shutdown, by 1500 on 3/23/18. This leakage is significantly less than 10 gallons per minute and therefore, does not meet the threshold for entry into the Emergency Action Plan. At the time of discovery, Unit 1 was in Mode 1 - Run. Shutdown began at 0500 CDT and the estimated completion to cold shutdown is 2000 CDT. All necessary shutdown equipment is available. There is no impact to Unit 2. NRC Resident Inspector was notified. |
| ENS 53192 | 1 February 2018 16:57:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram | At 1057 CST on February 1, 2018 with the unit in Mode 1 at approximately 27% power, a manual actuation of the Reactor Protection System (RPS) was initiated due to an unexpected trip of the B Recirc Pump with A Recirc Pump in fast speed. B Recirc Pump tripped during transfer from slow to fast speed resulting in single loop operation. Operators were unable to reconcile differing indications of core flow. This resulted in a conservative decision to initiate a manual scram. The cause of the B Recirc Pump trip and the apparent issues with core flow indication are under investigation. The plant is currently stable in Mode 3. The plant response to the scram was as expected. All control rods (fully) inserted as expected; the feedwater system is maintaining reactor vessel water level in the normal control band and reactor pressure is being maintained with steam line drains and main turbine bypass valves. The NRC Senior Resident (Inspector) has been notified.
This event was initially reported under 10 CFR 72(b)(2)(iv)(B) as a manual actuation of the RPS due to an unexpected trip of the B Reactor Recirculation Pump with the A Reactor Recirculation Pump running in fast speed (Single Loop Operations). Operations was unable to reconcile differing indications of core flow and made the conservative decision to perform a planned shutdown in accordance with normal operating procedures. Therefore, this event 'resulted from and was part of a pre-planned sequence during testing or reactor operation' as specified in 10 CFR 50.72(b)(2)(iv)(B), 10 CFR 50.73(a)(2)(iv)(A) and NUREG-1022 Section 3.2.6. Consequently, this event is not reportable as an actuation of RPS. The NRC Resident Inspector has been notified. R4DO (Groom) has been notified. |
| ENS 52795 | 8 June 2017 19:27:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 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 | Automatic Reactor Scram After Main Turbine Control Logic Loss of Power | At 1527 hrs (EDT) on June 8, 2017, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed due to a loss of Main Turbine Electro-Hydraulic Control (EHC) logic power causing a High Flux Reactor Power RPS (Reactor Protection System) trip. All control rods (fully) inserted and both reactor recirculation pumps tripped due to reaching reactor water level 2. Reactor water level lowered to -49 inches causing Level 3 (+13 inches) and Level 2 (-38 inches) isolations. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) automatically initiated and were overridden by control room operators after RPV (Reactor Pressure Vessel) water level was restored to the normal band with feedwater. HPCI and RCIC injected to the Reactor Coolant System during reactor level stabilization. All isolations and initiations occurred as expected. No main steam relief valves opened. Pressure was controlled via main turbine bypass valve operation. All safety systems operated as expected. Secondary Containment Zone 1, 2, and 3 differential pressure lowered to 0 inch WG (Water Gauge) due to a trip of the Reactor Building Ventilation system that resulted from Unit 1 Level 2 isolation. Differential pressure was restored to Zones 1, 2, and 3 by the initiation of Standby Gas Treatment System on the Unit 1 Level 2 initiation. Unit 1 reactor is currently stable in Mode 3. Investigation into the loss of Main Turbine EHC logic power is underway. The NRC Resident Inspector has been notified. A voluntary notification to PEMA and press release will occur. The suspected cause of the loss of power to the EHC logic circuit is ongoing maintenance on the system. |
| ENS 52727 | 30 April 2017 22:18:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Open Bypass Valve Causes Loss of Safety Function | On April 30, 2017, at 1818 (EDT), the main turbine steam bypass valve #1 partially opened. Power was incrementally lowered. While lowering power the bypass valve would shut and then reopen and power would again be lowered. When power was lowered to approximately 74 percent the bypass valve remained closed. During the transient the reactor protection system (RPS) Turbine Stop Valve Closure and Control Valve Fast Closure trip functions were declared inoperable due to the opening of the bypass valve which affects the bypass setpoint for those RPS trip functions. With the loss of these RPS trip functions a loss of safety function existed intermittently for approximately 37 minutes. The manual reactor trip function and other RPS functions remained operable. Both channels of the rod withdrawal limiter (RWL) and the end of cycle reactor recirculation pump trip (EOC-RPT) function were also declared inoperable. These functions are credited in accident analysis, this also resulted in a loss of safety function. Currently the bypass valve is closed and the RWL, EOC-RPT and RPS function are operable. Troubleshooting continues to determine the issue with the main turbine that caused the bypass valve to open. NRC Resident Inspector has been notified. |
