Semantic search
| Start date | Report date | Site | Reporting criterion | System | Event description | |
|---|---|---|---|---|---|---|
| ENS 40548 | 26 February 2004 16:51:00 | Oconee | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Steam Generator Feedwater Control Rod | While operating at power, a main turbine trip caused a reactor trip. The cause of the trip is unknown but is being investigated. All control rods inserted on the reactor trip, no primary or secondary system relief valves operated, and reactor temperature is being maintained using the turbine bypass valve to the condenser. Steam generator water levels are being maintained using main feedwater. The station electrical system is available and in a normal configuration. Current RCS pressure is 2149 psig and temperature is 552 deg F. Units 1 and 2 were not affected. The licensee notified the NRC Resident Inspector. | |
| ENS 40558 | 1 March 2004 11:22:00 | Farley | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | Reactor trip Unit 1 as a result of Turbine Trip from Hi-Hi Steam Generator level '1C' Steam Generator. Steam Generator Feed pump suction pressure initially dropped, deviation alarms were received on 'A' and 'C' Steam Generator levels. Third condensate (pump) was started, increasing feed to S/Gs. '1C' S/G went high and caused turbine trip/reactor trip. Autostart of 'A' and 'B' motor driven feed pumps occurred following the trip. All rods fully inserted following the reactor trip. Both motor-driven auxiliary feedwater pumps are currently supplying the steam generators with the steam dump system in-service to remove decay heat via the main condenser. Offsite power is stable with the EDGs in standby, if needed. All systems functioned as required. The licensee is conducting an investigation to determine the root cause. The licensee will inform the NRC Resident Inspector. | |
| ENS 40570 | 6 March 2004 05:50:00 | Millstone | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Auxiliary Feedwater Decay Heat Removal Control Rod | The licensee reported that the "B" Steam Generator Feed Pump tripped unexpectedly and would not reset causing lowering steam generator water levels. Operators manually tripped the reactor and all control rods properly inserted. The auxiliary feed water (AFW) system automatically initiated to restore steam generator water levels. The lowest steam generator water level observed during the event was 55% level as opposed to the normal level of 70%. No primary relief valves lifted. Operators established decay heat removal capability using AFW system and the atmospheric steam dump valves. The licensee initiated a post trip review to determine the cause of the feed pump trip. The NRC Resident inspector has been notified by the licensee. | |
| ENS 40589 | 15 March 2004 20:18:00 | Sequoyah | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Auxiliary Feedwater Main Turbine Control Rod | At 1518 on 3/15/04, Unit 1 Reactor tripped due to a Main Turbine trip. The Main Turbine tripped due to a Main Generator electrical fault, Auxiliary Feedwater (AFW) started when both Main Feedwater Pumps tripped on Low Tave Feedwater isolation. The AFW start was expected on the Reactor Trip. All safety systems performed as required. Investigation is in progress to determine and correct the cause of this trip. All control rods fully inserted; the electrical grid is stable; ECCS systems remain operable; decay heat is being removed via AFW and steam dumps. The licensee notified the NRC Resident Inspector. | |
| ENS 40591 | 16 March 2004 01:20:00 | Millstone | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Auxiliary Feedwater Decay Heat Removal Main Condenser Control Rod | The licensee reported that an automatic reactor trip occurred on 3/15/04 at 2020 EST due to the spurious trip of one main feed pump which caused a low steam generator water level trip signal. The plant has two main feed pumps. Operators reset the tripped main feed pump but steam generator levels didn't recover in time. The lowest level observed was in the "B" steam generator at 15% level as compared to the normal level of 65%. The trip setpoint is at 50% level. All control rods properly inserted into the core. The auxiliary feed water system automatically initiated as designed and expected. The plant remains stable in mode 3 while the licensee commenced the post trip review to determine the cause of the main feed pump trip. Decay heat removal was established using the AFW system to feed steam generators and bleed steam to the main condenser through the condenser dump valves. The licensee notified the NRC Resident Inspector. | |
| ENS 40601 | 20 March 2004 18:40:00 | Calvert Cliffs | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Feedwater | At 1340 on 3/20/2004, Calvert Cliffs Unit 1 Reactor automatically shutdown due to low Steam Generator Water Level. The low water level was caused by a loss of at least one Steam Generator Feed Pump. The loss was initially caused by a short or ground from Chart Recorder maintenance in Panel 1C29. 1C29 is a control panel in the control room. The chart recorder was a 500KV Bus Voltage Monitor. The post trip primary indications responded normally. The auto steam dump operation responded normally until the quick open signal was cleared at which time the Turbine Bypass Valves failed shut. It is unclear at this time why they failed shut. Steam dump continues through the use of the Atmospheric dump valves and feedwater is supplied via the Auxiliary Feedwater System with the use of 11 (Steam Driven Pump) & 13 AFW Pump (electric driven pump). Lowest Steam Generator Level was -210" in 11 S/G and -115" in 12 S/G level. Current Conditions are RCS pressure is 2250 PSIA and temperature is 532�F. All control rods properly inserted into the reactor core. The NRC Resident Inspector was notified. | |
| ENS 40604 | 23 March 2004 01:31:00 | Clinton | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | The plant had a reactor SCRAM, cause is under investigation with initial indications of a Main Generator over voltage alarm and Generator lockout #2 were indicated on the Generator first hit panel. All plant systems responded as expected to the generator trip and Reactor SCRAM. All rods fully inserted and no ECCS actuations or relief valves lifted. The plant is stable in Mode 3 with the recovery plan being implemented. The NRC Resident Inspector was notified. | ||
| ENS 40608 | 24 March 2004 08:31:00 | Crystal River | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Reactor Protection System Control Rod Emergency Feedwater System | At approximately 0331 EST the reactor tripped due to reactor protection system actuation. The Emergency Feedwater System actuated. It appears that a main feedwater perturbation (sensed low feedwater flow) occurred which actuated a main turbine trip. The main turbine trip actuated the reactor protection system which tripped the reactor. The reactor shutdown and all control rods fully inserted. Decay heat is being removed using the turbine bypass valves and emergency feedwater. The Licensee notified the NRC Resident Inspector. | |
| ENS 40615 | 28 March 2004 03:05:00 | Vogtle | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Auxiliary Feedwater Control Rod | On 3/27/2004, the Unit 1 Reactor was manually tripped when the 1B Main Feedwater Pump speed could not be controlled in automatic or manual. An unexpected increase in Main Feedwater Pump speed was observed which increased the Main Feedwater header pressure to higher than expected values. An attempt was made to manually control the speed of the 1B Main Feedwater Pump from the various controllers in the Main Control Room. Preparations to start the 1A Main Feedwater Pump were initiated. The speed of the 1B Main Feedwater Pump continued to increase to a point where level control of the Steam Generators and overspeed of the 1B Main Feedwater Pump became a concern. At this point the Reactor was manually tripped and the 1B Main Feedwater Pump was tripped. An Auxiliary Feedwater System automatic actuation occurred on the trip a the 1B Main Feedwater Pump as expected because the 1A Main Feedwater Pump was also tripped at the time due to being at a low power level. All systems responded as expected on the Reactor Trip. The plant is currently stable in Mode 3. An Event Review Team will be performing a review of the event and making recommendations related to restarting the Reactor. All control rods fully inserted. Decay heat is being removed using the steam dumps and auxiliary feedwater. Plant pressure is 2235 psig and temperature is 557 degrees F. The licensee notified the NRC Resident Inspector. | |
| ENS 40622 | 29 March 2004 19:04:00 | Cook | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Auxiliary Feedwater Control Rod | On March 29, 2004 at 1404 hours (EST), Unit 2 received an automatic actuation of the Reactor Protection (RPS). The RPS actuation occurred when the train B Reactor Trip Bypass Breaker was manipulated during the conduct of the Solid State Protection System (SSPS). Testing the Bypass Breaker had been racked in and closed, but correct indication was not received earlier during the test. The decision was made to restore from the testing and the Bypass Breaker was being racked out when the actuation occurred. The Unit was at full power with all system in normal alignment. Initial indication is that the actuation signal was a Power Range Rate Trip. The RPS actuation is believed to be related to the manipulation of the Bypass Breaker. Cause is under investigation. Following the RPS actuation, the Auxiliary Feedwater System automatically started on low-low Steam Generator levels. This constituted a valid unplanned actuation of the Engineered Safeguards Feature (ESF). Operators stabilized the Plant using Condenser Steam Dumps and Auxiliary Feedwater. No other ESF systems actuated. All other Plant systems responded normally. All control rods fully inserted into the core. The Plant is currently stable in Mode 3, and Steam Generator levels are being maintained using the Auxiliary Feedwater System. The Licensee notified the NRC Resident Inspector. | |
