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HomeNRC LibraryDocument CollectionsGeneric CommunicationsInformation Notices1992

# Information Notice No. 92-80: Operation with Steam Generator Tubes Seriously Degraded

```
                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                            WASHINGTON, D.C.  20555

                               December 7, 1992


NRC INFORMATION NOTICE 92-80:  OPERATION WITH STEAM GENERATOR TUBES SERIOUSLY 
                               DEGRADED


Addressees

All holders of operating licenses or construction permits for pressurized
water reactors (PWRs).

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to inform licensees of recent findings from steam generator (SG) tube
inspections and investigations at Arkansas Nuclear One Unit 2 (ANO-2).  The
Arkansas Power and Light Company, the licensee for ANO-2, found three tubes to
be degraded to the point where they no longer retained adequate structural
margins to sustain the full range of normal operating, transient, and
postulated accident conditions without rupture.  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

In 1978, the NRC licensed ANO-2 for operation.  ANO-2 is a two-loop PWR
designed by Combustion Engineering, Incorporated (CE).  On March 9, 1992, the
licensee shut down ANO-2 upon detecting a primary-to-secondary leak of 
0.95 liters per minute [0.25 gallons per minute]; half of the technical
specification limit.  The licensee conducted an eddy current inspection of the
SG tubes using a motorized rotating pancake coil (MRPC) probe and found the
source of the leak to be a circumferential crack in a tube at the hot leg
expansion transition location, which is near the top of the tubesheet.  The
licensee reviewed the eddy current test data from the previous refueling
outage inspection in 1991 and found that this tube had exhibited a bobbin coil
indication at that time.  Two independent data analysts had missed this
indication.  The licensee found six other bobbin coil indications that the
data analysts had also missed.  The licensee reports that, if these
indications had been correctly analyzed, the licensee would have evaluated
them further.  It is common industry practice to perform supplemental MRPC
inspections (and sometimes pulled tube examinations) to better characterize
low amplitude, ambiguous, or distorted bobbin coil indications.  In 1991, the
licensee did not perform MRPC inspections at the expansion transition
locations of these steam generators.

9212010198.

                                                            IN 92-80
                                                            December 7, 1992
                                                            Page 2 of 4


Because of the finding of the circumferential crack, the licensee conducted a
100 percent MRPC inspection of the expansion transition locations on the hot
leg side of both steam generators and a 20-percent MRPC inspection of the
expansion transition locations on the cold leg side of one steam generator. 
The licensee found indications (generally circumferential) on the hot leg side
of 488 tubes.  The licensee found no indications in the tubes on the cold leg
side.  Tubes with MRPC indications were also inspected with a bobbin probe;
however, this probe did not detect most of the MRPC indications.  The licensee
sleeved 448 of the tubes with indications and plugged and stabilized the
remaining 40 tubes.  The licensee pulled a number of tubes, including three
tubes with circumferential indications in the expansion transition location,
to measure the extent of damage and to determine the degradation mechanism.  

On August 7, 1992, the licensee reported to the NRC the preliminary results
from the examination of the pulled tube segments.  On three of the tubes, the
examination found circumferentially oriented intergranular stress corrosion
cracking (IGSCC) beginning on the outer diameter of the tubes.  The cracks
extended 360 degrees around each of the tubes and had average depths ranging
between 88 to 94 percent of the tube wall thickness.  The licensee reported
that all three tubes failed to satisfy the structural margin criteria of NRC
Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam Generator
Tubes."  Thus, at the time of the March 1992 shutdown, these tubes retained
inadequate structural margins to ensure their integrity for the full range of
normal operating, transient, and postulated accident conditions.

Discussion

The licensee attributed the missed bobbin indications from the 1991 inspection
to (1) a lack of training for the eddy current data analysts in guidelines for
tube damage mechanisms specific to ANO-2 and similar sites; (2) the lack of a
performance demonstration test of the data analysts using actual site data;
and (3) the inherent difficulties in analyzing signals at the expansion
transition locations caused by interference from the tubesheet, the expansion
transition geometry and deposits on the SG tubes.  The indications may also
have been missed if the entire tubesheet entry signal was not screened for
