05000339/LER-2013-002
| North Anna Power Station , Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(C), 50.54(x) TS Deviation 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3392013002R00 - NRC Website | |
1.0 DESCRIPTION OF THE EVENT On May 28, 2013, at 1507 hours0.0174 days <br />0.419 hours <br />0.00249 weeks <br />5.734135e-4 months <br /> with Unit 2 operating at 98 percent power (Mode 1), a manual reactor trip was initiated due to a main feedwater system transient.
North Anna Unit 2 was ramping up in power and had reached 98 percent power. Prior to the manual reactor trip, Abnormal Procedure 2-AP-31, Loss of Main Feedwater, was entered due to the feedwater system transient. The main feedwater system transient resulted from the spurious closure of the "C" Main Feedwater Pump (EIIS System-SJ, Component-P) Discharge Motor-Operated Valve (EIIS Component-V), 2-FW-MOV-250C, the auto-start of "B" Main Feedwater Pump, 2-FW-P-1B, and the subsequent opening of the main feedwater recirculation valves that resulted in the tripping of "A" Main Feedwater Pump, 2-FW-P-1A, on low suction pressure. Although the control room crew commenced a 2 percent per minute ramp down per 2-AP-31, steam generator (SG) (EIIS System-AB, Component-SG) levels continued to lower. A manual reactor trip was initiated to preclude an automatic reactor trip due to low-low SG levels.
All systems responded as expected during the event. All control rods (EIIS System-AA, Component-ROD) inserted into the core at the time of the trip and decay heat was removed via the condenser steam dumps (EIIS System-COND). The auxiliary feedwater (AFW) pumps (EIIS System-BA, Component-P) received an automatic start signal as designed following the reactor trip and provided makeup flow to the steam generators (SG). The SG levels were subsequently restored to normal operating level and the AFW pumps were secured and returned to automatic.
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS No significant safety consequences resulted from this event since the Reactor Protection System (RPS) and the Engineered Safety Feature (ESF) System equipment responded as designed. As such, the event posed no significant safety implications and the health and safety of the public were not affected by the event.
At 1809 hours0.0209 days <br />0.503 hours <br />0.00299 weeks <br />6.883245e-4 months <br />, a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report was made to the NRC in accordance with 10CFR50.72(b)(2)(iv)(B) for a Reactor Protection System (RPS) actuation and a 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report in accordance with 10CFR50.72(b)(3)(iv)(A) for a Auxiliary Feedwater System actuation. The event is reportable pursuant to 10CFR50.73(a)(2)(iv)(A) for a condition that resulted in the automatic actuation of the RPS and AFW Systems.
3.0 CAUSE The direct cause of the spurious closure of 2-FW-MOV-250C and the auto-start of 2-FW-P- 1B was a loss of conductivity (high resistance) across contacts 25 and 26 in the upper cell switch (EIIS Component-33) "AF" of breaker 2-EP-BKR-25C5 (EllS Component-BKR).
The root cause determined that adequate maintenance strategies were not implemented for mechanism operated cell (MOC) and truck operated cell (TOC) switches. This was due to the program owner's and supporting organization's failure to recognize and recommend industry known and documented strategies, and the satisfactory historical performance of these components at North Anna.
4.0 IMMEDIATE CORRECTIVE ACTION(S) The Control Room crew responded to the reactor trip in accordance with emergency procedure 2-E-0, Reactor Trip or Safety Injection. The post trip response progressed as expected and the Control Room crew transitioned to 2-ES-0.1, Reactor Trip Response. All equipment responded as designed.
5.0 ADDITIONAL CORRECTIVE ACTIONS The linkage to the MOC switch for breaker 2-EP-BKR-25C2 was lubricated and manually manipulated to verify the linkage was free to move. Breaker 2-EP-BKR-25C5 was replaced and the upper cell switch "AF" on breaker 2-EP-BKR-25C5 was replaced.
6.0 ACTIONS TO PREVENT RECURRENCE Corrective actions are being tracked under root cause evaluation RCE001101. Corrective actions include development of: 1) maintenance strategies for MOC and TOC cell switches in 4160 volt breaker cubicles, 2) preventative maintenance procedures based on the maintenance strategies, and 3) maintenance procedures for MOC and TOC switches that clean and inspect cell switches, verify the linkage is not binding, and verify the linkage is properly adjusted so the contacts are properly engaged.
7.0 SIMILAR EVENTS to loss of main feedwater suction pressure during a secondary system transient and auxiliary feedwater system actuation.
8.0 ADDITIONAL INFORMATION Unit 1 was operating in Mode 1, 100 percent power on May 28, 2013 and was not affected by this event.