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| Entered date | Event description | |
|---|---|---|
| ENS 51729 | 11 February 2016 19:38:00 | At 1504 EST on February 11, 2016, with the plant shutdown in a forced outage, the Division 1, 4.16 Kv Safety Bus (EH11) lost power. Division 1 Shutdown Cooling was in service at the time and the Division 1 Shutdown Cooling pump A tripped. The Division 1 Emergency Diesel Generator (EDG) started and loaded EH11 as designed. However, the Emergency Service Water (ESW) A pump, which supplies cooling water to the EDG did not start. Due to the absence of cooling water to the EDG, operators took manual action to secure the Division 1 EDG. Division 2 Shutdown Cooling was operable during this transient and was subsequently started. The Division 1 Shutdown Cooling common suction isolation valve (1E12F0008) had previously been de-energized in the open position to support planned maintenance. The Division 2 Shutdown Cooling isolation valve was not affected by the loss of bus EH11. Shutdown Cooling was re-established at 1544 EST using the Division 2 Shutdown Cooling pump. Reactor coolant temperature rose from approximately 89 degrees Fahrenheit to 115 degrees Fahrenheit during the event. The cause of the loss of EH11 and subsequent failure of ESW A pump to start are currently under investigation. This event is being reported under 10 CFR 50.72(b)(3)(iv)(A) as a specific system actuation due to the auto start of the Division 1 EDG on a valid signal. The plant remains shutdown with Division 2 Shutdown Cooling in operation. The plant is in a normal electrical line up with the exception of bus EH11 being de-energized. The licensee notified the NRC Resident Inspector. |
| ENS 51084 | 22 May 2015 09:24:00 | This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a containment isolation signal affecting more than one system. At 1629 EDT on March 26, 2015, the plant received a Division 1 balance of plant outboard containment and drywell isolation signal. The isolation signal was received while removing fuses to establish a clearance for the replacement of an average power range monitor bypass switch. The removal of two fuses removed power to the manual initiation logic resulting in an isolation signal. The following component actuations occurred: valves 1P51F0150 and 1P51F0652, isolating service air to the containment and drywell; valves 1G61F0155 and 1G61F0170, isolating the containment and drywell floor drain sumps; valves 1D17F0071A and 1D17F0079A, isolating the drywell radiation monitor; valves 1D17F0081A and 1D17F0089A, isolating the containment radiation monitor; valve 1P11F0080, isolating the containment pools drain; valves 1P50F0060 and 1P50F0150, isolating the containment vessel chilled water system; valves 1P53F0070 and 1P53F0075, isolating the upper and lower airlock local leak rate air supply; valves 1P52F0160 and 1P52F0170, isolating the upper and lower airlock air supply; valve 1P22F0015, isolating mixed bed water to the drywell; and valve 1P54F0395, isolating fire protection carbon dioxide to the drywell. This event is considered an invalid system actuation reportable under 10 CFR 50.73(a)(2)(iv)(A). The isolation was not initiated in response to actual plant conditions or parameters, and was not a manual initiation. Therefore, this notification is provided via a 60-day optional phone call in accordance with 10 CFR 50.73(a)(1) instead of submitting a written Licensee Event Report. The event meets reporting criteria specified in 10 CFR 50.73(a)(2)(iv)(B)(2) as a general containment isolation valve signal affecting containment isolation valves in more than one system. All affected systems functioned as expected in response to the loss of power to the manual initiation logic. The valves were reopened in accordance with plant procedures. The inadvertent isolation signal was the result of a human performance error. The NRC Resident Inspector has been notified. |
| ENS 42552 | 4 May 2006 17:19:00 | The Perry Nuclear Power Plant (PNPP) Operating License requires that a report be made within 24 hours to the NRC operations center via the emergency notification system in the event that a violation of the PNPP Fire Protection Program occurs. On May 2, 2006, it was discovered that a specific set of Division 1 Emergency Diesel Generator (EDG) Control Room Control Switch contacts were not designed to isolate the Control Room from the local Division 1 EDG controls in the event of a control room fire. At 1430 on May 4, 2006 it was determined that this deficiency violated the PNPP fire protection program and could adversely affect plant shutdown in the case of a control room fire. A potential fire induced hot short in the diesel generator logic circuit(s), may result in a failure to start or a spurious trip of the Diesel Generator even if control is transferred to remote control. Although repairs and operator actions could have been taken to restore the Division 1 EDG if the Control Room fire caused a failure to start or a spurious trip of the Diesel Generator, these activities were not specifically identified in the Fire Protection safe shutdown analysis or associated operating procedures. The ability to achieve and maintain safe shutdown in the event of a fire in the Unit 1 Control Room could be adversely affected. Compensatory actions (procedure changes) have been completed to address this issue. A written report will follow in accordance with Technical Specification 5.6.6.a and Operating License Condition 2.F. The licensee discovered this discrepancy while researching a modification. This condition has existed since 1987. The licensee informed the NRC Resident Inspector. |