05000280/LER-2005-001
| Docket Numbersequential Rev | |
| Event date: | 02-07-2005 |
|---|---|
| Report date: | 04-04-2005 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 2802005001R00 - NRC Website | |
1.0 DESCRIPTION OF THE EVENT On February 7, 2005 at 2002 hours0.0232 days <br />0.556 hours <br />0.00331 weeks <br />7.61761e-4 months <br />, Unit 1 commenced a reactor startup following a planned feedwater heater [EIIS-SJ, HX] maintenance outage. On February 7, 2005 at 2017 hours0.0233 days <br />0.56 hours <br />0.00333 weeks <br />7.674685e-4 months <br />, while withdrawing Control Bank "A", control rod [EIIS-AA, ROD] B-10 indicated a rapid drop from approximately 42 steps to 17 steps on the Computer Enhanced Rod Position Indication (CERPI) panel [EIIS-AA]. Withdrawal of Control Bank "A" was stopped and the CERPI indication for control rod B-10 remained at 17 steps. The remaining CERPIs in Control Bank "A" were observed to be indicating 40 to 45 steps, which is within normal tolerance. The operating team initiated 0-AP 1.00, "Rod Control System Malfunction," and subsequently determined that the reactor should be tripped. Therefore, at 2024 hours0.0234 days <br />0.562 hours <br />0.00335 weeks <br />7.70132e-4 months <br /> the reactor was manually tripped and procedure 1-E-0 "Reactor Trip Response" was initiated.
The operating staff acted promptly and appropriately. Proper response of the automatic protection systems following manual actuation of the reactor trip was verified. All required systems functioned as designed during the trip. Unit 1 was stabilized at hot shutdown with reactor coolant system temperature approximately 547 degrees Fahrenheit. Decay heat was removed via steam generator blowdown [EIIS-WI] and the Main Steam Dump Valves [EIIS-SB].
At 0014 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> on February 8, 2005 a non-emergency, eight-hour notification was made to the NRC pursuant to 10 CFR 50.72(b)(3)(iv)(A).
This report is being submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A) for a manual actuation of the reactor protection system (RPS).
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS This event resulted in no significant safety consequences or implications. At the time of the manual trip, reactor startup was in progress with reactor power in the source range at approximately 210 counts per second. All required systems functioned as designed during the manual reactor trip and there were no radiation releases due to this event. A risk impact concluded that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> integrated core damage probability increase was low at 1E-10 and the increase in potential large early release frequency was negligible. This core damage estimate is slightly conservative because it is based upon a more severe transient risk from a trip at full power. Therefore, the health and safety of the public were not affected.