| ENS 40625 | 30 March 2004 13:40:00 | Quad Cities | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Primary containment Reactor Water Cleanup Main Condenser Control Rod | At 0740 hours (CST) during testing of the turbine thrust bearing wear detector, a main turbine trip occurred. This resulted in an automatic reactor scram due to turbine stop valve closure. Following the scram all Group II (Primary Containment) and Group III (Reactor Water Cleanup) isolations occurred as expected. All essential equipment functioned as required. Unit 2 remains in Mode 3 with reactor water level in the normal level band. An investigation into the Unit 2 turbine trip is in progress. Unit 1 was unaffected by the event and remains at 85% power. All control rods fully inserted. Decay heat is being removed via steam to the main condenser using the bypass valves. The Licensee notified the NRC Resident inspector. | |
| ENS 40628 | 30 March 2004 16:29:00 | Summer | 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)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Steam Generator Reactor Coolant System Feedwater Main Turbine Control Rod | On March 30, 2004, while at power, VCSNS (V.C. Summer Nuclear Station) personnel were performing a reactor building inspection to identify the source of reactor coolant system unidentified leakage that was within Technical Specification (TS) limits. At 1129 hours, a pressure boundary leak was identified at the seal injection line to reactor coolant pump 'C'. Pursuant to TS 3.4.6.2, Action a., VCSNS commenced a controlled reactor shutdown at 1410 on March 30, 2004. During the shutdown, the main turbine experienced higher than normal vibration. At 1516, the turbine was manually tripped at approximately 43% reactor power. Subsequent to the turbine trip, feedwater regulating valve IFV-498 failed in the closed position while in automatic with a full open demand signal. The cause of this failure is not known. The reactor automatically tripped at 1520 due to lo-lo level in the 'C' steam generator at 10% reactor power. All control rods fully inserted and all safety systems responded normally. Both motor driven emergency feedwater pumps started as required. The plant stabilized in mode 3. The licensee notified the NRC Resident Inspector. | |
| ENS 40660 | 9 April 2004 02:16:00 | Cook | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Main Condenser Control Rod Main Steam | On April 8, 2004, at about 2215, Cook Unit 2 experienced a feedwater flow transient resulting in oscillating flows to Nr. 22, 23, & 24 steam generators. At 2216, the high level turbine trip setpoint was reached in Nr. 24 steam generator resulting in (a) main turbine trip signal. The main turbine trip signal caused a turbine trip resulting in an RPS (Reactor Protection System) actuation (reactor trip). At the time of the RPS actuation, Cook Unit 2 was at about 50 percent power and lowering power to facilitate turbine control maintenance. The cause of (the) feedwater transient is unknown and under investigation at this time. Following the RPS actuation, the Auxiliary Feedwater System automatically started on low-low Steam Generator levels. This constituted a valid unplanned actuation of the Engineered Safeguards Feature (ESF). Operators stabilized the Plant using Condenser Steam Dumps and Auxiliary Feedwater. Unit 2 is currently in Mode 3 with Reactor Coolant System conditions stable at normal operating temperature and pressure. No other ESF systems actuated. During the response to the RPS actuation by the operating crew, it was noted that the Main generator output breaker remained closed requiring a manual trip signal to open the output breakers. Another condition identified was a leak from a crack in the side of the 'C' South condenser near the condensate booster pumps' recirculation line inlet. The operating crew removed the condensate booster pumps from service to stop condenser hotwell outleakage; the main condenser remains in service. The causes for the above items are unknown and (are) under investigation at this time. All control rods fully inserted into the core in response to the automatic reactor trip and heat sink is currently been maintained using Auxiliary Feedwater Pumps and the Main Steam Dumps. Except as noted, all other systems functioned as required. The licensee notified the NRC Resident Inspector. | |
| ENS 40664 | 10 April 2004 18:50:00 | San Onofre | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Main Condenser Control Rod | San Onofre Unit 2 was manually tripped from 97% power due to the loss of feed water pumps. After both feedwater pumps tripped, the operators manually tripped the reactor. Both trains of the emergency feed water system actuated as expected due to the initial power level. The plant is stable at NOP/NOT. The cause of the feed water pump trips is being evaluated. All control rods fully inserted and all safety systems worked properly. The steam generators are discharging steam to the main condenser using the turbine bypass valves. AFW pumps are running to maintain steam generator water levels. The Licensee notified the NRC Resident Inspector. | |
| ENS 40666 | 11 April 2004 16:05:00 | Farley | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Control Rod | At 1247 EDT on 04/11/04, the licensee reported that at 1105 CDT on 04/11/04, control room operators were performing low power physics testing in accordance with FNP-2-STP-101 during startup of Unit 2 following a refueling outage. With reactor power at 10E-8 amps in the intermediate range in Mode 2, the 'B' reactor trip breaker opened for unknown reasons. All control rods inserted completely. The licensee is investigating the cause. The licensee notified the NRC Resident Inspector. | |
| ENS 40667 | 12 April 2004 08:47:00 | Farley | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Unit 2 reactor tripped during low power physics testing. Trip appeared to be from an invalid source range (SR) trip signal in one train of solid state protection system. All systems responded properly. The Unit is currently stable in mode 3. The licensee informed the NRC Resident Inspector. | ||
| ENS 40695 | 21 April 2004 20:35:00 | Susquehanna | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Main Turbine Main Steam Line Main Condenser Control Rod | At 1635 on 04/21/2004, Susquehanna Unit 1 was returning to service from its 13th Refueling and Inspection Outage which included main turbine replacement. During start-up turbine testing with the generator off line, several main turbine bearings experienced high vibration. In response to these high vibrations, the reactor was manually scrammed from approximately 17% power. The main steam line isolation valves were manually closed and main condenser vacuum was broken in order to more rapidly slow the main turbine speed. All control rods fully inserted on the SCRAM, a level 3 containment isolation signal was received as expected. RCIC was manually initiated to control reactor water level. Lowest reactor water level reached was approximately 3" narrow range. There were no radioactive releases. This RPS actuation is reportable under 10CFR50.72(b)(2)(iv)(B) as an 'Unplanned RPS Actuation with the Reactor Critical.' The RPS Actuation and the RCIC injection are reportable per 10CFR50.72(b)(3)(iv)(A) 'Unplanned Actuations of Systems that Mitigate the Consequences of Significant Events.' Investigation into the high main turbine bearing vibrations is ongoing. The main turbine was tripped prior to the manual scram and no SRVs have lifted due to low decay heat level. The electric plant is in a normal lineup. The licensee notified the NRC Resident Inspector and the PEMA representative. | |
| ENS 40702 | 24 April 2004 11:03:00 | Dresden | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Main Steam Isolation Valve Safety Relief Valve Control Rod | This report is being made in accordance with 10 CFR 50.72 (b)(2)(iv)(B) . On 04/24/2004 at 06:03 (CDT), Dresden Unit 2 experienced an automatic Scram from 20% Reactor power due to Main Steam Isolation Valve closure, cause is under investigation. There were no Electromatic Relief or Safety Relief Valve actuations and the Isolation Condenser was initiated manually for pressure control. There were no ECCS initiations. PCIS Group 2 and Group 3 Isolations occurred as expected due to normal reactor water level decrease following the scram. All other systems responded as expected. All control rods fully inserted on the automatic scram. The electric plant is in a normal lineup and being supplied from offsite power. The licensee notified the NRC Resident Inspector. | |
