United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 96-26: Recent Problems with Overhead Cranes

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

                                April 30, 1996


NRC INFORMATION NOTICE 96-26:  RECENT PROBLEMS WITH OVERHEAD CRANES


Addressees

All holders of operating licenses or construction permits for nuclear power
reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to recent problems with overhead cranes.  It is
expected that recipients will review the information for applicability to
their facilities and consider actions, as appropriate, to avoid similar
problems.  However, suggestions contained in this information notice are not
NRC requirements; therefore, no specific action or written response is
required.

Description of Circumstances

Failure of Overhead Crane Bridge Rail

At the Trojan Nuclear Plant on July 7, 1995, a section of the reactor building
polar crane bridge rail failed.  The rail had a crack across the top of the
top flange and a piece of the flange had been displaced.  The end of one
section of the rail had failed through the plane of the rail joint bar bolts
extending up through the top flange.  Visual and metallographic examination of
the failure plane indicated that much of the failure was preexisting.  Rust on
the failure surfaces and "peening" of some areas indicated that the initial
crack could extend back to the plant's construction.

The licensee research of construction records determined that a nonconformance
report, dated July 26, 1972, noted that the rails were not slotted for bolts
in accordance with the drawings.  The corrective action recommended was to
"burn the slots in the field."  The licensee determined the cause of the
failure to be torsional shear and bending at the stress risers from the flame-
cut holes.  Flame cutting the slots left residual stresses in the material
because of the lack of careful preheating and controlled cooling.  Also, sharp
notches, noted in the area of the flame cutting, concentrated the stresses.  

The inappropriate use of a cutting torch created an untempered martensitic
heat-affected zone in the high-carbon steel rail.  This zone was especially
sensitive to hydrogen cracking and subsequent brittle crack propagation.  The
crack inducing and propagating loading was primarily due to bending of the 


9604260095.                                                            IN 96-26
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rail head to the outside during episodes of rail misalignment.  The licensee
had observed rail misalignment to be a continuing problem that had caused or 
contributed to 19 bridge truck wheel bearing failures over 23 years of
operation.  

The root cause of the failure was the inappropriate use of a cutting torch to
enlarge drilled holes to slots in the web of the rail.  This practice created
an untempered, martensitic, heat-affected zone in the rail material that was
sensitive to hydrogen cracking and subsequent brittle crack propagation.

Actuation of Overhead Crane Safety System

At the Prairie Island Nuclear Generating Plant on May 13, 1995, while lifting
a loaded spent fuel storage cask from the spent fuel pool for transfer to the
transport bay, the single-failure-proof overhead crane handling system
automatically stopped on overload, approximately 13 cm [5 inches] from the
high hook point (peak lift point).  The bottom of the cask was above the water
but approximately 8 cm [3 inches] below the operating deck of the spent fuel
pool.  Upon investigation of the event, the licensee, Northern States Power
Company (NSP) determined that the cause of the event was premature actuation
of the crane overload-sensing system.  The setpoint on the overload-sensing
system was set too low.  Upon actuation of the overload-sensing system,
control power is automatically removed from the hoist motor and the
conventional holding brakes are activated.  Subsequent to the actuation on
May 13, the cask remained in the hoisted position until a safety evaluation
was made that supported bypassing the sensing system and resuming the cask
lift.  The lift was resumed about 16 hours after it was stopped, and the cask
was placed in the decontamination area of the transport bay.  NSP initiated a
root-cause analysis to identify the cause of the actuation.  The conclusion of
this analysis was that the overload-sensing system was inaccurately
calibrated.

This event raises a concern for similarly designed overload-sensing systems
associated with single-failure-proof cranes.  As noted in the analysis
reports, this event was a "nuisance trip" that resulted from inaccurate
initial calibration during load cell setting adjustment.  Improved load cell
accuracy can help to reduce any unbalanced loading condition in the system.  

Additional details of these events can be found in Inspection Report 
No. 50-344/95-06 issued on September 18, 1995, and Inspection Report 
No. 50-282/95-06 issued on June 27, 1995.  .                                                            IN 96-26
                                                            April 30, 1996
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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.  


                                       signed by

                                    Brian K. Grimes, Acting Director
                                    Division of Reactor Program Management
                                    Office of Nuclear Reactor Regulation

Technical contacts:                 

Robert J. Pate, Region IV           Brian E. Thomas, NRR
(510) 975-0246                      (301) 415-1210
Internet:rjp1@nrc.gov               Internet:bet@nrc.gov

David B. Pereira, Region IV         Eric J. Benner, NRR
(510) 975-0307                      (301) 415-1171
Internet:dbp@nrc.gov                Internet:ejb1@nrc.gov

Russell L. Bywater, Region III
(612) 388-8209
Internet:rlb3@nrc.gov



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