Bulletin 88-08: Supplement 2, Thermal Stresses in Piping Connected to Reactor Coolant Systems
OMB No.: 3150-0011
NRCB 88-08, Supplement 2
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
August 4, 1988
NRC BULLETIN NO. 88-08, SUPPLEMENT 2: THERMAL STRESSES IN PIPING CONNECTED
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.
Although the actions requested in NRC Bulletin 88-08 are unchanged, reliable
examination of stainless steel piping requires specialized UT techniques.
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
Martin Hum, NRR
Attachment: List of Recently Issued NRC Bulletins
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