| ENS 52566 | 18 February 2017 21:37:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Loss of Control Building Chilled Water | At 1537 CST on February 18th, 2017, while the plant was in MODE 5 for a scheduled refueling outage, the main control room experienced a loss of Control Building chilled water and the associated ventilation systems while attempting to alternate divisions for testing. An equipment malfunction in a breaker supplying a Main Control Room air handling unit caused a loss of both divisions of Control Room and Control Building chilled water systems and associated ventilation systems until 1737 CST. During the period between 1537 and 1737, neither division of Control Building chilled water was able to perform the support function for cooling Division 1 and 2 AC and DC power distribution systems or the support function for the Division 1 and 2 Control Room Fresh Air systems. Shutdown Cooling remained in service throughout this event. There were no apparent effects on any plant equipment from the loss of chill water and ventilation. The Division 1 Control Building chill water and ventilation system was returned to service at 1737 on February 18, 2017. Actions were initiated to terminate the OPDRV (operations with potential to drain the reactor vessel) that was in progress at the time of the event by installing the reactor recirculation pump seal. As a conservative measure, actions were initiated to set containment and containment was set at 2145. Troubleshooting and analysis is ongoing to confirm and correct the problem which caused the loss of the Control Building chill water and ventilation system. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(v)(B). The NRC Senior Resident Inspector has been notified. |
| 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. |
| ENS 52042 | 24 June 2016 16:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Reactor Recirculation Pumps Degradation | At 1215 (EDT) on 6/24/2016, James A. FitzPatrick (JAF) was at 100% power when Breaker 710340 tripped and power was lost to L-gears L13, L23, L33, and L43. These provide non-vital power to Reactor Building Ventilation (RBV), portions of Reactor Building Closed Loop Cooling (RBCLC), and 'A' Recirculation pump lube oil systems. Off-site AC power remains available to vital systems and Emergency Diesel Generators (EDG) are available. Due to the loss of RBV, Secondary Containment differential pressure increased. At 1215 (EDT), Secondary Containment differential pressure exceeded the Technical Specifications (TS) Surveillance Requirement SR-3.6.4.1.1 of greater than or equal to 0.25 inches of vacuum water gauge. The Standby Gas Treatment (SBGT) system was manually initiated and Secondary Containment differential pressure was restored by 1219 (EDT). The 'A' Recirculation pump tripped at 1215 (EDT) and reactor power decreased to approximately 50%. 'B' Recirculation pump temperature began to rise due to the degraded RBCLC system. At 1236 (EDT), a manual scram was initiated. Reactor Pressure Vessel (RPV) water level shrink during the scram resulted in a successful Group 2 isolation. All control rods have been inserted. The RPV water level is being maintained with the Feedwater System and pressure is being maintained by main steam line bypass valves. A cooldown is in progress and JAF will proceed to cold shutdown (Mode 4). Due to complete loss of RBCLC system, the Spent Fuel Pool (SFP) cooling capability is degraded but the Decay Heat Removal system remains available. SFP temperature is slowly rising and it is being monitored. The time (duration) to 200 degrees is approximately 117 hours. The initiation of reactor protection systems (RPS) due to the manual scram at critical power is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The general containment Group 2 isolations are reportable per 10 CFR 50.72(b)(3)(iv)(A). In addition, the temporary differential pressure change in Secondary Containment is reportable per 10 CFR 50.72(b)(3)(v)(C), as an event that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector and the State of New York. |
| ENS 51983 | 6 June 2016 09:56:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Isolable Leakage Identified from Seal Water Line Weld Inside Rcs Pressure Boundary | Susquehanna Unit 1 commenced a manual shutdown on 06/05/2016 for a maintenance outage. At 2202 hours (EDT) on 06/05/2016, operators began reducing power in accordance with plant procedures. At 0352 hours on 06/06/2016, the Main Turbine was tripped with reactor power at approximately 15%. The Mode switch was taken to 'STARTUP/HOT STANDBY' (Mode 2) at 0515 hours on 06/06/2016. Manual insertion of control rods was paused as scheduled for entry into the drywell for inspections. There were no ESF actuations. At 0556, the licensee identified leakage from a weld on seal water line piping connected to the 1B reactor recirculation pump seal area. The location is within the reactor recirculation loop isolation valves, therefore is isolable from the reactor vessel. The piping is ASME Class 2 and is reactor coolant pressure boundary. The reactor was in Mode 2 at the time of discovery. This event is being reported as a plant shutdown required by technical specifications pursuant to 10CFR50.72(b)(2)(i) and degraded condition pursuant to 10CFR50.72(b)(3)(ii)(A). Activities are continuing to achieve cold shutdown. The licensee informed the Commonwealth of Pennsylvania and the NRC Resident Inspector. |
| ENS 51681 | 24 January 2016 08:57:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Average Power Range Monitoring System Inoperable | At 0357 hours (EST) on January 24, 2016, during a shut down required by plant Technical Specifications (see EN#51679), the 3A Feedwater heater isolated while performing a Reactor Recirculation Pump downshift. A consequence of this Feedwater heater isolation was that all 8 of the Average Power Range Monitors (APRM) became inoperable due to a calibration set point being out of tolerance. The APRM's are relied upon for the reactors high neutron flux trips. The inoperable APRM's resulted in a loss of RPS Trip Capability and a loss of safety function. The manual reactor trip function and other RPS functions remained available. RPS Trip capability for the APRM's was restored at 0445 hours on January 24, 2016. This notification is being made under (10 CFR) 50.72(b)(3)(v)(A) for an event or condition that could have prevented the fulfillment of a safety function affecting the ability to shut down the reactor. The licensee has notified the NRC Resident Inspector. |
