Semantic search
| Discovered date | Reporting criterion | Title | Description | LER | |
|---|---|---|---|---|---|
| ENS 43511 | 23 July 2007 13:32:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center (Tsc) Outage for Planned Maintenance | Planned preventive maintenance activities are being performed today (July 23, 2007) on the Hatch Nuclear Plant's Alternate Supply Breaker (5C) to the Diesel Generator Building Motor Control Center (MCC) (MPL # 1R24-S026). This MCC feeds the Technical Support Center (TSC) 480V AC distribution panel (MPL# 1R25-S102) which supplies power to the TSC HVAC. The work activities affecting the TSC are planned to be performed and completed expeditiously within one work shift (less than or equal to 6 hours). During this work activity the TSC HVAC system will be removed from service. If an emergency condition occurs that requires activation of the Technical Support Center during the time these work activities are being performed, it will take no more than three hours to return the equipment back to functional status, dependent on the stage of the work activity at the time the emergency occurs. Plans are to utilize the TSC for any declared emergency during the time these work activities are being performed, as long as radiological conditions allow. Procedure 73EP-EIP-063-0, Technical Support Center Activation provides instructions to direct TSC management to the Control Room and TSC support personnel to the Simulator Building to continue TSC activities if it is necessary to relocate from the primary TSC. This event is reportable per 10CFR50.72(b)(3)(xiii) as described in NUREG-1022, Rev 1 since this work activity affects an emergency response facility for the duration of the evolution. The licensee informed the NRC Resident Inspector.
Update: On 07/23/07 at 1935 EDT, the HVAC System for the TSC was returned to functional status following the replacement of the alternate supply breaker to MCC 1R24S026. 1R24S026 was energized restoring power to the TSC HVAC System." R2DO (M. Lesser) notified. The NRC Resident Inspector was notified of this event by the licensee. | |
| ENS 43541 | 1 August 2007 06:05:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Fire in Protected Area Greater than 10 Minutes | An Unusual Event was declared based on a fire lasting > 10 minutes within the Protected Area. Fire was on a Pole Mounted Transformer located on the 230 KV Supplemental Power System in the North East corner of the Protected Area. Helper Cooling Tower Fans were lost and sequenced back on. Unit 1 maintained 100% Rated Thermal Power. Unit 2 reduced power to 91.2% due to increasing Circulating Water Temperatures per procedure. Fire was extinguished at 0213 using dry chemical fire extinguishers. Fire was initially spotted at 0152 EDT. It is believed that a snake may have caused the transformer to arc and that the arcing caused the wood transformer pole to catch on fire. Other than the brief loss of helper cooling to the circ water, there was no other impact on the facility. No offsite assistance from the fire department was called or needed. The licensee terminated the Unusual Event at 0250 EDT. The licensee notified local and State Authorities and the NRC Resident Inspector. | |
| ENS 43552 | 7 August 2007 19:06:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram on Low Reactor Water Level Following Partial Loss of Condensate Feedwater Flow | Unit 2 RPS actuation / unplanned scram occurred at 1506 eastern time on low reactor water level scram initiation. The unit experienced a partial loss of condensate feedwater due to the trip of 2D 4160 volt station service bus (non-safety related). The 2C 4160 volt bus remained in service supplying power to one condensate pump and one condensate booster pump. Investigation of the 2D 4kv bus trip is in progress. Also received a Group II isolation signal due to low reactor water level. The reactor is currently stable with water level at 37 inches. Normal feedwater has been used to makeup water level and decay heat is being discharged to the condenser via turbine bypass valves. All rods fully inserted. No SRVs lifted during the transient. The lowest water level reached was -5 inches. The Unit remained in a normal electrical lineup. There were no significant LCOs in effect at the time of the scram. There was no impact on Unit 1. The licensee notes that I&C activities had been in progress on the 2D 4kv bus about the time that it tripped however there is currently no specific connection between these activities and the bus trip. The licensee notified the NRC Resident Inspector. | |
| ENS 43607 | 30 August 2007 12:58:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Planned Electrical Maintenance Removes Tsc Hvac from Service | Planned preventive maintenance activities are being performed today (August 30, 2007) on the Hatch Nuclear Plant's Normal Supply Breaker (1C) to the Diesel Generator Building Motor Control Center (MCC) (MPL # 1R24-S026). This MCC feeds the Technical Support Center (TSC) 480V AC distribution panel (MPL# 1R25-S102) which supplies power to the TSC HVAC. The work activities affecting the TSC are planned to be performed and completed expeditiously within one work shift (<12 hours). During this work activity the TSC HVAC system will be removed from service. If an emergency condition occurs that requires activation of the Technical Support Center, during the time these work activities are being performed, it will take no more than four hours to return the equipment back to functional status, dependent on the stage of the work activity at the time the emergency occurs. Plans are to utilize the TSC for any declared emergency during the time these work activities are being performed as long as radiological conditions allow. Procedure 73EP-EIP-063-0, Technical Support Center Activation provides instructions to direct TSC management to the Control Room and TSC support personnel to the Simulator Building to continue TSC activities if it is necessary to relocate from the primary TSC. This event is reportable per 10CFR50.72 (b)(3)(xiii) as described in NUREG-1022, Rev. 2 since this work activity affects an emergency response facility for the duration of the evolution. The licensee will notify the NRC Resident Inspector.
The planned maintenance was completed successfully and the TSC HVAC system was returned to service at 1748 EDT on 08/30/07. The licensee notified the NRC Resident Inspector. Notified the R2DO (Shaeffer) | |
| ENS 43624 | 8 September 2007 16:02:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Prompt Notification System Not Available | The licensee was informed by the Jacksonville, FL National Weather Service that the NOAA Prompt Notification System was not operational. Plant Hatch utilizes the NOAA system to broadcast prompt emergency notifications. Hatch will use radio and phone notifications as a backup (if needed) until the NOAA system is restored. No information was available on the cause of the NOAA system outage or when it would be returned to service. The license has notified the Georgia Emergency Management Agency, the local authorities, and will notify the NRC Resident Inspector.
The prompt notification system was restored at 1745 EDT on 9/8/07. The license has notified the Georgia Emergency Management Agency, the local authorities, and the NRC Resident Inspector. R2DO (Shaeffer) notified. | |
| ENS 43658 | 22 September 2007 07:15:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Jacksonville National Weather Service Land Lines | Jacksonville National Weather service has notified Plant Hatch that the capability to broadcast prompt notification messages is NOT available at this time. Plant Hatch security notified the Operations Shift Manager at 0315 hours to notify the NRC. Also, site Emergency Preparedness on call person has been contacted and he in conjunction with security has notified the state and local agencies. Information Technology (IT) has been notified of problem with 'land lines.' Compensatory measures are in effect. There is no estimate at this time for restoration of the Jacksonville "land lines. The licensee will inform the NRC Resident Inspector.
The licensee received notification from the Jacksonville National Weather service that the 'land lines' have been restored. Plant Hatch has regained the capability to broadcast prompt notification messages. The licensee will inform the NRC Resident Inspector. Notified R2DO (Bonser). | |
| ENS 43698 | 7 October 2007 05:15:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Prompt Notification Message System | Jacksonville National Weather Service does not have the capability to broadcast prompt notification messages due to phone lines or system problem at this time. Plant Hatch security notified the Operations Shift Manager at 0115 hour to notify NRC. Also, site Emergency Preparedness on call person has been contacted and security has notified the state and local agencies. Information Technology (IT) has been notified of problem. Compensatory measures performed by the local emergency response agencies will be used if needed. The licensee will notify the NRC Resident Inspector.
The Jacksonville National Weather Service has restored the capability to broadcast prompt notification messages effective 0344 EDT on 10/07/07. The licensee notified the NRC Resident Inspector and the state and local emergency response agencies. Notified R2DO (Evans). | |
| ENS 43884 | 6 January 2008 16:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Hpci Declared Inoperable Due to Failed Surveillance Test | During HPCI pump operability surveillance in preparation for a system outage, the system failed to achieve rated flow and pressure in the time required by procedure and Tech Specs. The procedural requirement is <49 seconds and the Tech Spec requirement is < 50 seconds. The system achieved rated flow and pressure in 54 seconds. The system outage has been delayed until troubleshooting plans can be developed and implemented. The licensee notified the NRC Resident Inspector.
