United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 97-26: Degradation in Small-Radius U-Bend Regions of Steam Generator Tubes

WASHINGTON, D.C. 20555-0001

May 19, 1997

                               REGIONS OF STEAM GENERATOR TUBES


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


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to disseminate information about recent degradation affecting small-
radius (rows 1 and 2) U-bend regions of tubes in recirculating steam
generators (SGs), in order to alert utilities to potential problems in
ensuring the integrity of the small-radius U-bends, and to provide information
about action taken by certain licensees to ensure adequate integrity.  It is
expected that recipients will review the information for applicability to
their facilities and consider this information, as appropriate, in their SG
inspection programs.  However, suggestions contained in this information
notice are not NRC requirements; therefore, no specific action or written
response is required.

Description of Circumstances

Licensees that use Westinghouse-designed recirculating SGs have for many years
identified indications in the U-bend regions of tubes with small radii. 
During the late 1970s and early 1980s, many units, as a  preventative measure,
plugged small-radius U-bend tubes to avoid potential forced outages due to
leakage.  However, some licensees subsequently unplugged these tubes and
performed in situ stress relief to reduce the susceptibility for degradation. 
Also SG designs evolved over time and a number of different material
conditions are represented in currently operating PWRs.  These include mill-
annealed alloy 600, mill-annealed alloy 600 in situ stress relieved, thermally
treated alloy 600, and thermally treated alloy 690.  The following discussion
of experience at four plants represents recent operating experience regarding
U-bend degradation that involved various tube material conditions.
During a 1996 inspection, Commonwealth Edison Company (ComEd) identified a
total of 64 axially oriented and 2 circumferentially oriented indications in
the U-bends of the row 1 SG tubes at Zion Unit 2.  ComEd characterized the
indications as primary water stress-corrosion cracking.  The tubes at Zion
Unit 2 were fabricated with mill-annealed alloy 600 material, and the U-bends
had not been heat treated.  As a result of the inspection findings, ComEd
preventively plugged all the row 1 tubes at Zion Unit 2.

9705140336.                                                            IN 97-26
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                                                            Page 2 of 4

During a 1996 inspection, the Tennessee Valley Authority (TVA) identified
axial indications in 17 small-radius U-bends of SG tubes at Sequoyah Unit 2
and characterized the degradation as primary water stress-corrosion cracking. 
The tubes were fabricated with mill-annealed alloy 600 material and the
U-bends were in situ heat treated during the cycle 6 outage in 1994.  TVA
plugged the 17 row 1 tubes that contained the U-bend indications.
During 1992 and 1995 inspections, Pacific Gas & Electric Company (PG&E)
identified  circumferential indications having relatively small arc angles in
the small-radius U-bends  of SG tubes at Diablo Canyon Unit 1.  The tubes were
fabricated with mill-annealed alloy  600 material and the small-radius U-bends
were in situ heat treated after the second refueling outage in 1988.  PG&E
plugged the degraded tubes.

During a 1996 inspection, ComEd identified a single axial indication in the
U-bend of one of the SG tubes at Braidwood Unit 2.  The Braidwood Unit 2 tubes
were fabricated with thermally treated alloy 600 tubes and the U-bends in the
first seven rows received additional thermal stress relief after bending
during the manufacturing process.  ComEd plugged the degraded tube.   

A small number of axial indications originating on the outside diameter of the
tubes have been reported in the small-radius U-bend regions of the SGs at Palo
Verde 1, 2, and 3 and St. Lucie 1.  These SGs were designed by Combustion

U-bend degradation has occurred in mill-annealed alloy 600 tubes irrespective
of whether they have been heat treated.  Tubes with thermally treated alloy
600 material are less susceptible to degradation than mill-annealed alloy 600
tubes.  However, thermally treated alloy 600 tubes have also begun to
experience U-bend degradation.  None of the degraded thermally treated alloy
600 tubes have been removed from SGs for confirmation of the degradation
mechanism.  Reports of U-bend degradation have been based on eddy current
inspection results.  The susceptibility to cracking in small-radius U-bends
and the findings of recent field inspections have emphasized the importance of
inspection of this area of SGs with techniques capable of accurately detecting
U-bend degradation.   

