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| Entered date | Site | Region | Reactor type | Event description | |
|---|---|---|---|---|---|
| ENS 39459 | 18 December 2002 15:22:00 | Turkey Point | NRC Region 2 | Westinghouse PWR 3-Loop | An unauthorized entry into the protected area occurred by a contract employee. Immediate compensatory measures were taken upon discovery. The licensee has notified the NRC Resident Inspector. Contact the HOO for additional details.
After further review of this event the Licensee has concluded that this event is not reportable and therefore it should be retracted. The NRC Resident Inspector will be informed. Notified Reg 2 RDO(Ogle) |
| ENS 40049 | 6 August 2003 19:12:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | On 8/06/2003 at 1630 hours, the Beaver Valley Power Station (BVPS) Unit 2 control room was notified by Engineering that the manual operator actions specified in Unit 2 post-fire procedures are not adequate for defeating a potential fire-induced spurious operation of the Power Operated Relief Valves (PORVs). The procedure step de-energizes the PORVs by opening breakers at the d-c distribution panel. This action alone would not be sufficient to prevent a cable-to-cable hot short from re-energizing the circuit since the de-energized circuit is routed in cable trays with other energized circuits in the affected fire areas. Preliminary reviews have identified the following potentially affected fire areas: CB-1 Control Building Instrumentation and Relay Area CB-2 Control Building Cable Spreading Room CB-3 Control Building Main Control Room CB-6 Control Building West Communication Room CT-1 Cable Tunnel CV-1 Cable Vault and Rod Control Area SB-1 Emergency Switchgear Room (Orange) RC-1 Reactor Containment An hourly roving fire watch patrol has been established for the affected fire areas as compensatory measures, with the exception of the Main Control Room and the Reactor Containment area, until the condition is fully evaluated and resolved. The Main Control Room is continuously manned and does not require an hourly fire watch patrol. The Reactor Containment area is not accessible during normal power operations and, as such, compensatory measures for this area will include a once-per-shift verification of remote instrumentation by operations personnel to confirm that there are no abnormal conditions or indications for this area. This condition was discovered during the review of manual operator actions for fire-induced spurious operations to confirm the safe shutdown circuit analysis is consistent with the manual actions identified in the procedures. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii) as an unanalyzed condition that significantly degraded plant safety since the failure to assure the PORVs remain in the closed position could result in the failure to meet the fire protection safe shutdown criteria. BVPS Unit 2 was licensed to the NUREG 0800 Standard Review Plan 9.5.1 "Fire Protection Program", and License Condition 2.F requires compliance to the provisions of the approved fire protection program as described in the Final Safety Analysis Report. UFSAR Section 9.5A.1.2.1.6 states that the safe shutdown capability should not be adversely affected by a fire in any plant area which results in spurious actuation of the redundant valves in any one high-low pressure interface line. Even though there is a very low likelihood of multiple shorts causing a spurious PORV actuation in an ungrounded DC circuit, the existing actions to prevent re-energizing the circuit and causing a spurious actuation of the PORVs is not consistent with the NRC guidance for high-low pressure interface valves. The NRC resident inspector has been notified.
On 8/07/03 at 1515, the following additional areas were identified as being potentially affected by this condition: CV-2 Cable vault and Rod Control Area (East) CV-3 Cable vault and Rod Control Area (Elev 755'6") SB-2 Emergency Switchgear Room (Purple) SB-3 Service Bldg. Cable Tray Area ASP Alternative Shutdown Panel Room Interim compensatory measures in the form of hourly roving watch patrols have been expanded to include the above noted areas until the condition is fully evaluated and resolved. The NRC Resident Inspector has been notified. Notified the Region 1 Duty Officer (C. Anderson)
The licensee is retracting this report based on the following: Beaver Valley Power Station (BVPS) Unit No. 2 retracts the notification made on 08/06/2003 at 19:12 hrs regarding the event reported under 10 CFR 50.72(b)(3)(ii)(B) (ENS #40049). The previous notification identified a potential unanalyzed condition that significantly degraded plant safety since the failure to assure the PORVs remain in the closed position could result in the failure to meet the Fire Protection Program safe shutdown criteria. The postulated event for spurious PORV opening due to an external cable-to-cable hot short from a fire was subsequently analyzed. It was determined that since the reactor core does not experience an unrecoverable plant condition and the radiological conditions remained bounded by analyzed DBAs, the plant's safe shutdown capability for a fire event is maintained. Although the now-credible spurious opening of a PORV is a deviation from the current Fire Protection Program criteria, the results of the spurious PORV opening does not adversely impact the plant's safe shutdown capability and hence is not a significant degradation of plant safety. Therefore this condition is not reportable pursuant to either 10 CFR 50.72(b)(3)(ii)(B) nor 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degraded plant safety. Compensatory measures in the form of shiftly fire watch patrols for the affected fire areas, with the exception of the Main Control Room and Reactor Containment area (as described in prior notifications), will remain in place until the Fire Protection Program non-conformance issue is dispositioned. The licensee informed the NRC Resident Inspector. Notified R1DO(Gray). |
| ENS 40080 | 16 August 2003 00:28:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | At 0831 hrs on 08/15/2003 at Beaver Valley Power Station, a person licensed under 10 CFR Part 55 was randomly tested for fitness for duty and found to have a blood alcohol content (BAC) level of 0.01. It was determined that this person's BAC did not exceed the limit of 0.04 BAC when it was back calculated for their time on site prior to the random test. Therefore, this was not a positive test for alcohol. However, it was the opinion of the medical review officer that the tested person was not fit to remain on scheduled duty and the person was provided transportation off site property. This is being reported pursuant to 10 CFR 26.73(a)(2)(iv). The individual's license was not in active status and therefore was not performing licensed duties on 08/15/2003. The NRC Resident Inspector was notified of this event by the licensee. |