The initial notification was made as a result of the failure of HPCI to meet its response time of 50 seconds as defined in the Technical Requirements Manual (TRM) due to a degraded component. Failure to meet the operability procedure requirements resulted in the high pressure coolant injection (HPCI) system being considered inoperable. Since HPCl is a single train system, its inoperability was the event that warranted a notification to meet the following reporting requirement: 10 CFR 50.72(b)(1)(v) Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems needed to: (D) Mitigate the consequences of an accident. An additional review and evaluation of the licensing basis was performed and demonstrated that the procedural required response time was set conservatively. As expected, the licensing basis does not require or assume any specific start time and HPCl is not credited in the accident analyses. The acceptance criterion contained in the TRM is within licensee control via the 10 CFR 50.59 process. Consequently, the TRM criterion was revised to 75 seconds to retain a value to assure continued monitoring and trending of HPCl performance in order to recognize and prevent continued performance degradation. Additionally, it is reasonable to conclude that HPCl would have completed its mission time of 4 hours despite the degraded condition of the EGR (Electronic Governor Remote) that caused the initial slower response time. This is based on the fact that HPCl started and ran at rated flow and pressure for approximately 39 minutes prior to shutdown with no problems identified. This removes any questions regarding its ability to restart and run based on demand during its mission time of 4 hours. An industry expert on this system and the engineer from the vendor that supports this system concurred with that conclusion. Since this subsequent review and evaluation determined that the slower response time did not render HPCl inoperable, no single train failure of HPCl occurred. The system was fully capable of performing its intended safety functions during the event timeline. Based on this information this notification serves to retract notification # 43884 made on 1/06/08. The licensee notified the Resident Inspector of this retraction. Notified the R2DO (Musser). | |
| ENS 43968 | 7 February 2008 08:40:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Osc Hvac System Did Not Start as Required | In preparation for scheduled Motor Control Center (MCC) cleaning activities, the Health Physics normal HVAC system had to be shutdown and the Health Physics emergency HVAC was to be started while the MCC that supplies power to the normal HVAC would be out of service. When the emergency HVAC system was started per the system operating procedure (34S0-Z41-006-0), the air handling unit (1Z41-B100) started, but neither compressor units (1Z41-B101) started. Maintenance has been notified and is currently working to restore the system to operable status. This event is reportable per 10CFR50,72 (b)(3)(xiii) as described in NUREG-1022, rev. 2 as an emergency facility (Plant Hatch Operational Support Center) is affected with the Health Physics emergency HVAC out of service. Update: maintenance cleaned electrical contactors and have both compressors running at 0436 hours. The licensee notified the NRC Resident Inspector. | |
| ENS 43969 | 7 February 2008 07:30:00 | 10 CFR 50.72(b)(3)(xii), Transport of a Contaminated Person Offsite | Injured Carpenter Transported to Appling County Hospital | At 0230 EST an injured contract worker who was potentially contaminated was transported by ambulance to an offsite medical facility. The contract employee fell approximately 15 feet from a scaffold in the Unit 1 condenser bay. Unit 1 is presently shutdown (Mode 5) for a refueling outage. The worker was placed on a stretcher for transport and was partially frisked prior to release. However, due to concerns for a potential back injury, the worker was not removed from the stretcher for a back survey and was transported, with HP escort, as a contaminated person. After arriving at the hospital, a complete survey was performed and the worker was found to be free of contamination. The licensee notified the NRC Resident Inspector. | |
| ENS 43992 | 18 February 2008 06:20:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to #2 Turbine Bypass Valve Failure | Unit two reduced power to 90% to perform monthly turbine testing surveillance on 2/18/08. All tests were completed satisfactorily with the exception of the main turbine #2 Bypass Valve (BPV). At 0120 hours, the #2 BPV stroked fully open per procedure, but the last 10% of travel to full open did not yield the expected response of BPV fast open from 90% to 100%. The fast acting solenoid did indicate expected state change to 'energized' at 90% valve open, but the BPV did not indicate fast open. Therefore the Main Turbine Bypass System has been declared inoperable and associated actions of Tech Spec 3.7.7 have been invoked. This spec requires compliance with LCO 3 .2.2 MINIMUM CRITICAL POWER RATIO (MCPR) limits for an inoperable main turbine bypass system as specified in the COLR (Core Operating Limit Report), are made applicable (within 2 hours) or reduce THERMAL POWER to <24% RTP within the following 4 hours. The MCPR limits were calculated by reactor engineering and installed in the process computer at 0347 hours. A decrease in reactor power was not required once the MCPR limit was installed. The COLR (ref.: TRM Appendix A) states Unit Two can be operated with EITHER the End- of- Cycle Recirc Pump Trips (EOC-RPTs) out of service OR the Turbine Bypass Valves inoperable, but not both. The EOC-RPTs were already out of service as allowed for current conditions of the operating cycle. Upon discovery of the inoperable Bypass Valve, it was recognized that this placed Unit Two in an unanalyzed condition for fuel thermal limit. The shift crew took immediate actions to confirm the surveillance was current for EOC-RPTs and placed them in service per approved plant procedures. This was accomplished at 0315 hours, which returned the unit to an analyzed condition. The licensee notified the NRC Resident Inspector. * * * UPDATE FROM J. ANDERSON TO P. SNYDER ON 3/20/08 AT 0943 * * * During the surveillance testing of the main turbine bypass valves, one of the three main turbine bypass valves did not function as expected. The function of the main turbine bypass system was degraded but not lost. Upon discovery the required action statement (RAS) in Technical Specifications 3.7.7 and 3.2.2 were properly entered, and the required actions were taken within the allowed out of service time of two hours. Based on the initial review of the condition and the fact that the core operating limits report (COLR) described operation in the condition with the EOC-RPT out of service concurrent with loss of the main turbine bypass capability as an unanalyzed condition, a notification was made in accordance with the following reporting requirement: '10 CFR 50.72 (b)(3)(ii)(B) The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' The initial notification was a conservative action taken shortly after the condition was discovered. The condition was immediately identified during surveillance testing of the main turbine bypass valves, the required Technical Specifications RAS was entered and the actions completed within the allowed two hour time frame of 0120 - 0315 EST on 2/28/08. This prompt action prevented continued operation with EOC-RPT out of service and main turbine bypass inoperable and eliminated this potential to be in a condition where the design basis may not have been met. A more detailed review was subsequently performed which determined that the minimum critical power ratio thermal limit for having EOC-RPT out of service and main turbine bypass inoperable as calculated during the reload analysis was 1.42. At the time of the event the actual MCPR at that point in core life was 1.57. Even though prompt actions were taken as required, there was actual margin to the calculated MCPR limit of 1.42. Had a design basis transient occurred, the MCPR Safety Limit would not have been exceeded. Based on this information the determination has been made that the unit was not in an unanalyzed condition that significantly degraded plant safety. This notification serves to retract the previous event notification # 43992 made on 2/18/2008 at 0508 EST. The licensee notified the NRC Resident Inspector. | |
| ENS 43996 | 20 February 2008 20:05:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Rhr Service Water System Declared Inoperable Due to Inability to Meet Fsar Requirement | At 1505 EST, it was determined by an engineering analysis that a leak on a 3/4 inch vent line of the Residual Heat Removal Service Water (RHRSW) System common discharge line would prevent continuous operation of the system for 30 days as required by the Unit 2 FSAR. Specifically, if the line were to completely shear, the resulting leak would have exceeded the capacity of the building sumps and over a prolonged period of operation would have resulted in the excessive water levels in the Reactor Building Torus room area. In order to mitigate the condition, supplemental pumping capacity would be needed in the building sumps or the RHRSW system would have to be temporarily secured. The RHRSW system has been tagged out for repair of the line. The appropriate 8-hour RAS in Tech Spec has been entered. The line will be repaired and the system returned to service. The licensee identified the crack in the socket weld during a system walkdown and anticipates completion of repairs in approximately 6 hours. The licensee informed the NRC Resident Inspector.
