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| Entered date | Event description | |
|---|---|---|
| ENS 53381 | 4 May 2018 12:50:00 | On March 8, 2018, an invalid system actuation occurred while preparations were underway to perform Safety Features Actuation System (SFAS) integrated response time surveillance testing during the recent Davis Besse Nuclear Power Station refueling outage. Several minutes after connecting a data recorder to monitor the Emergency Diesel Generator (EDG) 1 start signal, at 1323 hours (EST), the EDG started with no valid actuation signals or test inputs present. The EDG successfully came up to speed and voltage as expected. The associated essential 4160 volt electrical bus remained energized from the normal power supply, therefore, the EDG output breaker did not close to supply power to the bus. Troubleshooting determined the inadvertent actuation was due to a short in the test lead wires at the recorder connection caused by a faulty test lead. The test lead was replaced and the SFAS surveillance testing completed satisfactorily. This event is being reported as an invalid system actuation per 10 CFR 50.73(a)(2)(iv)(A); this 60-day optional telephone notification is being made per 10 CFR 50.73(a)(i) in lieu of submitting a written Licensee Event Report. The NRC Resident Inspector was notified of the inadvertent EDG start at the time of the event and has been notified of this invalid specified system actuation notification. |
| ENS 51837 | 31 March 2016 00:14:00 | On March 30, 2016, at 1715 EDT, with the Unit shutdown and in Mode 6 for refueling, evidence of leakage was identified on a 3/4-inch flexible braided piping connection on Reactor Coolant Pump (RCP) 1-1, and this issue was determined to be reactor coolant system pressure boundary leakage. This flexible piping is for RCP 1-1 first stage seal cavity vent line, and is categorized as ASME Section III Class 2 piping. The leakage was identified due to the discovery of a small amount of boric acid (approximately 1/2 teaspoon) on the welded end connection of the flexible piping. No active leakage was identified at the time of discovery with the Reactor Coolant System depressurized and approximately 110 degrees F. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.13, 'RCS Operational Leakage,' does not apply in the current plant condition (Mode 6). The cause and resolution of the leakage are under evaluation. This event is reportable within 8 hours per 10 CFR 50.72(b)(3)(ii)(A). The NRC Resident Inspector has been notified. |