| ENS 40713 | 28 April 2004 20:36:00 | Dresden | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Safety Relief Valve Control Rod | On 4/28/04 at 1536 (CDT), Dresden Unit 2 experienced a trip of the 2A Recirc MG set and associated recirc pump. This placed the unit in the immediate scram region of the power to flow map. The Unit NSO manually scrammed the reactor in accordance with the immediate operator actions of DOA 202.01. Troubleshooting is in progress to determine the cause of the 2A Recirc MG set trip. There were no Electromatic Relief or Safety Relief Valve actuations. There were no ECCS initiations. PCIS Group 2 and Group 3 Isolations occurred as expected due to normal reactor water level decrease following the scram. All systems responded as expected following the reactor scram. This report is being made in accordance with 10 CFR 50.72 (b)(2)(iv)(B) for an actuation of the RPS and 10 CFR 50.72(b)(3)(iv)(A) for Group 2 and 3 actuations. All control rods fully inserted, the electrical grid is stable, ECCS and the EDGs remain operable. The licensee notified the NRC Resident Inspector. | |
| ENS 40719 | 2 May 2004 06:17:00 | Nine Mile Point | 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)(i), Tech Spec Required Shutdown | High Pressure Coolant Injection Shutdown Cooling Control Rod | Operators inserted a manual scram at 0217 (EDT) in anticipation of torus bulk temperature approaching 110 degrees F following an Electromatic Relief Valve (ERV) failing to close during ERV testing. Technical Specification 3.3.2.e requires the insertion of a manual scram prior to torus bulk temperature reaching 110 degrees F. The maximum average torus temperature was 104 degrees F. Operators opened ERV 123 at 0209 for post maintenance testing from approximately 19% power. The ERV subsequently stuck open. Operators performed the off-normal procedure for a stuck open ERV, but this failed to shut the valve. This ported steam from the reactor vessel to the torus resulting in an uncontrolled torus temperature rise. Operators placed torus cooling in service but this did not stop the rising torus temperature. Following the manual scram, cooldown, and depressurization, shutdown cooling is being placed in service. Immediately following the manual scram, a turbine trip signal actuated the logic for high pressure coolant injection. High pressure coolant injection actuated and operators controlled level above 53 inches. With the ERV stuck open, the cooldown rate could not be controlled. The cooldown rate was approximately 190 degrees F during the first hour following the scram. All control rods fully inserted into the core. Decay heat is being removed by shutdown cooling (in service at the time of the report). The operators intend to cooldown to cold shutdown. The electrical buses are stable. Nine Mile Point Unit 2 was not affected. The licensee notified the NRC Resident Inspector. | |
| ENS 40730 | 6 May 2004 16:52:00 | Harris | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Auxiliary Feedwater Emergency Core Cooling System Control Rod | The following information was received from the licensee via facsimile: On May 6, 2004, with the reactor at 100 percent power in MODE 1, an unplanned actuation of the reactor protection system occurred. At 1252 (EDT) the reactor was automatically tripped from a power range negative flux rate trip signal. The auxiliary feedwater system actuated as expected to stabilize steam generator levels. All systems functioned as required and no other safety systems were actuated. All control rods inserted on the reactor trip. The operations staff responded to the event in accordance with applicable plant procedures. The plant stabilized at normal operating no-load reactor coolant system temperature and pressure following the reactor trip. Steam generator water levels are being maintained using normal main feedwater. All emergency core cooling system equipment is available. The plant electrical system is available and in a normal configuration. The cause of the plant trip is under investigation. This condition is being reported as an unplanned reactor protection system actuation and specified system actuation in accordance with 10 CFR 50.72(b)(2)(iv)(B) and10 CFR 50.72(b)(3)(iv)(A) . During the transient, a steam generator power-operated relief valve lifted momentarily and then re-seated. No reportable radiological release occurred during the event. The licensee notified the NRC Resident Inspector. | |
| ENS 40737 | 9 May 2004 03:39:00 | Palo Verde | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | On May 8, 2004, at approximately 20:39 Mountain Standard Time (MST) Palo Verde Unit 1 operations manually tripped the reactor when a Control Element Assembly (CEA) slipped approximately 6 inches (CEA # 89) while conducting physics testing (at 10E-02 percent power) following Unit 1's eleventh refueling outage. Unit 1 was at normal operating temperature and pressure prior to the trip. All CEAs inserted fully into the reactor core (Initial Conditions: Regulating groups 1, 2 & 5 were fully withdrawn, regulating group 3 was fully inserted, regulating group 4 was being inserted when CEA # 89 slipped approximately 6 inches. Shutdown groups were fully withdrawn). This was an uncomplicated reactor trip. No emergency classification was required per the Emergency Plan. No automatic ESF actuations occurred and none were required. Safety related buses remained energized during and following the reactor trip. No LCOs (Limiting Conditions of Operations) have been entered as a result of this event. No major equipment was inoperable prior to the event that contributed to the event. Unit 1 is stable at normal operating temperature and pressure in Mode 3 (Hot Standby). The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public. The Senior Resident Inspector was informed of the Unit 1 reactor trip. | ||
| ENS 40752 | 14 May 2004 21:29:00 | Turkey Point | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Main Condenser Control Rod | At 1728 EDTon 05/14/04, while performing maintenance on Turbine 1st stage steam pressure transmitter (PT-447), feedwater regulating valve (FRV) 4A was observed to close. At 1729, an automatic actuation of RPS occurred on Low Steam Generator level in the 4A Steam Generator. This was followed shortly by a manual reactor trip by the operator. All control rods fully inserted. All 3 turbine driven AFW pumps started as expected and are supply steam generators through both feedwater supply headers. The electric plant remains in a normal lineup. RCPs are in operation transferring decay heat to the steam generators. The MSIVs are open with the steam generators discharging steam to the main condenser using the condenser steam dump valves. All steam generator atmospheric dump valves opened on the reactor trip and closed. The plant is investigating the cause of the FRV closure and intends to stay in Mode 3 until the cause is determined and appropriate repairs are made. There was no impact on the other operating unit. The licensee will notify the NRC Resident Inspector. | |
| ENS 40754 | 15 May 2004 16:54:00 | Point Beach | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Protection System Auxiliary Feedwater Circulating Water System Emergency Core Cooling System Control Rod | A Unit 2 manual reactor trip was initiated when the control room was notified that a diver was entangled in the intake crib. Divers were being used to inspect the intake crib, install buoys, and the fish deterrent system. The diver's umbilical cord became snagged and attempts to free it were unsuccessful. The Unit 2 circulating water system was secured to aid in removing the diver from the water. The diver still had breathing air available during the transient. The diver was subsequently removed from the water unhurt. Plant systems functioned as required, including the Reactor Protection and Auxiliary Feedwater Systems. There was no Emergency Core Cooling System actuation. Note: The condenser was unavailable because circulating water was secured. This caused a loss of condenser vacuum and its use as a heat sink. The atmospheric steam dumps are currently being used for heat removal from the steam generators. The circulating water system was subsequently restored to service. This event is reportable pursuant to 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 and 50.72(b)(3)(iv)(A), PWR auxiliary feedwater system. All control rods inserted into the core. The electrical busses are in a normal shutdown line up. The licensee notified the NRC Resident inspector. | |
| ENS 40769 | 22 May 2004 01:08:00 | Surry | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Main Condenser Control Rod | Unit 2 Reactor tripped at 2108, first out is F-E-2, 500KV leads differential lockout relay trip. This notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) for 4-hour notification of RPS Activation and 10 CFR 50.72(b)(3)(iv)(A) for 8-hour notification of automatic actuation of AFW. Plant responded as expected with the exception of 'C' SG AFW flow indication. The 'C' SG level responded normally to Unit conditions. The NRC resident has been notified of this event and is on site. Notification was made to NRC duty officer upon declaration on NOUE(See EN#40768), which was declared based on EAL Tab K-6 for 'Confirmed report of unplanned explosion within Protected Area or Switchyard.' An electrical fault was observed in the 500 KV switchyard and is currently being investigated. No further damage has been detected in the switchyard. The NOUE, which was declared at 2116, has been terminated as of 2256. Reactor shutdown with all control rods fully inserted. Electric plant is in a normal shutdown mode lineup. The other operating unit was unaffected by this event. Decay heat is being removed to the main condenser using the condenser steam dump valves. See Surry Unit 2 event # 40770 & 40771. | |
