Event Notification Report for July 25, 2001

ADDITIONAL INDICATIONS DISCOVERED ON CRDM HOUSINGS

"Nondestructive examination of welds on control rod drive mechanism pressure housings has resulted in the discovery of additional indications to those reported earlier (reference event notification numbers 38083, 38103 and 38111). The weld examinations are being conducted as part of an extent-of-condition evaluation resulting from a discovery on June 21, 2001 of a through-wall indication on CRDM 21. Extent-of-condition and cause evaluation is ongoing. Regular updates with NRC Region III and NRR will continue."

The licensee stated that no further notifications will be made to the NRC Operations Center regarding this issue.

The NRC resident inspector has been informed of this notification by the licensee.

OFFSITE NOTIFICATION REGARDING ACTUATION OF THE CIVIL DEFENSE SIRENS WHEN NO ACTUAL EMERGENCY EXISTED

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

"During an emergency preparedness drill, a local agency (Claiborne County Civil Defense) activated the civil defense sirens. The sirens should not have been sounded for the drill. This report is being made as an event that for which notification to other agencies has been or will be made (50.72(b)(2)(xi)). Plant operation was not affected, and no release of radioactivity occurred."

The licensee stated that the sirens sounded briefly (for probably less than 1 minute). At the time of this event notification, the site had not received any calls from the public.

It was reported that Claiborne County Civil Defense notified local radio stations and requested them to make an announcement stating that the actuation of the sirens was unintentional and that there was no emergency. Claiborne County Civil Defense also plans to issue a press release. Applicable State, local, and other government agencies were involved in the exercise and are aware of the event. The licensee notified the acting NRC resident inspector (Peter Alter).

AUTOMATIC TURBINE TRIP/REACTOR SCRAM DURING TESTING

"During performance of CIV [combined intercept valve] testing per 2-OI-47 section 6.7, the unit 2 reactor scrammed due to Turbine Generator Load reject. Emergency control procedure EOI-1 was entered due to low reactor water level and high reactor pressure. PCIS [primary containment isolation system] groups 2,3, 6 and 8 were received, A, B and C SBGT [standby gas treatment] trains started and A CREV [control room emergency ventilation] started as expected. Initial indications show that some Safety Relief Valves (SRVs) momentarily opened during the pressure spike following closure of the main turbine control and stop valves. The unit remains in hot shutdown (mode 3) to continue troubleshooting the exact cause of the EHC [electrohydraulic control] logic control system causing the main turbine trip.

"This event is reportable as a 4 hour ENS report per 10CFR50.72(b)(2)(iv)(B) as 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor Operation.' And an 8 hour report per 10CFR50.72(b)(3)(iv)(A) as 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(8), ie: RPS'.

"This is also reportable per 10CFR50.73 (a)(2)(iv)(A) for the above conditions as a 60 day written report."

The NRC resident inspector has been informed of this event by the licensee.

HPCI SYSTEM DECLARED INOPERABLE

"Unit 3 high pressure coolant injection (HPCI) system inoperable. This constitutes an 8-hour ENS report in accordance with 10 CFR 50.72(b)(3)(v) and 10 CFR 50.36(c)(1)(ii)(A). The unit 3 HPCI system was considered inoperable while aligned to the suppression pool. This condition was determined during the performance of an approved system operating procedure that would ensure the suction line of the HPCI system is filled and vented while aligned to the suppression pool. Adequate suction pressure could not be developed during the performance of the procedure, therefore Operations placed the auxiliary oil pump in pull-to-lock to prevent automatic starts of the system.

"Unit 3 HPCI was aligned to the suppression pool in order to comply with Technical Specification Section 3.3.5.1, Required Action D.2.2. Required Action D.2.2 was performed due to a failure of the condensate storage tank (CST) level switches. The failed CST level switches caused an unplanned automatic swap of the HPCI suction line from the CST to the suppression pool.

"The problem is being investigated and the cause is unknown at this time."

The NRC resident inspector has been informed of this event by the licensee.

POTENTIAL TO DRAIN AFW STORAGE TANK DURING PLANT FLOODING SCENARIO

"STPNOC Engineering identified a condition outside the station's design basis that has been determined to be reportable under 10CFR50.73(a)(2)(ii) as an unanalyzed condition that significantly degraded plant safety. Notification is required by 10CFR50.72(b)(3)(ii)(B).

"During review of design calculations, a new internal flooding condition was identified that could have resulted in depletion of the AFW Storage Tank to the point where the plant would not be able to transition from AFW to RHR as designed. Each of STP's four trains of AFW is enclosed in its own water-tight compartment directly beneath its associated MFW line RCB penetration. Operator response to a main feedline break includes isolation of the faulted steam generator, including AFW. If the MFW break is postulated to occur in the MFW penetration area above AFW, the AFW cubicle beneath the break will flood. In the case of the D train steam-driven AFW, the water level will submerge the turbine-driven AFW pump, its trip/throttle valve and AFW isolation valves in about 6 to 30 minutes, depending on break size. As a consequence of the accident, the submerged motor-operated valves are assumed to fail as-is, supplying steam to the turbine-driven pump and allowing AFW flow. In addition, the analysis assumes the single failure of one unsubmerged steam supply isolation valve. The steam-driven pump will continue to function while submerged and continue to take suction from the AFWST and expel it out the break. Unless the pump is secured, its continued operation could accelerate the depletion of the AFWST to the point that the plant would not be able to transition from AFW to RHR as designed. This condition affects only the D train steam-driven AFW. Trains A, B, and C are motor driven and can be readily secured at their power source if necessary.

"Compensatory action is being taken to implement a temporary modification to allow operator action to isolate the AFW supply to the steam-driven AFW pump at the AFWST."

The NRC resident inspector has been informed of this condition by the licensee.

***** RETRACTION FROM WAYNE HARRISON TO L. TROCINE RECEIVED AT 1507 EDT ON 08/09/01 *****

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

"[...] This notification is RETRACTED. The break described above is in a break exclusion zone and is the non-mechanistic break required by the Standard Review Plan to be postulated for the purpose of defining the environmental conditions for qualification of equipment. Consideration of a single failure with this non-mechanistic break is not required. Consequently, the unsubmerged steam supply valve may be assumed to function to isolate steam to the turbine-driven AFW pump and there is not a condition that significantly degrades plant safety."

The licensee plans to notify the NRC resident inspector. The NRC operations officer notified the R4DO (Charles Marschall).

MEDICAL MISADMINISTRATION REPORT INVOLVING ADMINISTERED DOSAGE > PRESCRIBED DOSAGE

At 1700EDT on 7/25/01 a patient undergoing treatment for hyperthyroidism received a 25.7 millicurie dose of I-131 instead of the 20 millicuries prescribed (28% greater than prescribed). The licensee attributes the error to inattention to detail on the part of the technologist involved, i.e., the prescribed dosage was specified on the form. Both the patient's mother and the referring physician have been informed. Since the administered dosage is within the normal range allowed for this treatment, no adverse effects are anticipated. The licensee will determine appropriate corrective actions. Call the NRC Operations Center for contact information.