| ENS 51560 | 23 November 2015 16:40:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram Due to a Reactor Recirculation Pump Lockout | At 1040 CST, with the plant at 100% power, a lockout of the 11 recirculation pump occurred. Following the 11 recirculation pump lockout, at 1041 CST, a reactor scram and a Group 1 isolation occurred. All Main Steam Isolation Valves closed as a result of the Group 1 isolation signal. HPCI (High Pressure Core Injection) has been placed in service to control RPV (Reactor Pressure Vessel) pressure. HPCI did not inject into the RPV and was not needed to control RPV level. At 1104 CST, a Group 2 containment isolation signal was received due to RPV level less than +9 inches. The Group 2 isolation signal has been reset. The cause(s) of the 11 recirculation pump lockout, the reactor scram, and the Group 1 isolation are currently not known and are under investigation. This event is being reported under 50.72(b)(2)(iv)(B) due to the actuation of the Reactor Protection System when the reactor is critical. For the following reasons, this event is also being reported under 50.72(b)(3)(iv)(A): 1) This event resulted in a valid Group 2 containment isolation signal, 2) Since the cause of the Reactor Protection System actuation is not known, the event is being reported as a valid actuation of the Reactor Protection System, and 3) Since the cause of the Group 1 isolation is not known, the event is being reported as a valid primary containment isolation signal affecting multiple Main Steam Isolation Valves. All systems have responded as expected, all control rods fully inserted following the Reactor Protection System actuation. The plant is currently shutdown in mode 3, RPV pressure and RPV level are stable. This event did not result in any radiological release from the plant. This event did not challenge the health and safety of the public. The NRC Resident Inspector has been notified. The plant is in its normal shutdown electrical lineup. HPCI is in pressure suppression mode with RHR cooling the suppression pool. |
| ENS 51538 | 13 November 2015 22:45:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition of Leakage from Pump Lower Seal Weld | On 11/13/15 at 1745 EST, Unit 1 drywell entry was performed during an unplanned Unit 1 outage. The licensee identified leakage from a weld on the 3/4 inch lower seal vent piping connected to the 1B reactor recirculation pump lower seal area. The location is within the reactor recirculation loop isolation valves, therefore isolable from the reactor vessel. The piping is ASME Class 2 and is a reactor coolant pressure boundary. The reactor was in mode 3 at the time of discovery. Control Room determined at 2110 EST on 11/13/15, that requirements for 10CFR50.72(b)(3)(ii)(A) were not met. This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii)(A). The Licensee notified the NRC Resident Inspector. |
| ENS 51532 | 12 November 2015 16:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram | At 1132 hours (EST) on November 12, 2015, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed due to one Main Steam Isolation Valve (MSIV) unanticipated closure causing a High Pressure RPS (Reactor Protection System) trip. All control rods inserted and both reactor recirculation pumps tripped due to reactor water level 2. Reactor water level lowered to -37 inches causing Level 3 (+13 inches) and level 2 (-38 inches) (Division 2 only) isolations. RCIC (Reactor Core Isolation Cooling) automatically initiated and was overridden by control room operators after RPV (Reactor Pressure Vessel) water level was restored to the normal band with feedwater. All isolations and initiations occurred as expected. No main steam relief valves opened. Pressure was controlled via main turbine bypass valve operation. All safety systems operated as expected. The Unit 1 reactor is currently stable in Mode 3. Investigation into the cause of the MSIV closure is underway. Unit 2 was unaffected and continues power operation. The NRC Resident Inspector has been notified. A voluntary notification to PEMA (Pennsylvania Emergency Management Agency) and press release will occur. The plant is in its normal shutdown electrical lineup. |
| 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. |
| ENS 51300 | 7 August 2015 18:40:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Due to Pressure Boundary Leakage on Vent Line Welded Joint | This notification is being provided in accordance with 10 CFR 50.72(b)(3)(ii)A, Degraded Condition. At 1340 CDT on 8/7/15, on LaSalle Unit 2, a through-wall (welded joint) leak was identified on a 3/4 inch vent line off of the bonnet of the 2B33-F067B, 2B Reactor Recirculation Pump Discharge Valve. This condition qualifies as pressure boundary leakage, which requires entry into Technical Specification 3.4.5, Reactor Coolant System Operational Leakage, Required Action C, to be in Mode 4, Cold Shutdown, by 0140 on 8/9/15. This leakage is significantly less than 10 gpm (leak rate is 0.2 gpm) and therefore, does not meet the threshold for entry into the Emergency Action Plan. At the time of discovery, Unit 2 was in Mode 3 - Hot Shutdown, heading into Cold Shutdown for a planned maintenance outage. This event does not affect Unit 1. The licensee notified the NRC Resident Inspector. |
| ENS 51128 | 24 May 2015 23:30:00 | 10 CFR 50.73(a)(1), Submit an LER | 60-Day Optional Telephonic Notification for an Invalid Actuation of Containment Isolation Valves | The following was received via phone call and email: This 60-day report, as allowed by 10CFR 50.73(a)(1), is being made pursuant to 10CFR 50.73(a)(2)(iv)(A) to describe an unplanned, invalid actuation of containment isolation valves. At 1930 EDT on May 24, 2015, a loss of power to Reactor Protection System (RPS) Train B occurred. Initial investigation found the RPS Motor Generator (MG) Set B not running, with its Motor Off light illuminated caused by both Normal EPA breakers and MG Set B output breaker being tripped. Visual inspection at the distribution cabinet was inconclusive at the time and revealed no abnormalities and no abnormal odors in the area. Further investigation of the RPS MG Set B verified normal voltages on all fuse clips, and all power and control power fuses were operational. As a result of the loss of RPS B, the following containment isolation valves closures occurred: Reactor Water Cleanup (RWCU) Outboard Isolation valves, Torus Water Management System (TWMS) Outboard Isolation valves, Division 2 Drywell Pneumatics Inboard and Outboard Isolation valves, Primary Containment Radiation Monitoring System Inboard and Outboard Isolation valves, Reactor Recirculation Pump Seal Purge Flow Outboard Isolation valves, and Drywell Floor and Equipment Drain Sump Inboard Isolation Valves. The Resident Inspector has been notified. |
