05000382/LER-2005-005
| Docket Numbersequential Revmonth Day Year Year Month Day Year Nanumber 'No. 05000 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3822005005R00 - NRC Website | |
condenser vacuum caused by a loss of all Circulating Water Pumps [KE]. Subsequently Emergency Feedwater Actuation Signals EFAS-1 and EFAS-2 were received due to low Steam Generator levels actuating the Emergency Feedwater System [BA]. The condition was reported to the NRC Operations Center within eight hours. This notification was followed up with a corrected notification on November 14, 2005 after it was subsequently confirmed that the condition was reportable under both four and eight hour reporting criteria. The event was reportable within four hours under criteria 10CFR50.72(b)(2)(iv)(B) for a manual trip of the plant in anticipation of receiving a Reactor Protection System (RPS) [JC] trip with the Reactor critical. The condition was also reportable within eight hours in accordance with 10CFR 50.72(0(2)(iv)(A) for the automatic actuation of EFAS upon low Steam Generator levels. The failure to report within four hours was entered in Waterford 3's Corrective Action Program (CR-WF3-2005-04599). The reactor manual trip is reportable in writing (Licensee Event Report) within 60 days in accordance with 10CFR50.73(a)(2)(iv)(A) due to the manual actuation of the RPS and due to the automatic actuation of the Emergency Feedwater system.
INITIAL CONDITIONS
Just prior to the initiating events, the plant was operating in Mode 1 at 100% power. There were no procedures being implemented specific to this condition. There were no Technical Specification Limiting Conditions of Operation specific to this condition in effect. Three of the four Circulating Water Pumps were running.
EVENT DESCRIPTION
The CW System provides cooling water to the Main Condenser to condense steam exhausted from the main turbine, the feedwater turbines [SJ], and condensate drain sources. The CW System also provides cooling to the Reactor Coolant System [AB] by cooling steam dumped to the Condenser through the bypass valves during plant startup and shutdown. The CW System contains four large pumps. Plant design allows for a minimum of 3 pumps operating during winter and four pumps operating during summer to obtain 100% power based on river temperature.
On November 11, 2005 at 20:34, while operating at 100% power with three of the four Circulating Water (CW) Pumps running, the "A" CW Pump discharge valve began to close and at 90% closure the pump tripped per design. Within 45 seconds, the "B" CW Pump tripped due to automatic discharge valve closure. At this time the Control Room Supervisor (CRS) directed removing 200 MWs of load from the Main Turbine [TA] at 20 MW per minute. While the Main Turbine was being adjusted for load removal, CW Pump "D" also tripped due to automatic discharge valve closure. The CRS directed a manual reactor trip due to no CW Pumps running and rapidly lowering Condenser vacuum. The reactor was manually tripped at 20:34 and Operations procedure OP-902-000, "Standard Post Trip Actions", was entered. Emergency Feedwater actuation signals EFAS-1 and EFAS-2 initiated due to low Steam Generator levels. The Operations crew entered Operations � Control System [JI] was unavailable, bf design, dice to a condenser interlock. As a result of the full load rejection, steam generatorpressure rose to the Atmopheric Dump Valve setpoint. During the initial pressure rise, the steam ge6erator wide range jand narrow range level indications deviated (believed to be due to velocity effects in the downcorner region) resulting in the Emergency Feedwater backup flow control valvesV opening prior to e'Emergency Feedwater Actuation Signal.
Normally, it is expected that the Einergency Feedwater ACtuation Signal occurs first and then the flow control valves open. Limit0J1 16itial surveillar,ice tesii16 and engineering analyses confirmed that no malfunction occurred and the sirstem responded operly for the sequence of inputs to which it was subjected. Steam Gener tor pressure was coptro! eI I ' ' l ' by the mosp enc Dump Valves,h .1 d b At ' which were both set between 9 0 -1- 992 psig. TP6 pIntli ■is maintained in Mode 3 with both Steaml11 , 1Generators being fed from the non-safety Auxiliary Feedwater [BA] System with Steam Generator levels in the normal operationaliband for Mode 3. The EFAS was reset. The plant was maintained in Mode 3 while a Post Trip Review was performed.1 1
CAUSAL FACTORS
,i,la , i 6Ir I,� � , ,Results of condition investigations pd a lure modesFl a:analysis conclU e that the most likely cause of the condition was a synchronods motor driven I -OriCW SCAT) malfunction, causing the relayres fin to drop out when not energized, which Oiienti'ai Posed I he Cy(Pkimp Discharge Valves at about one minute intervals and secild 'all' ru rOng' CIAI Limp.1 he tirirl relay,was sent off-site1 m i 1 1 . fl. 1 ,_ ., ,(Southwest Research Institute) or eValuation, Wh re,it Oeteripin I 'd to have extensive thermal degradation. Additional troublesholoting is being evaluated Under the Corrective Action Program.
CORRECTIVE ACTIONS
Corrective actions taken include:
- CW synchronous motor dri■in reset ti er waS rep)aci ed plant startup, and
- the reset timer was sent to So 6t west search Irisiitute for fai ure analysis.
,Additional appropriate correctiv p-p!stered under the Corrective Action Program (CR-WF3-2606-45
SAFETY SIGNIFICANCE
The loss of all circulating water pumps resulted in a loss of, condenser vacuum. Operators took action to initiate a manual reactor trip prior to condenser vacuum dropping to the point at which an automatic turbine trip would occur. The operator actions mitigated the consequences of the loss of condenser vacuum.
t� ..., .� FSAR Chapter 15.2.1.3 analyzes a loss of condenser vacuum. The FSAR analysis is conservative, intended to bound actual plant operation and assumes no Operator action for 30 minutes into the event. As a result, reactor trip occurs on high pressurizer pressure and the RCS safety relief valves open. The peak RCS pressure for the FSAR loss of condenser vacuum analysis is 2732 psia, which is below the acceptance criteria of 2750 psia.
The event on November 11, 2005 was significantly less severe than the FSAR loss of condenser vacuum analysis. Prompt Operator action to initiate a manual reactor trip prior to an automatic turbine trip on loss of condenser vacuum prevented any significant RCS pressure rise and the RCS safety valves from opening. Safety equipment responded to the reactor trip. Emergency Feedwater response was appropriate to ensure decay heat removal from the RCS. At all times during the transient, adequate water level was maintained in the steam generator to ensure decay heat removal from the RCS.
The impact on core damage frequency is negligible since the RCS safety valves did not open (there was no loss of RCS inventory) and multiple additional failures would be necessary for core damage to occur.
In conclusion, this event is bounded by the loss of condenser vacuum analysis in FSAR Section 15.2.1.3. Prompt Operator action to initiate a manual reactor trip mitigated the consequences of the transient to be significantly less severe than that presented in the FSAR. Therefore, the safety significance of this event is small.
This condition is not a Safety Systemunctional Failure. Per NUREG-1022, Event Reporting Guidelines, 10 CFR 50.72 and 50.73, there are four safety system functions: ability to shut down the reactor and maintain it in a safe shutdown condition, ability to remove residual heat, ability to control the release of radioactive material, and ability to mitigate the consequences of an accident. The subject condition involved a manual reactor trip and plant shutdown that did not involve a failure of the plant's ability to achieve the four safety system functions.
SIMILAR EVENTS
A search was performed for other similar reported events at Waterford 3. No similar events were identified.
Condition Reports for the Entergy Nuclear South plants dating from 1993 were also searched. No similar events were identified.
Energy Industry Identification System (EIIS) codes are identified in the text within brackets [ ].