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Event Notification Report for May 15, 2006

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
05/12/2006 - 05/15/2006

** EVENT NUMBERS **


42562 42571 42572 42573 42574 42575

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General Information or Other Event Number: 42562
Rep Org: MA RADIATION CONTROL PROGRAM
Licensee: HERBERT HENKEN
Region: 1
City: MILTON State: MA
County: SUFFOLK
License #:
Agreement: Y
Docket: 03-6234
NRC Notified By: MIKE WHALEN
HQ OPS Officer: JOE O'HARA
Notification Date: 05/09/2006
Notification Time: 08:48 [ET]
Event Date: 11/01/2005
Event Time: [EDT]
Last Update Date: 05/09/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
ANTHONY DIMITRIADIS (R1)
GREG MORELL (NMSS)

Event Text

MASSACHUSETTS AGREEMENT STATE - ABANDONED RA-226 QUANTITY 1000 TIMES GREATER THAN LABELING REQUIREMENTS

The following information was received from the state via fax:

"Cause Description: During Clean-Out of Deceased Dermatologist, the family discovered a box with Ra-226.

"About 11/1/2005: [Deleted] contacted the Boston Regional (Office) of the EPA to dispose of Ra-226 discovered at family home. EPA recommended to [Deleted] that it would be cheaper if he disposed of the Ra-226 directly via a waste broker. EPA suggested Radiac Environmental Services of Brooklyn, NY.

"11/17/2005: Radiac (Environmental Services of Brooklyn, NY) visited the Milton, MA residence and estimated that the amount of Ra-226 was about 2.9mCi. In addition, Radiac placed the box of Ra-226 sources into a 5-Gallon DOT Type A steel drum with a ring bolt closure, placed a security seal was placed over the ring, and labeled the drum with yellow/magenta radioactive label. [Deleted] was going to Florida for the winter.

"4/27/06: The EPA informed the Mass. Radiation Control Program about the Ra-226 in Milton, MA.

"Corrective Actions: Material to be disposed of by licensed broker - Radiac Environmental Services of Brooklyn, NY - during the week of May 8, 2006."

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Power Reactor Event Number: 42571
Facility: CALLAWAY
Region: 4 State: MO
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP
NRC Notified By: JOHN DAMPF
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 05/12/2006
Notification Time: 05:11 [ET]
Event Date: 05/12/2006
Event Time: 00:53 [CDT]
Last Update Date: 05/12/2006
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):
DALE POWERS (R4)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 M/R Y 12 Power Operation 0 Hot Standby

Event Text

MANUAL REACTOR TRIP ON HIGH STEAM GENERATOR LEVEL

"While lowering turbine load to 45% for planned maintenance to replace an RCS Loop flow transmitter, Callaway Plant experienced high vibration on two main turbine bearings. The main turbine was manually tripped at 0047 in accordance with off-normal procedures. At 0052, received a Steam Generator High-High Level (P-14) on the "A" S/G resulting in a Feed Water Isolation Signal (FWIS) and Auxiliary Feed Water Actuation (AFAS). The reactor was manually tripped at 0053. Emergency Operating Procedures were completed and exited at 0115."

After receiving the main turbine high vibration alarm, the plant reduced power to below the reactor trip/turbine trip setpoint and manually tripped the main turbine. Steam generator level rose to the P-14 feedwater isolation setpoint at which time, the reactor was manually tripped.

All rods fully inserted on the trip. Decay heat is being removed by condenser steam dumps. Steam generator level is being maintained with the startup feed pump. The electrical grid is stable. No relief valves or safety valves lifted. The cause of the high vibration is being investigated.

The licensee notified the NRC Resident Inspector.

