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HomeNRC LibraryDocument CollectionsGeneral CommunicationsInformation Notices1994

# 

```
                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                             WASHINGTON, DC  20555

                               December 22, 1994


NRC INFORMATION NOTICE 94-87:  UNANTICIPATED CRACK IN A PARTICULAR HEAT 
                               OF ALLOY 600 USED FOR WESTINGHOUSE MECHANICAL
                               PLUGS FOR STEAM GENERATOR TUBES


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 operating experience affecting the
predicted service life of mechanical tube plugs fabricated from alloy 600 and
supplied by Westinghouse.  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.

Background

During the November 1994 refueling outage at St. Lucie Unit 1, licensee
personnel found 15 leaking steam generator tube plugs.  All the affected plugs
were Westinghouse mechanical plugs, fabricated from alloy 600 heat NX-2387,
and installed in 1986.  When the leaking plugs were removed for replacement,
the base of one plug was noted to have a 360 degree circumferential through-
wall crack, evidently due to primary water stress corrosion cracking (PWSCC). 
The crack was located just below the expander and, therefore, outside the
effective pressure boundary of the plug.  The cause of the leakage for the
cracked plug (and for the others, which have no visually detectable cracks) is
being investigated.  The cracked plug and four others will be sent to
Westinghouse for metallurgical examination.

In NRC Bulletin 89-01 and Supplements 1 and 2, the staff discussed the
susceptibility to PWSCC of Westinghouse mechanical plugs made from various
specific heats of alloy 600.  It also discusses the algorithm for determining
when a plug of a specific heat should be preventively removed and replaced
with a new plug.

Discussion

Although a crack in the location noted above is not in the pressure boundary,
its existence raises questions regarding the crack resistance of heat NX-2387. 
This heat of alloy 600 was previously shown, in special steam tests conducted 

9412160203.                                                            IN 94-87
                                                            December 22, 1994
                                                            Page 2 of 2


by Westinghouse, to be highly resistant to PWSCC compared to other heats of
alloy 600.  The algorithm for determining plug life predicted that the
recommended repair date for the St. Lucie plugs made from heat NX-2387 would
be about 10 years in the future.  The recommended repair date is a prediction
of when a plug top release could occur due to a circumferential crack through
the plug pressure boundary.  A pressure boundary crack is one in the portion
of the plug above the expander.  Cracking below the expander is considered a
precursor to cracking above the expander.

Westinghouse is evaluating this event and considering revising the plug life
algorithm.  The event suggests that the life of alloy 600 tube plugs may be
overstated by the current algorithm for all heats of alloy 600, not only for
heat NX-2387.

Cracking below the plug pressure boundary could lead to loose parts in the
steam generator inlet plenum.  The staff is evaluating this issue to determine
if further regulatory action is needed.

As further information concerning this matter becomes available, the staff may
issue revisions to this information notice.

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 Office of
Nuclear Reactor Regulation (NRR) project manager.

                                    /S/'D BY BDLIAW/FOR


                                    Brian K. Grimes, Director
                                    Division of Project Support
                                    Office of Nuclear Reactor Regulation

Technical contacts:  Geoffrey P. Hornseth
                     (301) 504-2756

                     Robert A. Hermann
                     (301) 504-2768

Attachment:
List of Recently Issued NRC Information Notices
```

*Page Last Reviewed/Updated Wednesday, February 16, 2011*

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