Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
February 12, 1996
NRC INFORMATION NOTICE 96-09: DAMAGE IN FOREIGN STEAM GENERATOR INTERNALS
Addressees
All holders of operating licenses or construction permits for pressurized
water reactors (PWRs).
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to recent findings of damage to steam generator
internals, namely support plates and wrapper, at foreign PWR facilities. It
is expected that recipients will 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 April 1995 during a routine eddy current inspection of the steam generator
tubing at a foreign facility, anomalous support plate signals were observed at
the uppermost support plate. The steam generators are similar but not
identical to Westinghouse model 51 steam generators. The support plates are
of the drilled hole type and fabricated from carbon steel. Video camera
inspections were conducted to investigate the anomalous signals and revealed
that a significant portion of the support plate had wasted away. Pieces of
the affected region of the support plate were found resting on the next lower
support plate.
Subsequent investigation has identified chemical cleaning performed in 1992 as
the cause of the support plate damage. Review of previous eddy current data
shows that the anomalous support plate signals were present in inspections
dating back to 1993 when the first inservice inspection following chemical
cleaning was performed. Support plate signals obtained immediately prior to
the chemical cleaning were normal. The foreign regulatory authority believes
that pipes used to direct the chemical solution into the steam generators were
installed incorrectly, too close to the upper support plate. This caused an
excessively high impingement velocity of the cleaning solution against the
support plate which is believed to have been sufficient to render ineffective
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the corrosion inhibitor in the cleaning solution. U.S. industry representa-
tives stated during recent meetings with the NRC staff that chemical cleanings
which have been performed in the U.S. involve different cleaning agents and
inhibitors than that used at the foreign facility and involve less risk for
producing similar damage.
The support plate damage at the foreign facility effectively eliminated
lateral support to tubes within the affected region. Lateral support provides
vibrational stability and the ability to sustain earthquake and loss-of-
coolant-accident loadings. Accordingly, all tubes found not to be supported
at the uppermost support plate were plugged.
Based on this experience, the foreign utility carefully examined the support
plate eddy current signals at other PWR facilities. At one of these units,
with steam generators similar but not identical in design to Westinghouse
model 51 steam generators, eddy current signals indicative of support plate
ligament cracks were found at the uppermost support plate. The support plates
are of the drilled hole type and are fabricated from carbon steel. Subsequent
visual inspection confirmed the presence of ligament cracks near the periphery
of the support plate. Part of the support plate periphery was observed to be
entirely broken away in the vicinity of a radial seismic support. The steam
generators at this facility have not been chemically cleaned. Review of past
eddy current results indicates that the indications of ligament cracks date
back at least 9 years. It is not clear whether the ligament cracks were
present prior to initial service or whether the cracks may have developed
shortly thereafter. The cause of these cracks is under investigation by the
foreign utility and steam generator manufacturer. Tubes whose lateral support
was potentially affected by these cracks have been plugged. Press reports
indicate that similar indications of support plate ligament cracks have
recently been found at other facilities in the same country with similar steam
generators.
Visual inspections conducted in June 1994 at a foreign PWR facility revealed
the bottom of the wrapper had dropped down by 20 millimeters in one steam
generator and by 5 millimeters in another steam generator. The steam
generators are similar but not identical to Westinghouse model 51 steam
generators. The visual inspections were performed through handholes located
above the tubesheet. Further investigation revealed that wrapper welds at
each of six vertical supports in the first steam generator and at three of six
vertical supports in the second steam generator had failed, allowing the
downward displacement of the wrapper. The cause of this occurrence is under
investigation by the foreign utility and the steam generator manufacturer.
Their preliminary assessment is that unanticipated axial restraint against
differential thermal expansion between the wrapper and steam generator
pressure vessel shell led to significant loading of the wrapper vertical
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supports. This unanticipated restraint between the wrapper and shell may have
been due to differential thermal expansion between support plate number 7 and
the shell, preventing relative axial motion between the wrapper and shell at
this elevation, during transients involving the auxiliary feedwater. Poor
quality of the wrapper welds at the vertical support may also have been a
contributing factor.
Implications of a complete fall of the wrapper have been assessed by the
foreign utility to include the potential for loss of feedwater, damage to the
largest radius tube u-bends, loose parts, and tube rupture. Accordingly, the
foreign utility has implemented temporary repairs to stabilize and monitor the
wrappers pending further investigation regarding long-term resolution of this
matter.
Discussion
As illustrated by the foreign experience, support plate signal anomalies
during eddy current testing of the steam generator tubes may be indicative of
support plate damage or ligament cracking. The signal anomalies at the
foreign units were present for several years before they were first identified
by the data analysts. The Electric Power Research Institute (EPRI) has
initiated an effort, in response to the foreign experience, to develop a
qualified procedure for detecting support plate ligament cracks.
The steam generator tube support plates function to support the tubes against
lateral displacement and vibration and to minimize bending moments in the
tubes during accidents. Damage and/or cracking of the support plates can
impair the ability of the support plates to perform this function and, thus,
may potentially impair tube integrity. In addition, the staff has recently
approved a 3 volt plugging criteria for two U.S. plants based, in part, upon
evidence from inspections using EPRI preliminary procedures that the tube
support plates are capable of locally constraining the tubes against tube
rupture.
Known instances of support plate cracking/damage in the U.S. have generally
involved support plates with significant denting. The potential for support
plate cracks has tended not to be of significant concern in recent years since
the steam generators most affected by denting have been replaced and, in
addition, the industry has been successful in controlling denting progression
in operating steam generators. The foreign experience serves to highlight
that there are other mechanisms which may lead to support plate damage and/or
cracking.
Based on the information available to the NRC staff, it is not yet known
whether steam generators in the U.S. are vulnerable to the type of wrapper
damage observed at the foreign unit.
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February 12, 1996
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The staff will continue to monitor information on support plate and wrapper
damage as it becomes available from foreign authorities.
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.
signed by
Dennis M. Crutchfield, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts: Emmett L. Murphy, NRR
(301) 415-2710
Internet:elm@nrc.gov
Eric J. Benner, NRR
(301) 415-1171
internet:ejb1@nrc.gov