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Power Reactor Event Number: 42572
Facility: LIMERICK
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: ROY HARDING
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 05/12/2006
Notification Time: 10:35 [ET]
Event Date: 03/14/2006
Event Time: 17:40 [EDT]
Last Update Date: 05/12/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
ANTHONY DIMITRIADIS (R1)

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

Event Text

INADVERTANT CORE SPRAY PUMP ACTUATION

"The notification is being made pursuant to 10CFR50.73(a)(2) (iv) (A) and is reported per 10CFR50.73(a)(1)

"An invalid actuation of the 1 C Core Spray Pump start during D13 LOCA/LOOP Testing.
a) The specific train(s) and system(s) that were actuated. The 1C Core Spray Pump actuated automatically on an inadvertent Division 3 LOCA signal during LOCA/LOOP testing. No other ECCS train/system actuated.
b) Whether each train actuation was complete or partial. The 1C Core Spray Pump actuated in the minimum flow protection mode.
c) Whether or not the system started and functioned successfully. The 1 C Core Spray System started and functioned successfully but did not inject into the vessel. On March 14, 2006 at 5:40 PM during performance of the D13 LOCA/LOOP Test, ST-6-092-117-1, the 1 C Core Spray pump was inadvertently started. The pump started (but did not inject) when an I&C Technician attempted to demonstrate operation of the test switch. He inadvertently picked up the energized test switch which was next to 2 spare test switches and manipulated it which resulted in a Division 3 LOCA signal to the 1C Core Spray Pump. The 1C Core Spray Pump never injected and was running with min-flow protection and functioned as expected. The cause of the event was lack of attention to detail and self check."

The licensee notified the NRC Resident Inspector.

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General Information or Other Event Number: 42573
Rep Org: GENERAL ELECTRIC COMPANY
Licensee: GENERAL ELECTRIC COMPANY
Region: 1
City: WILMINGTON State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JASON POST
HQ OPS Officer: MIKE RIPLEY
Notification Date: 05/12/2006
Notification Time: 22:36 [ET]
Event Date: 04/24/2006
Event Time: [EDT]
Last Update Date: 05/12/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
ANTHONY DIMITRIADIS (R1)
JAMES MOORMAN (R2)
RICHARD SKOKOWSKI (R3)
OMID TABATABAI-EMAIL (NRR)
JACK FOSTER (EMAIL) (NRR)

Event Text

PART 21 NOTIFICATION - BWR CORE SHROUD TIE ROD UPPER SUPPORT CRACKING

"Summary:
GE Energy, Nuclear (GE) has provided core shroud repairs using tie rods to the US BWR plants identified in Attachment 1 [of the Part 21 notification]. Recently it was discovered during an in-vessel visual inspection (IVVI) that tie rod upper supports at Hatch Unit 1 experienced cracking. The apparent root cause is Intergranular Stress Corrosion Cracking (IGSCC) in the Alloy X-750 tie rod upper support material. Alloy X-750 material is susceptible to IGSCC if subjected to sustained, large peak stress conditions. GE opened an internal evaluation to determine if the potential IGSCC in the X-750 tie rod structural components of other BWR shroud repairs designed by GE could be a reportable condition under 10CFR21.

"GE used the criterion provided in the BWR Vessels & Internals Project (BWRVIP-84) for the IGSCC susceptibility assessment of the X-750 components in the tie rod vertical load path. GE has concluded that it is not a reportable condition for the plants that were found to be within or not significantly exceed the BWRVIP-84 criterion. These US plants are identified as 'NR' in Attachment 2 [of the Part 21 notification]. GE determined that two US plants exceed the BWRVIP-84 criterion for the upper supports (in addition to the Hatch Unit 1 as-found condition). GE has not completed the evaluation for these plants to assess if a substantial safety hazard (SSH) exists. These plants have been provided a 60-Day Interim Report Notification under 21.21(a)(2) and are identified as '60-Day' in Attachment 2 [of the Part 21 notification].

"Safety Basis:
Cracking in the tie rod components made of X-750 may render the tie rod ineffective in maintaining core shroud configuration integrity during postulated accident conditions. Loss of core shroud integrity could impact the ability to maintain adequate core cooling for postulated design basis accident conditions. This condition would be reportable under 10 CFR 21 as a substantial safety hazard.

