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Event Notification Report for May 17, 2004

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
05/14/2004 - 05/17/2004

** EVENT NUMBERS **


40748 40749 40751 40752 40753 40754

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Power Reactor Event Number: 40748
Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-2
NRC Notified By: FRANK CIGANIK
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 05/14/2004
Notification Time: 10:50 [ET]
Event Date: 05/14/2004
Event Time: 08:13 [EDT]
Last Update Date: 05/14/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
OTHER UNSPEC REQMNT
Person (Organization):
JOHN ROGGE (R1)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 100 Power Operation 100 Power Operation

Event Text

CHANGE IN THE METHODOLOGY USED BY GE/GNF TO DEMONSTRATE COMPLIANCE WITH ECCS PERFORMANCE REQUIREMENTS

"Oyster Creek has been informed of a change in its vendor's calculation of Peak Cladding Temperature (PCT) and local cladding oxidation that is based on a reanalysis of the oxygen available for injection into the reactor vessel that would then recombine with hydrogen produced during the postulated zinc-water reaction. Based on 10CFR50 Appendix K inputs and assumptions the additional heat released from this increased reaction rate would result in an estimated increase of 25 degrees F in PCT and 1.73% in maximum local oxidation. Using these results, the increased oxidation results in the analysis of record being non-conservative to the 17% limit as specified in 10CFR50.46(b)(2) by 1.23%. Maximum Average Planar-Linear Heat Generation Rate (MAPLHGR) limit adjustments will be applied to bring the oxidation limits below the 17% limit.

"However, sufficient margin exists under current operation to ensure the maximum local oxidation limit would not be exceeded. All other 10CFR50.46 ECCS criteria are not impacted. Additionally, based on a conservative evaluation using nominal inputs and assumptions significant margin to 10CFR50.48 ECCS criteria exists even under unlimited oxygen conditions.

"This voluntary notification is being made as a result of the 10CFR50.46(a)(3)(ii) requirement to report this issue in accordance with 10CFR50.72 and 10CFR50.73.

Background

"During a Loss of Coolant Accident (LOCA), oxygen is available in the fuel bundles due to evaporation of the Emergency Core Cooling System (ECCS) water and release of the dissolved oxygen. In addition, oxygen can also be drawn into the reactor vessel later in the LOCA when the vessel pressure has dropped to the Primary Containment (Drywell) pressure. Condensation of the steam in the reactor vessel upper plenum due to the injection of sub-cooled ECCS water causes a reduction in pressure that will result in drawing the non-condensable gases (including oxygen) from the Drywell.

"Hydrogen is generated, within the fuel bundles during a LOCA due to Zirconium metal - water reaction caused by high cladding temperatures. It has been postulated that, at the pressure and temperature conditions in the reactor core during a LOCA, free hydrogen and oxygen could combine and release heat that will increase the steam temperature. Since steam is the heat sink when the core is uncovered, an increase in the steam temperature can result in an increase in the PCT for non-jet pump plants including Oyster Creek (BWR/2). The LOCA scenario for a BWR/2 designed plant is different since the core remains uncovered and there is no period of reflooding for large breaks. The cladding will still be heating up when the oxygen from the containment gets into the vessel. This phenomenon also results in an increase in local oxidation.

"Based on current core thermal power level and existing margin to limits on power operation there is sufficient margin for analyzed accident scenarios requiring ECCS operation including appropriate compensation to restore the analytical oxidation within 10CFR50.46 criteria and there is no impact on safe operation. This is further supported by the results of conservative evaluation using nominal inputs. As a result of this change in the ECCS model and the consequential adjustment to the allowable MAPLGHR limit, the offsite dose for this scenario is still bounded by our current safety analysis. Therefore, this event is not significant with respect to the health and safety of the public.

Corrective Action(s):

"An administrative adjustment to our allowable MAPLHGR limits will be imposed to compensate for the estimated change in cladding oxidation and thereby restore the previously analyzed margin on core thermal power limits to within the new ECCS model prediction. Additionally deinerting of the Primary Containment will be constrained to less than 25% rated power-level until more specific analyses are complete.

"In accordance with 10CFR50.73, a voluntary LER will be submitted within 60 days with a more detailed discussion of the nature of this change in the ECCS evaluation model, its estimated effect of the limiting ECCS analysis, and a proposed schedule for providing a reanalysis or taking other action as may be needed to comply with this regulation.

"The licensee will notify the NRC Resident Inspector."

