Information Notice No. 96-38: Results Of Steam Generator Tube Examinations

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                         WASHINGTON, D.C.  20555-0001

                                 June 21, 1996


NRC INFORMATION NOTICE 96-38:  RESULTS OF STEAM GENERATOR TUBE EXAMINATIONS 


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 promulgate information about steam generator tube examinations.  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

Improved techniques and equipment are constantly developed to detect flaws in
steam generator tubes.  In addition, as nuclear power plants get older,
different degradation mechanisms of steam generator tubes occur.  This
information notice discusses recent experiences by licensees involving these
new techniques and equipment and different degradation mechanisms.

Recent steam generator tube examinations have revealed degradation at a number
of locations, such as in dented areas, the expansion transition region, the
freespan region, and in the tubesheet crevice.  The types of degradation
observed in these locations are discussed below.  In addition to identifying
several degradation mechanisms, these examinations raised a number of
technical issues with respect to classifying inspection results, periodicity
of examinations, and expanding the initial inspection scope.

Axial and circumferential indications at dented tube support plates were
identified at a number of plants, including Sequoyah Nuclear Plant Unit 1,
Diablo Canyon Nuclear Power Plant Unit 1, and Salem Generating Station Unit 1. 
These indications are associated with minor dents (i.e., dents that can be
inspected with a standard size probe).  These dented regions were examined
with Cecco probes or rotating probes with plus-point coils or pancake coils
(or both).  On the basis of the examinations, the axial indications appear to
have initiated from the inside diameter of the tube, and the circumferential
indications appear to have initiated from the outside diameter of the tube. 
However, at Diablo Canyon Unit 1, several circumferential indications have
initiated from the inside diameter of the tube (as evidenced by destructive
examination).


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Some plants that have Combustion Engineering and Westinghouse-designed steam
generators also reported circumferential indications at the expansion
transition region.  Among these are Sequoyah Nuclear Plant Unit 1, Diablo 
Canyon Unit 1, Salem Unit 1, Arkansas Nuclear One Unit 2, Braidwood Unit 1,
Byron Unit 1, and Callaway Unit 1.  At particular plants, from tens to
thousands of indications were reported.

The circumferential indications at the expansion transition have occurred at
roll expansions, kinetic/explosive expansions, and hydraulic expansions.  For
example, circumferential indications have been reported in mechanically roll-
expanded tubes at Farley Unit 1, Westinghouse explosively expanded (i.e.,
WEXTEX) tubes at Sequoyah Unit 1, Salem Unit 1, and Diablo Canyon Unit 1,
Combustion Engineering explosively expanded tubes (i.e., EXPLANSION tubes) at
Arkansas Nuclear One Unit 2, and in hydraulically expanded tubes at Callaway
Unit 1.  The majority of these indications were seen at the hot-leg expansion
transition; however, circumferential indications were reported at the cold-leg
expansion transition at Arkansas Nuclear One Unit 2.  The circumferential
cracks detected at these plants were all in Alloy 600 mill-annealed tubes.

Freespan degradation has been reported at a few plants.  Freespan degradation
is degradation observed above the sludge pile region at the top of the
tubesheet and is not located at any support structure (e.g., tube support
plates including eggcrates, anti-vibration bars, and batwings).  Historically,
moderate amounts of freespan degradation had been observed at McGuire Units 1
and 2 and at Palo Verde Units 1, 2, and 3.  During the fall outages, Arkansas
Nuclear One Unit 2, Farley Unit 1, and Point Beach Unit 1 reported freespan
tube degradation.  In addition, Oconee Units 1, 2, and 3 reported freespan
axial indications attributed to intergranular attack.

A few plants have tubes which are only partially expanded in the tubesheet. 
As a result, there is a crevice between the tube and the tubesheet for the
portion of the tube in the tubesheet that is not expanded.  Corrosion products
can accumulate in this crevice and can lead to tube degradation. Historically,
tubesheet crevice region defects have been observed with the bobbin coil and
repaired, accordingly; however, many of the indications detected during
outages this fall were not found with the conventional bobbin coil probe.  As
a result, extensive examinations using alternate techniques were performed
(e.g., rotating pancake coil examinations).  Extensive tube repairs were
performed, such as sleeving at Zion Unit 1 and tube rerolling at Point Beach
Unit 1.

