Event Notification Report for March 19, 2000

24 HOUR BULLETIN 91-01 REPORT

At approximately 0750 on 03/19/2000, a water leak was discovered above and adjacent to the Erbia master blender. The water leak was from the building outer roof and around an I-beam that penetrates the secondary roof. The blender was undergoing maintenance and was emptied of Special Nuclear Material (SNM). An open filter housing on the blending was immediately covered with plastic. The blender was currently empty and visually verified to contain less than a safe mass of uranium. Furthermore, no water was detected inside the blender. Thus, no unsafe condition existed. However, the blender will remain out of service until the moderation barriers are reestablished.

POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO(S) OF HOW CRITICALITY COULD OCCUR:

If the blender contained greater than a safe mass of uranium and greater than 11 liters of water entered the blender.

CONTROLLED PARAMETERS:

Moderation is the primary controlled parameter. Normally, there are at least three barriers to isolate the contents of the blender from moderators: two independent overhead roofs and the closed blender vessel. In this case
the blender was empty (<5 kg) and shutdown for maintenance. Therefore, mass control was also in place for the blender.

ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSE MATERIAL:

Approximately 5 kg of residual U02 powder was present in the filter housing of the blender. The maximum enrichment was 4.0% U-235 and the corresponding safe mass is 24 kg U02. The contents were poisoned with approximately 8% erbium.

CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEM AND WHEN EACH WAS IMPLEMENTED:

The system will remain out of service with less than a safe mass until the event investigation is completed and the overhead barriers are successfully repaired.

DISCOVERY OF REACTOR COOLANT SYSTEM LEAKAGE ON THE WITHDRAWAL LINE WELDS OF THREE CONTROL ROD DRIVE MECHANISMS DURING HYDRO TESTING

The following text is a portion of a facsimile received from the licensee:

"EVENT: During [the performance of procedure] OPT-80. 1, Reactor Pressure Vessel ASME Vessel ASME Section XI Pressure Test, reactor coolant system (RCS) pressure boundary leakage was observed on the withdrawal line welds of three control rod drive mechanisms (CRDM). The following CRDMS were noted with leakage: CRDM 06-15 (10 drops/minute at 1030 psig), CRDM 30-15 (20 drops/minute at 1030 psig) and CRDM 14-19 (20 drops/minute at 1030 psig). This leakage is outside the acceptance standard of zero leakage per ASME Section XI, Table IWB-3410-1."

"INITIAL SAFETY SIGNIFICANCE EVALUATION: Minimal; the leakage is well within the capacity of available injection systems."

"CORRECTIVE ACTION(S): The CRDMs will be removed, and the affected areas repaired prior to reactor startup."

The licensee notified the NRC resident inspector.

* * * RETRACTION 1522 3/21/2000 FROM MARK TURKAL TAKEN BY BOB STRANSKY * * *

"Upon further inspection and evaluation, CP&L had determined that the previously reported leaks from control rod drive mechanisms (CRDMs) 06-15, 30-15, and 14-19 do not represent a serious degradation of a principal safety barrier.

"Based on the location of the leaks, it was conservatively assumed to be pressure boundary leakage coming from the withdrawal line welds to the CRD housing flange; which was outside the acceptance criteria of zero leakage per ASME Section XI, Table IWB-3410-1. CP&L notified the NRC at 23:31 EST (i.e., Event Number 36812), in accordance with 10 CFR 50.72(b)(2)(i), that a condition on Unit 1 had been identified that had it been found while the reactor was in operation, would have resulted in the nuclear power plant, including its principal safety barriers, being seriously degraded.

"Subsequently, repressuization of the Unit 1 reactor pressure vessel was performed and the leaks were observed using advanced video inspection techniques. Based on this re-inspection, it was determined, on March 21, 2000, that the leakage was from the CRDM flange bolting and not the withdrawal line welds. As such, this is not pressure boundary leakage and the zero leakage acceptance criteria of ASME Section XI, Table IWB-3410-1 is not applicable. This condition does not represent a serious degradation of a principal safety barrier.

"CP&L has evaluated and dispositioned the observed leakage, in accordance with Code Case N-566-1, 'Corrective Action For Leakage Identified at Bolted Connections,' and determined it to be insignificant leakage, which is expected to seal during plant heatup."

The NRC resident inspector has been informed of this retraction. Notified R2DO (Decker).