Event Notification Report for December 20, 2000

AUTOMATIC REACTOR TRIP DUE TO TURBINE TRIP CAUSED BY TRANSFORMER GROUND

The apparent cause of the reactor trip was a turbine trip caused by a ground overcurrent relay actuation on the 2X01 transformer (main power out transformer). Heat removal is via the steam dumps to the main condenser with main feedwater supplying the steam generators. All systems operated as required. The licensee is investigating the cause of the ground indication.

The licensee notified the NRC Resident Inspector.

OFFSITE NOTIFICATION TO STATE DEP DUE TO OIL SPILL ON SITE

The licensee notified the Connecticut Department of Environmental Protection (DEP) about an oil spill due to operator error. The waste oil separator was in operation filling a 55-gallon drum when the operator overflowed the drum. The spill was contained on the concrete pad, there was no release offsite or to the water. Cleanup is now in progress.

The licensee notified NRC Region 1 (John Wray).

FITNESS-FOR-DUTY REPORT REGARDING A NON-LICENSED SUPERVISOR'S RANDOM POSITIVE TEST FOR ALCOHOL

A non-licensed supervisor tested positive for alcohol during a random test. This was a first time positive test for this individual. The individual was removed from site, and the individual's badge was placed on administrative hold.

The licensee made this 24-hour event notification in accordance with 10 CFR 26.73(a)(2)(iv).

The licensee notified the NRC resident inspector.

NRC BULLETIN 91-01 RESPONSE (24-HOUR REPORT) - LOSS OF ONE CONTROL IN THE DOUBLE CONTINGENCY

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

"The C-331 E and F surge drum room temperature was found to be at 103 degrees F in violation of NCSA.CAS-005. Requirement 3.2.6 of this NCSA states that when the surge drums contain UF6, they shall be operated at a minimum temperature of 105 degrees F. This temperature requirement prevents hydrogen fluoride (HF) from condensing at the maximum pressure rating of the drums. Should HF condense it could moderate any UF6 that might be present."

"SAFETY SIGNIFICANCE OF EVENTS:"

"HF cannot condense at 103 degrees F and 20 psia. The drum pressures were all less than 1 psia; therefore, it was not possible to condense HF."

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

"In order for a criticality to be possible, HF would have to condense. Therefore, the temperature of the drums would have to be less than 80 degrees F, and the drum pressures would have to exceed 20 psia."

"CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.):"

"[Since the double] contingency is based on two controls on moderation and [since] the temperature control was lost, double contingency was not maintained."

"ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS LIMIT AND % WORST CASE CRITICAL MASS):"

"The surge drums contain an estimated 1425 pounds of UF6 at less than 1 weight % U235 assay."

"NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES:"

"The first leg of double contingency is an administrative control, which limits the surge drum room temperature to a minimum of 105 degrees F. The control was violated, and the first leg of double contingency was lost."

"The second leg of double contingency is an administrative control to maintain the drum pressures below 20 psia, which is the maximum rated pressure for the drums. This control was not violated, and this leg of double contingency was maintained."

"CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMENTED:"

"Temperature loss was due to low outside ambient temperature. Building temperature was raised and surge drum room temperature rose above minimum required temperature. Additional heating for the area is being considered."

Paducah personnel notified the NRC resident inspector.

AUTOMATIC CONTROL ROOM VENTILATION SYSTEM SHIFT FROM NORMAL TO EMERGENCY RECIRCULATION MODE WHILE A PLANNED GAS DECAY TANK RELEASE WAS IN PROGRESS

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

"At 2014, with a gas decay tank release in progress (total activity 0.215 Ci, (Kr-85 = 0.214 Ci, Xe-133 = 0.001 Ci)), control room ventilation gas monitor R-36 went into alarm causing ventilation to shift to mode F. During the [2-second] spike, R-36 counts went from 30 cpm to 856 cpm and back to 30 cpm. Local air samples taken by the RP Tech both in the control room and in the area of the control room ventilation intake revealed no activity."

The licensee stated that the gas decay tank release was planned and that the counts going through the plant stack were considered normal. (These counts peaked at 120 cpm.) The licensee stated that the planned gas decay tank release was within both the expected and required release limits. The wind was blowing in a favorable direction, and the stability class for meteorological conditions was stable. Although counts have returned to normal, the licensee plans to leave control room ventilation in the emergency recirculation mode until tomorrow morning.

The licensee notified the NRC resident inspector.

* * * UPDATE ON 1/12/01 @ 1113 BY GOMEZ TO GOULD * * * RETRACTION

Rochester Gas and Electric (RG&E) is retracting this notification. This event was originally thought to be in response to a valid radiological condition because a waste gas decay tank release was in progress at the same time as this event. RG&E has confirmed that the alarm on Control Room Noble Gas radiation monitor
R-36 that occurred at 20:14 EST on December 20, 2000, causing the shift in Control Room ventilation to "Mode F" (emergency recirculation mode), was the result of a spurious spike on R-36. The alarm was not in response to any radiation condition,

Local samples taken by radiation protection technicians on December 20, both in the Control Room and in the area of the Control Room ventilation intake, revealed no activity. The Plant Process Computer (PPCS) plots show that R-36 counts were steady just before and immediately after the spike. Further engineering evaluation of this event has determined that the alarm on R-36 was caused by a sudden spike and was not in response to any radiation condition. Therefore, this event is not reportable and is retracted.

