05000311/LER-2011-004
| Salem Generating Station Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3112011004R01 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
Westinghouse — Pressurized Water Reactor (PWR/4) Reactor Coolant System / Reactor Coolant Pump {AB/P} * Energy Industry Identification System {EMS} codes and component function identifier codes appear as {SS/CCC}
IDENTIFICATION OF OCCURRENCE
Event Date: June 26, 2011
CONDITIONS PRIOR TO OCCURRENCE
Salem Unit 2 was in Mode 1. No additional structures, systems or components were inoperable at the time of the discovery that contributed to the event.
DESCRIPTION OF OCCURRENCE
At approximately 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> on June 26, 2011, the 23 Reactor Coolant Pump (RCP) {AB/P} tripped resulting in an automatic reactor trip on low flow in one reactor coolant loop above the P-8 permissive (36% power permissive). As expected, the 21, 22 and 23 Auxiliary Feedwater (AFW) pumps started on low steam generator level following the unit trip. Unit 2 was stabilized in Mode 3 at normal operating temperature and pressure with the 21, 22 and 24 RCPs in-service.
CAUSE OF OCCURRENCE
Salem Unit 2 tripped on low reactor coolant flow in the 23 reactor coolant loop. The low flow condition was the result of the trip of the 23 RCP due to two separate phase to ground faults. Both ground faults occurred at the termination to the motor where the cable shield interfaces with the stress tube of the termination kit.
The insulation at this location was found to be brittle. The reason for the severe hardening of the cable insulation under the stress relief tube was unable to be determined by laboratory testing. Existing cable testing practices performed in accordance with industry standards would not detect this cable condition.
Failure analysis determined that a partial discharge test or potentially destructive testing would have to be performed to detect the cable insulation degradation.
PREVIOUS OCCURRENCES
A review of LERs at Salem Station dating back to 2008 did not identify any prior similar occurrences as a result of medium voltage cable degradation.
SAFETY CONSEQUENCES AND IMPLICATIONS
The loss of forced reactor coolant flow is a Condition II Event analyzed in Section 15.2.5 of the UFSAR.
The UFSAR analysis states that the departure from nucleate boiling ratio (DNBR) will not decrease below the limiting value at any time during the transient and therefore no core safety limit is violated.
The UFSAR analysis assumes a reactor trip on low RCS loop flow within 1.6 seconds of reduction in RCS loop flow with control rods beginning to drop in 2.6 seconds. Based on the review of post trip data the low flow reactor trip occurred in less than 1 second from the opening of the 23 RCP breaker and control rod drop occurred in less than 2 seconds. These response times were within the analyzed values therefore there were no safety consequences as a result of the reactor trip.
A review of this event determined that a Safety System Functional Failure (SSFF) as defined in NEI 99- 02, Regulatory Assessment Performance Indicator Guidelines, did not occur. This event did not result in a condition that would have prevented the fulfillment of a safety function of a system needed to shutdown the reactor and maintain it in a safe shutdown condition, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.
CORRECTIVE ACTIONS
1. The 23 RCP motor was tested satisfactorily. The damaged sections of the RCP motor leads were removed and new terminations were made.
2. Megger testing and visual examination of the 21, 22, and 24 RCP motor leads was performed with no immediate concerns identified.
3. Salem Unit 1 RCP motor feeder cables for the 11, 12, 13 and 14 RCP motors will be replaced during the 1R21 refueling outage.
4. Salem Unit 2 RCP motor feeder cables for the 21, 22, 23 and 24 RCP motors will be replaced during the 2R19 refueling outage
COMMITMENTS
No commitments are made in this LER.