| ENS 40084 | 18 August 2003 00:40:00 | Harris | NRC Region 2 | Westinghouse PWR 3-Loop | At 3:51 PM EDT, on August 17, 2003, with the reactor at 100% in Mode 1, the reactor was manually tripped in response to a trip of the A condensate pump and subsequent trip of the A main feed pump. Both motor-driven auxiliary feedwater pumps and the turbine driven auxiliary feedwater pump started automatically due to Lo-Lo steam generator level. The operations crew responded to the event in accordance with the applicable plant procedures. The plant was stabilized at a normal operating no-load Reactor Coolant System (RCS) temperature and pressure following the reactor trip. The condensate pump electrical supply breaker tripped due to instantaneous overcurrent, possibly related to a severe electrical storm in the area at the time. The feed pump tripped due to the loss of the condensate pump. Offsite power remained available throughout the transient. A local fire department responded to a downed power line in the vicinity of the Harris plant but the response was not related to any onsite activities. This condition is being reported as actuation of the reactor protection system and AFW in accordance with 10CFR50.72(b)(2)(iv)(B), and 10CFR50.72(b)(3)(iv)(A). 10CFR50.72 requires an 8-hour report for "Any condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation." In this case, the AFW pumps start signal was due to Lo-Lo Steam Generator Level. A root cause team is being formed to identify the cause and corrective actions Due to low decay heat in the core, the Main Steam Isolation Valves were closed and the Steam Generator PORVs are being used to maintain the plant in a Hot Standby condition. No known leaking steam generator tubes are known. Both motor-driven and the turbine driven auxiliary feedwater pumps were secured after main feedwater was returned to service. All Emergency Core Cooling Systems and the Emergency Diesel Generators are fully operable if needed. The electrical grid is stable. The NRC Resident Inspector was notified of this event by the licensee. |
| ENS 40142 | 9 September 2003 16:28:00 | Summer | NRC Region 2 | Westinghouse PWR 3-Loop | At 1530 EDT on September 8, 2003, maintenance personnel opened the 'A' train Control Room charcoal filter plenum (XAA-0029A-AH) to replace the charcoal. At 0730 EDT on September 9, 2003, a question was raised about the work being performed on the 'A' train Control Room ventilation Charcoal Filter Plenum. The Operating shift asked Engineering if the Control Room was capable of maintaining the Technical Specification required pressure under the current conditions, should the ventilation system go into the emergency mode of operation. Prior to getting the answer, the conservative decision was made to cease work on the charcoal filter plenum and restore the integrity of the 'A' train ventilation system. This was accomplished at 0805 EDT on September 9, 2003. At 0915 EDT on September 9, 2003, Engineering determined that with the isolation dampers for the 'A' train charcoal filter plenum open, due to the associated fan being tagged de-energized, the 'A' and 'B' trains of the control room ventilation system were effectively crossed. Breaching the charcoal filter plenum would have prevented the control room ventilation system from maintaining a greater than 1/8th inch positive pressure during the emergency mode of operation. Replacement of the charcoal has been placed on hold until a methodology can be formulated to allow the ventilation system to maintain the design pressure. The licensee informed the NRC resident inspector. |
| ENS 40168 | 18 September 2003 09:16:00 | Surry | NRC Region 2 | Westinghouse PWR 3-Loop | At 0900 EDT the licensee declared an Unusual Event as a precautionary measure due to expected severe weather from hurricane Isabel. Current wind speed and direction is 10-15 mph sustained with gusting to 20 mph from the NNE direction. No additional staff other than those for the planned outage have been brought onsite. The licensee informed state/local agencies and the NRC Resident Inspector.
As of 1600 ET, Surry Unit 1 commenced a power reduction of 200 mwe (100% to approximately 75% power) in anticipation of grid instabilities due to heavy weather. One of eight circ water screens is degraded due to river debris and its associated circulating water pump has been secured. On-site meteorological panels are out of service. Additionally, the Emergency Operating Facility (EOF) is out of service due to a loss of power. Notified R2DO (A. Boland).
The following was received via fax from the licensee: With Surry Power Station Unit 1 at approximately 75% reactor power and Unit 2 at 100% reactor power, a manual reactor trip was initiated on Unit 1 at 1728 and on Unit 2 at 1732, due to the loss of the '1G' and `2G' buses which supply power to all eight circulating water pumps for both units. The loss of the buses occurred during high winds and rain associated with Hurricane Isabel, although no hurricane force winds were onsite prior to or at the time of the event. A discretionary Notification of Unusual Event had been declared earlier at 0900 due to the severe weather. On Unit 1, the 'B' inadequate core cooling monitor (ICCM) failed. The 'A' ICCM remained operable throughout. All control rods fully inserted and the shutdown margin for Unit 1 was determined to be satisfactory. On Unit 2, eight rod position indicators showed greater than 10 steps but less than 20 steps following the reactor trip. Emergency boration was initiated. All Unit 2 IRPIs (Individual Rod Position Indications) are currently indicating less than five steps and the shutdown margin was satisfactory. Auxiliary feedwater automatically initiated as designed for both units on low low steam generator level following the trip. Primary RCS temperature decreased to approximately 543 and 538 degrees on Unit 1 and 2 respectively and was stabilized at 547 degrees. No primary safety or relief valves were actuated during the event on either unit. As a result of the loss of the circulating water pumps, the emergency service water pumps were started to conserve intake canal level. After start, 2 out of the 3 emergency service water pumps tripped on high water temperature. The pumps were checked and restarted 14 minutes later. Also due to the need to conserve intake canal level, the main steam trip valves on both units were closed and primary temperature is being maintained by use of the secondary power operated relief valves. No indication of primary to secondary leakage exists on either unit, therefore no adverse radiological consequences resulted from this event. Both units are currently stable at hot shutdown. At 1845, the 2G bus was energized and circulating water pumps were started. The emergency service water pumps were secured and are in standby. This 4 hour and 8 hour report is being made for Surry Units 1 and 2 in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). Notified R2DO (A. Boland), NRR EO (J. Calvo), DIRO (R. Wessman), FEMA (Chuck Bagwell), DHS (Colleen Wilson).