At 2120 EST on 02/20/08 the RHR SW System was declared OPERABLE after completion of repairs and testing. The failure of the socket weld at 2E11-FV001 is preliminarily attributed to fatigue cracking caused by vibration. A root cause investigation is ongoing. A system walkdown did not identify any additional deficiencies. The licensee informed the NRC Resident Inspector. Notified R2DO(Robert Haag). | |
| ENS 44005 | 23 February 2008 21:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Unacceptable Flaw in Rpv Cap-To-Pipe Weld | At 1600 EST, it was determined by a phased array ultrasonic examination that an unacceptable flaw existed in reactor pressure vessel penetration N9. This penetration leads to a capped line and the flaw is in the cap-to-pipe weld. The flaw is 2.3 inches in length and reaches a maximum depth of 43.6 percent through-wall on a 5.75 inch OD pipe. The flaw was discovered during the routine ISI (In-service Inspection) examination of this penetration. The flaw has been found unacceptable per paragraph IWB-3514.4 of the 2003 Addenda of ASME Section XI and is therefore reportable. Upon discovery of the flaw, all penetrations of this type were examined with no further findings. A weld overlay repair is planned and should be completed by 3/6/08. The licensee notified the NRC Resident Inspector. * * * UPDATE FROM J. ANDERSON TO P. SNYDER ON 3/20/08 AT 0943 * * * On 2/23/2008 event number 44005 was made to report an unacceptable flaw in reactor pressure vessel penetration N9. The assumption made at that time was that the flaw seriously degraded the affected control rod drive (CRD) return line that was capped. This was based on the fact that the flaw was unacceptable from an ASME Section XI perspective and therefore reportable. Based on this assumption the event was reported in accordance with the following reporting requirement. '10 CFR 50.72 (b)(3)(ii)(A) Any event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded;' The depth of the flaw was initially reported to be 43.6% through wall and the thickness of the wall was measured to be 0.75 inches. Phased array ultrasonic examinations in accordance with Appendix VIII of Section XI were used to fully interrogate and characterize the circumferentially oriented flaw. The final dimensions were 2.3 inches in length on the inner diameter and a maximum depth of 60% through wall. A flaw evaluation of the Unit 1 CRD return line nozzle cap weld was subsequently performed by Structural Integrity Associates, Inc., to evaluate the 'as-found' condition and to determine the implications of the flaw on the affected system at the time of shutdown. The evaluation considered the flaw size and the appropriate stresses for operation. No crack growth was required to be assumed since the 'as-found' flaw was compared against the allowable flaw size which was used to determine the acceptability of the flaw during plant operation. Since the weld is associated with the capped CRD return line nozzle, the applied stress at the affected location was due to pressure loading only. The flaw evaluation concluded that the requirements of Section IWB-3640 were satisfied for a flaw less than 75% through-wall. The 'as-found' depth of 60% through wall is less than the allowable depth of 75%. Based on this information, the Structural Integrity Associates, Inc. evaluation demonstrated that the flaw was acceptable per the ASME Code Section Xl, 2001 Edition through 2003 Addenda requirements. Since the as- found flaw depth at the N9 weld is less than the allowable flaw depth, the required safety factors were met at all times during plant operation. Based on this updated information the conclusion has been reached that this flaw did not seriously degrade the plant or its principal safety barriers. Since the condition does not meet a reporting requirement, this notification serves to retract Notification # 44005 made on 2/23/2008. The licensee notified the NRC Resident Inspector. | |
| ENS 44007 | 25 February 2008 10:40:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Operations Support Center Rendered Non-Functional During Planned Maintenance | Planned preventive and corrective maintenance activities are being performed today (2/25/08) on the Hatch Nuclear Plant's HP/Chemistry area emergency air handling unit. The HP/Chemistry area emergency HVAC is required for functionality of the Operations Support Center (OSC), as this area may be utilized during declared emergencies. These planned work activities are to be completed today within (15) hours. During the time these activities are being performed, the HP office emergency HVAC will not be available for operation and will be rendered non-functional during the performance of the work activities. If an emergency condition requiring activation of the OSC occurs during the time these work activities are being performed, then contingency plans call for continued utilization of the HP/Chemistry areas as long as radiological conditions allow. Habitability surveys will be performed more frequently in the areas while they are being utilized. If required, respiratory protection equipment will be utilized, as appropriate, for emergency responders utilizing this area until such time as the power is restored to allow the HP/Chemistry emergency HVAC to be returned to service. This event is reportable per 10 CFR 50.72 (b)(3)(xiii) as described in NUREG-1022. Rev. 2, since this work activity affects an emergency response facility for the duration of the evolutions. The licensee has notified the NRC Resident Inspector. | |
| ENS 44038 | 6 March 2008 09:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Pin Hole Leak Discovered in 1 Inch Instrument Line During Rpv Pressure Test | With Unit 1 in Mode 4 for a planned refueling outage, a pin hole leak was discovered on a 1 inch line between the 'A' Main Steam Line (MSL) and MSL flow instrument condensing chamber (1B21-D006B) in a weld at a 45 degree elbow. Leakage was identified as approximately 2 gallons per hour (GPH). This leakage was identified during RPV pressure test while test pressure was 1050 psig. This elbow is near the 1B21-D006B condensing chamber and is located in the Unit 1 drywell (primary containment). This item constitutes a primary coolant boundary leak discovered while shutdown. The licensee notified the NRC Resident Inspector. | |
| ENS 44046 | 7 March 2008 19:46:00 | 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 with Hpci/Rcic Actuation Due to Loss of Condensate Feedwater | Unit 2 RPS actuation / unplanned scram with subsequent ECCS discharge to the RCS at 1446 hrs. on 3/07/08. Unit 2 scrammed on Low RPV water level of 3 inches above instrument zero as a result of a loss of condensate feedwater. Water level decreased to approximately 60 inches below instrument zero as a result of the loss of feedwater. (Top of active fuel is approximately 150 inches below instrument zero.) The cause of the loss of feedwater is presently under investigation. At 35 inches below instrument zero, HPCI and RCIC actuated and restored water level. HPCI oscillations were experienced and the system was taken to manual control, at which time the flow oscillations abated. All other systems functioned as required. A team has been assembled to investigate and determine the cause of the initiating event of the loss of feedwater. During the scram, all rods inserted into the core. There were no safety relief valve actuations as a result of the transient. RPV level was restored and is being maintained using control rod drive flow. The electrical grid is stable with normal offsite power supplying safety loads. Decay heat is being removed using the turbine bypass valves to condenser. The licensee has notified the NRC Resident Inspector. | |
| ENS 44093 | 25 March 2008 19:22:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Health Physics Ventilation System Used for Emergency Response | Shift received alarms which lead to the discovery of MCC 1 R24-S030 being de-energized. This MCC supplies power to the Health Physics emergency HVAC air handling unit (1Z41-B100). Maintenance has been notified and is currently working to restore the system to OPERABLE status. This event is reportable per 10CFR50.72 (b)(3)(xiii) as described in NUREG-1022, rev. 2 as an emergency facility (Plant Hatch Operational Support Center) affected due to the Health Physics emergency HVAC out of service. The licensee notified the NRC Resident Inspector. | |
| ENS 44201 | 19 March 2008 04:00:00 | 10 CFR 21.21, Notification of failure to comply or existence of a defect and its evaluation | General Electric Hitachi Cr120A Relay/Relay Coil Wire Clamp Cracking | Southern Nuclear Company made this notification of a potential defect in components received by its Edwin I. Hatch Nuclear Plant. The potential defect may exist in ninety-six (96) subject components: ninety-four (94) General Electric Hitachi (GEH) CR120A relays and two (2) GEH CR120A relay coils. The wire clamps supplied with these relays and relay coils may be susceptible to cracking. The wire clamps serve to secure the wire connections to the relays and relay coils. Currently there are thirty-seven (37) CR120A relays and one (1) CR120A relay coil installed. The components not currently installed have been placed on hold. The licensee has followed the recommendations of GEH by visually inspecting all installed components for cracks. No cracks were found. The licensee will replace the clamps on one subject relay installed in the Main Control Room Environmental Control System. The clamps on the components installed in the Source Range Monitoring and Intermediate Range Monitoring instruments show no signs of cracking will be left installed as is. Replacement wire clamps will be ordered from GEH. The licensee plans to replace the clamps on the other stock components. | |
| ENS 44301 | 17 June 2008 09:43:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Tsc Hvac System Rendered Nonfunctional | Work Order Number 1050889203 will perform maintenance on 1R24-S026 - 600/208V MCC. The specific work is to remove the LA 600 Breakers in frames 1C & 5C to take measurements for replacement breakers. The removal and measurement of the two breakers will be performed by two work crews simultaneously to minimize out of service time. In order to perform this work activity, the power will be removed from 1R24-S026 which also supplies power to the TSC HVAC rendering it nonfunctional during the performance of this work activity. This work activity is planned to be performed and completed within a 12 hour work shift with 2 hours scheduled for establishing and removing the clearances for a total of 14 hours. During the time this activity is being performed, the TSC air handling unit, TSC condensing unit, TSC filter train, and the fan unit for the TSC filter train will not be available for operation. As such, the TSC HVAC will be rendered nonfunctional during the performance of this work activity. If an emergency declaration is made requiring activation of the TSC during the time this work activity is being performed it will take approximately (3-6) hours to return the equipment back to an operable status dependent on the stage of the work activity at the time the emergency occurs. Plans are to utilize the TSC for any declared emergency during the time these work activities are being performed, as long as radiological conditions allow. Procedure 73EP-EIP-063-0, Technical Support Center Activation, provides instructions to direct TSC management to relocate to the Control Room and TSC support personnel to relocate to the Simulator Building to continue TSC activities if it is necessary to relocate from the primary TSC. Licensee notified NRC Resident Inspector.