| ENS 40778 | 27 May 2004 04:33:00 | Oyster Creek | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Main Condenser Control Rod | During a reactor shutdown for scheduled maintenance, a full reactor scram was generated from the Nuclear Instrumentation System. IRMs (Intermediate Range Monitoring detectors) 13, 14 and 18 spiked causing scram signals in both RPS (Reactor Protection Systems). All systems functioned properly post scram, all operator actions were correct. Holding reactor pressure at 900 psig to perform leakage inspection. Following (the leakage) inspection (operations intend) to proceed to cold shutdown. All control rods inserted fully into the core and excess decay heat is currently being diverted to the main condenser, as necessary. The licensee notified the NRC Resident Inspector. | |
| ENS 40791 | 4 June 2004 11:45:00 | San Onofre | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Control Rod | San Onofre Nuclear Generating Station (SONGS) experienced a large influx of seaweed that required securing one of the four running Circulating Water (CW) pumps. When conditions continued to degrade, the Operations Department decided to manually trip the unit. During the trip, all control rods inserted into the core. No manual or power-operated relief valves lifted during the transient. Decay heat is being removed via the atmospheric dump valves. There is no known primary-to-secondary leakage. The Operations staff is maintaining the unit at normal operating pressure and temperature. The electrical grid is stable. Currently, only one CW pump is running. Unit 2 is unaffected and is not experiencing a seaweed influx at this time. The licensee has informed the NRC Resident Inspector | |
| ENS 40795 | 7 June 2004 21:58:00 | Palo Verde | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Coolant System Feedwater Emergency Diesel Generator Auxiliary Feedwater Main Turbine Steam Bypass Control System | On June 7, 2004, at approximately 14:58 Mountain Standard Time (MST) while at 99% RTP (rated thermal power), Palo Verde Unit 3 experienced an apparent electro-hydraulic control (EHC) system fault resulting in Combined Intercept Valve (CIV) closure. This plant upset was followed by a Reactor Power Cutback System (RPCS) initiation. Several seconds later the Reactor automatically tripped on Lo DNBR from approximately 65% RTP. Unit 3 was at normal temperature and pressure prior to the trip. All CEAs (control rod assemblies) inserted fully into the reactor core. This was an uncomplicated reactor trip. No emergency classification was required per the Emergency Plan. No automatic ESF actuations occurred and none were required. Safety related buses remained energized during and following the reactor trip. The Emergency Diesel Generators did not start and were not required. The offsite power grid is stable. No significant LCOs have been entered as a result of this event. No major equipment was inoperable prior to the event that contributed to the event. Unit 3 is stabilized at normal temperature and pressure at approximately 565 degrees F and 2250 psia in Mode 3. The reactor coolant system remains in normal forced circulation with heat removal via the steam bypass control system to the condenser and feedwater from the non-essential auxiliary feedwater system. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public. The Senior Resident Inspector was informed of the Unit 3 reactor trip and this notification. The Senior Resident Inspector was on-site at the time of the reactor trip. | |
| ENS 40804 | 10 June 2004 17:13:00 | North Anna | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Coolant System Reactor Protection System Auxiliary Feedwater Control Rod Main Steam | A Unit 2 automatic reactor trip occurred while the licensee was performing planned periodic testing on train "A" solid state protection. All control rods fully inserted into the reactor core. The Auxiliary Feedwater Pumps automatically started as expected immediately following the reactor trip due to low-low level in the steam generators. The unit is being maintained stable in mode 3 and heat sink is being performed via steam dump to the condensers. All other systems functioned as required. The cause of the reactor trip is under investigation. The licensee notified the NRC Resident Inspector.
This is an update to event notification 40804. At 1313 hours on June 10, 2004, North Anna Unit 2 experienced an automatic trip from 100 percent during the performance of 2-PT-36.1A (Train 'A' Reactor Protection and ESF Logic Actuation Logic Test). The cause of the reactor trip, was determined to be an incorrect configuration of the cell switch (52h contract) on 'A' Reactor Bypass Breaker, 2-EP-BKR-BYA. The incorrect cell switch configuration resulted in a turbine trip signal being generated during testing which resulted in a reactor trip signal being generated in the 'B' train Reactor Protection System. The Auxiliary Feedwater System actuated in response to the event. Control room personnel responded to the event in accordance with emergency procedure E-0, Reactor Trip or Safety Injection. The control room team stabilized the plant using ES-0.1 Reactor Trip recovery. The lowest Reactor Coolant System (RCS) pressure during the event was 1988 psig and the lowest RCS temperature was 549 degrees. No human performance issues were identified during this event. A non-emergency four-hour report was made to the NRC operations center at 1611 hours pursuant to 10CFR50.72(b)(2)(iv)(B) for an actuation of the Reactor Protection System while critical. An eight-hour report was also made to the NRC in accordance with 10CFR 50.72(b)(3)(iv)(A) due to the Auxiliary Feedwater Pump starts (Engineering Safety Features Actuation). The Reactor Protection System, AMSAC (ATWAS Mitigating System Actuation Circuit), and the Auxiliary Feedwater System operated properly in response to the event. During the Unit 2 reactor trip, a blown output fuse on a logic card (that feeds the permissive for arming the Steam Dumps from loss of load) prevented the Main Steam Dump Valves from opening in Tavg Mode as expected. The Steam Generator Power Operated Relief Valves (PORVs) lifted and operated to control RCS temperature until transferring Steam Dump control to the Steam Pressure Mode. The fuse was replaced. A post trip review was conducted at 1500 hours on June 10, 2004. The cell switches on the Reactor Trip Bypass breakers have been repaired and post maintenance testing has been completed. Management approval was granted to start-up Unit 2. North Anna Unit 2 is currently in Mode 1 and is preparing to be placed on-line. The licensee notified the NRC Resident Inspector. Notified R2DO (Lesser) and NRR EO (Bateman). | |
| ENS 40814 | 14 June 2004 14:44:00 | Palo Verde | 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(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator Main Steam Safety Valve Main Steam | On June 14, 2004, at approximately 07:44 Mountain Standard Time (MST) all three units at the Palo Verde Nuclear Generating Station experienced automatic reactor trips coincident with a grid disturbance and loss of offsite power in the Palo Verde Switchyard. Unit 2 declared an ALERT Emergency Plan classification at approximately 07:54 due to a loss of AC power to essential buses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in a station blackout. Subsequently, at 09:51 Unit 2 downgraded the Emergency Plan classification to a NOTIFICATION OF UNUSUAL EVENT when AC power was restored from a single essential bus to both essential buses. Units 1 and 3 declared a NOTIFICATION OF UNUSUAL EVENT at 07:53 MST due to a loss of offsite power to essential buses for greater than 15 minutes. The NOTIFICATION OF UNUSUAL EVENT was terminated for all 3 units at 12:07 MST. Unit 1 and 2 manually initiated a Main Steam Isolation System ESF actuation by procedure. Unit 3 received an automatic Main Steam Isolation System ESF actuation. Due to the loss of offsite power, the Emergency Plan Technical Support Center (TSC) was unavailable. The Unit 2 Satellite TSC was to be staffed by the Emergency Response Organization in response to the loss of assessment capability. Power to the TSC has since been restored. The Emergency Plan ALERT declaration includes staffing of the Joint Emergency New Center to address expected media interest. All three units were at normal operating temperature and pressure prior to the trip. All CEAs inserted fully into the reactor cores. All Emergency Diesel Generators (EDGs) (2 per unit) associated with each of the 3 units started as expected in response to the loss of offsite power to their safety buses. Unit 2's train "A" EDG started, but did not indicate volts or amps and was manually shutdown. The offsite power grid had several perturbations for approximately one hour following the event but has been stable since. LCO 3.8.1, AC Sources - Operating, was entered in each unit as a result of this event. Heat removal is to atmosphere via atmospheric dump valves in natural circulation. Main steam safety valves may have lifted for a brief time. Restoration of forced reactor coolant circulation is pending assurance that the offsite power grid can reliably support the load. No major equipment was inoperable prior to the event that contributed to the event. All 3 units are stable at normal operating temperature and pressure in Mode 3. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant NRC Resident Inspector was notified. FBI Jeff Muller and Mr. Rosales (Mexican National Commission of Nuclear Safety and Safeguards (CNSNS)) were notified.
