Palisades Suffers Reactor Coolant Pressure Boundary Leakage

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Palisades nuclear power plant owned and operated by Entergy Nuclear in Covert Michigan, suffered yet another black eye Sunday, August 12, 2012. On that morning operators determined that an Unidentified Reactor Coolant leak existed, amounting to .3 Gallons Per Minute. Technical Specifications were invoked which forced the operators to expeditiously shutdown the Reactor and Cooldown to Mode 5. A Containment Entry Team found steam issuing forth from a Control Rod Guide Tube, indicating that this was the smoking gun.

Control Rod Guide Tubes enclose the Control Rod Drive Mechanisms which allow Control Rods to insert or withdraw from the Reactor Core. They are an extension of the Reactor Coolant System Pressure Boundary. As such, NO leakage is tolerated.

Repair to this component is likely to be complicated because the Event Report states that the leak is in a section of the Control Rod Guide Tube where there are no welds or threaded joints.  A detailed Failure Mechanism investigation will have to be performed in order to pin point the cause of the leak. Further, an Extent of Condition Analysis will have to determine how many other of the 45 Control Rod Guide Tubes may be susceptible to the same Failure. Other Combustion Engineering Plants will have to apply the results of the Investigation to their plants to determine whether they could occur there.

Recently, Palisades was found to have a leak in a very critical part of the Safety Injection System. The Safety Injection Refueling Water Tank was found to be leaking through cracks in nozzles in the lower part of the tank. The tank contains borated water from which Emergency Cooling Water Pumps would take suction to inject into the Reactor Coolant System in the event of a LOCA.

Both, the Safety Injection Refueling Water Tank and the Reactor Coolant System Leaks may be indications of woeful maintenance and operation practices.

Control Rod Guide Tube Background: FromPalisadesFSAR Section 3.3

There are 45 CRDMs mounted on flanged nozzles on top of the reactor

vessel closure head, located directly over the control rods in the reactor core. Each CRDM is connected to a control rod by a locked coupling. The weight of the CRDMs is carried by the vessel head. In order to provide lateral stability, particularly in resisting horizontal earthquake forces, the CRDMs are supported in the horizontal direction by interconnection. The interconnecting structure permits limited vertical movement due to thermal expansion, but restricts bending deflection so as to limit stresses to allowable values in the lower housing and nozzle areas.

The CRDM is designed to handle a control rod weighing 215 pounds (dry). The total stroke of the drive is 132 inches. The speed of the drive is

46 inches per minute. For a reactor trip, the time from receiving a trip signal to 90% of the full-in position of the rod is less than 2-1/2 seconds. The rod is allowed to accelerate to about 11 ft/s and is decelerated to a stop at the end of the stroke.

The CRDM is of the vertical rack-and-pinion type with the drive shaft  running parallel to the rack and driving the pinion gear through a set of bevel gears. The design of the drive is shown in Figure 3-13. The rack is driven by an electric motor operating through a gear reducer and a magnetic clutch. By de-energizing the magnetic clutch, the control rod drops into the reactor under the influence of gravity. The drive assembly is equipped with a magnetic brake and an antireversing clutch which maintain the position of the rod with the drive in the holding condition and prevent upward movement of the rod when in the scrammed condition. For actuating partial length control rods which maintain their position during a reactor trip, the CRDM is modified by replacing the magnetic clutch with a solid shaft assembly which eliminates the trip function. Otherwise, this CRDM is the same as those attached to the full-length control rods. The drive shaft penetration through the pressure housing is closed by means of a face-type rotating seal.

Power Reactor Event Number: 48178
Region: 1 State: PA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: RON FRY
Notification Date: 08/10/2012
Notification Time: 17:13 [ET]
Event Date: 08/07/2012
Event Time: 17:34 [EDT]
Last Update Date: 08/10/2012
Emergency Class: NON EMERGENCY
10 CFR Section:
Person (Organization):
DRY FUEL STORAGE TRANSFER TRAILER DETECTED WITH REMOVABLE CONTAMINATION”On Aug 7, 2012 at 1734 EDT, PPL Susquehanna personnel were in the process of releasing the Dry Fuel Storage Transfer Trailer from the Unit 1 Reactor Building 101 rail bay and detected removable contamination on the trailer.

“PPL Susquehanna Health Physics personnel identified removable surface contamination up to 30,000 dpm/100 sq cm on the transfer trailer. Gamma spectroscopy performed on smears identified the presence of Cs-137. No other radionuclides were identified on any of the analyzed smears. Cs-137, by itself, is not a nuclide characteristic to Susquehanna due to Susquehanna’s high fuel integrity performance. In addition, no loose surface alpha contamination was identified.

“The area around the trailer, located in the 101 bay, has been posted and controlled as a contaminated area. Decontamination of the transfer trailer is in progress. Onsite surveys of areas that were occupied by the transfer trailer, indicate no removable surface contamination. In addition, no Susquehanna personnel contamination events have been attributed to the contamination found on the transfer trailer.

“Although the receipt of this transfer trailer was not identified as an incoming radioactive shipment to Susquehanna from its’ supplier, this event is immediately reportable to the NRC Operations Center in accordance with 10 CFR 20.1906(d), since the Department of Transportation acceptance limits identified in 49 CFR 173.433 for this type of container are 22,000 DPM/100 sq. cm and PPL Health Physics personnel identified removable radioactive surface contamination in excess of the limits of 10 CFR 71.87(i) which refer to the DOT limits of 49 CFR 173.433.

“The final delivery carrier and NRC Senior Resident Inspector have been notified.” The licensee will be notifying the Commonwealth of Pennsylvania.

Source: NRC Event Reports


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