Information Notice No. 84-49: Intergranular Stress Corrosion Cracking Leading to Steam Generator Tube Failure
SSINS No.: 6835
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, DC 20555
June 18, 1984
Information Notice No. 84-49: INTERGRANULAR STRESS CORROSION CRACKING
LEADING TO STEAM GENERATOR TUBE FAILURE
All pressurized water power reactor facilities holding an operating license
(OL) or construction permit (CP).
This information notice is provided, as a notification of potentially
significant problems pertaining to operation and inservice inspections of
steam generators in pressurized water reactor systems. It is expected that
recipients will review their facilities and consider actions, if
appropriate, to minimize similar problems occurring at their facilities.
However, suggestions contained in this information notice do not constitute
NRC requirements and, therefore, no specific action or written response is
Description of Circumstances:
In February 1984, 3 weeks before a scheduled refueling outage, Fort Calhoun
detected a primary leak rate of approximately 0.2 gpd in steam generator B.
In a concerted effort to locate the leak during the outage, the licensee
conducted helium mass spectroscopy indicator tests before and after sludge
lancing. Both tests were unsuccessful in identifying the location of the
leak. A hydrostatic test with a dye indicator also was unsuccessful in
locating the leak.
During the outage, extensive eddy current testing (ECT) was conducted as
part of the licensee's planned inservice inspection program and in support
of a rimcut modification program. The Fort Calhoun Station has two steam
generators, each containing 5,005 Inconel-600 tubes which are 0.75 inch
outside diameter and 0.048 inch minimum wall thickness. Full length
examinations were made of 1,454 tubes in steam generator A and 1,034 tubes
in steam generator B. At the time of the testing, data evaluation detected
only one previously known flaw in steam generator B, A total of nine tubes
were plugged because they would not pass the 0.540 inch ECT probe.
On May 16, 1984, the unit was conducting a hydrostatic test in preparation
for returning to power operation. The cold-leg temperature was 398F.
The reactor coolant system pressure was 1,800 psi and the steam generator
pressure was 200 psi. While plant personnel were closely watching steam
generator B for indications of the small leak experienced before shutdown,
an unanticipated increase
June 18, 1984
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in water level indicated a tube failure. The maximum leak rate was later
estimated at 112 gpm. A high leak rate persisted for approximately 10
minutes, while the RCS pressure was decreased and the main steam line
isolation valve associated with steam generator B was closed.
The failed tube was found in the second peripheral row from the outside. The
failure was a 1 1/4-inch-long- axial "fishmouth" opening along the tube
bottom on the hot-leg side of the horizontal run at the top of the "U". It
was located between the scallop bars in the vertical batwing support.
Sections of the failed tube and adjacent tube were removed for laboratory
Analysis revealed the failure mode to be intergranular stress corrosion
cracking (IGSCC) from the outside, through 95% of the wall thickness, with
the remaining 5% evidencing ductile tearing. The tube cross section was
ovalized, with elongation by 0.046 to 0.122 inch on the major axis (along
the plane of the fracture) and compression of 0.045 to 0.070 inch on the
minor axis. An additional defect, through approximately 50% of the wall, was
found 1/4 inch from the hot leg end of the fishmouth failure. This was
similar to the first defect, except that it was oriented 45 to the tube
axis. Modified Huey tests indicated that the material was not sensitized.
Microstructure was typical of mill annealed Inconel-600. Scanning electron
microscope energy dispersive spectrometry failed to reveal corrosive
chemical deposits, even in the crack tips. There was no evidence of fretting
or wall thinning.
The failed tube was one that had been the subject of eddy current testing
(ECT) in both 1982 and 1984. Review of the ECT tapes of those tests showed
no flaw in 1982 but revealed an indication of a defect through 99% of the
wall in 1984. Although this indication was unambiguous and not affected by
interference, it was missed by the analyst who evaluated the 1984 tapes
before the hydrostatic test. The second defect also was apparent in the 1984
ECT tapes and also was missed.
Prior to restart, the licensee is performing ECT of all tubes in both steam
generators which are accessible with the remote probe insertion machine and
which were not tested in 1984. The licensee will reevaluate, with
independent verification, the ECT data tapes for the tubes already tested in
1984. The licensee has presented test results which indicate that tubes
sufficiently ovalized to obscure serious defects from detection by ECT are
sufficiently restricted to prevent passage of the 0.540 inch ECT probe.
These tubes would be plugged on the basis of their restriction.
Fort Calhoun has always operated with an all-volatile-treatment secondary
chemistry program. ECT examinations were conducted in 1975, 1976, 1977,
1978, 1981, and 1982. Very few degraded tubes were detected over this
period, and the failed tube is the first defective tube detected. ECT
conducted after the tube failure has revealed another tube in steam
generator B with a defect through 42% of the wall in the batwing section on
the hot-leg side of the horizontal run at the top of the bundle. In
addition, two tubes in steam generator A were found to have defects on the
cold-leg side, near the tube sheet: one showed a defect through 39% of the
wall, about 10 inches above the tube sheet;
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the other showed 2 defects 27% and 50% through the wall, 4 inches above the
Although it is likely that the failed tube is the one which was leaking
before the outage, this cannot be known with certainty, until the reactor
returns to power operation. Investigations by Combustion Engineering and
the licensee are continuing in an effort to identify the cause of the IGSCC.
The Nuclear Regulatory Commission is continuing to review the results.
If you have any questions regarding this matter, please contact the Regional
Administrator of the appropriate NRC regional office or this office.
Edward L. Jordan Director
Division of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: S. Long, IE
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