| ENS 50976 | 11 April 2015 13:58:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Drywell Leakage Point Identified | On 4/11/15, Unit 2 drywell entry was performed during a planned Unit 2 refueling inspection outage. At 0958 EDT, the licensee identified leakage from a weld on the 3/4 inch seal vent piping connected to the 2A reactor recirculation pump seal area. The location is within the reactor recirculation loop isolation valves, therefore it is isolable from the reactor vessel. The piping is ASME Class 2 and is a reactor coolant pressure boundary. The reactor was in mode 3 at the time of discovery. This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii)(A). The licensee notified the NRC Resident Inspector. |
| ENS 50973 | 11 April 2015 03:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Caused by Main Turbine Trip Due to Degrading Main Condenser Vacuum | At 2346 EDT on April 10, 2015, Susquehanna Steam Electric Station Unit 2 reactor automatically scrammed due to a main turbine trip caused by loss of turbine steam seals and degrading main condenser vacuum. Unit 2 reactor was being shutdown for a refueling outage. At approximately 37 percent power, turbine steam seals were lost resulting in a degrading vacuum. The vacuum degraded quickly, resulting in a main turbine trip before the reactor operator could insert a manual scram. At 37 percent power, the turbine trip caused an automatic scram. This occurred during a transfer from normal steam seal supply to the auxiliary boiler supply. All control rods (fully) inserted. Reactor water level lowered to +2 inches causing Level 3 (+13 inches) isolation. No ECCS actuations occurred. The Operations crew subsequently maintained reactor water level at the normal operating band using RCIC (Reactor Core Isolation Cooling). No steam relief valves opened. The reactor recirculation pumps tripped on EOC-RPT due to the turbine trip at power. The reactor is currently stable in Mode 3. Investigation into the cause of the loss of turbine steam seals is underway. The NRC Resident Inspector was notified. A voluntary notification to PEMA (Pennsylvania Emergency Management Agency) and press release will occur. Unit 2 is in a normal shutdown electrical lineup. Turbine steam seals were restored to the normal steam supply and condenser vacuum was restored. Decay heat is being removed via the steam bypass valves to the condenser. Unit 2 is proceeding with their cooldown to support the scheduled refueling outage. |
| ENS 50903 | 19 March 2015 11:02:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to an Oscillation Power Range Monitor Upscale Actuation | At 0702 EDT on March 19, 2015, Fermi 2 received an automatic scram due to actuation of the Reactor Protection System (RPS) function of Oscillation Power Range Monitor (OPRM) Upscale. The plant had recently transitioned to Single Loop Operation after securing the 'A' Reactor Recirculation Pump due to loss of normal and emergency cooling water supply. The lowest reactor water level was 134 inches above top of active fuel. Reactor water level is being maintained in the normal band by the Feedwater and Control Rod Drive Systems. No Safety Relief Valves (SRV) actuated. Reactor pressure is being maintained via the Main Turbine Bypass Valves and Main Condenser. Reactor Pressure Vessel Level 3 isolation occurred. No additional safety system actuations occurred. All off-site power sources were available throughout the event. The plant is currently in Mode 3 and in a stable condition. Investigation into the cause of the event is ongoing. This event is being reported under the four hour Non-Emergency reporting criteria of 10CFR50.72(b)(2)(iv)(B). The NRC Resident Inspector has been notified. |
| ENS 50686 | 13 December 2014 22:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Reactor Coolant Pressure Boundary | On 12/13/14 at 1700 (EST) hours, during a planned Unit 1 maintenance outage the licensee identified leakage from a weld on the (.75 inch) lower seal vent piping connected to the 1B reactor recirculation pump lower seal area. The location is within the reactor recirculation loop isolation valves, therefore isolable from the reactor vessel. The piping is ASME Class 2 and is a reactor coolant pressure boundary. The reactor was in mode 3 at the time of discovery. Control Room notified at 1345 (EST) on 12/16/14, that requirements for 10CFR50.72(b)(3)(ii)(A) were not met. This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii)(A). The unit was taken to Mode 4 and the seal vent piping repair was completed. The NRC Resident Inspector has been notified. |
| ENS 50546 | 17 October 2014 08:03:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram on High Average Power Range Monitor Flux | At 0303 (CDT), River Bend Nuclear Station sustained a reactor scram due to high Average Power Range Monitor (APRM) flux, suspected due to a malfunction of the Electrohydraulic Control System. Reactor recirculation pump 'B' tripped, reactor recirculation pump 'A' responded appropriately. All other systems responded appropriately except for loss of feed water due to low suction pressure trip from isolating the condensate demineralizers. Reactor water level did not get out of level band. Condensate demineralizers and feedwater were restored to service. Level 3 (isolation) was initiated due to scram. (One) system, Suppression Pool Cooling isolated accordingly due to level 3 signal. Currently the plant is in mode 3, hot shutdown. Plant will remain in mode 3 until investigation of scram is complete. During the scram, all rods inserted into the core. No relief valves lifted as a result of the transient. All safety equipment is available although reactor core isolation cooling is functional but inoperable due to an earlier issue discovered during a surveillance test. The reactor is at normal pressure and temperature for Mode 3. The cause of the high APRM flux and the identified anomalies are under investigation. The licensee has notified the NRC Resident Inspector.