The licensee terminated their unusual event due to the following conditions: - severe weather has left the area - both units are stable at hot shutdown on offsite power - 2 of 3 ESW pumps are operable, and the third ESW pump is being prepared for an operational test - 45 of 67 sirens remain inoperable, but the sirens will be restored when power to the grid is restored The licensee will notify the NRC Senior Resident Inspector, as well as state and local officials Notified R2DO (Boland), NRR EO (Calvo), DIRO (McGinty), FEMA (M. Eaches), DHS (D. Licastro) |
| ENS 40170 | 18 September 2003 15:43:00 | Harris | NRC Region 2 | Westinghouse PWR 3-Loop | The following was received via fax from the licensee: As of 2:33 PM EDT, more than 20% of the offsite emergency sirens were inoperable for greater than one hour due to loss of power caused by Hurricane Isabel. Currently 27 of 81 sirens are out of service. The State of North Carolina and all four counties within the 10-mile emergency planning zone were notified and are in stand-by to implement mobile route alerting if needed. At this time, Harris cannot estimate the time of siren recovery. This requires an 8-hour non-emergency notification per 10CFR 50.72(b)(3)(xiii) due to the loss of a significant portion of the offsite notification system. The NRC Senior Resident Inspector was informed.
The licensee is updating this report to include that less than 20% of the emergency sirens are inoperable as of 2000 hours 9/18/03. The licensee informed the NRC Resident Inspector. Notified R2DO(Boland). |
| ENS 40179 | 19 September 2003 11:25:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | At 0930 on September 19, 2003 it was identified that there was a major loss of offsite response capability. 43 out of 63 early warning sirens were determined inoperable due to loss of power as a result of Hurricane Isabel. The State and NRC Resident Inspector will be notified of this event.
On September 24, 2003, at 0715 hours, a siren poll test was conducted to verify availability of the early warning sirens (EWS). The test identified that 57 of 63 (90%) sirens were available. The current availability of the EWS no longer constitutes a major loss of emergency assessment capability, offsite response capability, or offsite communication capability pursuant to 10CFR50.72(b)(3)(xiii) The NRC Resident Inspector was informed. Notified R2DO (C. Julian) |
| ENS 40181 | 19 September 2003 15:45:00 | Surry | NRC Region 2 | Westinghouse PWR 3-Loop | A poll of the Surry EWS sirens was conducted to determine operability of the siren system (Reference Hurricane Response Plan - Nuclear, Section 5.6). The poll was conducted from the back-up status logger at Innsbrook due to the unavailability of the primary status logger in the Surry LEOF (see Plant Issue - S-2003-4152). 45 of the 67 sirens expected to respond to the siren polling are considered inoperable at this time due to power failure as a result of Hurricane Isabel (the status of 2 of these was indeterminate). 22 of the 67 sirens responded satisfactorily and are considered operational. This is a reportable event per VPAP - 2802 Attachment 3 d. Early Warning System second bullet "More than 25 percent of all sirens are unavailable" and the third bullet "The capability to alert a large segment of the population does not exist. This discrepancy was discovered at 1130 EDT on 9/19/03. The licensee notified the NRC Resident Inspector, as well as state and local government officials.
Currently Surry Unit 1 is at 100% power and Unit 2 is in Cold Shutdown. Surry EWS sirens have been returned to operable status on September 29, 2003 following repairs due to weather related effects of a Hurricane Isabel. Presently 63 of 67 sirens have been returned to service (94%). All sirens located within the 5 miles portion of the EPZ are operable. The current availability of the EWS no longer constitutes a major loss of emergency assessment capability, offsite response capability, or offsite communication capability pursuant to 10 CFR50.72(b)(3)(xiii). Information related to the return to service of these sirens is being transmitted to the State of Virginia and FEMA. This is the final report that will be issued relative to siren system restoration efforts following Hurricane Isabel. Normal siren system status monitoring practices will remain in effect going forward." R2DO (Mark Lesser) notified. The NRC Resident Inspector was notified of this update by the licensee. |
| ENS 40189 | 22 September 2003 15:07:00 | Turkey Point | NRC Region 2 | Westinghouse PWR 3-Loop | A licensed employee was determined to be under the influence of alcohol upon entry into the plant access control point. Access to the facility was denied and the employee's access to the facility has been suspended. Contact the Headquarters Operations Officer for additional details. The licensee has notified the NRC Resident Inspector. |
| ENS 40243 | 13 October 2003 17:53:00 | Turkey Point | NRC Region 2 | Westinghouse PWR 3-Loop | The control room was notified by Land Utilization at 1700 EDT that they had contacted the Florida Fish and Wildlife for information on where to take the injured crocodile found on the plant access road. The crocodile was taken to the Miami Zoo. The NRC Resident Inspector was notified along with the State |
| ENS 40247 | 15 October 2003 01:56:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | Automatic Trip due to low (21%) B Steam Generator Level (all rods fully inserted into the core). While decreasing power for heater drain pump repairs, the B main feedwater regulating valve did not respond properly. Manual attempts to restore level were not effective. Steam generator level decreased to the low level trip setpoint. Auxiliary feedwater systems automatically started due to low post trip steam generator levels. This includes the Steam Driven Auxiliary Feedwater Pump. The steam discharge of this pump is considered to a gaseous release due to trace amounts of tritium in the secondary system. These trace amounts are well below limits. The pump was shutdown after 16 minutes of operation, terminating the discharge. The Rad Waste & Effluents Section will assess the impact during the event response review. Plant is currently stable in Mode 3 with steam generator levels restored to post shutdown values. The steam dumps and main unit condenser are available for heat removal. All normal and emergency busses are energized from offsite power. The NRC Resident Inspector was notified of the event by the licensee. |
| ENS 40249 | 15 October 2003 14:07:00 | Summer | NRC Region 2 | Westinghouse PWR 3-Loop | Compensatory measures taken upon discovery. The NRC Resident Inspector was notified. Refer to the HOO for additional details |
