Event Notification Report for March 09, 2007

MAIN STEAM LINE FLOW INSTRUMENT LINE PRESSURE BOUNDARY LEAKAGE

"With Unit 2 in Mode 4 for a planned refueling outage, a one-inch socket weld elbow on an instrument line off the 'D' main steam line was identified as leaking water. Leakage was quantified as less than 20 drops per minute (with the main steam lines full of water).

"This elbow is on a steam line flow instrument (just below its condensing chamber) and is located in the Unit 2 drywell (primary containment).

"This item constitutes a primary coolant boundary leak discovered while shutdown."

A pressure test is planned to be performed after the repair of the instrument line elbow is completed.

The licensee notified the NRC Resident Inspector.

SECONDARY CONTAINMENT BYPASS LEAKAGE RATE EXCEEDED

"Appendix J Local Leak Rate Testing has determined that Tech Spec Acceptance Criteria for Secondary Containment Bypass Leakage has been exceeded. During the performance of SE-259-044 the combined SCBL limit of 9scfh was exceeded. Test results were within Administrative and Acceptance Criteria for the 10CFR50 Appendix J limits of 0.6La. The leakage limit is 9scfh. The identified degraded condition is reportable as a condition of the nuclear power plant, including its principle safety barriers being seriously degraded per 10CFR50.72(b)(3)(ii)."

The NRC Resident Inspector was notified of this event by the licensee.

UNUSUAL EVENT DECLARED DUE TO RCS LEAKAGE GREATER THAN TECH SPEC LIMIT

"A Notification of Unusual Event has occurred on Unit 2 due to Reactor Coolant System (RCS) unidentified leakage greater than 10 gallons per minute."

The licensee notified NRC Resident Inspector. Notified USDA (Brzostels).

* * * UPDATE AT 1754 ON 3/9/06 FROM T. PETRAK TO P. SNYDER * * *

"No actual RCS leakage occurred and the RCS always remained intact. The additional water injection occurred and the RCS always remained intact. The additional water injection into the Reactor Coolant System was required to makeup for gases that came out of solution and accumulated in the reactor vessel head and were being vented from the RCS during a planned drain down."

The licensee terminated the Unusual Event at 1751 on 3/9/07. The licensee notified the NRC Resident Inspector, state, and local government agencies. The licensee will remain in Mode 5 for their planned outage.

Notified R2DO (G. Hopper), NRR EO (M. Case), IRD (T. Blount), NRR (J. Dyer), DHS (McKentry), FEMA (Biscoe), DOE (Wyatt), EPA (Burgess), HHS (Dalviel), USDA (Giles).

* * * UPDATE AT 1531 EDT ON 5/11/07 FROM ROBERT DORMAN TO GERRY WAIG * * *

This report is retracted based on the following:

"On 3/9/07, an NOUE was declared while Unit 2 was in Mode 5, draining the RCS to support Rx Vessel disassembly, due to what was then diagnosed as an RCS Unidentified Leak greater than 10 gpm. Subsequent review of the event revealed that no leak existed and the mismatch between charging and letdown, which was the basis of the leak determination, was caused by a nitrogen bubble being vented from the Rx Vessel Head and the additional charging volume was displacing the nitrogen volume as it was being vented. As a result, Vogtle Unit 2 is retracting the NOUE classification that was made on 3/9/07."

The licensee informed the NRC Resident Inspector. Notified R2DO (Ernstes).

MANUAL REACTOR TRIP AT REDUCED POWER DUE TO INADEQUATE FEEDWATER CONTROL

"At approximately 0803 Central Standard Time (CST) on 3/9/2007, Callaway Plant received indications of a condenser tube leak. At approximately 0826 CST, a power reduction from 100 percent to less than 5 percent was initiated in accordance with the Secondary Water Chemistry Program. At 1043 CST, with the reactor at approximately 30 percent power, a manual reactor trip was initiated due to inadequate feedwater control to the C steam generator. There was indication of feedwater flow to and increasing level in the C steam generator with both the main feedwater reg valve and its bypass valve closed. Cause of the inadequate feedwater control is under investigation.

"All safety systems functioned as designed with the exception of the digital rod position indication for control rod K10 which failed during the trip due to an existing problem. Control rod K10 is believed to be fully inserted, but is being treated as indeterminate at this time. Main feedwater isolation and motor-driven auxiliary feedwater actuations occurred as expected. Both trains of the emergency diesel generators and the offsite power supplies are available.

" Callaway plant is currently stable in Mode 3 at normal reactor coolant system parameters.

"This notification is being made in accordance with 10CFR50.72(b)(2)(iv)(B) for manual reactor trip (RPS actuation), 10CFR50.72(b)(2)(xi) for offsite notification and 10CFR50.72(b)(3)(iv)(A) Specified System Actuation.

"The licensee informed the resident inspector of the event and will be making State notifications. A press release is expected."


* * * UPDATE ON 03/12/07 AT 2116 EDT FROM S. GANZ TO MACKINNON * * *


"Repairs were made to the Digital Rod Position Indication (DRPI) for control rod K10. The cause was determined to be a bent pin on a cable connector. When DRPI was restored, control rod K10 was verified to be fully inserted. Post maintenance testing was completed satisfactorily.

"The cause of the inadequate feedwater control for the C steam generator was a failure of the positioner for the C main feedwater reg valve. The positioner was replaced and tested satisfactory.

"The plant is currently in Mode 2 (Startup)

"The resident inspector was informed of the update."

R4DO (Mike Shannon) notified.