At 1500EDT the MCC was re-energized and the HVAC equipment confirmed operable at 1518EDT on 06/17/08. The licensee will inform the NRC Resident Inspector. Notified R2DO (Bonser). | |
| ENS 44337 | 4 July 2008 12:39:00 | 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 Turbine Trip | A Turbine Trip greater than 30% power caused a Reactor Trip (Scram), both Recirculation Pumps tripped. A low level (Reactor Vessel Water Level) of approximately 2 inches caused a Group 2 containment valve isolation signal, all valves closed as required. The cause of the Turbine Trip is under investigation All control rods fully inserted with no ECCS actuations. Unit 1 is currently stable in mode 3 (Hot Shutdown) with decay heat being removed via the bypass. Following the scram, one SRV lifted and reseated. At the time of the transient, an EHC pump autostart was in progress, however, there is no indication that this was the cause of the turbine trip. Unit 1 is in a normal shutdown electrical lineup. The licensee informed the NRC Resident Inspector. | |
| ENS 44646 | 8 November 2008 13:35:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Osc Inoperable Due to Planned Maintenance on Motor Control Center | Planned maintenance activities are being performed today (November 8, 2008) on the Hatch Nuclear Plant's 1C 4160V Bus (1R22-S003) Frame 5 under work order number 1041640301. The planned work will remove power from 1R24-SO34 and 1R24-SPO60 which will affect the Unit 1 Service Building and Unit 2 Service Building Annex. This will deenergize power to the Primary Operations Support Center (OSC) rendering the OSC non-functional. (TRM - Section T 3.10.1 EMERGENCY RESPONSE FACILITIES) during the performance of this work activity. The specific work is to replace the breaker in frame 5 of 1C 4160V Bus (1R22-S-003) and perform routine preventative maintenance on 1R24-S034 and 2R24-S050 (Work Order # 1041641101) while they are de-energized to minimize future out of service time. This work activity is planned to be performed expeditiously and will be completed within 10 hours. If an emergency were to occur, it is estimated that power can be restored in approximately 1 hour. If an emergency condition occurs that requires activation of the OSC, the Alternate OSC in the Simulator Building will be used in accordance with plant procedure 73EP-EIP-021-0, Alternate Operations Support center (OSC) Activation, until power is restored. This event is reportable per 10CFR50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 1 since this work activity affects the emergency response facility. The licensee notified the NRC Resident Inspector.
At 1602 EST, power was restored to the OSC Motor Control Center. Notified the R2DO (Widmann). | |
| ENS 44654 | 13 November 2008 06:02:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of National Weather Service Prompt Notification Messages | At approximately 0102 eastern time, Jacksonville National Weather service has notified Plant Hatch that the capability to broadcast prompt notification messages is NOT available at this time due to problems with the phone service. Emergency Preparedness personnel (DELETED) have been contacted to ensure proper notifications have been made. Security has notified state and local agencies along with Information Technology (IT) group of the problem. At 0145, Jacksonville National Weather service has reported that the problem with the phone service has been resolved and that the broadcast system has been restored. All agencies have been made aware of the NWS system return to service. The licensee notified Georgia EMA, Appling County, Jeff Davis County, Tattnall County, and Toombs County. The licensee will notify the NRC Resident Inspector. | |
| ENS 44658 | 14 November 2008 01:31:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Prompt Notification Capability | At approximately 2031 Eastern time, Jacksonville National Weather service has notified Plant Hatch that the capability to broadcast prompt notification messages is not available at this time due to problems with the phone service. Emergency Preparedness personnel have been contacted to ensure proper notifications have been made. Security has notified state and local agencies along with Information Technology (IT) group of the problem. The licensee has notified NRC Resident Inspector.
At 2315 on 11/13/08, the National Weather Service notified the licensee that the notification network is available. However, the National Weather Service cannot guarantee that the system will remain operational due to on-going maintenance activities on the system. The licensee is monitoring the notification network and will provide an update once the maintenance is complete and system reliability is fully restored. Notified R2DO(Bonser)
Troubleshooting the system has been completed. Although the system was restored at 2315 EST on 11/13/08, the verification that the system would maintain operability was not completed until 1440 EST on 11/14/08. The licensee has notified NRC Resident Inspector. Notified R2DO (Bonser). | |
| ENS 44679 | 22 November 2008 15:18:00 | 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 | Manual Reactor Scram After Feed Pump Trip | Manual Rx Scram initiated due to a loss of condensate/feedwater. Condensate Booster Pump 1A tripped due to low suction pressure and then both Reactor Feed-pumps tripped. Both HPCI and RCIC initiated on low level and restored reactor water level to normal band 5 to 50 inches. Both Reactor Water Recirculation Pumps tripped due to low level at RWL (Reactor Water Level) - 60". Lowest RWL was approximately minus 70 inches and a group two isolation occurred at RWL 3 inches. All group two 2 valves closed as required. The cause of the low condensate booster suction pressure is under investigation. Rods fully inserted on the scram. No safety or relief valves lifted after the scram. Reactor water level is being maintained with normal feed and decay heat is being removed to the main condenser. The plant is in its normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector. | |
| ENS 44815 | 29 January 2009 13:50:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event - Toxic Gas Discharge in Waste Gas Treatment Building | An unisolable pipe break on an in-service waste gas treatment building chiller has resulted in a toxic release of asphyxiant freon gas. All personnel were evacuated and no injuries occurred. An entry by operations personnel with self contained breathing apparatus verified that all site personnel have evacuated the building. There was no impact to offgas processing equipment in the area and actual gas concentration was not known at the time of declaration so the assumption of toxic gas concentration is conservative. The Licensee notified the NRC Resident Inspector, Georgia Emergency Management Agency, Appling County, Jeff Davis County, Tattnall County, and Toombs County.
Air samples in the area indicated no detectable freon. Licensee terminated the event at 0946. Notified R2DO(McCoy), NRR EO(Ross-Lee), DHS(Inzer), and FEMA(LaForte) | |
| ENS 44849 | 12 February 2009 17:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection Suction Source Instrument Malfunction | Unit 2 HPCI (High Pressure Coolant Injection) system is being considered inoperable due to the following information: (Condition Report 2009101257) The instrumentation associated with the automatic suction swap for Unit 2 HPCI was reviewed as a result of CR 2009100480 to confirm the set points that determine the condensate storage tank (CST) level at which the suction swap would occur. During the course of this review, the corporate design engineer contacted the level switch vendor to review the configuration of the level switches and to confirm the expected operation of the switches (2E41-N002 & 2E41-N003) given their configuration. Based on the configuration of the instrument lines and physical location of the level switches, the vendor reported that either liquid or gas would most likely be entrapped in the external cage of the Magnetrol level switches. This would prevent the instruments from performing their automatic swap function. Based on this information the 'as found' condition of the switches indicate that this condition has been present since the installation of the switches when implementing the DCP in 1991 which affects the operability of this instrumentation. Even though the suction swap instrumentation on low CST level is considered inoperable, there is no apparent actual adverse impact on nuclear safety. However, the instrumentation is included in the Technical Specifications and its inoperability would make HPCI inoperable if it is aligned to the CST rather than being aligned to the suppression pool. The normal system alignment is with its suction source to the CST, therefore HPCI is being considered as inoperable. Until the configuration of the level switches has been addressed, these Magnetrol level switches must be considered inoperable, the appropriate Technical Specification RAS (Required Action Statement) will be entered and the suction source for HPCI should be aligned to the suppression pool when HPCI is required to be operable. This condition only applies to Unit 2. The licensee has notified the NRC Resident Inspector.