Other occurrences of spiking on R-36 took place during December 2000 and January 2001. Some of these spikes resulted in alarms on R-36 and subsequent shift in Control Room ventilation to "Mode F". These alarms were recognized at the time of alarm as invalid actuations of the HVAC System for the Control Room, and did not result in NRC notification. Trouble-shooting and corrective maintenance is on-going on the R-36 monitor.

The licensee has notified the NRC Resident Inspector.

The Region 1 RDO (Lew) was notified.

FAILURE OF A REACTOR CORE ISOLATION COOLING (RCIC) BYPASS TO CONDENSATE STORAGE TANK VALVE TO FULLY STROKE DURING STROKE TIME TESTING

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

"EVENT: At 2008 on 12/20/00 during performance of OPT-10.1.8 (RCIC System Valve Operability Test), the 2-E51-F022 (RCIC bypass to condensate storage tank) failed to fully stroke during stroke timing of the valve. A valve thermal overload annunciator was received. This rendered the Brunswick Unit 2 RCIC system inoperable. RCIC was not in operation at the time of this failure."

"INITIAL SAFETY SIGNIFICANCE EVALUATION: Minimal safety significance. Remaining ECCS systems are operable."

"CORRECTIVE ACTION(S): Determine the cause of the 2-E51-F022 valve failure. Return the Unit 2 RCIC system to operable following repair of 2-E51-F022."

The licensee stated that this event did not result in any radiological release or reactor coolant leaks. The licensee also stated that all systems functioned as required and that there was nothing unusual or misunderstood. This event placed the unit in a 14-day technical specification limiting condition for operation.

The licensee notified the NRC resident inspector.

* * * RETRACTION ON 01/11/01 AT 1059ET BY CHARLES ELBERFELD TAKEN BY MACKINNON* * *

Upon further evaluation, it has been determined that the conditions resulting in the RCIC system being declared inoperable for Technical Specification LCO 3.5.3 did not result in a loss of the system's function to remove residual heat. The function of the RCIC system is to respond to transient events by providing makeup coolant to the reactor. The RCIC system is not an Engineered Safety Feature system, and no credit is taken in the Updated Final Safety Analysis Report (UFSAR) for RCIC system operation mitigating the consequences of a postulated accident. The RCIC system is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of normal coolant flow from the reactor feedwater system to provide adequate core cooling and control of the RPV water level. Its operational purpose is to provide an alternate source of reactor coolant to the vessel and to provide sufficient cooling to remove residual heat following reactor shutdown and loss of feedwater flow without requiring depressurization of the reactor. Neither the 2-E51-F022 being de-energized in the open position, nor the questionable capability of the 2-E51-F029 resulted in the loss of the ability of the RCIC system to remove residual heat.

In the case of the RCIC Bypass to Condensate Storage Tank valve 2-E51-F022, with the valve de-energized in the open position, injection flow into the RPV is not diverted to the Condensate Storage Tank, or into the High Pressure Coolant Injection system due to additional closed valves in the test flow path. Therefore, if the RCIC system had been called upon to perform its needed function to remove residual heat, the system would have performed as required.

In the case of the RCIC Torus Suction valve 2-E51-F029, although the ability for the RCIC system sump suction to align to the Torus must be available to meet its Technical Specification LCO Operability requirements, the RCIC system has the capacity to meet its function to remove residual heat when aligned to the Condensate Storage Tank. Although the RCIC system is a single train injection system, it has redundant suction flow paths that are fully capable of meeting these injection requirements. Loss of one suction flow path does not prevent the RCIC system from performing its required function to remove residual heat.

Subsequently, it has been determined that the malfunction of the RCIC system components did not adversely impact that system function to remove residual heat. Carolina Power and Light Company has determined that this event does not meet 10 CFR 50.72 or 10 CFR 50.73 reporting criteria and this notification is being retracted.
NRC R2DO (Chuck Ogle) notified.

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

LOST HUMBOLDT 5001EZ NUCLEAR GAUGE

This is an initial notification of a lost nuclear gauge. The gauge was lost on December 20, 2000, while being transported in an unsecured manner and not in it's transport case. The gauge was a one month old Humboldt 5001EZ. The gauge was probably lost along Highway 90 and 36th Street in San Antonio, Texas. A Police Report was filed (Report # 00805997/01). The operator performed an extensive search for the gauge along the highway. A notice was posted in the San Antonio Express News with a reward offered. The gauge has not been recovered to date. Investigation is on going.

Further details will be sent as they are received at the Texas Department of Health.

* * * UPDATE 0818 1/18/2001FROM OGDEN TAKEN BY STRANSKY * * *

A representative of the Texas Department of Health called the Operations Center to correct and clarify the initial report. The gauge is a Humboldt 5001EZ, not a 5001EX as initally reported. Additionally, the Texas licensee is Reed Engineering Group of Houston. [HOO Note: the text of the initial report above has been corrected.]

Notified R4DO (Kennedy).