Notified Reg 4 RDO (Graves), NRR (Bateman), DHS (Lee), FEMA (Canupp), DOE (Sal Moroni), EPA (Stalcup), EPA (Crews), HSS (Davidson), and Mexico (Rosales) NOTE: See events 40815, 40816 and 40818 | |
| ENS 40815 | 14 June 2004 14:44:00 | Palo Verde | 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(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator Main Steam Safety Valve Main Steam | On June 14, 2004, at approximately 07:44 Mountain Standard Time (MST) all three units at the Palo Verde Nuclear Generating Station experienced automatic reactor trips coincident with a grid disturbance and loss of offsite power in the Palo Verde Switchyard. Unit 1 declared a NOTIFICATION OF UNUSUAL EVENT at 07:53 MST due to a loss of offsite power to essential buses for greater than 15 minutes. The NOTIFICATION OF UNUSUAL EVENT was terminated at 12:07 MST. Unit 1 manually initiated a Main Steam Isolation System ESF actuation by procedure. Due to the loss of offsite power, the Emergency Plan Technical Support Center (TSC) was unavailable. The Unit 2 Satellite TSC was to be staffed by the Emergency Response Organization in response to the loss of assessment capability. Power to the TSC has since been restored. The unit was at normal operating temperature and pressure prior to the trip. All CEAs inserted fully into the reactor cores. All Emergency Diesel Generators (EDGs) (2 per unit) associated with the unit started as expected in response to the loss of offsite power to their safety buses. The offsite power grid had several perturbations for approximately one hour following the event but has been stable since. LCO 3.8.1, AC Sources - Operating, was entered as a result of this event. Heat removal is to atmosphere via atmospheric dump valves in natural circulation. Main steam safety valves may have lifted for a brief time. Restoration of forced reactor coolant circulation is pending assurance that the offsite power grid can reliably support the load. No major equipment was inoperable prior to the event that contributed to the event. The Unit is stable at normal operating temperature and pressure in Mode 3. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant NRC Resident Inspector was notified. FBI Jeff Muller and Mr. Rosales (Mexican National Commission of Nuclear Safety and Safeguards (CNSNS)) were notified.
Notified Reg 4 RDO (Graves), NRR (Bateman), DHS (Lee), FEMA (Canupp), DOE (Sal Moroni), EPA (Stalcup), EPA (Crews), HSS (Davidson), and Mexico (Rosales) Note: see related events # 40814, 40816 and 40818 | |
| ENS 40816 | 14 June 2004 14:44:00 | Palo Verde | 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(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator Main Steam Safety Valve Main Steam | On June 14, 2004, at approximately 07:44 Mountain Standard Time (MST) all three units at the Palo Verde Nuclear Generating Station experienced automatic reactor trips coincident with a grid disturbance and loss of offsite power in the Palo Verde Switchyard. Unit 3 declared a NOTIFICATION OF UNUSUAL EVENT at 07:53 MST due to a loss of offsite power to essential buses for greater than 15 minutes. The NOTIFICATION OF UNUSUAL EVENT was terminated at 12:07 MST. Unit 3 received an automatic Main Steam Isolation System ESF actuation. Due to the loss of offsite power, the Emergency Plan Technical Support Center (TSC) was unavailable. The Unit 2 Satellite TSC was to be staffed by the Emergency Response Organization in response to the loss of assessment capability. Power to the TSC has since been restored. The unit was at normal operating temperature and pressure prior to the trip. All CEAs inserted fully into the reactor cores. All Emergency Diesel Generators (EDGs) (2 per unit) associated with the unit started as expected in response to the loss of offsite power to their safety buses. The offsite power grid had several perturbations for approximately one hour following the event but has been stable since. LCO 3.8.1, AC Sources - Operating, was entered as a result of this event. Heat removal is to atmosphere via atmospheric dump valves in natural circulation. Main steam safety valves may have lifted for a brief time. Restoration of forced reactor coolant circulation is pending assurance that the offsite power grid can reliably support the load. No major equipment was inoperable prior to the event that contributed to the event. The Unit is stable at normal operating temperature and pressure in Mode 3. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant NRC Resident Inspector was notified. FBI Jeff Muller and Mr. Rosales (Mexican National Commission of Nuclear Safety and Safeguards (CNSNS)) were notified.