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Power Reactor Event Number: 40749
Facility: NINE MILE POINT
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] GE-2,[2] GE-5
NRC Notified By: QUENTIN HICKS
HQ OPS Officer: BILL GOTT
Notification Date: 05/14/2004
Notification Time: 15:45 [ET]
Event Date: 05/14/2004
Event Time: 09:17 [EDT]
Last Update Date: 05/14/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
OTHER UNSPEC REQMNT
Person (Organization):
JOHN ROGGE (R1)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 100 Power Operation 100 Power Operation

Event Text

CHANGE IN THE METHODOLOGY USED BY GE/GNF TO DEMONSTRATE COMPLIANCE WITH ECCS PERFORMANCE REQUIREMENTS

"Nine Mile Point has been informed of a change in its vendor's calculation of Peak Cladding Temperature (PCT) and local cladding oxidation that is based on a reanalysis of the oxygen available for injection into the reactor vessel that would then recombine with hydrogen produced during the postulated zinc-water reaction. Based on 10CFR50 Appendix K inputs and assumptions the additional heat released from this increased reaction rate would result in an estimated increase of 25 degrees F in PCT and 1.73% in maximum local oxidation during five reactor recirculation loop operation, with containment inerted. The increased oxidation results in the analysis of record being non-conservative to the 17% limit as specified in 10CFR50.46(b)(2). Maximum Average Planar-Linear Heat Generation Rate (MAPLHGR) limit adjustments will be applied to bring the oxidation limits below the 17% limit.

"Notification is being made as a result of the 10CFR50.46(a)(3)(ii) requirement to report this issue in accordance with 10CFR50.72 and 10CFR50.73.

Background

"During a Loss of Coolant Accident (LOCA), oxygen is available in the fuel bundles due to evaporation of the Emergency Core Cooling System (ECCS) water and release of the dissolved oxygen. In addition, oxygen can also be drawn into the reactor vessel later in the LOCA when the vessel pressure has dropped to the Primary Containment (Drywell) pressure. Condensation of the steam in the reactor vessel upper plenum due to the injection of sub-cooled ECCS water causes a reduction in pressure that will result in drawing the non-condensable gases (including oxygen) from the Drywell.

"Hydrogen is generated, within the fuel bundles during a LOCA due to Zirconium metal - water reaction caused by high cladding temperatures. It has been postulated that, at the pressure and temperature conditions in the reactor core during a LOCA, free hydrogen and oxygen could combine and release heat that will increase the steam temperature. Since steam is the heat sink when the core is uncovered, an increase in the steam temperature can result in an increase in the PCT for non-jet pump plants. The LOCA scenario for a BWR/2 designed plant is different since the core remains uncovered and there is no period of reflooding for large breaks. The cladding will still be heating up when the oxygen from the containment gets into the vessel. This phenomenon also results in an increase in local oxidation.

"The change in the allowable MAPLGHR limit for five loop operation with the containment inerted results in the offsite dose for this scenario being still bounded by our current safety analysis. Therefore, this event is not significant with respect to the health and safety of the public.

Corrective Action(s):

"Reactor Engineering has implemented the 2% MAPLHGR Limit reduction through imposition of a .98 MAPRAT to restore compliance to the 10CFR50.46 limits.

"Operations will place administrative controls/ procedure changes to preclude operating without containment inerted at or above 25% power to limit oxygen available for this postulated phenomenon.

"In accordance with 10CFR50.73, a LER will be submitted with a more detailed discussion of the nature of this change in the ECCS evaluation model, its estimated effect of the limiting ECCS analysis, and a proposed schedule for providing a reanalysis or taking other action as may be needed to comply with this regulation."

The licensee notified the NRC Resident Inspector.

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Other Nuclear Material Event Number: 40751
Rep Org: U.S. ARMY
Licensee: U.S. ARMY
Region: 3
City: ROCK ISLAND State: IL
County:
License #: 10-00722-06
Agreement: Y
Docket:
NRC Notified By: THOMAS GIZICKI
HQ OPS Officer: BILL GOTT
Notification Date: 05/14/2004
Notification Time: 17:38 [ET]
Event Date: 05/13/2004
Event Time: 12:15 [CDT]
Last Update Date: 05/14/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
30.50(b)(1) - UNPLANNED CONTAMINATION
Person (Organization):
ANNE MARIE STONE (R3)
LINDA SMITH (R4)
ROBERT PIERSON (NMSS)

Event Text

CONTAMINATION EVENT WHILE CONDUCTING MAINTENANCE

"The incident occurred at Ft Carson, CO in the Direct Support (DS) maintenance Shop on 13 May 2004. It involved a M64A1 mortar sight unit containing 6.69 curies of tritium gas in sealed source form (NSN 124001-379-7953). The M64 mortar was being turned in by the 10th Special Forces for routine maintenance. It was discovered that a mechanical pin needed to be removed from the M64 in order to fix it. After fixing the pin the device was leak tested. Leak test results showed removable contamination of approximately 2.5 million dpm [disintegrations per minute] (1.12 microcuries). The RSO took immediate action upon receiving the wipe test result of the device. He double bagged the device and placed it in a secured area. Several area wipes were performed of the shop area and of areas outside the shop area. The shop area showed contamination ranging from 500 dpm to 900 K dpm. The RSO took action to remove items from the shop that were contaminated. Survey results showed that the tritium contamination did not spread outside of the machining area. Three persons were bioassayed. Bioassay results were not available at the time of this report."