The licensee is updating the notification to include an 8 hour notification for a specified system actuation due to the Level 3 isolation signal. Licensee is proceeding to cold shutdown to troubleshoot the EHC system. The licensee will notify the NRC Resident Inspector. Notified R4DO (Haire). |
| ENS 50030 | 12 April 2014 15:12:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Transformer Sudden Pressure Relay Actuation | At 1012 CDT on April 12, 2014, Dresden Unit 2 automatically scrammed on TR-2 sudden pressure relay actuation. All rods inserted to their full-in positions. Following the reactor scram, the 'A' reactor recirculation pump did not run back to minimum speed automatically. Operators took action to run the recirculation pump to minimum speed manually. All other systems operated as expected. Reactor vessel inventory and pressure are being maintained in automatic. The cause of the scram was due to a trip of the sudden pressure relay for Main Power Transformer 2. Troubleshooting is in progress to determine the cause (of the trip of the sudden pressure relay). This condition is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B), any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical. Normal offsite power remains available. There was no impact on Unit 3. The NRC Resident Inspector has been informed. |
| ENS 49868 | 4 March 2014 06:43:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Unit 2 Manual Reactor Scram Following Loss of a Uninterruptible Power Supply (Ups) | At 0137 EST Nine Mile Point Unit 2 experienced a loss of an uninterruptible power supply 2VBB-UPS3B which resulted in a half scram and half isolations. This caused a loss of cooling water to the Reactor Recirculation Pumps and other indications for the loss of power. At 0143 EST a Manual Reactor Scram was inserted due to the rise of temperatures on the Reactor Recirculation Pump seal cavity temperature and motor winding temperature. The reactor building ventilation radiation monitor went non-functional when the reactor building isolated on the loss of UPS power. The standby gas treatment system was started as required and restored the reactor building differential pressure. This is a 4-Hour report for 10CFR50.72(b)(2)(iv)(B) RPS Actuation and 8-Hour report for 10CFR50.72(b)(3)(xiii) Loss of Emergency Assessment Capability. The NRC Resident inspector has been notified. All systems functioned as required following the manual scram. All control rods fully inserted. The cause of the loss of the UPS is under investigation. |
| ENS 49593 | 2 December 2013 14:04:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Scram Following Trip of Both Reactor Recirculation Pumps | At 0904 (EST) on Monday, December 2, 2013, Nine Mile Point Unit 2 was manually scrammed from approximately 40% thermal power due to the loss of both reactor recirculation pumps during a planned downpower evolution. Manual scram of the unit is procedurally required upon loss of both recirculation pumps to avoid potential power/flow oscillations. The reactor recirculation pumps failed to transfer to the low frequency motor generators when downshifted from fast speed. The cause of the loss of both reactor recirculation pumps is not known at this time. (Nine Mile Point Unit 2) NMP2 has commenced cooldown in preparation for the forced outage to investigate and commence repairs. 10 CFR 50.72(b)(2)(iv)(B) requires reporting within 4 hours of any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical. All control rods fully inserted. No safety systems actuated. Decay heat is being removed via the main condenser. The "A" recirculation pump was restarted in low speed at 1045 EST. Unit 2 is in a normal shutdown electrical lineup. The licensee informed the NRC Resident Inspector and will inform the New York State Public Services Commission. |
| 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. |
| ENS 48936 | 17 April 2013 01:42:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation During Turbine Stop Valve Logic Testing While Shutdown | During outage main turbine stop valve RPS logic surveillance testing, an invalid RPS actuation occurred due to an error in executing main turbine surveillance testing procedures. A Turbine Stop Valve closure RPS signal occurred due to an error in the restoration sequence of restoring the RPS bypass signal and a subsequent manual trip of the main turbine. This resulted in a full scram and a trip of both reactor recirculation pumps. The site post-scram response procedure was entered, which required that the mode switch be placed in the locked SHUTDOWN position. This caused an expected but valid RPS actuation. No control rod motion occurred due to all control rods were inserted at the time of the invalid RPS actuation and subsequent valid RPS actuation. The license has notified the NRC Resident Inspector. |
| ENS 48607 | 19 December 2012 22:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Low Reactor Pressure Vessel Level | At approximately 17:31 hours on December 19, 2012, Susquehanna Steam Electric Station Unit Two reactor automatically scrammed on low RPV level (Level 3, +13 inches) while transitioning the 'A' reactor feed pump from discharge pressure mode to flow control mode. All control rods inserted and both reactor recirculation pumps tripped. Reactor water level lowered to approximately -29 inches causing Level 3 (+13 inches) isolations. An automatic trip of the reactor recirculation pumps occurred, but is not expected at this RPV level. There were no automatic emergency core cooling system initiations. No steam relief valves opened during the event. All safety systems operated as expected. The cause of the loss of feed water flow and trip of the reactor recirculation pumps is under investigation. This report is being made per 10CFR50.72(b)(2)(iv)(B) for a 4 hour report, and 10CFR50.72(b)(3)(iv)(A) for an 8 hour report. Decay heat is removed via steam to the main condenser using the bypass valves . On-site electrical power is in the normal configuration. The Unit 2 reactor is currently stable in Mode 3. Unit 1 was not affected and operates at 99% power. The licensee will inform the Commonwealth of Pennsylvania and make a press release. The NRC Resident Inspector was notified. |
