United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 95-40: Supplemental Information to Generic Letter 95-03, "Circumferential Cracking of Steam Generator Tubes"

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                         WASHINGTON, D.C.  20555-0001

                              September 20, 1995


NRC INFORMATION NOTICE 95-40:  SUPPLEMENTAL INFORMATION TO GENERIC LETTER
                               95-03, "CIRCUMFERENTIAL CRACKING OF STEAM
                               GENERATOR TUBES"


Addressees 

All holders of operating licenses or construction permits for nuclear power
reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to provide additional information on steam generator tube examination
results from Maine Yankee Atomic Power Station as previously discussed in
Generic Letter (GL) 95-03, "Circumferential Cracking of Steam Generator
Tubes."  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

The staff issued GL 95-03, to obtain information necessary to assess
compliance with requirements regarding steam generator tube integrity in light
of the inspection findings at the Maine Yankee plant.  In GL 95-03, the staff
requested that utilities (1) evaluate recent operating experience with respect
to the detection and sizing of circumferential indications, (2) develop a
safety assessment justifying continued operation until the next scheduled
steam generator tube inspections are performed, and (3) develop plans for the
next inspections of steam generator tubes as they pertain to the detection of
circumferential cracking.  Since the issuance of GL 95-03, additional
information pertaining to in situ pressure testing and destructive analysis
for the tubes removed from the Maine Yankee plant has become available.  In
addition, the wrong title given to NUREG-0844 in GL 95-03 was erroneously
indicated as, "Voltage-Based Interim Plugging Criteria for Steam Generator
Tubes."  The correct title is, "NRC Integrated Program for the Resolution of
Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube
Integrity."

Discussion

On July 15, 1994, Maine Yankee Atomic Power Company, the licensee for Maine
Yankee, shut down the plant when the measured primary-to-secondary leak rate
approached 189 liters [50 gallons] per day.  After shutting down the plant,

9509140386.                                                            IN 95-40
                                                            September 20, 1995
                                                            Page 2 of 3


the licensee tested for leaks and found four leaking tubes.  IN 94-88, 
"Inservice Inspection Deficiencies Result in Severely Degraded Steam Generator
Tubes," discusses in situ pressure testing performed by the licensee in 1994,
on tubes containing some of the largest indications, to assess their actual
burst integrity.  At that time, certain tubes could not be pressurized due to
a combination of leakage and pump capacity limitations, and the staff had not
reached a conclusion regarding the validity of the tests to simulate an actual
pressure transient in the steam generators.

In 1995, the licensee performed additional steam generator inspections.  Seven
tubes were subjected to in situ pressure testing, three of which were from the
sample subjected to in situ pressure testing in 1994 and four of which were
tubes containing some of the largest indications identified at the end of the
1994-to-1995 operating interval.  The testing indicated that the tubes were
capable of withstanding pressure loadings in excess of the loads for which
failure would be predicted on the basis of the size estimates with the
standard pancake coil.  Furthermore, the pressures to which the tubes were
subjected were greater than design-basis loads.  NRC Regulatory Guide 1.121,
"Bases for Plugging Degraded PWR Steam Generator Tubes," indicates that tubes
should be able to withstand "3 times operating pressure" and "1.4 times main
steam line break maximum pressure" without bursting.  At Maine Yankee, 3 times
operating pressure is approximately equal to 34.47 MPa [5000 psi] and 1.4
times main steam line break maximum pressure equals 27.97 MPa [4057 psi].  All
tested tubes at Maine Yankee were subjected to at least 39.30 MPa [5700 psi]
hydrostatic pressure.  Three tubes exhibited no defect leakage and no tubes
burst.  The staff has concluded that these tests adequately bound main steam
line break loads on steam generator tubes.

As stated in GL 95-03, three tubes were removed from the Maine Yankee steam
generators for destructive examination:  two tubes with marginal plus-point
coil responses (sized by the eddy current analysts as probably less than 
40 percent through-wall depth) and one with an intermediate response (sized by
the eddy current analysts as probably greater than 40 percent through-wall
depth).  Before the tubes were removed, they were examined with several
nondestructive methods, such as ultrasonic, fluorescent penetrant, and eddy
current techniques to confirm the nature of the indications.  The eddy current
methods included examination with a standard rotating pancake coil, a
plus-point coil, and a high-frequency pancake coil.  The indications were
sized with various techniques.  The size estimates for the high-frequency
pancake coil and the plus-point coil were obtained after calibration of the
probes on electric discharge-machined (EDM) notches contained within a
standard.  With the high-frequency pancake coil, the most sensitive of the
coils to the degradation at Maine Yankee, the indications on the pulled tubes
were sized with maximum through-wall depths of 36, 32, and 44 percent, and
average depths of 30, 21, and 27 percent, respectively.  The average depth
estimates obtained from the eddy current examination are calculated from the
maximum depth and the circumferential extent by assuming that the maximum
depth is the depth of the degradation over the entire measured circumferential
arc length and averaging this estimate over the entire tube circumference. 
The corresponding destructive examination results for these tubes indicated
that the maximum depths were 45, 37, and 57 percent, with average depths of  .                                                            IN 95-40
                                                            September 20, 1995
                                                            Page 3 of 3


24, 23, and 26 percent, respectively.  The destructive examination of these
tubes indicated that numerous small cracks had initiated at various locations
about the circumference and at various elevations (axial locations) within a
1.27 mm [0.05 inch] band in the "expansion" transition region of the tubes, 
noncorroded ligaments existed between some of the cracks.  The cracks
initiated at the inner diameter of the tubes.  The licensee compared the
sizing of several of the larger indications that were inspected with both a
standard pancake coil and the high-frequency pancake coil.  The high-frequency
pancake coil is, in general, more sensitive than the standard pancake coil to
cracks initiating at the inner diameter.  The results of this comparison
indicated that the maximum and average depths estimated by the high-frequency
pancake coil were consistently lower than the maximum and average depths
estimated with the standard pancake coil even though the length (i.e.,
circumferential extent) estimates were longer with the high-frequency coil.

The smaller depth estimates obtained with the high-frequency coil suggest that
many of the indications may not have been as structurally significant as the
standard pancake coil suggested and as was reported in IN 94-88.  Furthermore,
the destructive examination indicated that the cracks were not coplanar, but
rather of short circumferential length and staggered over a short axial
region.  There were, in fact, ligaments of material between the cracks.  Due
to the nature of this cracking (i.e., the spacing between the cracks), the
ligaments of sound material could not be distinguished by the nondestructive
examination (i.e., standard and high-frequency pancake coil and plus-point
coil) data; however, the nondestructive examination data are conservative in
that the tubes are most likely more structurally sound than estimated by the
eddy current examination.  The observed segmented character of these cracks is
consistent with the results of fluorescent penetrant examination results at
Maine Yankee and with the morphology of circumferential cracks observed on
specimens of tubes pulled from other plants.  

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.

                                    /s/'d by DMCrutchfield


                                    Dennis M. Crutchfield, Director
                                    Division of Reactor Program Management
                                    Office of Nuclear Reactor Regulation

Technical contacts:  Kenneth J. Karwoski, NRR
                     (301) 415-2754

                     Eric J. Benner, NRR
                     (301) 415-1171


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