Event Report 44849 Retraction: On February 12, 2009, a condition was discovered where the physical location of level switches relied upon for automatically transferring the suction of the Unit 2 high pressure coolant injection (HPCI) system from the condensate storage tank (CST) to the suppression pool on low CST level did not meet the setpoints given in the Technical Specifications. Based on the information available at that time HPCI would have to be considered inoperable based on the fact that the affected instrumentation was inoperable and with HPCI aligned to the CST. Since the unit was shutdown HPCI was not required to be operable. After further review the determination has been made that at the time of discovery the 'as found' plant configuration associated with the suction swap setpoint for the Unit 2 high pressure coolant injection (HPCI) system could NOT have prevented the fulfillment of the safety function since the unit was in Cold Shutdown, and HPCI was not required to be operable. Based on this information this condition did not require an NRC notification in accordance with I0CFR50.72 and as such is being retracted through this update response. The condition will be reported in accordance with I0CFR50.73(a)(2)(v), The licensee has notified the NRC Resident Inspector. Notified R2DO (Sykes) | |
| ENS 45049 | 8 May 2009 19:15:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Inadvertant Main Steam Isolation While Performing Maintenance | At 1515 on 5/8/09, a Group 1 isolation signal was received which resulted in all eight Main Steam Isolation Valves closing. The signal was received based upon a valid main condenser low vacuum signal coincident with a main turbine reset signal which opened the turbine stop valves. The reset of the turbine was an unanticipated result of ongoing Mark VI turbine control system processor repair work. The reactor was maintaining hot shutdown while conducting nuclear instrumentation repair. The Group 1 isolation was completed successfully with all MSlVs and small bore valves closing as designed. The condition causing the turbine reset has been cleared and the turbine tripped with all valves closed. The Group 1 isolation signal has been reset and the MSIVs have been re-opened. The cause of the turbine reset signal is being investigated by Station Engineering and the on-site GE representative. The reactor is being taken to cold shutdown for intermediate range nuclear instrument maintenance. The licensee notified the NRC Resident Inspector. | |
| ENS 45052 | 10 May 2009 14:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Intermediate Range Reactor Scram During Transition to Run | During Startup of HNP-1, after reaching greater than or equal to 7% RTP (rated thermal power), the crew placed the Reactor MODE switch to 'RUN,' in accordance with the Startup Procedure 34GO-OPS-001-1. Upon placing the MODE switch to 'RUN,' a full RPS actuation occurred due to upscale trip signals on the Intermediate Range Nuclear Instrumentation (IRM) 1C51K601A, 1C51K601D and 1C51K601H. Placing the Reactor MODE switch to 'RUN' bypasses the IRM inputs to the RPS system, so actuation of the RPS from the IRMS was not expected. All withdrawn control rods inserted properly upon receipt of the full SCRAM signal. All equipment functioned as expected, with the exception of the unexpected upscale trips of IRMs 1C51K601A, 1C51K601D and 1C51K601H and the subsequent RPS actuation from the IRMs. No reactivity changes were in progress at the time to cause the upscale trip signals. At this time, investigation is in progress, but the investigation and corrective action have not yet been completed. The shift crew is progressing to Cold Shutdown, MODE 4 at this time. The decay heat is being removed by main turbine sealing steam with makeup provided via the control rod drive system. Offsite power is provided by the normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector. | |
| ENS 45055 | 11 May 2009 20:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification to National Response Center | (The licensee was) Notified by chemistry, that while performing (the procedure for monitoring) the annual releases via unplanned routes for gamma sampling for groundwater well P17B, which is located near the southeast corner of the Diesel Generator (DG) building, a strong diesel / fuel oil odor in the groundwater being collected was detected by the sampling technician. In addition to the strong petroleum odor, the groundwater from the well had a reddish tint. This color is indicative of the red dye used to identify off-road diesel. The P17B is normally sampled annually for tritium and gamma. The last sample from this well was taken in January 2009 for the tritium sample surveillance. Southern Nuclear Company (SNC) Corporate Environmental Affairs was contacted to make the appropriate communications. Environmental Affairs has contacted the National Response Center. (The licensee's) Plans are to excavate in the area of well P17B to determine potential source of fuel oil to (the) ground water well. The licensee has notified the NRC Resident Inspector. | |
| ENS 45056 | 12 May 2009 11:16:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Elevated Tritium Levels in Ground Water Sample | (The licensee was) notified by Southern Nuclear Company (SNC) Chemistry that, while performing (the procedure for monitoring) the tritium activity results for ground water sample well number T3, elevated tritium levels were detected. On May 5, 2009, (a) data report from Georgia Power Central (GPC) Lab containing tritium activity results for ground water sample well T3 was received and reviewed for input into (the) Chemistry database. (The) activity for well T3 was observed to be elevated (36,500 pCi/L). This concentration had increased from (the) last sample activity (5400 pCi/L) on March 16, 2009. Chemistry personnel were directed to resample for verification (of the elevated activity levels). A sample was collected on May 6, 2009 and shipped to GPC Lab. Sample results for the verification sample were received May 11, 2009. These results were confirmed to be elevated (34,300 pCi/L) compared to previous levels. Chemistry Management has notified the resident NRC inspector as well as SNC environmental affairs. SNC Corporate is performing the necessary notifications to the Georgia State Environmental Protection Division (EPD). The licensee is developing a plan to ascertain the cause of this event. The licensee has notified the NRC Resident Inspector. | |
| ENS 45071 | 15 May 2009 09:19:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Group 1 Isolation of All Main Steam Valves While Performing a Special Procedure | At 0519 on 5/15/09, a Group 1 isolation signal was received which resulted in all eight Main Steam Isolation Valves closing. The signal was received based upon a valid main condenser low vacuum signal coincident with reactor mode switch placed in RUN position. The isolation was an unanticipated result of a special purpose procedure which was being performed as a functional test for maintenance work that had been performed on intermediate range nuclear instrumentation. The procedure had installed jumpers to bypass the Group 1 isolation for Mode Switch in Run, but did not account for low condenser vacuum isolation. The low condenser vacuum switches were in the bypass position, but this logic does not prevent Group 1 isolation in the Run mode. The Group 1 isolation was completed successfully with all MSIVs and small bore valves closing as designed. MSIV closure with Mode Switch in Run position also caused a RPS actuation / full scram. The reactor was subcritical and all control rods were already fully inserted as the reactor was being maintained in Cold Shutdown. The licensee has notified the NRC Resident Inspector. | |
| ENS 45114 | 5 June 2009 13:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Techncial Support Center Ventilation System Out of Service | Planned maintenance activities are being performed today (June 5, 2009) on the Hatch Nuclear Plant's Technical Support Center (TSC). This work activity is planned to be performed and completed expeditiously within one work shift (less than 12 hours). This maintenance activity is to repair or replace solenoid for valve 1X75F003 on the TSC HVAC System which is not always reliably opening. During this work the TSC HVAC will be rendered non-functional during the performance of the work activity. If an emergency condition occurs that requires activation of the Technical Support Center during this work activity, it will take no more than two hours to return the equipment back to functional status, dependent on the stage of the work activity at the time the emergency occurs. Plans are to utilize the TSC for any declared emergency during the time the work activity is being performed, as long as radiological conditions allow. Procedure 73EP-EIP-063-0, Technical Support Center Activation, provides instructions to direct TSC management to the Control Room and TSC support personnel to the Simulator Building to continue TSC activities if it is necessary to relocate from the primary TSC. This event is reportable per 10CFR50.72 (b)(3)(xiii) as described in NUREG-1022, Rev. 2 since this work activity affects an emergency response facility for the duration of the evolution. The licensee has notified the NRC Resident Inspector.
The planned maintenance activities are complete and the TSC ventilation system has been returned to service. The licensee has notified the NRC Resident Inspector. Notified R2DO (Lesser). | |
| ENS 45145 | 20 June 2009 18:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram from High Reactor Pressure Scram Signal | Plant Hatch Unit 2 experienced a full reactor scram from the main generator protection circuitry (generator runback circuit). Preliminary indications are that a main generator high temperature signal was received, initiating the generator protection (runback) circuitry and a high reactor pressure scram signal was received during the turbine/generator runback. Investigations into the cause of the generator high temperature signal are ongoing. Reactor water level was recovered using the reactor feed system, and reactor pressure was controlled using main turbine bypass valves. All control rods inserted, as expected, during the scram. Other than the cause of the main generator high temperature signal, all systems functioned as expected. Unit is currently at 837 psig; 540 degrees F in Mode 3. Electrical system is in a normal lineup. The licensee informed the NRC Resident Inspector. | |
| ENS 45148 | 23 June 2009 07:51:00 | 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 High Reactor Water Level | (The reactor automatically scrammed) on a Main Turbine Trip >27.6% rated thermal power. The main turbine trip was due to reactor high level. Post scram, reactor level decreased to approximately -26 inches. Reactor water level was restored with the condensate system. Both reactor recirc pumps tripped as required on EOC RPT Logic when the main turbine tripped. Both pumps have been restarted. A Group 2 isolation was received at +3 inches reactor water level with all valves closing as required. Investigation as to the cause of the transient is underway. All rods inserted during the scram. No relief valves actuated during the transient. Decay heat is being removed via turbine bypass valves to the main condenser. The plant is within normal shutdown temperature and pressure limits. The electrical grid is stable and the plant is in a normal shutdown electrical lineup. The Group 2 has been reset. There was no effect on Unit 1. The licensee has notified the NRC Resident Inspector. | |
| ENS 45371 | 21 September 2009 12:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center Unavailable Due to Planned Maintenance | Planned preventive maintenance and testing activities are being performed on the Hatch Nuclear Plant's Technical Support Center (TSC) HVAC system on September 21, 2009. These maintenance activities include the performance of preventive maintenance on the TSC air handling unit, condensing unit and fan and testing of the filter train. These work activities are planned to be completed within the (12) hour day shift on 9/21/2009. During the time these activities are being performed, the TSC air handling unit, TSC condensing unit, TSC filter train and the fan unit for the TSC filter train will not be available for operation. As such, the TSC HVAC will be rendered non-functional during the performance of this work activity. If an emergency condition occurs during the time these work activities are being performed which requires activation of the TSC, the contingency plan calls for utilization of the TSC, as long as radiological conditions allow for habitability of the facility. Procedure 73EP-EIP-063-0, Technical Support Center Activation, provides instructions to direct TSC management to the Control Room and TSC support personnel to the Simulator Building to continue TSC activities if it is necessary to relocate from the primary TSC so that TSC functions can be continued. This event is reportable per 10CFR50.72 (b)(3)(xiii) as described in NUREG-1022, Rev. 1 since this work activity affects an emergency response facility for the duration of the evolution. The licensee notified the NRC Resident Inspector.
