United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 96-09, Supplement 1: Damage in Foreign Steam Generator Internals

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

                                 July 10, 1996



All holders of operating licenses or construction permits for pressurized-
water reactors (PWRs). 


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to the most recent findings of damage in 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. 


The NRC issued Information Notice (IN) 96-09 to alert addressees to findings
of damage in steam generator internals, namely support plates and wrapper, at
foreign PWR facilities.  NRC has since learned of additional reports of damage
mechanisms affecting the tube support plates in foreign steam generators.

Description of Circumstances

Foreign authorities have reported different tube support plate damage
mechanisms affecting foreign units.  The affected steam generators are similar
but not identical to Westinghouse Model 51 steam generators.  As previously
documented in IN 96-09, the first of these damage mechanisms involved wastage
of the uppermost support plate at a foreign facility, which was caused by
misapplication of a chemical cleaning process.  Also, U.S. industry
representatives previously stated that chemical cleanings performed in the
U.S. involve different cleaning agents and inhibitors than those used at the
foreign facility and involve less risk for producing similar damage.  

The second damage mechanism, also previously reported, involves broken tube
support plate ligaments which affect the uppermost and, sometimes, the next
lower tube support plates.  The staff of at least eight foreign facilities
have identified this damage mechanism in steam generators similar but not
identical to Westinghouse Model 51 steam generators.  These broken support
plate ligaments occur near a radial seismic restraint and near an antirotation
key and appear to date back to the initial startup of the affected plants. 
According to foreign authorities, the broken ligaments may be due to excessive 

9607030266.                                                            IN 96-09, Supp. 1
                                                            July 10, 1996
                                                            Page 2 of 3

stress in the ligaments during the final thermal treatment of the monobloc
steam generators, which is due, in turn, to a lack of clearances for
differential thermal expansion between the support plates, wrapper, and
seismic restraints.  

IN 96-09 did not previously report a third damage mechanism to tube support
plates.  This mechanism appears to involve wastage that is not associated with
chemical cleaning but that affects the tube support plates at a variety of
elevations, including the flow distribution baffle plate.  This phenomenon has
affected support plate ligaments at four foreign plants with steam generators
similar but not identical to Westinghouse Model 51 steam generators.  The
number of affected tube-to-support plate intersections ranged from 4 at one
plant to 1500 at another.  This damage mechanism is active (progressive) and
apparently involves a corrosion or erosion-corrosion mechanism of undetermined

The staff of potentially affected foreign units are currently inspecting steam
generators for the various damage mechanisms, both visually and with eddy
current.  Tubes without adequate lateral support are being plugged.  

NRC has also learned that cooling transients involving injection of a large
quantity of auxiliary feedwater may be a key causal factor that led to the
steam generator wrapper drop phenomenon observed at a foreign PWR facility and
discussed in IN 96-09.  These cooling transients are believed to have been
particularly severe for two specific foreign units as a result of the use of a
special operating procedure to accelerate the transition from hot to cold
shutdown.  The weight of the wrapper assembly and support plates is borne by
six tenons mounted on the steam generator shell.  The wrapper is nominally
free to expand axially relative to the shell.  However, it is postulated that
an interference fit developed between the wrapper and the seismic restraints
(mounted to the shell) as a result of differential thermal expansion
associated with the cooling transients at the seventh support plate elevation. 
This interference fit prevented axial expansion of the wrapper, which led to
excessive vertical bearing loads at the tenon supports, thus causing localized
wrapper failure at this location and downward displacement of the wrapper 
(20 millimeters, maximum).  Poor quality wrapper support welds may also have
contributed to this failure.  Temporary repairs have been implemented at the
affected foreign PWR facility, and visual inspections are being performed at
similar steam generators.


As previously noted in IN 96-09, tube support plate signal anomalies found
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.  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.  Vibration-induced fatigue .                                                            IN 96-09, Supp. 1
                                                            July 10, 1996
                                                            Page 3 of 3

could be a potential problem if tube support plates are lost, particularly in
areas of high secondary side cross flows. 

The foreign experience highlights the potential for degradation mechanisms
that may lead to support plate damage and/or cracking.  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.  Video camera inspections on the secondary side of the
steam generators may also provide useful information concerning the nature of
support plate degradation.

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
there are 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

                                    Brian K. Grimes, Acting Director
                                    Division of Reactor Program Management
                                    Office of Nuclear Reactor Regulation

Technical contacts:  Sheri R. Peterson, NRR
                     (301) 415-1193
                     E-mail:  srp@nrc.gov

                     Eric J. Benner, NRR
                     (301) 415-1171
                     E-mail:  ejb@nrc.gov
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