| ENS 40261 | 20 October 2003 13:21:00 | Summer | NRC Region 2 | Westinghouse PWR 3-Loop | On October 19, 2003, during performance of surveillance test procedure, STP 250.001B, Boric Acid Inspection Program, boron was observed around the seal injection piping and on the pump case housing of reactor coolant pump (RCP) 'C.' Inspection personnel contacted the Control Room and Outage Management at 1350 hours. Evaluation of the condition was initiated and a dye penetrant examination (PT) was scheduled for the morning of October 20. Evaluation of the PT results determined at 0700 hours on October 20 that the seal injection piping at the RCP pump case housing had through-wall leakage. This condition is being reported in accordance with 10CFR50.72(b)(3)(ii)(A)(2). The licensee performed a similar inspection in April 2002 and found no indication of leakage at that time. Inspections of RCPs 'A' and 'B' have been completed during this outage with no indication of leakage observed. The licensee has corrective actions under review which will include additional UT inspections. The licensee informed the NRC Resident Inspector. |
| ENS 40292 | 3 November 2003 17:36:00 | Surry | NRC Region 2 | Westinghouse PWR 3-Loop | The design data used to support Dominion's methodology to cope with a loss of (RCP) seal cooling (which was based on Westinghouse WCAP-10541 Rev. 2) may be non-conservative. The NRC issued an SER on Westinghouse report WCAP-15603 that indicates data from this new model must consider a 20% probability that the hot RCP seals may result in up to 182 gpm leakage per pump if seal cooling is lost for more than 13 minutes. Assuming this higher leak rate, the requirements of Appendix R, Section III.L, cannot be met since Pressurizer level could not be maintained. Current methodology indicates Surry will lose charging and seal cooling for greater than 13 minutes from a fire in the Emergency Switchgear Room. This report is being made pursuant to 10 CFR 50.72 (b)(3)(ii)(B). The NRC resident (inspector) was notified of this condition. |
| ENS 40302 | 7 November 2003 14:57:00 | Surry | NRC Region 2 | Westinghouse PWR 3-Loop | At 0830 hours on 11/07/03, a phosphoric acid/sodium hydroxide/potassium hydroxide diluted chemical mix overflowed from a holding tank, which resulted in a discharge of the chemicals into state waters through the plants storm drain system. This discharge was a result of two un-isolated filter vents that allowed the chemical mixture to overflow to the ground. It is estimated that approximately 70 gallons of the chemical mixture entered the plants discharge canal, which enters state waters, before it was isolated. The chemical mixture is considered non-hazardous. There were no radiological consequences associated with the discharge of the chemical mixture. Cleanup efforts are in progress. This event is being reported in accordance with 10CFR50.72(b)(2)(xi) due to the fact that the Virginia Department of Environmental Quality was notified at 1330 hours EDT of the discharge into state waters. The NRC resident was notified of this condition. |
| ENS 40309 | 10 November 2003 12:41:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | Reactor Protection System actuation in response to an indicated (not an actual) 2A RCP (Reactor Coolant Pump) breaker open position signal. All reactor protection and support systems operated as expected. The Aux Feedwater System started as required in response to the tripping of both Steam Generator Feed Pumps. All 3 RCPs are running; none have tripped. Not understood is the indication of the 2A RCP breaker open when the breaker has remained closed. The licensee reported that all control rods fully inserted; decay heat is being rejected to the condenser via the steam dumps; a steam generator atmospheric relief may have momentarily lifted during the transient; and that the electrical grid is stable. The licensee will be notifying the NRC Resident Inspector. |
| ENS 40314 | 11 November 2003 17:11:00 | Summer | NRC Region 2 | Westinghouse PWR 3-Loop | On November 10, 2003, at 1740 hours it was determined that a Fitness-For-Duty program condition existed which required a 24-hour report to the NRC. Personnel with unescorted access are placed into different groups within the Fitness-For-Duty program. One of these groups, the Immune group, is for personnel not eligible for random test selection at our site, such as the NRC. The remaining groups become part of the random eligible population. During a review of the Fitness-For-Duty program it was discovered that the total number of personnel in the Immune population group was higher than normally expected (133 vs. 24). Some personnel required to be in the random eligible population were discovered to be in the Immune group, which rendered them ineligible for selection. (This primarily affected personnel transferring from an approved program within 30 days.). This resulted in non-compliance with 10 CFR 26.24. Security personnel are manually correcting the group listings, identifying the affected individuals, scheduling individuals still on site for testing, and notifying the plants where individuals potentially affected may have gone. A root cause investigation is presently being conducted to identify the extent, time frame, and cause of this event. If relevant information is uncovered as a part of the root cause investigation that affects the reportability determination, the NRC will be notified. This condition is being reported in accordance with 10CFR26.73. The licensee has notified the NRC Resident Inspector
Based on subsequent research, the licensee determined that the above event is not reportable as a significant event under 10 CFR 26.73. The licensee will notify the NRC Resident Inspector. Notified R2DO (Bonser) |
| ENS 40320 | 13 November 2003 13:48:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | Unit 1 was at 100% power when an automatic reactor trip occurred. The Auxiliary Feedwater systems automatically started due to low post trip steam generator levels. This includes the steam driven Auxiliary feedwater pump. The steam discharge of this pump is considered to (be) a gaseous release due to trace amounts of tritium in the secondary system. These trace amounts are well below limits. The pump was shutdown after 31 minutes of operation, terminating the discharge. The Rad Waste & Effluents Section will assess the impact during the event response review. The plant is currently stable in mode 3 with steam generator levels restored to post shutdown values. The steam dumps and main unit condenser are available for heat removal. All normal and emergency busses are energized from offsite power. All rods fully inserted, no ECCS actuation and no primary or secondary relief valves lifted. The NRC Resident Inspector was notified. |
| ENS 40328 | 18 November 2003 10:03:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | This is notification of a Special Report in accordance with Technical Requirement 7.1.1 of the North Anna Technical Requirements Manual. On November 17, 2003, at 1150 hours, the Main Control Room observed the start of all fire pumps and receipt of associated alarms. An investigation determined that a significant leak in the fire protection header occurred outside the protected area. The cause for the fire protection header leak has not been determined. In an effort to isolate the leak, two of the three fire pumps had to be shut down and the third pump was isolated. The leaking portion of the fire protection header was isolated by 1313 hours. The running fire pump was then used to refill and pressurize the fire protection header. By 1420 hours, the fire protection system had been returned to normal. The licensee notified the NRC Resident Inspector. |