"Corrective Action:
The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action (note, these are actions specifically associated with the identified deviation or failure to comply):
1. A preliminary cause evaluation has been performed. The apparent cause of the cracking is Intergranular Stress Corrosion Cracking (IGSCC). A material sample is being shipped to the GE Vallecitos Nuclear Center for examination to confirm the apparent cause. GE will report the results of the examination by August 21, 2006.
2. The issue has been communicated to the industry through the BWR Owners' Group and the Electric Power Research Institute (EPRI)/BWR Vessel and Internals Project (BWRVIP). The NRC was informed in a NRC management meeting with EPRI and the BWRVIP Executive Oversight Committee at the NRC offices, Rockville, on March 15, 2006.
3. GE has completed an evaluation of the susceptibility to IGSCC using the BWRVIP-84 criterion. Determination of whether any possible cracking could lead to a substantial safety hazard (i.e., loss of core shroud configuration integrity during a design basis accident condition) depends upon many factors, including the actual extent of cracking in the repair components. Until inspections are completed, the actual extent of cracking is not known. GE is developing a model to predict the postulated extent of tie rod upper support cracking for tie rods with upper supports made of Alloy X-750. For upper supports that exceed the BWRVIP-84 criteria significantly, the model will be used to postulate the extent of cracking. This prediction will be used to determine if a substantial safety hazard could exist. GE will report the results of the evaluation by October 9, 2006.
4. The original design basis stress reports will be reviewed to assess the available margin in the primary membrane + bending stress intensities of the upper supports with respect to ASME code allowable values. Where reasonable margin exists in the original design basis code evaluation (an existing margin of approximately 25 % will be considered as reasonable margin), the existing margin is deemed adequate to offset any engineering assumptions or judgments used in the original analysis. Where the original margin is less than 25%, further review will be performed (including finite element analysis, if necessary) to confirm that the upper support remains qualified. This review will be completed by October 9, 2006."


Affected US Plants per Attachments 1 and 2 of the Part 21 notification: Clinton, Nine Mile Point 1, Pilgrim, Dresden 2 & 3, Quad Cities 1 & 2, Hatch 1 & 2.

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Power Reactor Event Number: 42574
Facility: QUAD CITIES
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: ERIC MARKS
HQ OPS Officer: MIKE RIPLEY
Notification Date: 05/14/2006
Notification Time: 16:53 [ET]
Event Date: 05/14/2006
Event Time: 09:57 [CDT]
Last Update Date: 05/14/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
RICHARD SKOKOWSKI (R3)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N N 0 Refueling 0 Refueling
2 N Y 97 Power Operation 97 Power Operation

Event Text

EMERGENCY DIESEL GENERATOR AUTOMATIC START

"During restoration from a scheduled LCO, the Unit 1 EDG auto started when its control switch was placed in the auto position. The auto start occurred as a result of a 4KV Bus Feed Breaker being open. This breaker was in the open position as part of scheduled maintenance.

This notification is being made in accordance with 10CFR50.72(b)(3)(iv)(A) due to an actuation that was the result of a valid signal."

The 4KV Bus was subsequently restored and the EDG was returned to standby. The licensee notified the NRC Resident Inspector.

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Power Reactor Event Number: 42575
Facility: SALEM
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: TOM BYYKKONON
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 05/15/2006
Notification Time: 01:37 [ET]
Event Date: 05/14/2006
Event Time: 23:20 [EDT]
Last Update Date: 05/15/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
Person (Organization):
ANTHONY DIMITRIADIS (R1)
RICK LAYMAN (EPA)
STU BAILEY (DOE)
AMANDA (USDA)
KEN SWEETSER (FEMA)
JANET DUZMAN (HHS)

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

SODIUM HYPOCHLORITE LEAK INTO THE DELAWARE RIVER

"This 4-Hour notification is being made in accordance with 10CFR50.72 (B)(2)(xi).

"A notification was made to the New Jersey Department of Environmental Protection of a discharge of approximately 1000 gallons of water containing 3000 parts per million Sodium Hypochlorite to the Delaware river, via a permitted outfall. The source of the water was a leak on the chlorination injection line in the Unit One Service Water Bay. The water was pumped to the river via the building sump pump. On discovery the leak was isolated and the building sumps turned off to prevent further discharge. Follow up investigation to determine the cause of the leak is in progress.

"There was no equipment out of service that contributed to this event and there were no personnel injuries or radiological occurrences associated with this event."

The leak occurred when changing the pump in service. The licensee will notify the NRC Resident Inspector.

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