Event Notification Report for October 11, 2008

STEAM GENERATOR WATER LEVEL TRIP ACTUATION INITIATED REACTOR TRIP SYSTEM

"During plant shutdown to begin a scheduled refueling outage, with the plant at 0% power in Mode 3, an actuation of the reactor trip system (RTS) occurred when a steam generator (S/G) Low-Low water level trip inadvertently occurred during recovery from a feedwater isolation actuation that had previously occurred in response to a steam generator high water level trip condition.

"At approximately 0500 hours, plant operators were closing the main steam isolation valves (MSIVs) for plant shutdown. Once the MSIVs were closed, the reactor coolant system (RCS) cooldown rate decreased significantly. The 'A' atmosphere steam dump valve (ASD) was then opened. Shortly after the ASD was opened, the MSIV was also opened. With both the MSIV and ASD open, however, the 'A' S/G water level swelled until the P-14 S/G Hi Level protective interlock setpoint was reached, resulting in a feedwater isolation (FWIS) actuation at 0505 hours.

"Subsequent to the FWIS actuation, plant operators took action to recover from the FWIS in accordance with off-normal operating procedure OTO-SA-00001, 'Engineered Safety Feature Actuation Verification and Restoration.' The motor-driven auxiliary feedwater pumps were started to restore S/G water level. However, S/G water level lowered rapidly in response to the cold auxiliary feedwater flow to the S/G. A S/G Low-Low water level trip signal was then reached on the 'A' S/G, thus resulting in a reactor trip system actuation at 0508 hours. All systems functioned normally in response to plant conditions."

The NRC Senior Resident Inspector has been informed.

* * * UPDATE AT 0239 EDT ON 10/12/08 FROM FRED BIANCO TO VINCE KLCO * * *

"This report is an update to information reported under ENS notification 44561. The original notification reported a valid reactor trip system actuation as a 4-hour notification under 10CFR50.72(b)(2)(iv)(B). The correct reporting requirement is 10CFR50.72(b)(3)(iv)(A), an 8-hour notification, as the plant was in Mode 4 with the reactor subcritical at the time of the event.

"The [NRC] Resident Inspector will be notified."

Notified R4DO (Powers).

UNANALYZED CONDITION DUE TO VENDOR CALCULATION ERROR

"Turkey Point received notification of an error in a vendor calculation methodology regarding the containment building heat transfer. This calculation is used to support containment temperature and pressure as well as heat input to the Component Cooling Water system under design basis accident conditions.

"While validating a new analysis model, it was discovered that the Turkey Point containment steel liner was not included in the analysis of record. This is a non-conservative error in that less heat would be expected to be transferred to the external environment with the presence of the liner. The initial evaluation of the error indicates that at the maximum allowable Ultimate Heat Sink (Intake) temperature, Containment pressure, long-term Containment temperature and Component Cooling Water return temperature could exceed licensing basis limits during a design basis accident.

"Preliminary analysis indicates that under current Ultimate Heat Sink conditions, operability requirements are met."

The licensee notified the NRC Resident Inspector.

* * * UPDATE AT 1642 ON 3/16/09 FROM MCKEE TO HUFFMAN * * *

"The purpose of this update is to retract EN #44562 submitted to the NRC Operations Center on October 11, 2008.

"FPL has performed a detailed evaluation of the impact of errors identified by Westinghouse in the Loss of Coolant Accident (LOCA) containment analysis of record on the operation of Turkey Point Units 3 and 4. The evaluation took into consideration actual ultimate heat sink (intake cooling water - ICW) temperatures, and removed conservatisms used in the LOCA containment integrity analysis and Component Cooling Water (CCW) system analysis. The evaluation confirmed that the Units 3 and 4 containment structures were capable of performing the safety function required by the Technical Specifications. Additionally, after removal of conservatisms in conjunction with a minimum CCW heat exchanger heat removal capability consistent with an ICW temperature of 95 degrees F, and an assumed CCW heat exchanger tube resistance, the long term containment temperature and CCW return temperature would not have exceeded licensing basis limits during a design basis accident.

"The evaluation determined that Turkey Point Units 3 and 4 were not in an unanalyzed condition that significantly degraded plant safety as a result of the containment integrity analysis error and this event is not reportable under 10CFR50.72(b)(3)(ii)(B) criteria. Therefore, EN #44562 is retracted.

"The evaluation described above is documented in the Corrective Action Program in report 2008-31338 ."

The licensee will notify the NRC Resident Inspector. Notified R2DO(Nease).

CONTROLLED SUBSTANCE FOUND IN THE PROTECTED AREA

Licensee personnel found an unattended controlled substance inside the protected area. Contact the Headquarters Operations Officer for additional details.

The licensee notified the NRC Resident Inspector.

TURBINE TRIP/ AUTOMATIC REACTOR TRIP ON LOW STEAM GENERATOR LEVEL

At 2236 on 10/11/08 while reducing power for a planned refueling outage, "C" Steam Generator water level decreased and the turbine tripped. The reactor tripped on the turbine trip. Auxiliary feedwater initiated as expected and decay heat is being removed via the condenser steam dumps. All control rods fully inserted.

No significant safety equipment is out of service and all safety buses are being supplied by offsite power. Emergency Diesel generators are available if needed. No PORVs or primary/secondary relief valves lifted.

The licensee is evaluating cause of steam generator level excursion. Unit 2 is unaffected.

The licensee notified the NRC Resident Inspector.

DISCOVERY OF AN AFTER-THE-FACT NOTIFICATION OF UNUSUAL EVENT

"At 0049 on 10/11/08 Letdown Relief valve BG8117 lifted prior to set point, resulting in Pressurizer water inventory being diverted to the Pressurizer Relief Tank. Letdown was isolated within 13 minutes, isolating the leakage. Upon review of the level trends of the Pressurizer Relief Tank level changes, approximately 574 gallons were diverted from the Reactor Coolant system. This would result in an RCS leakage rate of 44 gpm of RCS Identified Leakage. At the time that the calculation of leakage rate was discovered, it was immediately recognized that this met the requirements for EAL 4R, 'RCS Identified Leakage greater than 25 gpm'. As a result of identification that the criteria of an EAL was exceeded, and no longer existed, a notification was made to the NRC Operation Center in accordance with 10 CFR50.72(a)(1)(i)."


Licensee has notified NRC Resident Inspector and will notify state and local agencies on the next business day.