Maintenance to the TSC has been completed and the TSC is fully functional. The licensee notified the NRC Resident Inspector. R2DO (Bonser) notified. | |
| ENS 45529 | 1 December 2009 14:18:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Health Physics Control Point and Chem Lab Hvac Out of Service for Planned Maintenance | On 12/1/2009 the emergency HVAC (heating, ventilation, and air conditioning) for the Hatch Nuclear Plant's Health Physics (HP) Control Point/Chemistry Lab area will be removed from service for preventative maintenance, planned corrective maintenance and glycol addition to the cooling coil. The Technical Response Manual (TRM) section 3.10.1 lists the HP Control Point/Chemistry Lab area emergency HVAC system as being required for functionality of the Operations Support Center (OSC). This area is normally utilized by HP/Chem for analysis of radiological samples during normal as well as emergency conditions. These work activities are planned to be performed and completed within a 12 hour work shift. During the time these activities are being performed the HP emergency HVAC and filter train will not be available for operation for approximately eight (8) hours. As such the OSC will be rendered non-functional during the performance of this work activity. The OSC facility working area itself, however, does not lose EP functionality throughout this evolution. If an emergency were to occur, plans are to continue utilize the Health Physics Control Point/Chemistry Lab areas during the time required to restore the HVAC system, as long as environmental and radiological conditions allow. Habitability surveys will be performed in the areas while they are being utilized. If required respiratory protection equipment can be utilized as appropriate for emergency responders utilizing this area until such time as the emergency HVAC is returned to service. This event is reportable per 10CFR50.72 (b)(3)(xiii) as described in NUREG-1022, Rev. 1 since this work activity affects an emergency response facility. The licensee will notify the NRC Resident Inspector. | |
| ENS 45541 | 5 December 2009 15:59:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Hpci Out of Service Due to the Loss of the Suction Auto-Swap Function | On 12-04-09 at 16:00, condensate storage tank (CST) level switch (2E41 N002), was declared inoperable due to not actuating at the correct set point. A RAS (Required Action Statement) was entered to align high pressure coolant injection (HPCI) suction to the suppression pool within 24 hours as required by Tech Spec (Technical Specification) action statement 3.3.5.1.D 2.2. The switch was found stuck and was freed up after manual manipulation and tripped correctly during a functional test. It was decided on 12-05-09 to perform a calibration on the level switch prior to declaring it operable. During performance of calibration procedure 57CP-CAL-012-2 at 1059 EST on 12-05-09, a jumper was installed that rendered the HPCI and reactor core isolation cooling (RCIC) auto-transfer suction swap from CST to suppression pool inoperable. This is a loss of function for the initiation capability of HPCI and RCIC CST low level suction swap instrumentation. This loss of function was not discovered until 1330 on 12-05-09, at which time TS (Technical Specification) 3.3.5.1.D was entered for HPCI and TS 3.3.5.2.D for RCIC, until HPCI and RCIC suction were manually aligned to the suppression pool which allowed the plant to exit the required actions to declare HPCI and RCIC inoperable within 1 Hour. HPCI and RCIC were aligned to the suppression pool at 1341 EST. The RCIC suction was realigned to the suppression pool as required by the Tech Specs in order to restore its operability. It should be noted that no credit is taken for RCIC in the safety analysis nor is this system considered an ESF system. For this reason there are no reporting requirements associated with the inoperability of RCIC. HPCI was declared inoperable in accordance with the instrumentation Tech Specs, but during this time frame HPCI was capable of performing its safety function. However- additional review will be needed to confirm that HPCI could have operated for the duration of its mission time of 4 hours while aligned to the condensate storage tank. Absent that information this report is being made due to HPCI being declared inoperable until its suction was realigned to the suppression pool. This assumed loss of function for HPCI is being reported in accordance with 10CFR50.72(b)(3)(v)(D) since a final determination has not been made that HPCI would have continued to perform its safety function for the required mission time while aligned to the CST. The licensee notified the NRC Resident Inspector. | |
| ENS 45775 | 18 March 2010 22:16:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Tone Alert Radios | At approximately 1816 EDT, loss of prompt notification system (tone alert radios) occurred. Security notified Jacksonville National Weather Service and Information Technology to investigate. Jacksonville National Weather Service determined a problem with their equipment and prompt notification system (tone alert radios) was returned to service at approximately 1845 eastern time. Security has notified state and local agencies of prompt notification system (tone alert radio) outage and return to service. The licensee has notified the NRC Resident Inspector. Licensee also notified State and local agencies. * * * RETRACTION FROM FRANK GORLEY TO PETE SNYDER AT 1535 ON 3/25/10 * * * On 3/18/10 at approximately 11:40 am we started receiving an 'off-air' alarm for the Plant Hatch NOAA Weather Radio. Initially we determined that a loss of the prompt notification system had occurred due to the alarm and consultation with Jacksonville National Weather Service, Ref. 10 CFR 50.72 (b)(3)(xiii). NOAA personnel subsequently provided information to Plant Hatch personnel at which time it was learned that even though 'off-air' alarms were received for the NOAA Weather Radio, the radio was never off the air. While NOAA personnel were troubleshooting the system the broadcast was switched over to a digital backup system that continued to have the capability to warn the public. There was a problem noted in the broadcast that was causing several seconds of silence in the broadcast signal. While on the digital backup, the notification capability was maintained. The broadcast had been returned to the primary system today. During the time these activities were underway, broadcast capability to the public was never lost. Based on this information, Plant Hatch is entering a Notification of Retraction for the 8 hr. report documented to the NRC Event Number 45775, as entered on 3/18/10. The licensee will notify the NRC Resident Inspector. Notified R2DO (Franke). | |
| ENS 45835 | 23 February 2010 14:50:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of Standby Gas Treatment System | This report is being made under 10CFR50.73(a)(2)(iv)(B)(2). On February 23, 2010 at 1050 EST procedure 52PM-C71-001-0, RPS M/G Set System Preventative Maintenance, was being performed. During restoration of RPS Buses on Unit 1, only the Unit 1 required logic was reset. The Unit 2 logic was also required to be reset but was not. Procedure 52PM-C71-001-0, RPS M/G Set System Preventive Maintenance, did not clearly require the reset of both Unit 1 and Unit 2 logic. The procedure has been revised to make this requirement clear. Continuation of steps in the procedure required links to be closed which resulted in SBGT starting on Unit 1 and Unit 2 from Unit 2 logic. Unit 1 and Unit 2 Reactor Building ventilation isolated. This was not due to a valid signal. The automatic actuation of the standby gas treatment system (SBGT) and the isolation of Unit 1 and 2 secondary containment isolation dampers is considered an invalid actuation since the parameters that cause this actuation to occur had not been exceeded. For this reason the actuation is considered invalid and a report to the NRC is not required by 10CFR50.72(b)(3)(iv); however, because the secondary containment isolation signals affected containment isolation valves in more than one system (Unit 1 and 2 components affected) the event is reportable as required by 10CFR50.73(a)(2)(iv)(B)(2). A licensee event report (LER) is required, but can be a telephone notification as allowed by 10CFR50.73. In the case of an invalid actuation reported under 10CFR50.73(a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. The four Standby Gas Treatment (SBGT) fans auto started and both Unit 1 and Unit 2 reactor building and refueling floor normal ventilation systems automatically shutdown and isolated. The SBGT Initiation and the ventilation system shutdown were both complete actuations. The licensee notified the NRC Resident Inspector. | |
| ENS 45850 | 16 April 2010 23:17:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Cooling Accident Signal Due to High Drywell Pressure Signal | On April 16, 2010 at 1917 hrs., Unit 1 received an ECCS (emergency core cooling system) loss of cooling accident (LOCA) signal on high drywell pressure. Based on plant data, drywell pressure reached a maximum pressure of approximately 1.25 psig, which is below the LOCA and RPS (reactor protection system) signal actuation pressure of 1.85 psig. At this time, the cause of the drywell pressure increase is under investigation. RPS logic did not initiate due to drywell pressure not reaching the actuation setpoint of 1.85 psig. Although the ECCS logic prematurely actuated, the signal is being treated as 'valid' for the ECCS actuation until further investigation is completed. All expected ECCS actions occurred as a result of the signal. The LOCA logic has been reset and all affected systems have been returned to normal or standby configuration. As a result of the LOCA system actuations, several cooling tower fans tripped and condenser vacuum began to decrease. Reactor power was reduced to approximately 86 percent as a result of decreasing condenser vacuum. Power is being maintained at approximately 86 - 88 percent at this time. There are no other plant issues or concerns at this time. The licensee notified the NRC Resident Inspector. According to the licensee, normal drywell operating pressure is .5 to 1.2 psig. Prior to the event, drywell pressure had been steady at approximately 1.0 psig. Current drywell pressure is .6 psig. According to the licensee, ECCS systems that started (but did not inject) included: Core Spray pumps, Residual Heat Removal pumps, High Pressure Coolant Injection pump, and the Diesel Generators. An event review team is assessing this event to determine the root cause.