| ENS 48598 | 16 December 2012 06:56:00 | 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 | Reactor Scram from 98% Power During Turbine Control Valve Testing | At approximately 01:56 hours on December 16, 2012, Susquehanna Steam Electric Station Unit 2 reactor automatically scrammed while performing testing of the number 2 control valve per the station surveillance testing program. The number 2 control valve closure initiated a 1/2 scram in the reactor protection system as designed; specifically the 'B1' channel. Evaluation of plant data indicates that an 'A' scram channel signal was activated during the time period the number 2 control valve scram signal was active, thereby causing a full reactor scram. The cause of the 'A' scram channel signal is not understood at this time and is under investigation. A second reactor scram signal was received at approximately 02:10 hours due to reactor water level lowering to 13 inches. Reactor water level was restored above the trip setpoint. All control rods inserted and both reactor recirculation pumps tripped at -38 inches. Reactor water level lowered to -48.5 inches causing Level 3 (+13 inches) and level 2 (-38 inches) isolations. HPCI and RCIC both automatically initiated and were overridden by control room operators after water level was restored. All isolations and initiations at this level occurred as expected. No steam relief valves opened. Pressure was controlled via turbine bypass valve operation. All safety systems operated as expected. This report is being made per 10 CFR 50.72(b)(2)(iv)(A) and 10 CFR 50.72(b)(2)(iv)(B) for 4 hour reports, and 10 CFR 50.72(b)(3)(iv)(A) for an 8 hour report. The unit 2 reactor is currently stable in Mode 3. An investigation into the cause of the reactor scram is underway. Unit 1 continued power operation. The NRC resident inspectors were notified. A press release will occur." The licensee will be notifying the State authorities. |
| ENS 48496 | 9 November 2012 06:18:00 | 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 | Unit 2 Manual Scram Due to Loss of the Integrated Control System | At approximately 0118 hours (EST) on November 9, 2012, Susquehanna Steam Electric Station Unit Two reactor was scrammed by plant operators due to a loss of ICS (Integrated Control System; which controls the reactor feed and reactor recirculation systems). The reactor operator placed the mode switch in shutdown when reactor water level reached +25 inches and lowering. All control rods inserted and both reactor recirculation pumps tripped at -38 inches. Reactor water level lowered to -52 inches causing Level 3 (+13 inches) and level 2 (-38 inches) isolations. HPCI and RCIC both automatically initiated. HPCI was overridden prior to injection and RCIC was utilized to restore reactor water level to the normal band. All isolations and initiations at this level occurred as expected. No steam relief valves opened. Pressure was controlled via turbine bypass valve operation. All safety systems operated as expected. The (Unit 2) reactor is currently stable in Mode 3. An investigation into the cause of the loss of ICS is underway. Unit One continued power operation (at 78% power). The NRC Resident Inspectors were notified. A press release will occur. The licensee will inform the State of Pennsylvania. Decay heat removal is being maintained through the main condenser. On-site electrical power is in the normal configuration. |
| ENS 48341 | 25 September 2012 15:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram During Maintenance on 4160V Bus 12 Ammeter | During maintenance on 4160V Bus 12 ammeter, a Bus 12 lockout occurred. The station power was from 1R Reserve transformer for work on the 2R Auxiliary transformer. Net effect was Bus 12 locked out, removing power from 12 Reactor Feed Pump and 12 Reactor Recirculation pump. Reactor level lowered to +23 inches then began to rise. With both Main Feed Reg Valves in AUTO, the level transient reached +48 inches, the Reactor Water Level Hi Hi setpoint. The Main Turbine and 11 Reactor Feed Pump tripped as designed, and a Reactor SCRAM occurred. Reactor water level began to drop, and C.4.A Abnormal Procedure for SCRAM was used to restart 11 Reactor Feed Pump and recover water level. Minimum water level reached was -26 inches. Reactor Low Level SCRAM signal and Group 2 Primary Containment isolation occurred at +9 inches as designed, No Safety Relief valves lifted during this transient. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) did not receive an initiation signal due to not reaching their setpoints. There were no Emergency Core Cooling Systems initiation setpoints reached. Prior to the event, both divisions of Standby Liquid Control were inoperable as part of planned maintenance. All control rods fully inserted. Decay heat is being removed through the turbine bypass to the main condenser. The plant is in a normal shutdown electrical lineup and stable in Mode 3. The licensee has notified the NRC Resident Inspector and will notify the State and local governments. |
| ENS 47172 | 19 August 2011 14:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram During Quarterly Surveillance Test | At 1046 hours on August 19, 2011, Susquehanna Steam Electric Station Unit 2 reactor scrammed while (operators were) performing a quarterly functional test of reactor water high level trip channels for feedwater / main turbine. The main turbine tripped when a single channel high reactor water level signal was inserted, which was unexpected. Actual water level was within the normal band when the main turbine tripped. The main turbine trip resulted in a reactor scram. Reactor recirculation pump trips as designed and all control rods inserted. Reactor water level lowered to +2 inches causing Level 3 (+ 13 inches) isolations. Reactor water level was restored to the normal operating band using the feedwater system. Six main steam relief valves opened for a short duration, as expected, due to the turbine trip transient. Subsequently, reactor pressure was controlled via turbine bypass valve operation. All safety systems operated as expected. No ECCS or RCIC initiations occurred or were required. The reactor recirculation pumps were subsequently restarted to re-establish forced core circulation. The reactor is currently stable in Mode 3. An investigation into the cause of the scram is underway. Unit 1 was unaffected and continued power operation. The NRC Resident Inspector was notified. A press release will occur. |
| ENS 46495 | 20 December 2010 13:05:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Fish Kill Caused by Rate of Temperature Change in Discharge Canal | At 0938 hours (EST), the New Jersey Department of Environmental Protection (NJDEP) was notified of a fish kill. The fish kill is due to the rate of temperature change in the discharge canal, associated with an unscheduled and non-routine plant shutdown for a reactor recirculation pump repair. As of 0930 hours, seven (7) Speckled Sea Trout and two (2) Black Drums were reported dead west of the route 9 bridge in the stations discharge canal. The (New Jersey Pollutant Discharge Elimination System) NJPDES discharge to surface water permit No. 0005550 part IV-B/C item 11, rate of temperature change states, "The rate of temperature change shall not cause mortality of fish or shellfish. This report is being made in accordance with 10 CFR 50.72(b)(2)(xi) for the occurrence of any event or situation related to protection of the environment, for which a news release is planned or notification to other government agencies has been made. The licensee notified the NRC Resident Inspector. |
| ENS 46443 | 1 December 2010 19:55:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Excessive Rcs Leakage from "B" Recirc Pump Seal | Oyster Creek has declared an Unusual Event (MU7) due to Reactor Coolant System leakage greater than 10 gpm. Leakage has been determined to be from the 'B' reactor recirculation pump seals. The 'B' Reactor Recirculation Loop was isolated and leakage stopped. The plant remained in the normal electrical lineup and no safety systems actuated during this evolution. The licensee notified the NRC Resident Inspector.