| ENS 40340 | 19 November 2003 14:23:00 | Robinson | NRC Region 2 | Westinghouse PWR 3-Loop | At approximately 1330 hours (EDT) on November 19, 2003, during reviews being conducted on H. B. Robinson Steam Electric Plant (HBRSEP), Unit 2 No. 2, Appendix R fire scenario safe shutdown circuit analysis, it was concluded that two failure modes not previously postulated could be introduced by a fire in either Fire Zone 19, Cable Spreading Room, or Fire Zone 20, Emergency Switchgear Room. The suction source to the charging pump could spuriously close during a fire in one of these zones, which could cause a loss of Reactor Coolant System (RCS) make-up capability of the "A" Charging Pump were operating at the time if the event. The other failure mode involves spurious operation of both pressurizer power operated relief valves (PORVs), which could cause a loss of RCS inventory. Each of these failure modes could result in loss of safe shutdown capability. Therefore, this condition is being reported in accordance with 10 CFR 50.72(b)(3)(v)(A), any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are required to shut down the reactor and maintain it in safe shutdown condition. Short term corrective actions for this situation include procedure changes to establish preemptive actions that will provide reasonable assurance that safe shutdown can be achieved and maintained. The procedure changes will include rapid verification at the onset of a fire scenario that the "A" Charging Pump is not operating, and will direct closure of the pressurizer PORV block valves. Additionally, the Fire Protection Program (FPP) requirements for loss of detection or suppression in these zones were reviewed. The FPP requires establishment of a continuous fire watch in these areas for loss of detection or suppression capability. Additional administrative controls will be implemented or transient combustibles to further reduce the probability of these failure modes occurring. Long term corrective actions are being developed via the Corrective Action Program and will be implemented to further reduce the likelihood of these failure modes The NRC Resident Inspector was notified of this event by the licensee. |
| ENS 40348 | 21 November 2003 15:46:00 | Summer | NRC Region 2 | Westinghouse PWR 3-Loop | On Friday November 21, 2003, the V.C. Summer Nuclear Station was performing rod position testing in preparation for plant start-up, following refueling outage-14. The testing is in accordance with STP-106.002 Rod Position Indication Operational Test. Control Bank 'C' was being withdrawn. Per STP-106.002, RCS boron concentration was verified adequate to ensure Keff less than or equal to 0.95. When Control Bank 'C' reached 36 steps Digital Rod Position Indication system (DRPI) indication for rod M-4 went to 18 steps. At 0835, Control Rod motion was stopped and Abnormal Operating Procedure (AOP)-403.5, Stuck or Misaligned Control Rod, was entered after reviewing the AOP applicability versus current plant condition. At 0910 while taking action per the AOP, it was determined that both channels of DRPI were not functioning properly and per Tech Spec 3.10.5 the Rx Trip Breakers were immediately opened (manual actuation of the Reactor Protection System - reactor scram). Emergency Operating Procedure (EOP)-1.0, Reactor Trip, was entered. At 0915 EOP-1.0, was exited with the plant stable in Mode 3 (Hot Standby). At this time it appears there was failure in the DRPI system. This event is being reported in accordance with 10 CFR 50.72(b)(3)(IV)(A), a valid actuation of the Reactor Protection System (RPS). The NRC Resident Inspector was notified of this event by the licensee. |
| ENS 40364 | 1 December 2003 14:37:00 | Summer | NRC Region 2 | Westinghouse PWR 3-Loop | Failure of Master Relay K507 (MidTex Part Number 15614D200) in the Solid State Protection System (SSPS) was attributed to corrosion caused by off-gassing of chlorine contamination from paper used to wrap the relay coil, or the adhesive used to attach the paper wrap to the coil. Master Relays must change state, upon energization, to actuate various safety related equipment in the plant. The failure of Master Relay K507 is considered to be a manufacturing defect in a basic component that could cause a loss of safety function. The majority of MidTex relays investigated had cloth wrapped coils and these relays showed no evidence of corrosion. The number of affected relays and installed locations was not available at the time of the initial notification. The licensee will notify the NRC Resident Inspector. |
| ENS 40379 | 9 December 2003 17:12:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | At 1559 EST on 12/09/03, Seismic activity was felt at North Anna Power Station (NAPS). Both units continue to operate at 100 percent power with no problems. Seismic monitoring systems did not pick up the event. The earthquake was centered 15 miles Southeast of Colombia, VA (29 miles from NAPS) and registered 4.5 on the Richter Scale. Initial plant walk downs have identified no issues. Operations and engineering are continuing to walk down the units to identify problems that are undetectable from the control room. The licensee notified the NRC Resident Inspector. |
| ENS 40389 | 12 December 2003 14:33:00 | Surry | NRC Region 2 | Westinghouse PWR 3-Loop | Surry Power Station has determined that an unanalyzed condition related to Auxiliary Feedwater (AFW) isolation during a steam generator tube rupture (SGTR) event exists. The existing accident analysis for a SGTR requires that AFW to the ruptured steam generator (SG) be isolated prior to the water level in that SG rising into the main steam pipe. Failing to take these actions could result in a radioactive release or failure of equipment important to safety, which has not been previously analyzed. The current configuration of the AFW system has six motor operated valves (MOVs), two for each SG, which are used to control flow from one of two AFW headers. Three MOVs are powered from "H" emergency bus and three MOVs are powered from "J" emergency bus. The MOVs are maintained normally open. With a loss of emergency power to either train of AFW MOVs, the control room operators are not able to close the three MOVs from the de-energized emergency bus without manual action inside containment. A review of the plant safety analysis design basis indicates that with this AFW configuration, isolation of AFW to a ruptured SG would not be possible within the time frame specified in the analysis. Unit One and Unit Two entered a TS 3.0.1 clock at 1119 hours on December 12, 2003 due to the AFW systems being declared inoperable. The station took immediate actions to address the configuration control issue in order to meet the assumptions made in the SGTR accident analysis and both Unit One and Unit Two exited the TS 3.0.1 clock at 13:41 hours on December 12, 2003. The NRC Resident Inspector was notified of this event by the licensee. |