On April 16, 2010, Hatch Unit 1 received a LOCA ECCS initiation from a high drywell pressure signal. Based on the information available at that time, a notification was made to the NRC assuming the signal to be valid until further investigation could be completed. After further review, the determination has been made that the initiation signal originated from a faulted ATTS (Analog Transmitter Trip System) card and not from a valid high drywell pressure condition. Based on this information, this condition did not require an NRC notification in accordance with 10CFR50.72 and, as such, is being retracted through this updated response. The licensee notified the NRC Resident Inspector. Notified the R2DO (Hopper) | |
| ENS 45947 | 23 May 2010 20:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Nonfunctional Technical Support Center Ventilation | The Technical Support Center (TSC) ventilation system was found to be non-functional on 5/23/10 at 1630 EDT. The system repair was complete at 1747 EDT on 5/23/10. The time period that the TSC ventilation system was non-functional exceeded the 30 minute time limit that is delineated in the Technical Requirements Manual section T 3.10.1. This event is reportable per 10 CFR 50.72 (b)(3)(xiii) as described in NUREG 1022 Rev. 1 since this issue affected an emergency response facility. Also, this 8 hour notification is being made in accordance with the Technical Requirements Manual section T 3.10.1.B.2. The alternate TSC facility remained functional during this entire time period. The licensee notified the NRC Resident Inspector. | |
| ENS 45981 | 16 April 2010 23:15:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Specified System Actuation | Invalid actuation of ECCS for boiling water reactors (BWRs) including: high-pressure and low-pressure core spray systems; high pressure coolant injection system; low pressure injection function of the residual heat removal system. The following information is provided as a 60 day telephone notification to NRC under 10 CFR 50.73(a)(1) in lieu of submitting a written LER to report a condition that resulted in an invalid actuation of the 10CFR50.73(a)(2)(iv)(B) system checked above. NUREG1022 Revision 2 identifies the information that needs to be reported as discussed below. This report is being made under 10CFR50.73(a)(2)(iv)(A). On April 16, 2010, at 1717 EDT, Unit 1 received a false ECCS (Emergency Core Cooling System) Loss Of Cooling Accident (LOCA) signal on high drywell pressure. Based on plant data, all plant conditions were normal prior to event. Investigation identified an ATTS Slave Card failed in one of the Card File Racks. The Card File Rack contains ten (10) cards with two (2) being Master Trip Units (MTUs) that actuate on high drywell pressure. For initiation of an ECCS LOCA signal, both of the MTUs in the card file rack would have had to actuate. Cause of the event appears to be a momentary drop in supply voltage on the card file rack due to the failed slave card that lowered the set point of the MTUs to a point where they tripped. The set point is an internal voltage setting within the MTU. The failed ATTS cards have been replaced. Actuations: Core Spray A and B pumps started; Residual Heat Removal A, B, C, D pumps started; Diesel Generator 1A, 1B, and 1C started; Standby Plant Service Water Pump started; Plant Service Water 1A pump started (it had been in standby); Plant Service Water Valves 1P41-F310A, B, C, D closed (These are the Turbine Building isolation valves); Turbine Building Chiller B tripped; Main Control Room Environmental Control trains A and B started; HPCI Started but was secured prior to injection; Cooling Tower fans on towers A, B, C tripped. Reactor Water Level was normal at +37 inches and the highest Drywell Pressure indicated was 1.25 psig. The licensee has notified the NRC Resident Inspector. | |
| ENS 46093 | 14 July 2010 17:55:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Tsc Out of Service for Planned Maintenance | Planned preventive maintenance activities are being performed on the Hatch Nuclear Plant's Technical Support Center (TSC) Emergency Ventilation System on July 14, 2010. These maintenance activities include the performance of preventive maintenance on the TSC air handling unit and the TSC condensing unit. These work activities are planned to be completed within the (12) hour day shift on 7/14/2010. During the time these activities are being performed, the TSC air handling unit, TSC condensing unit, TSC filter train and the fan unit for the TSC filter train will not be available for operation. As such, the TSC HVAC will be rendered non-functional during the performance of this work activity. If an emergency is declared requiring activation of the TSC during the time these work activities are being performed, then the contingency plans call for utilization of the TSC, as long as habitability and radiological conditions allow. Procedure 73EP-EIP�063-0, Technical Support Center activation, provides instructions to direct TSC management to the Control Room and TSC support personnel to the Simulator Building to continue TSC activities if it is necessary to relocate from the primary TSC so that TSC functions can be continued. This event is reportable per 10CFR50.72 (b)(3)(xiii) as described in NUREG-1022, Rev. 1 since this work activity affects an emergency response facility for the duration of the evolution. The NRC Resident Inspector has been notified. | |
| ENS 46115 | 29 May 2010 12:49:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of Primary Containment Isolation Valves Due to a Faulty Transistor | This report is being made under 10CFR50.73(a)(2)(iv)(A). On May 29,2010, at 0849 EDST Unit 2 received a trip of the 'B' RPS alternate supply. This trip was due to a faulty transistor on the internal circuit board for the voltage regulator. The transistor was replaced, the voltage regulator tested and replaced in the system. The trip of the 'B' RPS alternate supply resulted in isolation of Primary Containment Isolation Valves in more than one system. The systems that were isolated due to this event were Reactor Water Clean-up and Fission Products Monitoring. Except for critical scrams, invalid actuations are not reportable by telephone under � 50.72. NUREG 1022 defines by examples given that actuation of the reactor protection system constitutes actuation of RPS full scram signal when the reactor is critical. In the condition that occurred during day shift on Saturday, May 29, 2010 only one channel of RPS actuated on failure of the Unit 2 'B' RPS alternate supply. That being the case there was not sufficient logic made up to result in a reactor scram or RPS actuation. An RPS actuation would be considered a reportable event; however, the trip of only the 'B' alternate RPS supply would not be considered an RPS actuation. This condition did involve the initiation of some containment isolation signals resulting in containment isolation valves in more than one system since RWCU and fission product monitor isolation valves closing as a result of the loss of the 'B' RPS system. Valid signals are those signals that are initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the safety function of the system. In this case the isolation signal was caused by the loss of power to the affected instrumentation rather than response to required conditions or parameters which makes this an 'invalid' isolation signal. For this event, Plant Hatch has chosen to make this telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. All actuations above were complete. No system actuations resulted from the trip of the 'B' RPS alternate supply. The licensee has notified the NRC Resident Inspector. | |
| ENS 46128 | 26 July 2010 15:27:00 | 10 CFR 26.719, FFD Reporting requirements | 24-Hour Fitness-For-Duty Report | A licensed (not active) employee supervisor had a confirmed positive during a random fitness-for-duty test. The employee's unescorted access has been suspended. Contact the Headquarters Operations Officer for additional details. The licensee informed the NRC Resident Inspector. | |
| ENS 46145 | 3 August 2010 10:50:00 | Other Unspec Reqmnt | Discovery of After-The-Fact Emergency Condition Unusual Event Due to a High Level Water Alarm | This is a one-hour report for the discovery of a condition that met an emergency action level (EAL) for a Notification of Unusual Event (NOUE) but was not determined classifiable until after conditions meeting the EAL level for the NOUE no longer existed. A detailed review of the conditions that existed at 0650 (EDT), on August 3, 2010 during a planned system drain down of the 2A loop of RHR to the northeast diagonal sump met the criteria for an NOUE: EAL HU1 - Natural and Destructive Phenomena Affecting the Protected Area. Threshold Value 6 states: 'Exceeding Max Normal Operating Values specified in EOP 31EO-EOP-014-1(2) SC - Secondary Containment Control Table 5 Secondary Containment Operation Water Levels.' At 0650 (EDT) on August 3. 2010 the Hatch Unit 2 crew received a HIGH-HIGH-HIGH alarm which is the Max Normal Level described in 31EO-EOP-014-1(2). This occurred during a planned draindown evolution when a valve (2T45-F004) which is required to transfer water from the sump to radwaste failed to open. At 0656 (EDT) on August 3, 2010 the RHR drain valve (2E11- F069A) was closed discontinuing draining into the northeast diagonal sump. It is estimated that no more than 800 gallons of water was in the system and internal flooding was not a significant concern. However, despite the low safety significance, the EAL threshold criteria was technically met and therefore this report is being made. This event is a one hour report based on the guidance in NUREG-1022 Section 3.1.1 for a condition that met an EAL and the condition for the classification no longer existed at the time of discovery. A courtesy notification will be made to state and local agencies. The licensee notified the NRC Resident Inspector. | |