At 1537 EST, "Oyster Creek has terminated from an Unusual Event. Event terminated due to isolating leakage from the 'B' Reactor Coolant Loop. The licensee notified the NRC Resident Inspector. Notified R1DO (Schmidt), IRD (Gott), NRR EO (Nelson), DHS and FEMA. |
| ENS 46042 | 24 June 2010 00:51:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Following Loss of Both Recirculation Pumps | Limerick Unit 1 was manually scrammed from 100% power on 6/23/10 at 2051 hours in accordance with plant procedure OT-112 'Recirculation Pump Trip' when both 1A and 1B recirc pump MG set drive motor breakers were observed to have tripped, resulting in a loss of both reactor recirculation pumps. Preliminary indication is a loss of power to 114A Load Center, caused by 'A' phase overcurrent trip of 13.2 KV feeder breaker (11-BUS-07) to the 114A Transformer and Load Center. The cause of the MG set drive motor breaker trips is under investigation at this time. All Control Rods inserted as required. No ECCS or RCIC initiations occurred. No Primary or Secondary Containment Isolations were received. The plant is currently in Hot Shutdown maintaining normal reactor level with feedwater in service. All systems functioned as required during the transient. The manual scram was characterized as uncomplicated. No PORVs or Safety Relief valves lifted during the transient. Decay heat is being discharged to the condenser via turbine bypass valves. The unit is in a normal shutdown electrical lineup and there was no impact on Unit 2. The electrical supplies for the recirc pump MG sets has been walked down by the licensee and no indication of any damage or electrical faults has been found at this time. The NRC Resident Inspector has been notified and the licensee indicated a media or press release will be made. |
| ENS 45930 | 15 May 2010 03:01:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Increasing Reactor Water Level | At approximately 2301 hours EDT on May 14, 2010, Susquehanna Steam Electric Station Unit One reactor scrammed while performing a condensate pump trip test. The reactor operator placed the mode switch in shutdown when reactor water level reached +51 inches and rising. The main turbine tripped due to high reactor water level. All control rods inserted and both reactor recirculation pumps tripped. Reactor water level lowered to -30 inches causing Level 3 (+13 inches) isolations. The Operations crew restored reactor water level to the normal operating band using RCIC (Reactor Core Isolation Cooling) and subsequently the feedwater system. All isolations at this level occurred as expected. No steam relief valves opened. Pressure was controlled via turbine bypass valve operation. All safety systems operated as expected. The reactor is currently stable in Mode 3. An investigation into the cause of the shutdown is underway. Unit Two continued power operation. The NRC Resident Inspectors were notified. A press release will occur. The licensee was performing testing on the digital feedwater control system which was installed during their recent refueling outage when the loss of level control occurred. It appears that the control system did not respond fast enough to control water level. This resulted in the reactor operator inserting a manual scram at +51 inches prior to reaching the reactor automatic scram setpoint of +54 inches for water level. Currently, the plant is removing decay heat via main steam line drains to the condenser. The plant is in its normal shutdown electrical lineup with all safety equipment available. The licensee has notified the Pennsylvania Emergency Management Agency. |
| ENS 45902 | 5 May 2010 15:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Reactor Feed Pump Trip | On May 5, 2010, at 1144 hours Eastern Daylight Time (EDT), an automatic reactor scram occurred on Unit 1 following a trip of the 1B Reactor Feed Pump (RFP). Following the 1B RFP trip, the reactor recirculation pumps did not run back as expected. The resulting water level shrink caused level in the Reactor Pressure Vessel (RPV) to drop to Low Level 1, causing the activation of the Reactor Protection System (RPS) and the Primary Containment Isolation System (PCIS). All control rods properly inserted. PCIS Group 2 (i.e., Drywell Equipment and Floor Drain, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems), and Group 8 (i.e., RHR Shutdown Cooling) isolation signals were received on Low Level 1. Actuations of the Primary Containment Isolation Valves (PCIVs) were completed and the affected equipment responded as designed. Due to the expected RPV level reduction following a reactor scram, water level in the RPV momentarily reached Low Level 2. This initiated the High Pressure Coolant Injection (HPCI) System, the Reactor Core Isolation Cooling (RCIC) System, and a partial Group 3 PCIS (i.e., RWCU) isolation. The HPCI and RCIC systems did not inject. The 1-G31-F001 isolated (i.e., inboard isolation) but 1-G31-F004 (i.e., outboard isolation) did not automatically isolate. Based on a preliminary assessment, this response appears to be in accordance with plant design. Further assessments of plant response are on-going to validate plant response. The licensee has notified the NRC Resident Inspector. The scram was uncomplicated. No SRVs lifted. Decay heat removal is via the 'A' feed water pump via the turbine bypass valves to the condenser. The electrical line-up of Unit 1 is normal. Brunswick Unit 2 was not affected. |
| ENS 45440 | 16 October 2009 04:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Reactor Recirculation Pump Trip | On October 16, 2009 at 0048 a manual reactor scram was inserted at the Perry Nuclear Plant. The plant was conducting a planned shutdown due to the Division 2 Emergency Service Water inoperability. While shifting reactor recirculation pumps to slow speed the 'A' pump failed to transfer and tripped off. Following stabilization from this event a manual reactor scram was inserted from approximately 30% power. This was different from the initial planned shutdown sequence. Following the scram all systems operated as expected. The plant is stable in Mode 3. The plant will transition to Mode 4 in accordance with Technical Specification 3.7.1 (Emergency Service Water Inoperability) required actions. All control rods fully inserted and the plant electrical power is in a normal line-up. The licensee notified the NRC Resident Inspector. |