| ENS 40434 | 10 January 2004 04:40:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | During a surveillance of the Beaver Valley Power Station Unit No. 1 Solid State Protection System (SSPS) train "B" on 01/09/2004, a master relay contact exhibited unacceptable high resistance and was declared inoperable at 2042 hours. Corrective action was initiated to replace the relay, which requires that the "B" train of the reactor trip system be transferred from the reactor trip breaker to the bypass reactor trip breaker and the "B" train of SSPS be de-energized. Action 40.b of Technical Specification 3.3.1.1, Table 3.3-1, Item 21 requires the Unit to be in at least Hot Standby within 6 hours with one reactor trip breaker inoperable as a result of something other than an inoperable diverse trip feature. Bypassing the reactor trip breaker in order to replace the SSPS master relay makes the reactor trip breaker inoperable as a result of something other than an inoperable diverse trip feature. The 'B' train Reactor Trip System reactor trip breaker was bypassed at 00:28 hours on 01/10/2004. Beaver Valley Power Station Unit No. 1 initiated a plant shutdown at 01:27 hours on 01/10/2004 in response to entering Action 40.b of Technical Specification 3.3.1.1, Table 3.3-1, Item 21. Action to replace the SSPS master relay was completed and the reactor trip breaker was returned to operability at 02:55 hours and the plant shutdown was terminated at 72% power. The Unit is scheduled to return to full power operation. Source Range N-31 re-energized when source range high voltage switch was restored to "Normal" position. Investigation is currently underway to determine the cause of this. The licensee notified the NRC Resident Inspector. |
| ENS 40443 | 13 January 2004 17:42:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | General FTS Failure recognized by abnormal ring (long drawn out ring) coming from ENS. System operation verification performed after ring noted. At this time it was realized all lines had no dial tones and were inoperable. In addition to the ENS other FTS phones/lines inoperable noted were the MCL, HPN, PMCL, RSCL, and possibly the LAN line, ERDS 1 and 2. (ERDS, and LAN line not operability tested). ENS Failure- Discovered at 12:55 CST it was inoperable and reportable per FNP-0-EIP-8.0 (non emergency report) para. 12.6.2 at 13:55 CST. Plant Commercial lines not effected. The licensee notified the NRC Resident Inspector |
| ENS 40477 | 26 January 2004 03:50:00 | Robinson | NRC Region 2 | Westinghouse PWR 3-Loop | At approximately 22:36 hours EST, on January 25, 2004, the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, Emergency Response Facility Information System (ERFIS) computer system became inoperable due to a loss of the uninterruptible power supply (UPS). The ERFIS computer system provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), Meteorological Data link system, FAX modems for Emergency Preparedness functions, and the Inadequate Core Cooling Monitor (ICCM). The loss of ERFIS requires alternate methods as described in plant procedures to be used for these functions, as necessary. Therefore, it is expected that appropriate assessment of plant conditions, notifications, and communications could still have been made as necessary, if required, during the time that the ERFIS was inoperable. The ERFIS computer system was restored at approximately 01:47 hours EST on January 26, 2004, using a back-up power supply. The cause of the loss of the ERFIS UPS has not yet been determined. Plant personnel are continuing to investigate the loss of the power supply and are in the process of repairing the UPS. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. As previously stated, alternate means remained available to assess plant conditions, make notifications, and accomplish required communications, as necessary. The Licensee notified the NRC Resident Inspector. |
| ENS 40483 | 27 January 2004 13:33:00 | Summer | NRC Region 2 | Westinghouse PWR 3-Loop | The licensee reported intermittent problems with the ENS telephone and Early Warning Siren System (EWSS) capability. EWSS capability is less than 75% and is currently 58%. The ENS phone system is repaired. Ice storms are the cause of the communication problems. The licensee notified the NRC Resident Inspector.
The licensee is updating this report to indicate that greater than 75% of the EWSS capability has been restored. The licensee informed the NRC Resident Inspector. Notified R2DO(Kuzo). |
| ENS 40497 | 2 February 2004 16:41:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | A licensed operator was determined to be under the influence of alcohol during a random test. However, it was below the 10CFR26 cut-off level but could have resulted in the individual being unfit for scheduled work activities. The employee's access to the plant has been put on fitness-for-duty hold pending resolution. Contact the HOO for additional details. The licensee notified the NRC Resident Inspector. |
| ENS 40534 | 21 February 2004 02:04:00 | Robinson | NRC Region 2 | Westinghouse PWR 3-Loop | At 2220 hours EST, on February 20, 2004, the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, Emergency Response Facility Information System (ERFIS) computer system became inoperable. The ERFIS computer system provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), Meteorological Data link system, FAX modems for Emergency Preparedness functions, and the Inadequate Core Cooling Monitor (ICCM). Actions were completed to restore the ERFIS computer system to an operable status at 0053 hours on February 21, 2004. The loss of ERFIS requires alternate methods, as described in plant procedures, to be used for the above described functions. Therefore, it is expected that appropriate assessment of plant conditions, notifications, and communications could still have been made, if required, during the time that the ERFIS computer system was inoperable. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. As previously stated, alternate means remained available to assess plant conditions, make notifications, and accomplish required communications, as necessary. The licensee notified the NRC Resident Inspector. |
| ENS 40558 | 1 March 2004 07:23:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | 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 40590 | 15 March 2004 18:40:00 | Turkey Point | NRC Region 2 | Westinghouse PWR 3-Loop | At 1330 ET on 3/15/04, it was determined that a contract supervisor had tested positive for drugs on an access reinstatement urinalysis test. Access has been denied. The individual had only been badged for one day. A work review was conducted and no safety-related activities had been performed. Contact HOO for additional details. The licensee has notified the NRC Resident Inspector. |