Notified Reg 4 RDO (Graves), NRR (Bateman), DHS (Lee), FEMA (Canupp), DOE (Sal Moroni), EPA (Stalcup), EPA (Crews), HSS (Davidson), and Mexico (Rosales) Note: see related events 40815 , 40814 and 40818 | |
| ENS 40832 | 22 June 2004 17:13:00 | Limerick | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Main Turbine Decay Heat Removal Safety Relief Valve Main Condenser Control Rod | Unit 2 reactor scrammed at 13:13 (EDT) following an electrical yard manipulation. The reactor scram occurred as a result of an RPS (Reactor Protections System) actuation following an automatic trip of the main turbine from a generator lockout. All control rods inserted (fully). The plant is stable (in mode 3). No ECCS (Emergency Core Cooling System) or safety relief valve actuations have occurred. This report is being made pursuant to 50.72(b)(2)(iv)(b). At this time the plant is stable, as expected. An investigation is in progress to determine the reason for the electrical fault. All plant systems functioned as required and decay heat removal is being performed via bypass to the main condenser. The licensee notified local and State authorities and the NRC Resident Inspector. | |
| ENS 40858 | 9 July 2004 03:32:00 | Browns Ferry | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Secondary containment Reactor Protection System Primary Containment Isolation System Primary containment Main Turbine Main Condenser Control Rod Main Steam | The following information was received from the licensee via facsimile: On 7/08/2004 at approximately 2232 hours CDT, Browns Ferry Unit 2 scrammed due to a Turbine Generator Load Reject Signal. All systems responded as required to the scram signal. No ECCS initiations occurred as a result of this event. Reactor Water Level lowered to the Low Level setpoint which generated a redundant SCRAM signal and initiated the Primary Containment Isolation System (PCIS) function for PCIS groups 2 (Primary Containment), 3 (RWCU) (Reactor Water Clean Up), 6 (Secondary Containment), and 8 (TIP system) (Transverse Incore Probe). As expected, Main Steam Relief Valves (MSRVs) opened due to the high reactor pressure (maximum value observed was 1137 psig) as a result of the Main Turbine / Generator Trip. Reactor Water level was restored to normal via the Reactor Feedwater system and all PCIS isolations have been reset. The cause of the Load Reject Signal is still under investigation at this time. This event is reportable under 10CFR50.72(b)(2)(iv)(B), 'Any event or condition that results in a valid actuation of the Reactor Protection System (RPS) when the reactor is critical' and also under 10CFR50.72(b)(3)(iv)(A), 'Any event or condition that results in a valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B)'. This event also requires a 60 day written report in accordance with 10CFR50.73(a)(2)(iv)(A). All control rods inserted during the scram. All Main Steam Relief Valves have properly reset. Decay heat is being removed from the reactor via main turbine bypass valves to the main condenser. Pressure is being maintained at normal operating pressure. The electrical grid is stable. Unit 3 was not affected by the scram on Unit 2. The licensee has notified the NRC Resident Inspector. | |
| ENS 40862 | 11 July 2004 03:35:00 | Browns Ferry | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Intermediate Range Monitor Control Rod | On 07/10/2004 at 2235 (CST), during Browns Ferry Unit 2 startup activities, as IRMs (Intermediate Range Monitors) were being ranged up, an upscale trip on IRM E (RPS (Reactor Protection System) A Channel) and IRM F (RPS B Channel) was received, resulting in a full reactor scram. Mode Switch was in STARTUP, Mode 2 at time of trip. IRMs were on ranges 6 and 7, and reactor pressure was approximately 950 psig. All systems responded as designed, all control rods are at full-in. No ECCS (Emergency Core Cooling System) or PCIS (Primary Containment Isolation System) actuation set points were reached. This is reportable as 4 hour ENS (Emergency Notification System) report per 10CFR50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of pre-planned sequence during testing or reactor operation.' It is also reportable as an 8 hour ENS report per 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). (1) Reactor protection system (RPS) including: Reactor scram and reactor trip.' Also reportable as a sixty day written report per 10CFR50.73(a)(2)(iv)(B). Mode Switch is presently in shutdown, Mode 3. Investigation is still on going. NRC Resident (Inspector) was notified at approximately 2310 (CST). | |
| ENS 40865 | 13 July 2004 17:05:00 | Salem | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Coolant System Emergency Diesel Generator Auxiliary Feedwater Emergency Core Cooling System Main Condenser Control Rod | An automatic Reactor Trip occurred on 21 Steam Generator low low level. The cause is currently under investigation. All systems responded as required. All control rods fully inserted. All auxiliary Feedwater pumps started as expected, on the low level signals and the steam driven auxiliary feed (23) pump was secured by procedure. Currently the steam dump system is rejecting heat to the main condenser for reactor coolant system temperature control, currently at normal operating temperature and pressure. No major equipment was unavailable at the time of the trip. All Emergency Core Cooling Systems and the Emergency Diesel Generators are fully operable if needed and the electrical grid is stable. The Licensee is investigating the cause of Steam Generator "21" low low water level. The NRC Resident Inspector was notified of this event by the licensee. | |
| ENS 40868 | 13 July 2004 21:10:00 | Clinton | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Protection System Reactor Recirculation Pump Reactor Pressure Vessel Main Condenser Control Rod | On July 13, 2004 at 1610 CDT, the Clinton Power Station 345 KV unit output breakers GCB 4506 and 4510 opened resulting in a turbine control valve fast closure. This caused a reactor protection system actuation and reactor scram. Clinton was in Mode 1, operating at 95% rated thermal power when the event occurred. A tornado warning was in effect at the time of the event and weather conditions were degrading in the area surrounding the plant. The exact cause of the trip of the unit breakers is not yet known. All systems responded as expected following the scram with the exception of Reactor Recirculation pump A which tripped off instead of downshifting to slow speed on a reactor water level 3 signal. The unit is currently in Mode 3 with reactor pressure vessel level and pressure in their normal bands. Recirculation Pump "A" remains secured until the licensee completes their investigation. All control rods fully inserted following the scram. The Main Condenser is in service removing decay heat. All ECCS equipment including the EDGs are available, if needed. The licensee informed the NRC Resident Inspector. | |
| ENS 40870 | 14 July 2004 08:35:00 | Palo Verde | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Emergency Diesel Generator Steam Bypass Control System | The following information was received from the licensee via facsimile: The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73. On July 14, 2004, at approximately 01:35 Mountain Standard Time (MST) Palo Verde Unit 2 experienced a Main Generator Trip immediately followed by an automatic Reactor Trip. The reactor was at approximately 100% power and normal operating temperature and pressure prior to the event. The cause of the Main Generator Trip was most likely the result of electrical storm conditions present at the site at the time of the trip. Unit 2 was at normal operating temperature and pressure prior to the trip. All CEAs inserted fully into the reactor core. Heat removal was maintained to the condenser via the steam bypass control system. This was an uncomplicated reactor trip, No emergency classification was required per the Emergency Plan. No automatic ESF actuations occurred and none were required. Safety related buses remained energized during and following the reactor trip. The Emergency Diesel Generators did not start and were not required. The offsite power grid is stable. No LCOs have been entered as a result of this event. No major equipment was inoperable prior to the event that contributed to the event. Unit 2 Is stable at normal operating temperature and pressure in Mode 3. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public. The (NRC) Senior Resident Inspector was informed of the Unit 2 reactor trip. No primary or secondary power-operated or manual relief valves lifted as a result of the plant transient. Units 1 and 3 were unaffected by the trip on Unit 2. Offsite power was maintained to Unit 2 safety busses throughout the event. | |
| ENS 40875 | 15 July 2004 23:44:00 | Salem | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Auxiliary Feedwater Main Condenser Control Rod Main Steam | During plant start up, a manual Reactor trip was initiated in response to lowering water level on 23 Steam Generator. All systems responded as required. All control rods fully inserted and no ECCS actuated or relief valves lifted. Both motor driven auxiliary feedwater pumps (21 and 22) started as expected on the low level signals from 23 Steam Generator. The steam driven auxiliary feed (23) pump was not required to start and remained in standby. Decay heat is being removed via the main steam dump system to the main condenser. The reactor is currently at normal operating temperature and pressure. No major equipment was unavailable at the time of the trip. No personnel injuries occurred as a result of this event. The cause of the low steam generator level is being investigated. The NRC Resident Inspector was notified along with Lower Alloway Creek Township. The States of New Jersey and Delaware will be notified. | |
| ENS 40910 | 30 July 2004 17:00:00 | Columbia | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Decay Heat Removal Rod Worth Minimizer Main Condenser Control Rod | The following information was obtained by the licensee via facsimile: Reactor SCRAM received at 0924 (hrs) PDT. Initial indications are the scram signal was caused by RPS (Reactor Protection System) High Pressure. Following the scram, 2 (two) control rods did not immediately indicate (fully inserted). Control room staff entered (procedure) PPM5.1.2 and took the required actions. All control rods subsequently indicated (fully inserted). An ALERT Classification was declared at 1000 (hrs) based on PPM 13.1.1 Criteria 2.2.A.1, 'RPS Setpoint exceeded and automatic actions failed to result in a rod pattern which alone assures reactor shutdown'. Manual actions resulted in all rods (fully inserted) and reactor power (less than or equal to) 5 percent. All other plant systems responded as expected with the exception of a wetwell-to-drywell vacuum breaker which indicates open. Investigation into the cause of the scram and actual control rod position is ongoing. Further details will be provided when available. The licensee also reports that no relief valves lifted during the transient. Decay heat removal is via the main condenser. Reactor level is steady at 36 inches. Reactor temperature and pressure are 507 degrees and 710 psig respectively. Offsite power is available. All emergency systems are available in standby. The licensee has notified the NRC Resident Inspector of the incidents. The NRC entered Monitoring mode at 1327 hrs EDT with Region IV leading. The NRC exited Monitoring at 1530 hrs. EDT and returned to Normal mode. Notified HHS (Ayles) as well as others noted in notification block.
