Information Notice No. 93-20: Thermal Fatigue Cracking of Feedwater Piping to Steam Generators

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                            WASHINGTON, D.C.  20555

                                March 24, 1993

                               PIPING TO STEAM GENERATORS


All holders of operating licenses or construction permits for pressurized
water reactors (PWRs) supplied by Westinghouse or Combustion Engineering.


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to inform addressees of cracks found in the feedwater piping to steam
generators at the Sequoyah Nuclear Power Plant, Units 1 and 2 and the
Diablo Canyon Nuclear Power Plant, Unit 1.  Recipients are expected to 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 1992, cracks were found in the feedwater lines at Diablo Canyon Unit 1.  In
addition, a through-wall crack and other cracks were found in the feedwater
lines at Sequoyah.  The cracks were attributed to thermal fatigue.  

The NRC staff first learned of cracks in feedwater lines to steam generators
which resulted from thermal fatigue in 1979 when the Indiana and Michigan
Electric Company, the licensee for the Donald C. Cook Plant, reported leaks. 
In dealing with this problem, the NRC staff issued the following documents:

o    a letter to PWR licensees pursuant to Paragraph 50.54(f) of Title 10 of
     the Code of Federal Regulations, May 25, 1979

o    Office of Inspection and Enforcement (IE) Bulletin 79-13, "Cracking in
     Feedwater System Piping," June 25, 1979, Revision 1, August 30, 1979, and
     Revision 2, October 17, 1979

o    NUREG/CR-5285, "Closeout of IE Bulletin 79-13," 1991

In Bulletin 79-13, the NRC requested that licensees perform radiographic and
ultrasonic examinations of feedwater lines.  As a result of these
examinations, cracks were found at 18 of the 54 facilities inspected.  The
staff closed the bulletin on the basis of the results of the one-time
inspection.  The industry had taken the actions recommended in the bulletin 


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and instituted the augmented inservice inspection programs, which appeared to
provide for reliable detection of cracks in feedwater piping.  

Other technical evaluations were documented in the following reports:

o    "Investigation of Feedwater Line Cracking in PWRs" (Westinghouse, 1980)

o    NUREG-0691, "Investigation and Evaluation of Cracking Incidents in Piping
     in PWRs," (PWR pipe crack study group, 1980)

These studies showed that thermal fatigue was the main cause of the cracks. 
Modifications to minimize the effects of thermal stratification in feedwater
lines and augmented licensee inservice inspections were recommended in the

Recently, some licensees have again reported cracks in feedwater piping.  The
licensee for the Sequoyah units reported an actual leak despite augmented
inservice inspections.  The augmented inspections using ultrasonic techniques
showed indications that might earlier have revealed the cracks, but the
licensee misinterpreted these as resulting from the geometric configuration of
the pipe.  After finding the leak, the licensee performed radiography on all
feedwater nozzles of both units and found cracks in five of the eight nozzles. 

The licensee for Diablo Canyon Unit 1 reported indications with cracklike
ultrasonic signal characteristics in feedwater piping to all four steam
generators.  The indications varied in length up to 20 cm [7-3/4 inches] in a
circumferential direction, and many were intermittent.  Some intermittent
indications extended the full circumference with segments up to 5 cm 
[2 inches] long.  The licensee tried to verify the indications by radiography
but failed.  Later, metallurgical analysis showed the indications to be

Only one nozzle at Diablo Canyon had been scheduled for an inspection.  This
inspection was performed in accordance with Section XI of the American Society
of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).  However,
information on the leak at Sequoyah led the licensee for Diablo Canyon to use
enhanced ultrasonic techniques to inspect all four lines.  These techniques
were considered more appropriate for finding small cracks from thermal fatigue
than techniques specified by the ASME Code, which may not be adequate to
detect these types of defects.  Flaw sizing by ultrasonic techniques proved to
be overly conservative at Diablo Canyon, however, presumably because
inclusions in the material led to inaccurate results:  cracks sized at 8.9 mm
[0.35 inch] deep by ultrasonic inspection were shown by cross sectioning to be
cracks 0.94 mm [0.037 inch] deep.


Cracks from thermal fatigue in PWR feedwater lines have proved to be a
recurring problem.  The main cause of crack growth appears to be fatigue
induced by stresses from thermal stratification during cold, low-flow,
feedwater injections.  Other factors that contribute to crack growth are a
high oxygen content, counterbore weld preparation geometry, and thermal

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                                                            March 24, 1993
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conditions during heatup, hot standby, and low-power operation.  A favored
solution has been to replace the degraded piping.  However, replacing the
degraded piping in kind without other corrective actions to eliminate or
minimize the factors which cause the cracks can leave the piping susceptible
to the same problem.

Inspection techniques specified in the ASME Code Section XI, do not appear
adequate to find cracks of this type.

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.

                                      ORIGINAL SIGNED BY

                                   Brian K. Grimes, Director
                                   Division of Operating Reactor Support
                                   Office of Nuclear Reactor Regulation 

Technical contacts:  Lee Banic, NRR
                     (301) 504-2771

                     Robert A. Hermann, NRR
                     (301) 504-2768

Attachment:  List of Recently Issued NRC Information Notices

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