| ENS 40617 | 28 March 2004 15:00:00 | Robinson | NRC Region 2 | Westinghouse PWR 3-Loop | At approximately 1343 hours on March 27, 2004, during a routine hand - rotation check of the "C" high pressure safety injection pump (HPSI), it was discovered the pump shaft was exhibiting some binding. At the time, the "C" HPSI pump was in service as the "B" Train HPSI pump. The "C" HPSI pump was declared inoperable and Condition A of LCO 3.5.2 was entered. This LCO condition requires restoration of the inoperable safety injection train within 72 hours. At that time, the "B" HPSI pump was in service as the Train "A" HPSI pump. The "A" HPSI pump was out of service due to a previously discovered condition of minor leakage observed near two of the casing bolts. The H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, HPSI system has three safety injection pumps. The "A" HPSI pump is the normal Train "A" pump, the "C" pump is the normal Train "B" pump, and the "B" pump is capable of serving as either the Train "A" or the Train "B" pump in the place of the "A" or "C" HPSI pump. At the time of discovery of the binding in the "C" HPSI pump, the "B" HPSI pump was in service as the Train "A" HPSI pump. It was determined that the "A" pump could be restored to operable status and placed back in service to restore two Trains of HPSI. In order to do so, the "B" HPSI pump was removed from service as the Train "A" pump and placed in service as the Train "B" pump. Therefore, at 1026 hours on March 28, 2004, for approximately 25 minutes, during the process of placing the "B" HPSI pump in service on Train "B," which was necessary to allow the "A" HPSI pump to be returned to service, there was no HPSI pump automatically available to provide HPSI to the Reactor Coolant System, if an accident were to occur. Therefore, this condition is reportable in accordance with 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function of systems needed to mitigate the consequences of an accident. An additional 8-hour reporting criterion associated with the plant being in an unanalyzed condition that significantly degrades plant safety 10 CFR 50.72(b)(3)(ii) has also been identified due to the inoperability of the HPSI system. It was known prior to the switching the "B" HPSI pump to Train "B" that this would cause inoperability of both trains of HPSI. It was also known that LCO 3.0.3 would be entered due to this circumstance. This situation was not avoidable, based on the sequence of events. The time in this condition was minimized, the operators were fully aware of plant conditions, and no other system inoperabilities were known of that would have complicated the situation. Both trains of HPSI are operable, although the "C" HPSI pump remains inoperable and out-of-service, pending investigation and repair. The licensee notified the NRC Resident Inspector |
| ENS 40628 | 30 March 2004 17:38:00 | Summer | NRC Region 2 | Westinghouse PWR 3-Loop | 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 40637 | 3 April 2004 13:17:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | Notification made to State Department of Environmental Quality (DEQ) for a chemical spill of Calgon H-901-g going into onsite waters." Approximately 500 gallons of chemicals containing Bromine used to treat service water spilled onsite. A portion of the spill entered the drainage system and was released offsite. The licensee attributes the spill to a mechanical failure on their brominator system and is in the process of cleaning up the material onsite. The licensee informed the NRC Resident Inspector. |
| ENS 40666 | 11 April 2004 12:47:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | 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 05:25:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | 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 40688 | 20 April 2004 13:56:00 | Robinson | NRC Region 2 | Westinghouse PWR 3-Loop | On April 20, 2004, a non-work related fatality occurred at H. B. Robinson. At 0730 EDT, a Progress Energy employee assigned to the main turbine maintenance crew for the current maintenance outage that began on April 20, suffered a condition that required immediate medical attention. Medical assistance was provided by onsite plant employees trained as first responders. The employee was transported by ambulance to a nearby medical facility where additional medical treatment was rendered. At 1310, site personnel were notified that attempts to revive the employee were not successful and that the employee had died. The employee was working in a non-radiological controlled area of the plant and no radioactive material or contamination was involved. Notification of the South Carolina Occupational Safety and Health Agency is planned. The Licensee notified the NRC Resident Inspector. |
| ENS 40698 | 22 April 2004 13:48:00 | Surry | NRC Region 2 | Westinghouse PWR 3-Loop | At 1045 hours on 04/22/04, the Safety Parameter Display System (SPDS) portion of the Emergency Response Facility Computer System (ERFCS) was noted to be inoperable due to all outputs being displayed in "magenta" color. Attempts were made to reboot the system, but were unsuccessful. At 1145 hours on 4/22/04, SPDS was returned to service. Therefore, the SPDS had been out of service for 1 hour, which is considered a major loss of emergency assessment capability. This report is being made in accordance with 10CFR50.72(b)(3)(xiii). SPDS is not 100% operable, but it does meet the licensee's operability criteria to be declared operable. The licensee is analyzing the problem. The licensee notified the NRC Resident Inspector. |
| ENS 40730 | 6 May 2004 16:25:00 | Harris | NRC Region 2 | Westinghouse PWR 3-Loop | 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 40739 | 10 May 2004 00:11:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | 2-EE-EG-2H was out of service for scheduled maintenance. During an inspection, 4 shims on the exhaust stack seismic support on the south wall were found missing. The shims provide baseplate bearing on the concrete wall near each anchor bolt and had fallen to the floor. In addition, both lower anchor bolt nuts had worked loose and had fallen to the floor and one of the upper anchor bolt nuts had worked loose and was partially disengaged. Following this discovery 2-EE-EG-2J was inspected and 3 of the 4 supports were not properly fastened. 2J EDG was declared inoperable based on this discovery. Due to both U-2 Emergency Diesel Generators being in operable the safety functions needed to maintain the reactor in a safe shutdown condition and remove residual heat would have been imparted in the event of a loss of off site power 2-EE-EG-2J was repaired and returned to operable at 2313 EST on May 09, 2004. The NRC Resident Inspector was notified of this event by the licensee. |
| ENS 40752 | 14 May 2004 18:26:00 | Turkey Point | NRC Region 2 | Westinghouse PWR 3-Loop | 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 40768 | 21 May 2004 21:36:00 | Surry | NRC Region 2 | Westinghouse PWR 3-Loop | At 2108 hours a main generator lead failed (exploded) causing a generator lead differential lockout/turbine trip/reactor trip. All rods fully inserted into the core. Both Motor Driven and the Turbine Driven Auxiliary Feedwater Pumps started as expected and are supplying water to the steam generators. Electrical power is being supplied from offsite and all systems are operating as expected. The reactor tripped at 2108 hours and a report of a fire in the switchyard was reported to Control Room at 2109 hours. Fire Brigade was dispatched and the fire was reported out at 2142 hours. Reflash fire watch has been set. All Emergency Core Cooling Systems and the Emergency Diesel Generators are fully operable if needed. Surry Unit 1 remains at 100% power. Licensee notified State and Local officials of their declaration of an Unusual Event. The NRC Resident Inspector was notified of the event by the licensee.