| ENS 45433 | 15 October 2009 10:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Reactor Recirculation Pump Trip | At 0537 hours on October 15, 2009, the 'B' Reactor Recirculation pump tripped. The reactor mode switch was placed in shutdown due to rising reactor water level (approx. 49 inches) prior to the Level 8 automatic scram setpoint (52 inches). All controls rods inserted as a result of the manual scram. All systems performed as expected. Reactor water level is being controlled by the motor driven feedwater pump. Main steam isolation valves were manually closed and decay heat was initially controlled through the main steam line drains to the main condenser via the main turbine bypass valves. Reactor Core Isolation Cooling (RCIC) was manually placed into service (tank-to-tank) to assist in RPV pressure control. RCIC is currently being used for decay heat removal. Investigation is underway to determine the cause of the Reactor Recirculation pump trip. The licensee will be contacting the state, and issuing a press release. The NRC Senior Resident Inspector has been notified. |
| ENS 45391 | 30 September 2009 04:23:00 | 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 | Unit 2 Manually Scrammed After Trip of an Operating Condensate Pump and Rapidly Lowering Rvwl | On 9/29/09, at 2323 (hours) Unit 2 was manually scrammed due to loss of one of the remaining two Condensate Booster Pumps due to low pump suction pressure. The cause for the Condensate Booster Pump low suction pressures is unknown at this time, but is under investigation. The operating crew was removing feedwater pump 2B from service when the condensate booster pump tripped. The condensate booster pump 2C was already out of service to support maintenance. After the reactor was scrammed manually, reactor water level lowered below the automatic scram set point (+2 inches) and below the automatic start for HPCI and RCIC (-45 inches). All expected Primary and Secondary Containment Isolation valves operated as required, isolation groups 2, 3, 6 and 8 were actuated. Both reactor recirculation pumps tripped due to the low reactor water level. HPCI and RCIC actuated as expected to restore reactor water level. Reactor pressure control was maintained on the turbine bypass valves, and no Main Steam Relief Valves (MSRVs) were opened as a result of the transient. At this time the unit is stable in mode 3. Reactor water level is being controlled using one Reactor Feedwater pump. HPCI and RCIC have been returned to standby readiness. Reactor pressure is being automatically maintained by the main turbine bypass valves. This event is reportable as a 4 hour non-emergency report due to 10CFR50.72(b)(2)(iv)(A) and (B) (ECCS discharge to the reactor and Reactor Protection System (RPS) actuation) and as an 8 hour non-emergency report due to 10CFR50.72(b)(3)(iv)(A) (specified system actuations). Lowest observed Reactor Vessel Water Level (RVWL) was -50 inches. Following actuation of HPCI level recovered to +51 inches and then returned to the normal operating band of +33 inches. Safety-related equipment out-of-service prior to the scram included Core Spray Loop 1. All control rods fully inserted. Unit 2 is in a normal post scram electrical lineup. The licensee informed the NRC Resident Inspector and does not plan a press release.
The initial notification made at 0409 hours ET on September 30, 2009, reported that the RCIC system actuated as expected in conjunction with the HPCI to restore Reactor Pressure Vessel (RPV) water level. However, during a review of plant data, BFN (Browns Ferry Nuclear) determined that after receiving a valid actuation signal, RCIC failed to inject to the RPV. The cause of the failure is under investigation. The licensee informed the NRC Resident Inspector of the update and does not plan a press release. Notified R2DO(Ernstes). |
| ENS 45369 | 20 September 2009 22:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Scram Due to Reactor Recirc Pump Trip | At 1747 hours on September 20, 2009, at River Bend Station, during plant shut down for a planned outage, an unplanned manual reactor scram was initiated by plant operators. As part of the planned shutdown, the reactor recirculation pumps were being transferred from fast to slow speed. This transfer did not occur as expected. Instead, the pumps tripped to off. After this occurred, the operators entered the manual reactor scram. Power level was approximately 23 percent at the time of the scram. All other plant equipment and systems performed as expected. Plant personnel are investigating the cause of the pump trip. The plant is proceeding with planned outage activities. All rods fully inserted. Decay heat is being removed via main steam drains and bypass valves to the condenser. Reactor pressure is at 200 psig. The electrical lineup is normal and all safety related equipment is available if required. No safety or relief valves lifted during the manual scram. The licensee has notified the NRC Resident Inspector. |
| ENS 45290 | 24 August 2009 23:50:00 | 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 | Condensate Booster Pump Trip Resulting in a Manual Reactor Scram | On 8/24/09, at 18:50 Unit 3 was manually scrammed due to loss of 2 of the 3 Condensate Booster Pumps due to low pump suction pressure. The cause for the Condensate Booster Pump low suction pressures is unknown at this time, but is under investigation. After the reactor was scrammed manually, reactor water level lowered below the automatic scram set point (+2 inches) and below the automatic start for HPCI and RCIC (-45 inches). All expected Primary and Secondary Containment isolation valves operated as required, isolation groups 2,3,6 and 8 were actuated. Both reactor recirculation pumps tripped due to the low reactor water level. HPCI and RCIC actuated as expected to restore reactor water level. Reactor pressure control was maintained on the turbine bypass valves, and no Main Steam Relief Valves (MSRVs) were opened as a result of the transient. At this time the unit is stable in mode 3. Reactor water level is being controlled using one Reactor Feedwater pump, HPCI and RCIC have been returned to standby readiness. The 3B Reactor Recirculation Pump has been returned to service. Reactor pressure is being automatically maintained by the main turbine bypass valves. This event is reportable as a 4 hour non-emergency report due to 10CFR 50.72(b)(2)(iv)(A) and (B) (ECCS discharge to the reactor and Reactor Protection System (RPS) actuation) and as an 8 hour non-emergency report due to 10CFR50.72(b)(3)(iv)(A) (specified system actuations). All rods fully inserted on the SCRAM. The plant is in its normal shutdown lineup. The licensee notified the NRC Resident Inspector. |