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Power Reactor Event Number: 40752
Facility: TURKEY POINT
Region: 2 State: FL
Unit: [ ] [4] [ ]
RX Type: [3] W-3-LP,[4] W-3-LP
NRC Notified By: GRANT MELIN
HQ OPS Officer: JEFF ROTTON
Notification Date: 05/14/2004
Notification Time: 18:26 [ET]
Event Date: 05/14/2004
Event Time: 17:29 [EDT]
Last Update Date: 05/14/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
KERRY LANDIS (R2)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
4 A/R Y 100 Power Operation 0 Hot Standby

Event Text

SCRAM DUE TO LOWERING STEAM GENERATOR LEVEL

At 1728 EDTon 05/14/04, while performing maintenance on Turbine 1st stage steam pressure transmitter (PT-447), feedwater regulating valve (FRV) 4A was observed to close. At 1729, an automatic actuation of RPS occurred on Low Steam Generator level in the 4A Steam Generator. This was followed shortly by a manual reactor trip by the operator. All control rods fully inserted. All 3 turbine driven AFW pumps started as expected and are supply steam generators through both feedwater supply headers. The electric plant remains in a normal lineup. RCPs are in operation transferring decay heat to the steam generators. The MSIVs are open with the steam generators discharging steam to the main condenser using the condenser steam dump valves. All steam generator atmospheric dump valves opened on the reactor trip and closed.

The plant is investigating the cause of the FRV closure and intends to stay in Mode 3 until the cause is determined and appropriate repairs are made. There was no impact on the other operating unit.

The licensee will notify the NRC Resident Inspector.

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Power Reactor Event Number: 40753
Facility: OCONEE
Region: 2 State: SC
Unit: [ ] [2] [ ]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: PHIL NORTH
HQ OPS Officer: JAMIE HEISSERER
Notification Date: 05/15/2004
Notification Time: 10:18 [ET]
Event Date: 05/15/2004
Event Time: 08:31 [EDT]
Last Update Date: 05/15/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xii) - OFFSITE MEDICAL
Person (Organization):
JOEL MUNDAY (R2)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
2 N N 0 Refueling 0 Refueling

Event Text

POTENTIALLY CONTAMINATED EMPLOYEE TRANSPORTED TO OFFSITE MEDICAL FACILITY

At 0831 EDT, an individual was transported to an offsite medical facility due to chest pains. The individual was preparing to leave a contaminated area when the chest pains occurred. His protected clothing was cut off, and he was transported via ambulance to an offsite medical facility. It was not known at the time of notification if the individual was in fact contaminated.

The licensee notified the NRC Resident Inspector.


* * * * UPDATED ON 5/15/04 AT 1134 EDT FROM P. NORTH TO M. RIPLEY * * * *

At 1134 EDT, the licensee reported that the individual was not contaminated.

Notified R2DO (Munday).

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Power Reactor Event Number: 40754
Facility: POINT BEACH
Region: 3 State: WI
Unit: [ ] [2] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: JIM WILSON
HQ OPS Officer: BILL GOTT
Notification Date: 05/15/2004
Notification Time: 15:12 [ET]
Event Date: 05/15/2004
Event Time: 11:54 [CDT]
Last Update Date: 05/15/2004
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
ANNE MARIE STONE (R3)
CHRISTOPHER GRIMES (NRR)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
2 M/R Y 100 Power Operation 0 Hot Standby

Event Text

MANUAL REACTOR TRIP DUE TO DIVER ENTANGLED IN INTAKE CRIB

"A Unit 2 manual reactor trip was initiated when the control room was notified that a diver was entangled in the intake crib. Divers were being used to inspect the intake crib, install buoys, and the fish deterrent system. The diver's umbilical cord became snagged and attempts to free it were unsuccessful. The Unit 2 circulating water system was secured to aid in removing the diver from the water. The diver still had breathing air available during the transient. The diver was subsequently removed from the water unhurt.

"Plant systems functioned as required, including the Reactor Protection and Auxiliary Feedwater Systems. There was no Emergency Core Cooling System actuation. Note: The condenser was unavailable because circulating water was secured. This caused a loss of condenser vacuum and its use as a heat sink. The atmospheric steam dumps are currently being used for heat removal from the steam generators. The circulating water system was subsequently restored to service. This event is reportable pursuant to 50.72(b)(2)(iv)(B), any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical and 50.72(b)(3)(iv)(A), PWR auxiliary feedwater system."

All control rods inserted into the core. The electrical busses are in a normal shutdown line up.

The licensee notified the NRC Resident inspector.

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