U-bend degradation can potentially impair tube integrity if not effectively
managed.  Concerns in this regard stem from limitations of eddy current
testing to detect and size U-bend cracks, the potential for some U-bend cracks
to have relatively long lengths, and the potential for high crack growth rates
for some of these cracks.  The industry standard bobbin coil has proven
unreliable for detecting U-bend cracks and, in addition, is not qualified for
this application under the Electric Power Research Institute (EPRI) technique
qualification protocol.  The industry has developed special probes for these
inspections.  The industry has qualified a rotating pancake coil and a Plus
Point coil for detecting indications in small-radius U-bends, in accordance
with enhanced qualification criteria developed by EPRI.    
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                                                            Page 3 of 4

There continues to be an absence of pulled tube information to confirm that
the detection threshold for these cracks is better than 40 or 50-percent
through wall.  In addition, available inspection techniques are not capable of
reliably sizing crack depths and, for this reason, it has been industry's
practice to "plug on detection" U-bend indications that are found.

Information available on crack growth rates being experienced in the field is
very limited by virtue of the inability to perform reliable crack depth
measurements and the resulting need to "plug on detection."  However, U-bend
cracks have led to leakage as early as the first cycle of operation and, thus,
crack growth rates may potentially be high for some cracks.  Given the
relatively high detection thresholds, the relatively long operating cycles,
and the potentially high growth rates, the depth of cracks may be in excess of
50-percent through wall when they are first detected.

In view of these concerns, effective management of the degradation of SG tubes
is important to ensure that adequate tube integrity is being maintained in
accordance with  10 CFR Part 50, Appendices A and B.  One such approach being
implemented by a number of licensees involves the use of tube integrity
assessments to ensure that inspection sensitivity to U-bend cracks and the
frequency and scope of inspection are sufficient to ensure that U-bend flaws
are being detected and removed from service before tube integrity is impaired. 

For example, ComEd performed in situ pressure tests at Zion Unit 2 on four
tubes having  the longest axial U-bend indications and on two tubes with
circumferential U-bend indica- tions using pressure loading consistent with
the margins recommended in Regulatory Guide  (RG) 1.121, "Bases for Plugging
Degraded PWR Steam Generator Tubes."  The two tubes  having circumferential
indications satisfied RG 1.121 margins without leaking.  Three of the four
tubes having axial indications leaked at a pressure of main steamline break
conditions but did not burst under a pressure loading of three-times-normal
operating pressure.  Because of the limitations of the test equipment, the
pressure in the fourth tube did  not reach the three-times-normal operating
pressure criterion of RG 1.121.  For this tube, ComEd performed analyses to
show that the tube would not burst under a pressure loading of three-times-
normal operating pressure.  These analyses are based on eddy current test
measurements.  Since these measurements may have large uncertainties, ComEd
conservatively assumed that the cracks were 100-percent through wall.  On the
basis  of the leakage measurements at main steamline break pressures, ComEd
was able to demonstrate that accident leakage would satisfy the requirements
of 10 CFR Part 100.  

For U-bend indications at Sequoyah Unit 2, TVA did not perform in situ
pressure testing; instead, it performed bounding analyses to show that the
three tubes having the largest U-bend cracking satisfied RG 1.121 criteria. 
However, it should be noted that in situ pressure testing provides more
definitive assurance of structural and leakage integrity than analyses.  

For axial indications in the small-radius U-bend regions of the SGs at Palo
Verde 1, 2, and 3 and St. Lucie 1, the licensees plugged the tubes..                                                            IN 97-26
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As shown by the examples discussed above, the integrity of the small-radius
U-bend regions can be more fully ensured by efforts that include performing
inspections of rows 1 and 2 U-bends using qualified eddy current techniques;
performing in situ pressure testing, as necessary, to assess the condition of
defective tubes; taking appropriate corrective actions, including plugging
defective tubes; and assessing the appropriate operating intervals until the
next SG tube inspection.  

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 contacts list below or the appropriate Office of Nuclear Reactor
Regulation (NRR) project manager.

                                          signed by S.H. Weiss for

                                    Marylee M. Slosson, Acting Director
                                    Division of Reactor Program Management     
                                    Office of Nuclear Reactor Regulation

Technical Contact:  John C. Tsao, NRR           
                    (301) 415-2702        
                    E-mail:  jct@nrc.gov        

                    Eric J. Benner, NRR
                    (301) 415-1171
                    E-mail:  ejb1@nrc.gov
Page Last Reviewed/Updated Tuesday, December 03, 2013