Event Notification Report for September 27, 2002

OFF-SITE NOTIFICATION

On September 27, 2002 at approximately 15:00 Mountain Standard Time, Palo Verde Nuclear Generating Station
(PVNGS) notified the Environmental Protection Agency (EPA) that a INITIAL NOTIFICATION OF CONTINUOUS RELEASE was being reported.

The following information was provided to the National Response Center and State of Arizona:

"Palo Verde Nuclear Generating Station is calling to make an initial notification of a continuous release in accordance
with CERCLA Section 103(f)(2). 40CFR 302.8 (d). This is not an accidental or episodic release.

Palo Verde Nuclear Generating Station will be performing a scheduled steam generator cleaning activity as part of its routine refueling outage for Unit 1. The steam generator cleaning activity will begin on Saturday, September 28 at approximately 18:00. During this activity, Palo Verde Nuclear Generating Station will exceed the Reportable Quantities for the release of ammonia and hydrazine. The same activity is scheduled to occur during the Unit 3 refueling outage, spring of 2003. The activity occurs approximately once per every five to seven years on each unit. The hazardous substances released are ammonia (CAS #7684-41-7) and hydrazine (CAS #302-01-2). PVNGS expects no adverse effects to health and welfare of the general public, nor to the environment.

The NRC Resident Inspector was notified along with the State.

* * * UPDATE ON 4/8/03 FROM DAN MARKS TO HOWIE CROUCH * * *

The following information was received from the Licensee via facsimile:

"The following event description is a follow-up notification to ENS 39226, placed at 15:12 Mountain Standard Time (MST) (18:52 ET) on 09/27/2002.

"ENS 39226 informed the NRC of a government notification to the National Response Center (EPA), the State of Arizona and Maricopa County related to a continuous release in accordance with CERCLA [Comprehensive Environmental Response, Compensation, and Liability Act] Section 103(f)(2), 40 CFR 302.8(d). This was not an accidental or episodic release. The planned release was the result of steam generator cleaning activities and would result in exceeding the Reportable Quantities for the release of ammonia (CAS #7664-41-7) and hydrazine (CAS #302-01-2).

"Included in the notification was the schedule to occur in Palo Verde Unit 3 during its refueling outage in the spring of 2003.

"Since that original notification, Palo Verde has implemented a condensation unit that captures a large amount of the effluent prior to release. Although the Reportable Quantities will still be exceeded, the notification to the National Response Center (EPA), [Maricopa County] and the State of Arizona will identify the lower levels.

"The contact name for information concerning this notification is [DELETED], Environmental Section Leader, at telephone number [DELTED].

"As during the previous notification, PVNGS [Palo Verde Nuclear Generating Station] expects no adverse effects to the health and safety of the public, nor to the environment."

The Licensee has notified the NRC Resident Inspector.

NOUE DECLARED DUE TO LOSS OF OFFSITE POWER SUPPLY

The licensee declared a notice of unusual event (NOUE),in accordance with EAL 3.1, due to the loss of two offsite electrical power supplies to the station for greater than 15 minutes. The cause of the loss of offsite power was a fire at a local hydro electric station. All systems functioned as required. All emergency diesels responded and supplied electrical power to the station shutdown electrical boards. The loss of two offsite power supplies places the station in a technical specification (TS 3.8.1 D) limiting condition for operation (LCO).

The licensee notified the NRC Resident Inspector, the State of Tennessee, and the local fire department.

At 0955 NRC entered Monitoring for Normal Phase Mode.

* * * UPDATE ON 09/28/02 AT 0311 FROM RICKY STOCKTON TO GERRY WAIG * * *

The licensee terminated the NOUE at 0308 after the second offsite electrical power supply was restored to the site at 0300 and exited technical specifications 3.8.1 limiting conditions for operation. The first offsite power supply was restored at 0125.

At 0311 NRC exited monitoring for the Unusual Event

Notified IRO (Wessman), R2DO (W. Rogers ), PAO (R. Virgilio ), NRR EO (M. Johnson), FEMA (K.C.), OHA Ron Brahm

DEVIATION FROM TECHNICAL SPECIFICATIONS PER 10 CFR 50.54(x)

The licensee reported a deviation from license condition 2 F, which requires compliance with the Watts Bar Fire Protection Report, part 2, section 9.1, per 10 CFR 50.54(x). This section requires the minimum fire brigade staffing to be restored within two hours. The Watts Bar fire brigade was dispatched to a fire at the nearby Watts Bar Hydro Generation Station (see NRC Event Number 39220) and replacement fire brigade members had not yet arrived on-site. The licensee stated that the fire brigade can respond to a fire at Watts Bar Nuclear but response time would be greater than normal.

The licensee notified the NRC Resident Inspector.

