Information Notice No. 90-77: Supplement 1:Inadvertent Removal of Fuel Assemblies from the Reactor Core

                                UNITED STATES
                        NUCLEAR REGULATORY COMMISSION
                    OFFICE OF NUCLEAR REACTOR REGULATION
                           WASHINGTON, D.C.  20555

                              February 4, 1991


Information Notice No. 90-77, SUPPLEMENT 1:  INADVERTENT REMOVAL OF FUEL  
                                                 ASSEMBLIES FROM THE REACTOR 
                                                 CORE


Addressees:

All holders of operating licenses or construction permits for 
pressurized-water reactors (PWRs).

Purpose:

This information notice supplement is intended to provide additional 
information to that previously provided in Information Notice No. 90-77, 
"Inadvertent Removal of Fuel Assemblies from the Reactor Core."  It is 
expected that recipients will review this information for applicability to 
their facilities and consider actions, as appropriate, to avoid similar 
problems.  However, suggestions contained in this information notice do not 
constitute NRC requirements; therefore, no specific action or written 
response is required.

Description of Circumstances:

On October 4, 1990, at Indian Point Station Unit 3, while the upper core 
internals package (UIP) was being lifted during preparations for defueling, 
two fuel assemblies (FAs) were inadvertently lifted from the reactor core.  
In response to this event, the NRC sent a Special Inspection Team to the 
site to monitor the licensee's recovery of these FAs.  The NRC issued 
Information Notice No. 90-77 on December 12, 1990, to report this event.  
The licensee's recovery activity was detailed in NRC Inspection Report 
50-286/90-19, dated December 13, 1990.

The NRC also sent an Augmented Inspection Team (AIT) to the site after 
recovery of the FAs to determine the probable cause and relevant facts of 
this event and to evaluate the licensee's proposed corrective actions.  The 
AIT concluded that the two FAs were stuck to the UIP because of bent guide 
pins on the upper core plate.  The guide pins were bent during movement of 
the UIP for reinstallation into the reactor vessel in May 1989.  The 
refueling crew had moved the UIP laterally before raising it to a height 
above its storage stand adequate to avoid bumping the guide pins against the 
stand.  Refueling personnel did not recognize that the UIP had inadvertently 
bumped against the storage stand and bent the guide pins.  The UIP was thus 
reinstalled with bent guide pins into the reactor vessel.  As a result, the 
bent guide pins severely damaged the top 



9101290331 
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                                                  IN 90-77, Supplement 1
                                                  February 4, 1991
                                                  Page 2 of 3


nozzles of two FAs.  In addition, the bent guide pins caused mechanical 
deformation of a portion of the fuel rods in one of the two FAs.  The 
outwardly bent fuel rods in this FA caused mechanical deformation of the 
fuel rods in an adjacent FA.  None of these component-related damages were 
identified by the licensee until the next refueling.  The detailed 
inspection findings are described in NRC Inspection Report 50-286/90-80, 
dated January 8, 1991.

Discussion:

Since 1985, the NRC and the Institute of Nuclear Power Operations (INPO) 
have provided the nuclear industry with three documents regarding the 
inadvertent lifting of FAs from the reactor core.  NRC Information Notice 
86-58, "Dropped Fuel Assembly," and two INPO Significant Event Reports 
(proprietary information) addressed events involving stuck FAs.  As a result 
of reviewing these generic communications, the licensee revised its 
refueling procedure before this refueling outage to require the placement of 
an underwater camera and extra lights on the reactor cavity floor to inspect 
the UIP for stuck FAs.  The procedure revision also included a step for 
performing a video inspection after the upper internals were raised 
approximately 1 foot above the reactor vessel flange.  However, the 
licensee's implementation of these procedural steps was not effective 
because, as noted in the AIT findings, (1) the operators did not place extra 
lights on the upper reactor cavity floor as directed by the note in the 
procedure and (2) the operators did not move the camera around the reactor 
vessel flange in order to view the underside of the upper internals from 
different angles.  At the time this step was performed, the lighting was 
particularly inadequate at this location because about half of the lamps in 
the reactor cavity were burned out, and the angle and distance of the lights 
to the UIP created a dark shadow underneath the UIP.  The problem of 
ineffectiveness was compounded by the fact that the camera was in place only 
on the eastern side of the vessel.  Under these circumstances, the camera 
apparently was not able to scan the distance across the vessel to detect the 
stuck FAs, which were located peripherally on the western side of the 
vessel.  As a result, the stuck FAs were lifted upward and transported 
horizontally for a short distance before being noticed by refueling 
personnel.   

The licensee utilized a refueling contractor (Westinghouse) to perform major 
steps of the refueling operations.  The licensee limited its overall 
supervisory control of these operations because of the contractor's 
extensive experience with the design of the plant and with refueling at 
similar plants.  This lack of oversight resulted in the licensee's failure 
to provide adequate control, either supervisory or technical, over key steps 
of the refueling operations.  For example, the AIT found that the licensee's 
refueling Senior Reactor Operator job tasks and responsibilities related to 
maintaining overall supervision and coordination of the safety related 
aspects of refueling operations were not identified or included in training.  
The AIT also found the procedure that controlled movement of the UIP during 
the previous refueling outage (May 1989) to be deficient in that (1) it did 
not contain the detailed information necessary to inform refueling person-nel 
on how to move the UIP without bumping it, (2) it contained action steps in 
the form of notes, and (3) information important to the proper completion of 
some procedural steps was included in notes located several pages away from 
the steps.  
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                                                  IN 90-77, Supplement 1
                                                  February 4, 1991
                                                  Page 3 of 3


The AIT also identified an apparent shortcoming in the design of the storage 
stand.  The storage stand has three guide studs, which extend 3 feet 10 
inches above the support stand flange seating surface (see Figure 1).  These 
guide studs provide guidance and protection for critical features of the UIP 
during the lifting and setting-down evolutions involving the storage stand.  
One of the most critical features of the UIP is the FA guide pins.  The 
guide pins are appended to the UIP upper core plate, which is located about 
13 feet below the UIP upper support plate and lifting rig guide bushing.  
Because of the short guide stud design (3 feet 10 inches), when the UIP is 
being lowered onto the storage stand, the guide pins are approximately 8 
feet below the stand flange by the time the lifting rig guide bushings 
engage the stand guide studs.  Similarly, when the UIP is being lifted from 
the storage stand, at the point the lifting rig guide bushings clear the 
stand guide studs, the guide pins, upper core plate, and some portion of the 
upper internal guide tubes are still beneath the stand flange.  
Consequently, the alignment of the stand guide studs and the UIP lifting rig 
guide bushings is not a factor in preventing interference between the 
storage stand flange and the guide pins.

This information notice requires no specific action or written response.  If 
you have any questions about the information in this notice, please contact 
one of the technical contacts listed below or the appropriate NRR project 
manager.




                              Charles E.  Rossi, Director
                              Division of Operational Events Assessment
                              Office of Nuclear Reactor Regulation


Technical Contacts:  Peter C. Wen, NRR
                     (301) 492-0832

                     James A. Prell, RI
                     (215) 337-5108


Attachments:
1.  Figure 1, Simplified Elevated View of West Side of Containment
2.  List of Recently Issued NRC Information Notices 
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