distortions that could indicate flaws at the expansion transition locations. 
Examining the entire tubesheet entry signal is important because the precise
location of the expansion transition in relation to the top of the tubesheet,
and thus the location of the cracks, varies among tubes.

Although the above may have been contributing factors, the NRC staff believes
that the failure to prevent the excessive loss of tube structural margin at
ANO-2 resulted primarily from using test probes inappropriate for tube
locations susceptible to circumferential cracking.  Based on experience at
other CE plants, the tube expansion transition locations in CE steam
generators are known to be susceptible to circumferential IGSCC.  Bobbin
probes are generally insensitive to circumferential IGSCC unless the cracks.

                                                            IN 92-80
                                                            December 7, 1992
                                                            Page 3 of 4


are large enough to have a significant axial component or crack opening. 
Before this condition could occur, the structural margins of the tube would
already be substantially reduced.  

Circumferential cracks can be detected satisfactorily only by specialized
probes such as the MRPC.  This has has been discussed extensively in industry
literature and reported by the NRC.  In Information Notice 90-49, "Stress
Corrosion Cracking In PWR Steam Generator Tubes," the NRC alerted licensees to
problems with such cracks at Millstone Unit 2 and at Maine Yankee.  During the
ANO-2 SG tube inspections in 1991, the licensee did not perform MRPC inspec-
tions of the expansion transition locations.  

Effective inspection programs to ensure early detection of SG tube degradation
can be achieved through the use of data acquisition equipment and procedures
and data analysis procedures appropriate for the detection of potential flaw
types, including circumferential IGSCC.  Training and performance
demonstration testing of the data analysts for all potential flaw types are
important parts of such programs.  Further, by reviewing relevant industry
experience, insights may be gained into the types of flaws that affect a
particular SG as a function of the design or fabrication process.  For
example:  

      Industry reports show that circumferential IGSCC indications have been
      observed at the tube expansion transition locations of SGs at four CE
      plants (including ANO-2) and at several Westinghouse plants.  Tubes in
      the SGs at the four CE plants were explosively expanded against the
      tubesheet using the CE "explansion" process.  Tubes in seven of the
      affected Westinghouse plants with Model 51 SGs were explosively expanded
      against the tubesheet using the Westinghouse WEXTEX process.  Tubes in
      two of the affected Westinghouse plants with Model D SGs were expanded
      using a mechanical rolling process over the full depth of the tubesheet.
      Tubes in another affected Westinghouse plant with Model 51 SGs were
      expanded over only the lower 2-1/2 inches of the tubesheet thickness
      (partial depth expansion) using a mechanical rolling process.

      Widespread circumferential IGSCC has also been observed at tube support
      plate intersections in Westinghouse Model 51 SGs at the North Anna Power
      Station Unit 1 and in the SGs that were replaced at the Palisades
      Nuclear Power Station.  Axial stress caused by denting is believed to
      have caused the cracks in these steam generators.  Denting at the top of
      the tubesheet crevices in Westinghouse Model 51 SGs at the Donald C.
      Cook Plant (Cook) Unit 1 appears to have caused circumferential SCC at
      these locations.  The tubes in the SGs at Cook were partial depth
      expanded against the tubesheet.  Isolated instances of circumferential
      cracks have also been observed at the upper "egg-crate" supports and 
      U-bends in the CE SGs that were replaced at 
.

                                                            IN 92-80
                                                            December 7, 1992
                                                            Page 4 of 4


      the Millstone Nuclear Power Station Unit 2 and at a row 1 U-bend
      in the Westinghouse Model 51 SGs at the Zion Nuclear Power Station 
      Unit 1.

This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
the technical contact listed below or the appropriate Office of Nuclear
Reactor Regulation (NRR) project manager.


                                       ORIGINAL SIGNED BY


                                    Brian K. Grimes, Director
                                    Division of Operating Reactor Support
                                    Office of Nuclear Reactor Regulation


Technical contact:  E. Murphy, NRR
                    (301) 504-2710

Attachment:  List of Recently Issued Information Notices
.
```

*Page Last Reviewed/Updated Friday, May 22, 2015*

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