FOUR HOUR 91-01 BULLETIN RESPONSE

At approximately 7:30 a.m. on September 27, 2002, a UO2 furnace operator reported an overload alarm condition at the Line 1 UO2 sintering furnace. Subsequent visual inspection of the furnace indicated a pellet boat "jam" had occurred inside the sintering furnace chamber. An estimated 4-5 boats are involved in the localized area in which the jam occurred. The stacking of the boats inside the furnace constitutes a loss of geometry control.

The independent control on moderation remained intact, thus no unsafe condition existed, or currently exists. The UO2 sintering furnace temperature is in excess of 1700 degrees C.

Following procedure, operations promptly notified nuclear safety. Safe recovery of the material and re-establishment of the geometry control will require a 2 week cool down. This condition is reported pursuant to NRC bulletin 91-01 (within 4 hours) due to a loss of geometry that requires in excess of 4 hours to re-establish. Implementation of corrective actions is pending completion of incident investigation,

SAFETY SIGNIFICANCE OF EVENTS:
Low safety significance - localized loss of safe slab geometry, moderation control remained intact.

POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO(S) OF HOW CRITICALITY COULD OCCUR):
Multiple failure modes required before a criticality accident could occur.

CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.):
Geometry and moderation.

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

Estimated 4-5 pellet boats involved in actual jam represent [ . . .] kgs UO2 in the form of dry fuel pellets, enriched to [ . . . ]

wt% U235. Under a loss of geometry, our analyses have demonstrated a fully reflected hemisphere mass of UO2 pellets at [ . . . ] enrichment containing 5% water is ~[. . . ]kgs. Thus, a large safety margin exists for this loss of geometry Condition.


NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES:
The geometry control provided by safe slab height of 3 was lost because of jam inside UO2 sintering furnace.

CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMENTED:
Controlled shutdown of sintering furnace (requires 2 weeks to cool down), recovery of material into favorable geometry containers under NSE approved procedure.


Investigation of event and evaluation of additional corrective actions pending.

VALID ENGINEERED SAFETY FUNCTION ACTUATION DUE TO PARTIAL LOSS OF OFF SITE POWER

"While operating at 100% power on September 27, 2002, Watts Bar experienced a loss of one off-site power source, followed about 15 minutes later by the loss of the second off-site power source. These two off-site power sources are fed from the Watts Bar Hydro plant, and are the primary power source for the 6900 V shutdown boards, via Common Station Service Transformers (CSST's) C & D. CSST C feeds shutdown bds[boards] 1A-A & 2A-A, CSST D feeds shutdown bds 1B-B & 2B-B. A fire at the hydo plant was the cause of the off-site power supply breakers opening. The breaker for the supply to the CSST C opened at 0824 (EDT). This deenergized shutdown bds 1A-A & 2A-A, initiating a valid engineered safety function (ESF) actuation that started all four standby diesel generators (DG's), two of which reenergized these shutdown bds. When the other off-site source was lost, the remaining 2 DG's then loaded onto their respective shutdown bds. The plant remain at full power since one train of shutdown power was always available. As designed, AFW started coincident with the EDG start. At this time, the 2 lines have been reenergized and the A, B, C and D Common Station Service Transformers are also energized. Preparations are being made to re-power nonsafety related loads from off-site power."

The Licensee has notified the NRC Resident Inspector.

INVALID SPECIFIED SYSTEM ACTUATION WHILE PERFORMING MAINTENANCE

"EVENT DESCRIPTION

"This report is being made in accordance with §50.73 (a)(1), which states, in part, 'In the case of an invalid actuation reported under §50.73 (a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER.' These invalid actuations are being reported under §50.73 (a)(2)(iv)(A). NUREG-1022, Rev. 2, states that the report should provide the following information:

"The specific train(s) and system(s) that were actuated

"Whether each train actuation was complete or partial

"Whether or not the system started and functioned successfully.

"On September 27, 2002, at 1042 hours, during the performance of Maintenance Surveillance Test (OMST-RWCU41R), 'RWCU System Isolation Logic System Functional Test,' technician actions to perform positive identification of a logic relay prior to performing visual verification of contacts on that relay as specified within the surveillance procedure instructions, resulted in invalid actuations. The actuations included the Primary Containment Isolation system (PCIS) Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems) and Division A (i.e., inboard) PCIS Group 2 valves (i.e., the Drywell Equipment and Floor Drains). Both Standby Gas Treatment (SBGT) system trains A and B started and the Reactor Building ventilation system isolated. The actuations of PCIS Group 6 and Division A Group 2 valves and Reactor Building ventilation were complete and the affected equipment responded as designed to the invalid signal (i.e., the valves and dampers that were open, at the time of the event, closed). By 1115 hours, the PCIS Group 2 and 6 isolation logic circuits were reset and associated valves reopened as required by plant condition, the Reactor Building ventilation system was returned to service, and both SBGT trains A and B were placed In standby.

"Discussion of the causes and corrective actions associated with this event are documented in the corrective action program in AR 72925."

The NRC Resident Inspector has been notified.