| ENS 46211 | 28 August 2010 21:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of the Operations Support Center Due to Inoperable Emergency Hvac | During the performance of procedure 34S0-Z41-006-0 section 7.3.1.8, which tests the operability of the HP/CHEM area emergency HVAC, the system failed to indicate proper flow of the system. On panel 1Z41-N030, the system flow gauge failed upscale high and could not be lowered. The decision was made to halt the procedure at 1700 (EDT) on step 7.3.1.8 and restore the normal HP/CHEM HVAC lineup. The HP/CHEM area emergency HVAC is required for functionality of the Operations Support Center (OSC), as this area may be used during declared emergencies. Alternate OSC facilities have been verified to be available. All efforts and resources have been made available to restore the HP/CHEM emergency HVAC to functional status. This event is reportable per 10CFR50.72(b)(3)(xiii) and TRM 3.10.1.B.2 The licensee has notified the NRC Resident Inspector. | |
| ENS 46227 | 3 September 2010 02:50:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of Unusual Event Declared Due to High Water Level in the Rcic Instrument Sump | This is a one-hour report for the discovery of a condition that met an emergency action level (EAL) for a Notification of Unusual Event (NOUE). EAL HU1 - Natural and Destructive Phenomena Affecting the Protected Area Threshold Value 6 states: 'Exceeding Max Normal Operating Values specified in EOP 31EO-EOP-014-1(2) SC - Secondary Containment Control Table 5 Secondary Containment Operation Water Levels.' On September 2, 2010 Plant Hatch Unit 2, as part of an on-going RCIC (Reactor Core Isolation Cooling) system outage, was restoring a tagout which un-isolated the CST (Condensate Storage Tank) from the RCIC system. The shift crew received several alarms in a short period of time for RCIC Northwest Diagonal Instrument Sump levels culminating with receiving the RCIC N-W DIAG INSTRSUMP LVL HIGH HIGH HIGH alarm at 2240. There was a report of water on the floor of the diagonal and 2E51-FV005 (a vent valve on the discharge of the barometric condensate pump) was identified as the source. The shift crew closed 2E51-F010 (suction isolation valve to the CST) to isolate the CST from the RCIC system. The Shift Manager declared an NOUE at 2250 for HU-1 Natural and Destructive Phenomena Affecting the Protected Area. At 2255 the Triple High alarm cleared. Plant personnel have visually confirmed that the sump level has been pumped down to normal level. The water level in the RCIC room did not rise to a level such that any equipment was adversely affected. Plant Hatch notified the following agencies: Georgia EMA, Appling County EMA, Jeff Davis County EMA, Tattnall County EMA, and Toombs County EMA. At the time of this notification, exit criteria for exiting the NOUE had been met and Hatch personnel were in the process of officially terminating the NOUE. The licensee will notify the NRC Resident Inspector.
The licensee officially terminated the NOUE at 0013 EDT on 9/3/2010. | |
| ENS 46258 | 18 September 2010 20:27:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to High-High Reactor Building Sump Level Alarm | This is a one-hour report for the discovery of a condition that met an Emergency Activation Level (EAL) for a Notification of an Unusual Event (NOUE). EAL HU-1 - Natural and Destructive Phenomena Affecting the Protected Area Threshold Value 6 states, 'Exceed Max Normal Operating Values specified in EOP 31 EO-EOP-014-1 SC - Secondary Containment Control Table 5 Secondary Containment Operating Water Levels.' Plant Hatch Unit 1 declared a NOUE on 9/18/10 at 1627 based on HU-1 Natural and Destructive Phenomena Affecting the Protected Area due to exceeding the Max Normal Operating Value for the Reactor Building NE Diagonal Floor Drain Sump. At 1615, both the Level High and Level High-High annunciators alarmed. Water has not overflowed the sump. However, due to its design, it is not possible to immediately visually confirm the sump level. The sump pumps were not running and would not run with their switches in START indicating either: (1) there is no water in the sump, (2) both pumps are OOC (Out of Commission), or (3) there is a problem in the sump level control system. Maintenance is assessing the situation. The licensee has notified the state and local agencies and will notify the NRC Resident Inspector.
At 0400 EDT, the NOUE was terminated. The termination criteria was that the NE Diagonal Floor Drain Sump level was in its normal operating range with no abnormal sump in-leakage detected. Compensatory measures are in place to measure level in the sump. An investigation into the abnormal level indication is still in progress. The licensee will be notifying state and local authorities as well as the NRC Resident Inspector. Notified R2DO (Henson), IRD (Gott), NRR EO (Howe), DHS (Inser) and FEMA (Via).
On September 18, 2010, a notification, Event Number 46258, was submitted to the NRC to report a circumstance at Plant Hatch Unit 1 which appeared to be a 'High-High' reactor building sump level alarm. Such an alarm would meet the designated Emergency Activation Level (EAL) for a Notification of an Unusual Event (NOUE). On the day of the event, it was thought that receipt of both the 'High' and 'High-High' level annunciators for the reactor building northeast diagonal floor drain sump was indicative of the sump level having reached its maximum normal operating level, and a NOUE was made. During a subsequent Investigation, engineering personnel determined that the 'High' and 'High-High' level switches for this sump are wired such that both alarms are actuated concurrently when the 'High' level in the sump is reached at approximately 52 Inches. The 'High-High' sump level annunciation (the level at which the NOUE must be made) Is supposed to alarm at approximately 60 inches. Engineering personnel also determined that the level logic is currently improperly configured to clear both the 'High' and 'High-High' level alarms when the sump level rises to 60 inches. The initial 'High' and 'High-High' level alarms in the main control room were received at approximately 1615 EDT on September 18, 2010. At that time the sump level would have reached an actual level of approximately 51 inches (albeit with a false indication of approximately 60 inches due to the configuration error noted above) based on the setpoint at which this alarm occurs. The normal leakage into this sump at the time of this event was at a rate of approximately 1 inch per hour. The sump pumps were started at approximately 1800 EDT to pump down the sump. This action resulted in a maximum actual sump level of approximately 53.5 inches, assuming the normal leakage into the sump was occurring during this period. At no time did the sump level actually reach the 'High-High' sump level criteria of 60 inches, nor did the annunciated level clear prior to pumping down the sump which also demonstrated that the 60 inch criteria was not met. It is now apparent that the initial NOUE was a conservative report that was properly made based on the information available to the operators at the time. The subsequent investigation provided additional information regarding actual sump level conditions including information that both 'High' and 'High-High' alarms annunciate at the 'High' sump level and would have cleared had the 'High-High' level been reached. Since the 'High-High' sump maximum normal operating level of 60 inches was never reached during this event, the determination has been made that the EAL was not appropriate given the actual sump conditions, thereby making the event non-reportable and an NOUE unnecessary. Based on the preceding Information, Southern Nuclear Company (SNC) hereby provides notification that Event Number 46256 is retracted. The licensee will notify the NRC Resident Inspector. Notified R2DO(McCoy). | |
| ENS 46717 | 1 April 2011 08:04:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Area Telephone Outage Affects Emergency Notification System Phone | Baxley phone lines were discovered out of service. (This affects the) 366 and 367 (prefix) exchange. Upon further investigation, (it was) determined that Emergency Notification System phone in the Main Control Room was out of service. According to the licensee, and AT&T circuit in the Baxley Central Office was disabled, possibly due to a fiber cable cut. The licensee will be notifying the NRC Resident Inspector. |