At 1358 (hrs.) EDT on 7/30/04, NRC was notified of an Alert at Columbia Generating Station (EN #40910). This is a follow-up to inform NRC that the event was terminated at 1457 (hrs) (EDT) (1157 PDT). All control rods are inserted. Reactor is shutdown and water level is normal. All required emergency systems are operable. All offsite and onsite power sources are operable. Reactor pressure is normal. The licensee has notified the NRC Resident Inspector, State of Washington and local authorities of the termination. The NRC Operations Center notified R4DO(Bywater), DHS, FEMA, DOE(NRC), USDA, EPA and CDC(HHS).
At 1358 EDT on July 30, 2004, NRC was notified of an Alert at Columbia Generating Station (EN #40910). A subsequent notification was made to inform the NRC that the event was terminated at 1157 PDT (1457 EDT) on July 30, 2004. This is a follow-up to inform the NRC that Columbia Generating Station is retracting its Alert Emergency Declaration due to the following reason. Following the RPS actuation, control rod position indication for two control rods was indeterminate for approximately two minutes to the control room staff. A subsequent review of control rod position indication from the Plant Data Information System (PDIS), Rod Worth Minimizer (RWM) logs, and Auto Scram Timer (AST) data by Columbia Generating Station personnel shows that all rods were successfully inserted to the 'Full-in' position following the initial RPS actuation, assuring that the reactor was shutdown under all conditions. The Emergency Action Level for this Alert classification requires that the following three conditions be met: Any RPS set point (including manual) has been exceeded per T.S. 3.3.1.1 AND RPS actuation failed to result in a control rod pattern which alone always assures reactor shutdown under all conditions AND Manual actions (mode switch in shutdown, manual push buttons, and ARI) result in reactor power LE 5%. Since all rods were successfully inserted without the assistance of any manual actions and within the Technical Specification required time, Columbia Generating Station staff now believes that no emergency classification should have been made, and we are retracting the Alert emergency classification from this event notification. The problem with control rod position indication following the scram is being addressed through our corrective action program. The licensee notified the NRC Resident Inspector and will notify local, State, and other Government Agencies of this update. Notified R4 DO (Runyan). | |
| ENS 40921 | 4 August 2004 14:24:00 | Davis Besse | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Steam Generator Main Turbine Decay Heat Removal Main Condenser Control Rod Main Steam | At 1024 EDT a reactor trip occurred during maintenance activities involving the control rod drive trip breakers. All control rods fully inserted. The unit is currently stable with decay heat removal via the main steam system through the turbine bypass valves to the main condenser. Post trip response was normal with the following exceptions noted: 1. #4 main turbine stop valve may not have fully closed. 2. A #2 Steam Generator Safety Valve may be lifting early (lifted at 1010 psi rather than the 1050 psi setpoint). 3. Turbine Bypass valve SP13A3 stuck slightly open - isolated. All offsite power lines have been verified operable and both EDGs are available in standby, if needed. The licensee informed the local sheriff's department as required whenever a Steam Generator Safety valve lifts and the NRC Resident Inspector. | |
| ENS 40954 | 14 August 2004 16:59:00 | Brunswick | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Emergency Diesel Generator Reactor Building Ventilation Control Rod | On August 14, 2004, at approximately 1259 hours (EDT), the unit scrammed due to a loss of offsite power. The cause of the loss of power has not been determined at this time. The four emergency diesel generators started as required and properly aligned to the emergency buses. All control rods properly inserted. The initial safety significance of this condition is considered to be minimal. All safety systems function properly with the exception of 1 B Standby Gas Train. The 1 B Standby Gas Train tripped due to an overheat condition. Safety significance of this failure is minimal since the 1A Standby Gas Train operated properly and Reactor Building Ventilation System isolated. The plant is in a stable condition. Troubleshooting activities are in progress to determine the cause of the event and corrective actions. Received Group II, VI, and VIII PCIS isolations due to the loss of power. The MSIVs are closed and the licensee manually started RCIC for reactor level control and HPCI for reactor pressure control. The licensee notified the NRC Resident Inspector. | |
| ENS 40957 | 15 August 2004 09:05:00 | River Bend | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Standby Gas Treatment System Main Condenser Main Steam | At 0405 on August 15, 2004, River Bend Station had a loss of 230Kv Line #353 (RSS #2, Port Hudson) feeding the plant's switchyard. This resulted in a generator trip and a reactor scram. Feedwater tripped on a Reactor vessel Level 8 signal. RCIC is operating to control reactor vessel level. SRVs were used to manually control pressure. Mechanical vacuum pump (ARC A) would not start to control condenser vacuum. The MSIVs were manually isolated in anticipation of a Group 6 isolation (low vacuum). There was no ECCS initiation. Power was restored to the mechanical vacuum pump ARC B. Vacuum was restored and main steam lanes are in use for pressure control. We are still investigating the root cause. The following ESF actuations occurred: - Division II Diesel Generator - Standby Gas Treatment System - Annulus Mixing System - Control Building Ventilation System. The licensee will inform the NRC Resident Inspector.
River Bend Station is currently in Mode 3, Hot Shutdown. During the event, offsite power was maintained through Reserve Station Service Line #1 to Division 1 safety related equipment and 'A' balance of plant (BOP) non-safety related equipment. The 230 KV line from the plant's switchyard (RSS Line #2) that was lost during the event has been recovered and the Division 2 bus has been restored to its normal power supply. The Division 2 diesel generator has been secured from operation (and has been placed in standby) During, the event, main condenser vacuum was restored from the main control room by starting the 'B' mechanical vacuum pump. Feedwater Pump 'A' remained available during the event. The circulating water supply to the main condenser remained in operation throughout the event. RCIC (Reactor Core Isolation Cooling) remains in service for reactor vessel level control. The ESF (Emergency Safety Feature) systems have been secured and placed in standby. A root cause team has been assembled to address the cause of the initial failure of the 230 KV Line #353. The licensee has notified the NRC Resident Inspector Notified R4DO (Linda Smith) and NRR EO (Cynthia Carpenter) | |
| ENS 40959 | 15 August 2004 20:03:00 | Columbia | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Feedwater Reactor Core Isolation Cooling Reactor Recirculation Pump Main Condenser Control Rod | With a reactor startup in progress at 1303 PDT, operators at Columbia Generating Station inserted a manual reactor scram when the running Reactor Feedwater (RFW-P-1A) pump tripped. Reactor power was approximately 18% at the time of RFW-P-1A trip. The cause of the RFW-P-1A trip was a high RFW Turbine Drain Tank (MD-TK-1) level; the cause of the MD-TK-1 high level is under investigation. The Reactor Core Isolation Cooling (RCIC) was used (manually started) to maintain reactor vessel water level until reactor pressure was reduced to within the capacity of the condensate booster pumps; the RCIC system has been returned to a standby lineup. The reactor is in mode 3 with both reactor recirculation pumps running at minimum speed (15 Hertz). Decay heat is being rejected to the main condenser via auxiliary steam loads. One control rod that indicated full-in immediately after the scram lost full-in indication eleven seconds after the scram. This control rod indicated full-in again 84 seconds after the scram. This ENS notification is made pursuant to 10 CFR 50.72(b)(2)(iv)(B). The licensee has notified the NRC Resident Inspector. | |
| ENS 40964 | 17 August 2004 12:28:00 | Columbia | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Feedwater Emergency Diesel Generator Reactor Core Isolation Cooling Reactor Recirculation Pump Safety Relief Valve Main Condenser | With a reactor startup in progress at 0528 PDT, operators at Columbia Generating Station inserted a manual reactor scram when the operating Reactor Feed Water (RFW) pump RFW-P-1A tripped. Reactor power was approximately 20% at the time of RFW pump trip. The cause of the RFW pump trip was due to low suction pressure; the cause of the low suction pressure is currently under investigation. The Reactor Core Isolation Cooling (RCIC) system (was manually started and) was used to maintain reactor vessel water level until reactor pressure was reduced to within the capacity of the condensate booster pumps (500 to 600 psi). The RCIC system has been returned to a standby lineup. The reactor is in Mode 3 (Hot Shutdown) with both reactor recirculation pumps running at minimum speed (15 Hertz). Decay heat is being rejected to the main condenser via auxiliary steam loads. All ECCS systems are operable. All emergency diesel generators are operable. No Safety Relief valves lifted during the scram. The NRC Resident Inspector was notified of this event by the licensee. See similar event number 40959 that occurred on 08/15/04. | |
| ENS 40974 | 22 August 2004 15:10:00 | Wolf Creek | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Coolant System Auxiliary Feedwater Control Rod | At 1010 CDT on 8-22-2004, during the performance of STS IC-211B, 'ACTUATION LOGIC TEST TRAIN B SOLID STATE PROTECTION SYSTEM,' Control Room operators received a reactor trip concurrent with power range flux lo (low) setpoint trip and intermediate range hi (high) flux reactor trip alarms. All safety-related equipment operated as required. The Steam Dump valves, which function to remove excess heat as the secondary systems shutdown, did not initially operate as expected, resulting in the operation of the Steam Generator Atmospheric Relief Valves (ARVs) to control Reactor Coolant pressure for approximately three minutes following the reactor trip, at which time the Steam Dump valves responded as expected and the ARVs closed. The Steam Dumps are currently operating correctly in automatic. The Auxiliary Feedwater System actuated as designed. The plant is currently stable in Mode 3 at Normal Operating Temperature and Pressure while plant personnel investigate the causes of the reactor trip and formulate the repair/ restart plan. A press release is planned. During the reactor trip, all control rods fully inserted. The electric plant is in a stable shutdown plant lineup. Plant personnel also intend to investigate the unusual response of the steam dump valves. The licensee has notified the NRC Resident Inspector.