Discussion

Steam generators with mill-annealed Alloy 600 steam generator tubes are
susceptible to such degradation as stress corrosion cracking.  Degradation has
been observed in the hot legs and cold legs of the steam generator tubes, in
the expanded portion of the tube, at the expansion transition, in the
tube-to-tubesheet crevice, in the sludge pile, in the freespan, and at tube
support structures such as the tube support plate, batwings, anti-vibration
bars, and vertical straps.  The severity of the degradation and the number of
tubes affected tend to be plant specific since these depend on many factors .                                                            IN 96-38
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such as temperature, operating time, water chemistry history, and tube
mechanical properties, including microstructure.  Inspections have illustrated
the importance of comprehensive steam generator tube examinations using
appropriate techniques to ensure tube integrity even if a specific type of
degradation has not been observed at a given location in the past.  Previous
inspection findings do not ensure that a location/tube is not susceptible to a
particular mechanism.  For example, before the inspections at Callaway Unit 1,
no circumferential cracking had occurred domestically at tubes which had been
hydraulically expanded within the tubesheet.  The inspections at Callaway 
demonstrate that continually assessing the condition of all portions of the
steam generator tube can ensure that new forms of degradation are detected.

The recent inspections also indicate the importance of comprehensively
examining all portions of the steam generator tubes using techniques and
equipment capable of reliably detecting degradation to which the steam
generator tubes may potentially be susceptible.  This experience calls into
question the effectiveness of the bobbin coil for detecting circumferential
indications or for detecting indications where significant interfering signals
exist (e.g., expansion transition locations, dented locations, and locations
with excessive deposits), as discussed in NRC Information Notice 94-88,
"Inservice Inspection Deficiencies Result in Severely Degraded Steam Generator
Tubes."  In addition, this experience further indicates that a generically
qualified technique may need to be supplemented to account for the testing
conditions at a specific plant.  Furthermore, optimizing such test variables
as probe design and frequencies for the type of degradation observed at the
plant such as inside-diameter initiated indications versus outside-diameter
initiated indications, and controlling such test variables as cable length and
capacitance within the range for which the technique was qualified can be
important in ensuring the reliable detection of degradation.

Several large indications were detected during the most recent examinations of
steam generator tubes.  As a result, several licensees took additional
measures to ensure that all tubes were capable of withstanding the pressure
loadings specified in Regulatory Guide 1.121, "Bases for Plugging Degraded PWR
Steam Generator Tubes."  These additional measures (in situ pressure testing
and removing tubes for destructive examination) were performed even though
many of these indications were repaired.  Although methods other than removing
tubes for destructive examination exist for evaluating tube integrity, tube
removal has the advantage of assessing inspection reliability, developing
additional confidence in the ability to size indications, determining the root
cause of the degradation, and possibly identifying corrective actions. 
Assessment of the inspection findings after every inspection assures that all
tubes are capable of performing their intended safety function for the planned
operating interval.  In some instances, these assessments have led to
mid-cycle inspections. 

When degraded tubes are left in service (i.e., for degradation mechanisms for
which qualified sizing techniques exist), assessment of the acceptable
operating interval typically involves a detailed knowledge of the growth rate
of the degradation, the scope of the examination, and the capabilities of the
inspection technique..                                                            IN 96-38
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For degradation mechanisms for which there is no qualified depth sizing
technique, a tube with an indication typically has been considered defective. 
In these instances, demonstrating that the largest indications detected during
an inspection were capable of withstanding specified pressure loadings
(through such techniques such in-situ pressure testing or burst and leakage
testing or both) can provide assurance that tubes currently without
indications will also be capable of withstanding specified pressure loadings
at the end of the next inspection interval, if the interval is of comparable
duration and operating parameters (e.g., water chemistry and hot leg
temperature) to the previous inspection interval.

Although only steam generators that contain tubes made from mill-annealed
Alloy 600 are discussed above, the information may have applicability to all
PWRs.  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.


                                          signed by C.L. Miller

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

Technical contacts:  Kenneth J. Karwoski, NRR
                     (301) 415-2754
                     Internet:kjk1@nrc.gov

                     Eric J. Benner, NRR
                     (301) 415-1171
                     Internet:ejb1@nrc.gov