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| Entered date | Event description | |
|---|---|---|
| ENS 51717 | 9 February 2016 09:54:00 | Planned maintenance of the Plant Computer System (PCS) will cause a loss of emergency assessment capability. Beginning February 9, 2016, PCS data will not be available to the following Comanche Peak Nuclear Power Plant (CPNPP) facilities due to planned PCS software modifications: -Emergency Operations Facility (EOF) -Backup EOF -Operations Support Center The Emergency Response Data System (ERDS) will also be unavailable. The planned maintenance of the PCS is being reported as a loss of assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii) because the duration is expected to be more than 72 hours and the data to the Backup EOF is also affected. CPNPP has compensatory measures in place to ensure timely emergency classification, protective action recommendation and emergency notification, as needed. The PCS modification is expected to be complete by February 18, 2016. A follow-up ENS (Emergency Notification System) communication will be made when the EOF assessment capability is restored. The Licensee has notified the NRC Resident Inspector.
The planned maintenance of the Comanche Peak Plant Computer System (PCS) that began on February 9, 2016 is complete. Assessment capability in the following facilities was restored effective 1300 (CST) February 18, 2016: -Emergency Operations Facility (EOF) -Backup EOF -Operations Support Center The Emergency Response Data System availability has also been restored. The Licensee has notified the NRC Resident Inspector. Notified the R4DO(Werner). |
| ENS 51444 | 3 October 2015 12:02:00 | During the scheduled Unit 2 down power to start 2RF15, Steam Generator 3 Feed Control Valve failed to control in Automatic or Manual resulting in excessive feedwater flow. Control Room Operators manually tripped the reactor and aligned Auxiliary Feedwater. No automatic RPS or ESF actuations occurred. Plant conditions are stable. All control rods fully inserted on the trip and no safety or relief valves lifted. The plant is supplying water to the steam generators with auxiliary feedwater and removing decay heat through the main condenser. The unit is in its normal shutdown electrical lineup. There was no effect on unit-1. The licensee informed the NRC Resident Inspector.
The licensee added 10 CFR 50.72(b)(3)(iv)(A) to the 10 CFR codes reporting section. No other text was modified. Notified the R4DO (Taylor). |
| ENS 50562 | 23 October 2014 16:42:00 | The following information is provided as a 60 day telephone notification to the NRC in accordance with 10 CFR 50.73(a)(1) reported under 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of the Unit 1, Train 'A' Emergency Diesel Generator on September 12, 2014, at 1342 CDT. Per NUREG-1022, R.3, page 33, the following information should be provided for an invalid actuation reported under 10 CFR 50.73(a)(1): (a) the specific train(s) and system(s) that were actuated, (b) whether each train actuation was complete or partial, (c) whether or not the system started and functioned successfully. (a) The specific train(s) and system(s) that actuated were: Unit 1, Train 'A' Emergency Diesel Generator. (b) Whether each train actuation was complete or partial: The actuation of the Unit 1, Train 'A' Emergency Diesel Generator was a complete actuation. (c) Whether or not the system started and functioned successfully: The Unit 1, Train 'A' Emergency Diesel Generator started and functioned successfully. The NRC Senior Resident Inspector has been notified. |
| ENS 50043 | 18 April 2014 14:35:00 | While implementing a multiple spurious operations (MSO's) modification during a refueling outage, it was identified that a fire safe shutdown cable routing location may have been in question. On 04/18/2014 at 1252 CDT, it was determined that this cable was routed through a cable tray that was not designed to have a fire safe shutdown barrier. This created an unanalyzed condition that significantly degrades plant safety, consistent with NUREG 1022, Rev. 3 guidance. This cable is associated with a Motor Operated Valve (MOV) used to isolate the containment sump from the Residual Heat Removal (RHR) pump suction during normal system operations. Compensatory measures in this area were already in effect for the resolution of MSO's scenarios. The compensatory measures assure the Systems, Structures, and Components (SSC's) associated with this cable remain operable. The licensee has notified the NRC Resident Inspector. |
| ENS 49419 | 8 October 2013 15:42:00 | A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits in the control room has determined the described condition to be applicable to Comanche Peak Nuclear Power Plant resulting in a potentially unanalyzed condition with respect to 10CFR50 Appendix R analysis requirements. The original plant wiring design and associated analysis for the Class 1 E batteries control room ampere indications do not include overcurrent protection features to limit the fault current. In the postulated event, a fire in the control room could cause one of the ammeter wires to hot short to the ground plane. Simultaneously, the fire causes another DC wire from the opposite polarity on the same battery to also hot short to the ground plane. This could cause a ground loop through the unprotected ammeter wiring. This event could result in excessive current flow (heating) in the ammeter wiring to the point of causing a secondary fire in the raceway system. The secondary fire could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10CFR50 Appendix R. This condition is reportable in accordance with 10CFR50.72(b)(3)(ii)(B) as an unanalyzed condition. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Resident Inspector. See related Event #49411. |