Information Notice 1995-40, Supplemental Information to GL-95-03, Circumferential Cracking of Steam Generator Tubes
| ML013100290 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 09/20/1995 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| +sunsimjr=200611, -RFPFR, FOIA/PA-2001-0256, GL-95-003 IN-95-040 | |
| Download: ML013100290 (3) | |
http://nrrIO.nrc.gov/gencoms/ins/in 9 5040.htm
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
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, the licensee tested for leaks and found four leaking tubes. IN 94-88, "Inservice
Inspection
Deficiencies Result in Severel
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
3 situ pressure testing, three of which were from the sample subjected to in situ pressure testing in 1994 and
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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 thettubes 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 pancakecoil; the most senisitiVe of the coils
to the degradation at Maine Yankee, the indications on the pulledtbesweresizd with maximum
through-wall depths of 36, 32, and 44 percent, and average depths of 30, 21 and 27 p*ercent, 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 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.Thehigh-frecquenýypancake coil is, in. generalm.
more sesitive.1hinuie staidad ancake coil
to cracks in-itiat-mingalt-th~e innr i~r..ý,The'reulfs~fthi
coiaionndii-ifdha h
hkruan
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
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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