Upon data review related to the operation of the Steam Dump valves and the Atmospheric Relief Valves (ARVs), plant staff has determined that the ARVs and Steam Dump valves operated as expected. The Steam Dump valves opened immediately following the reactor trip, as designed, due to the difference between their setpoint and the Reactor Coolant System (RCS) temperature at the time of the reactor trip. The Steam Dump valves remained open for approximately 20 to 30 seconds, at which time the RCS temperature had decreased to a value within the control range for the Steam Dump valves, causing them to close. The opening of the ARVs was in response, as designed, to the increase in Steam Generator pressure that resulted from the turbine trip shortly after the reactor trip." NRC R4DO (Tom Farnholtz) notified. The NRC Resident Inspector was notified of this update by the licensee | |
| ENS 40998 | 30 August 2004 12:35:00 | Nine Mile Point | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater High Pressure Coolant Injection Main Condenser Control Rod | Nine Mile Point, Unit One is initiating a 10 CFR 50.72 (b) (2) (iv) (B) 4-Hour Non-Emergency notification based upon insertion of a 'MANUAL' scram that occurred at 08:35 on Monday, August 30, 2004. At time of transient, plant was operating in Mode 1, Power Operating Condition, at 99.7% of rated power. At 08:25 on Monday, August 30, 2004, Operators noted oscillations on 13 Feedwater flow control valve (FCV) while in 'AUTOMATIC' mode of operation (normal mode of operation for this equipment). Operators took "MANUAL" control of 13 FCV per Plant Operating Procedures. 13 FCV oscillations continued while in the 'MANUAL' mode, and a decision was made to insert a 'MANUAL' scram at 08:35. All control rods fully inserted and the plant responded as designed to the scram. At 08:44, the scram signal was reset per procedure. Currently, plant is in Mode 2, Hot Shutdown Condition with cooldown in progress. Plant is transitioning to Mode 3, Cold Shutdown Condition, per Plant Operating Procedures. At the time the manual scram was inserted, Reactor Vessel Water Level (RVWL) was 67 inches and decreasing (automatic scram setpoint is 53 inches). The 13 FCV is on the discharge of the turbine-driven feedwater pump. Decay heat is currently being removed by the main condenser via the steam bypass valves. All ECCS and safety-related equipment is available, if needed. At the time of the transient there was no plant maintenance on-going which could have been a contributing factor. The licensee informed the NRC Resident Inspector.
The notification sent to the NRC on 8/30/04 at 11:42 was found to be incomplete. As a normal and expected response to a manual scram at high power, the High Pressure Coolant Injection System (feedwater) automatically initiated during the transient following the manual scram. This should have been reported as an 8 hour Non-Emergency 10CFR50. 72 (b) (3) (iv) (A) notification. The licensee notified the NRC Resident Inspector. Notified R1DO (Henderson). | |
| ENS 41002 | 31 August 2004 11:18:00 | Palisades | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Auxiliary Feedwater Main Condenser | At 0718 EDT, the reactor was manually tripped from approximately 95% power following notification to the control room of a fire associated with condensate pump 2B. Initially, upon notification of smoke at the condensate pump, a rapid down power had been commenced, wherein reactor power was reduced from 100% to approximately 95% power. An automatic actuation of the auxiliary feedwater system also occurred as designed to maintain steam generator water level following the reactor trip. The fire was extinguished in less than 10 minutes. The local fire department was notified, responded to the site as a precautionary measure, but was not used in extinguishing the fire. All systems functioned as designed. The reactor is stable in mode 3. Decay heat is being removed with the steam generators discharging steam to the main condenser. The licensee notified the NRC Resident Inspector. | |
| ENS 41003 | 1 September 2004 04:05:00 | Indian Point | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater Control Rod | On September 1, 2004, at 00:01 hours, operations noticed the 22 Main Feedwater (FW) regulating valve operating erratically and FW and 22 Steam Generator (SG) level oscillating. Operators placed the 22 Main FW regulating valve in manual and attempted to control FW flow and SG level but were unsuccessful. A Nuclear Plant Operator (NPO) in the plant reported water hammer in the Auxiliary Building that contains the FW regulating valves. At 00:05 hours operations initiated a manual trip of the reactor based on the degrading condition with FW flow and SG level. All control rods fully inserted and safety equipment operated as expected. Auxiliary Feedwater started as expected, the Emergency Diesel Generators did not start and offsite power remained available. The plant is in Hot (Standby) at Normal Temperature and Pressure and there was no radiation release. The 22 Main FW regulating valve was discovered to be partially stuck open. At 00:19 hours operations isolated the 22 FW flow path by closing the 22 FW line motor operated valve (MOV) BFD 5-1. The condition is under investigation and a post trip review is being conducted." State of NY was notified of this event by the licensee. The NRC Resident Inspector was notified of this event by the licensee.
After further review of the reporting guidelines and discussion with the NRC R1, Licensing concluded the event notification should be reported under 10CFR50.72 (b)(1)(iv)(B) for a 4-hour report. A Condition Report was initiated for the late notification. Notified NRR (Reis) and R1DO (Henderson). | |
| ENS 41017 | 4 September 2004 03:45:00 | Fermi | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | On September 3, 2004, at 2345, the Reactor Scrammed as the result of an (Automatic Voltage Regulator) AVR trip relay. AVR Channel A was not operating at the time due to an earlier fault (0541, 9/3/04). The AVR trip relay caused a Main Generator trip which caused a Main Turbine trip. The Main Turbine trip causes a direct Reactor SCRAM on Turbine Valve position. RPS functioned properly and all rods inserted. MSIVs remain open with Reactor Level maintained in the normal band of 173 to 214 inches. Reactor Pressure is being controlled with the Main Turbine Bypass valves at 600 to 1050 psig. Isolations occurred as expected for Reactor Level 3. The Licensee notified the NRC Resident Inspector. |