Bulletin 88-08: Supplement 2, Thermal Stresses in Piping Connected to Reactor Coolant Systems

                                                  OMB No.: 3150-0011
                                                  NRCB 88-08, Supplement 2

                                UNITED STATES
                           WASHINGTON, D.C.  20555

                                 August 4, 1988

                                       TO REACTOR COOLANT SYSTEMS


All holders of operating licenses or construction permits for 
light-water-cooled nuclear power reactors.


This supplement emphasizes the need for enhanced ultrasonic testing (UT) and 
for experienced examination personnel to detect cracks in stainless steel 
piping.  No new requirements are included in this supplement.

Description of Circumstances:

On the basis of changes in containment atmospheres at Farley 2 and Tihange 1, 
operators found leakage of reactor coolant from cracks in the first upstream 
elbow of emergency core coolant system (ECCS) piping connected to the reactor 
coolant systems.  The cracked pipe at both plants was fabricated from 6-inch, 
type 304, stainless steel components, except for a check valve body at 
Tihange 1 that was cast, type 316, stainless steel.  At Farley 2, the 
through-wall crack was in the upstream weld and in the heat-affected zones on 
both sides of the weld.  At Tihange 1, the through-wall crack was in the base 
metal of the elbow.  Other cracks at Tihange 1 were found in the pipe spool 
connected to one side of the elbow and in the body of the check valve 
connected to the other side.  The maximum depth of these cracks was 30 
percent of the wall thickness.  During repair of the piping, cracks in the 
check valve body were found by using dye-penetrant testing, and the depth was 
determined by grinding.

At Farley 2, the weld that failed had been examined on April 17, 1986, as 
part of the inservice inspection program using the UT technique required by 
Section XI of the ASME Boiler and Pressure Vessel Code.  No reportable flaw 
indications were found.  The same UT procedure was used again after the plant 
was shut down on December 9, 1987, and again no rejectable flaw indications 
were reported.  After supplementing the UT technique with a 60-degree shear 
wave transducer and increasing the gain with the 45-degree transducer by 8 
db, the through-wall crack was identified.  To detect the through-wall crack 
and other cracks in the Tihange 1 elbow and spool, an instrumentation gain 24 
db higher than ASME Code sensitivity was required.

.                                                  NRCB 88-08, Supplement 2 
                                                  August 4, 1988 
                                                  Page 2 of 2 


The experience at Farley 2 and Tihange 1 indicates that problems could exist 
with detection of thermal fatigue cracks in stainless steel piping, fittings, 
and welds.  For the UT procedure to reliably detect these cracks, the 
practices that were found to provide reliable detection include (1) using 
sufficient instrument gain so that cracks can be distinguished from 
non-relevant reflectors, (2) using multiple-angle beam transducers on 
surfaces that have geometric discontinuities or weld conditions that limit 
scanning, (3) recording any indication of a suspected flaw regardless of 
amplitude, and (4) using examination personnel with demonstrated ability to 
detect and evaluate cracked stainless steel welds.  

Personnel training and experience are important considering the elevated 
scanning sensitivity and the reliance on signal interpretation for reporting 
and characterizing flaws.  The examination procedure describes the acceptance 
standards and methodology for sizing flaw indications in order to establish 
actual or conservative flaw dimensions.  A UT procedure that has been shown 
to be capable of detecting and sizing intergranular stress corrosion cracking 
at boiling water reactors has been demonstrated to be effective in detecting 
thermal fatigue cracks. 

Actions Requested:

Although the actions requested in NRC Bulletin 88-08 are unchanged, reliable 
examination of stainless steel piping requires specialized UT techniques.  

Reporting Requirements:

The reporting requirements set forth in NRC Bulletin 88-08 remain unchanged.

If you have any questions regarding this matter, please contact one of the 
technical contacts listed below or the Regional Administrator of the 
appropriate regional office.

                               Charles E. Rossi, Director
                               Division of Operational Events Assessment
                               Office of Nuclear Reactor Regulation

Technical Contacts:  Roger W. Woodruff, NRR
                     (301) 492-1180

                     Martin Hum, NRR
                     (301) 492-0932

Attachment:  List of Recently Issued NRC Bulletins 


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