The licensee terminated the NOUE at 2256 EDT. Damage appears to be confined to the main generator leads with one lead missing and some damage to other leads. The plant is stable with RCS temperature at 547 degrees Fahrenheit, 2235 psia. The NRC Resident Inspector will be notified of the NOUE termination by the licensee. Notified R2DO (Jay Henson), NRR EO (Edwin Hackett), DIRO (Tom Andrews), FEMA (David Barden), DHS (Robert Bozzo). See EN 40769 , 40770 & 40771 for related Non Emergency notifications. |
| ENS 40769 | 22 May 2004 00:28:00 | Surry | NRC Region 2 | Westinghouse PWR 3-Loop | 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 40770 | 22 May 2004 15:06:00 | Surry | NRC Region 2 | Westinghouse PWR 3-Loop | At 2146 hours on 05/21/04, Unit 2 entered a 30 Hr. clock to Cold Shutdown (CSD) in accordance with T.S.3.0.1 due to Auxiliary Feedwater (AFW) not being in compliance for Continued Operation (JCO) SC 03-002 configuration. Auxiliary Feedwater was declared inoperable at 2146 hours. During post trip recovery, Emergency Procedure, 2-ES-0.1, directed that the six (6) AFW motor operated valves (MOVs) be checked opened after securing the AFW Pumps. This is contrary to the AFW JCO, which requires four (4) MOVs out of six (6) to be closed for AFW to remain Operable. At 0040 hours on 5/22/04 the correct valve alignment in accordance with the JCO was established and the 30 hour TS 3.0.1 clock was exited. This event is reportable in accordance with 10CFR50.72(b)(3)(ii)(B); any event or condition that results in the Station being in an unanalyzed condition that significantly degrades plant safety. The NRC resident has been notified. The licensee found out about this event today during event investigation. See Surry Unit 2 events 40769, 40768 and 40771.
The following information was received from the licensee via facsimile: At 1506 hours on 5/22/04, Surry Power Station made an 8-hour Non-Emergency Notification in accordance with 10CFR50.72 (b)(3)(ii)(B) for an unanalyzed condition due to Auxiliary Feedwater (AFW) not being aligned in accordance with Justification for Continued Operation (JCO) SC 03-002 for 2 hours and 54 minutes following a reactor trip. (Event Number 40770). Upon further analysis, this event was determined not to be immediately reportable. Based on calculations performed by Dominion Nuclear Safety Analysis in Engineering Transmittal ET NAF-04-0045, Rev 0, a risk-based period of time was calculated to allow operations personnel a defined period of time to realign the AFW system in accordance with this JCO. Operators must not exceed 150 hours to restore the acceptable JCO configuration during all modes where AFW is required to be operable. Since the amount of time the system was not in the JCO alignment was less than 150 hours, plant risk increase was maintained less than the limits specified in Regulatory Guide 1.177. This notification is being made to retract the report made on 5/22/04 based on the above discussion. The NRC resident has been notified. The NRC Headquarters Operations Officer notified R2DO (Ayres). |
| ENS 40771 | 22 May 2004 22:29:00 | Surry | NRC Region 2 | Westinghouse PWR 3-Loop | There is sufficient reason to believe that the buried auxiliary feedwater piping associated with Unit 2 is degraded and leaking and thus as a conservative measure it is declared inoperable. This operability determination is based on not having a reasonable assurance that the auxiliary feedwater piping can continue to perform its intended function under all design basis events including seismic events. Approximate leakrate is 8 gpm. Makeup is sufficient at this time to maintain desired level in the affected tank. Further efforts to determine the exact leak location are ongoing. This notification is being made pursuant to 10 CFR 50.72(b)(3)(v)(D) for the 8-hour notification of a condition that could have prevented the fulfillment of the safety function of systems or structures that are needed to mitigate the consequences of an accident. Emergency Condensate Storage Tank level was decreasing and water has been detected coming from a conduit in the Safeguard Building. Licensee is in Tech Spec 3.0.1 (30 hours to get below 350 degrees F and 450 psig). The NRC Resident Inspector was notified of this event by the licensee. See EN #'s 40770, 40769 & 40768 for Surry Unit 2 related problems. |
| ENS 40772 | 23 May 2004 06:21:00 | Surry | NRC Region 2 | Westinghouse PWR 3-Loop | Based on the inoperability of Unit 2 Auxiliary Feedwater (reference EN # 40771) due to an underground leak, Unit 2 has initiated a plant cooldown from Hot Shutdown (547F) to Cold Shutdown (<200F). This notification is being made pursuant to 10 CFR 50.72(b)(2)(i) for the 4-hour notification of a Technical Specification Required Shutdown. This event notification is required per licensee's Administrative procedures. The licensee has notified the NRC Resident Inspector. |
| ENS 40784 | 29 May 2004 09:55:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | During performance of Hot Rod Drop Testing (and) when withdrawing 'D' Control banks, a failure of Group 1 Position Indication was identified. Entered action of Technical Requirement Manual (TRM) 3.1.3 and opened the Reactor Trip Breakers within 15 minutes per action (statement) of TRM 3.1.3. After the Reactor Trip Breakers were opened all Group 1 "D" Control Rods fully inserted into the core. There were no reactivity concerns since the reactor was borated with adequate shutdown margin. The failure of the position indicator has been identified and repaired. There were no other issues associated with this incident and the licensee will proceed with Hot Rod Drop Testing while at 0% reactor power and Mode 3. The licensee will notify the NRC Resident Inspector. |
| ENS 40804 | 10 June 2004 16:11:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | 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). |