Information Notice 2002-21, Axial Outside-Diameter Crackling Affecting Thermally Treated Alloy 600 Steam Generator Tubing
| ML030900517 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 04/01/2003 |
| From: | Beckner W NRC/NRR/DRIP/RORP |
| To: | |
| Caldwell, R, NRC/NRR/DRIP/RORP, 415-1175 | |
| References | |
| IN-02-021, Suppl 1 | |
| Download: ML030900517 (6) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
April 1, 2003
NRC INFORMATION NOTICE 2002-21, SUPPLEMENT 1:
AXIAL OUTSIDE-DIAMETER
CRACKING AFFECTING
THERMALLY TREATED ALLOY
600 STEAM GENERATOR
TUBING
Addressees
All holders of operating licensees for nuclear power reactors, except those who have
permanently ceased operations and have certified that fuel has been permanently removed
from the reactor vessel.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this supplement to IN 2002-21 to
inform addressees of the root cause assessment for the axially oriented outside-diameter crack
indications in the thermally treated Alloy 600 steam generator (SG) tubing at Seabrook. 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.
Background
Seabrook is a four-loop Westinghouse 1198 MWe (PWR) unit. Commercial operation started
in August of 1990. The unit has operated for approximately 10 effective full-power years
(EFPY).
Seabrook has four Westinghouse Model F recirculating steam generators (A, B, C, D). Prior to
installation, the tubes in rows 1 through 10 were stress-relieved to relieve the stresses from
bending the tubes. Each steam generator contains eight stainless steel tube support plates
and six antivibration bars in the U-bend region. The first tube support plate is a partial plate, consisting of only a plate ring with drilled tube holes. The remaining seven plates have
quatrefoil broached tube holes.
During the eighth refueling outage, 42 eddy current indications in 15 low row tubes (tubes in
rows 1 through 10) were identified and classified as potential axially oriented outside diameter
stress corrosion cracks (ODSCC). All indications were in one steam generator and all
indications were located in the region where the tube passes through a TSP (i.e., tube-to-tube- support-plate intersection). Both hot and cold leg tubes were affected. No indications were
observed at the top of the tubesheet. This issue was discussed in NRC IN 2002-21, Axial
Outside-Diameter Cracking Affecting Thermally Treated Alloy 600 Steam Generator Tubing, issued June 25, 2002 (ADAMS Accession No. ML021770094).
IN 2002-21, Sup 1
Description of Circumstances
The licensee completed its root cause evaluation, including destructive examination of two
pulled tubes, confirmed that the indications were axially oriented ODSCC, and also identified
unusually high levels of residual stress in the straight leg sections of both the hot and cold legs.
Nonoptimal tube processing during SG manufacturing was strongly suspected to be the primary
cause of the high residual stresses and the principal factor increasing the susceptibility of the
affected tubes to stress corrosion cracking. The precise processing steps responsible for the
adverse stress state could not be conclusively determined from a review of the tube processing
records.
Although an aggressive environment, locally created by concentrating chemistry effects in the
crevice region between the tube and the tube support plate, is a necessary contributing factor
for stress corrosion cracking, evidence of abnormal chemistry was not identified and chemistry
is not believed to have been a significant factor in the early onset of stress corrosion cracking at
Seabrook. Seabrook has maintained secondary chemistry in accordance with EPRI guidelines
throughout plant life and has not experienced any major chemical excursions.
The Alloy 600 material in the pulled tubes complied with established chemical limits and the
microstructure, although not optimal, was considered to be representative of thermally treated
Alloy 600 material. Three material heats were identified as being affected (13 of the 15 cracked
tubes were from one heat). Tubes from the affected heats are used throughout the four steam
generators.
Prior to destructive examination, the pulled tubes were pressure-tested. One pulled tube, containing the largest flaw, was tested to 7000 pounds per square inch (psi) without signs of
leakage: the tube was not tested to burst pressure in order to save the flaw for fractographic
examination. Other tube portions, with and without flaws, were tested to burst pressures
averaging about 11,000 psi.
During the root cause investigation, the licensee noted that the eddy current signature of the
cracked tubes contained a unique offset or shift on the low-frequency (150 KHz) absolute
channel between the straight leg portion of the tube and the U-bend region. This offset was
attributed to changes in the residual stresses in the tube. No offset in the eddy current data
was expected in the low row tubes (i.e., rows 1 through 10) because the U-bend region is
stress-relieved after bending, resulting in consistently low levels of residual stress throughout
the tube. Since testing of the archived material for the heats of material affected by this
cracking found the expected low levels of stress, the licensee attributed the changes in residual
stress levels and the resultant eddy current offset in these tubes to nonoptimal tube processing.
Based upon the above findings, the licensee reviewed the eddy current data from the prior
outage to determine the number of tubes that may have high residual stresses (i.e., exhibit the
offset). This review included not only low row tubes, where the residual stresses are expected
to be consistent throughout the tube, but also the higher row tubes (i.e., those not receiving the
local U-bend stress relief), where the residual stresses are expected to be higher in the U-bend
region (when compared to the straight portion of the tube). Review of the eddy current data
from the tubes in all four steam generators identified 21 tubes, including the 15 tubes with
cracks, which exhibited the eddy current offset. The 15 degraded tubes (including the two
tubes pulled for destructive examination) have been plugged. The six additional tubes identified
IN 2002-21, Sup 1 as having the offset showed no signs of degradation and were also located in the low row tubes
(rows 1 through 10). The licensee indicated that the six tubes would be plugged during the next
outage. The 21 tubes identified with the offset were all located in SG D.
A summary of the licensees root cause analysis presentation to the staff and the root cause
analysis report may be found under ADAMS Accession Nos.: ML023300457 and
Discussion
The indications of axially oriented ODSCC in thermally treated Alloy 600 tubing at Seabrook, reported in IN 2002-21, have been confirmed through destructive examination.
Tube cracking at Seabrook was both unexpected and unusual. Thermally treated Alloy 600
material has been successfully used for over 20 years with no prior reports of ODSCC in the
United States. Seabrook has significantly less operating history, roughly 10 effective full-power
years, than other plants with Model F steam generators. The first signs of cracking were
observed not in the top of the tube sheet region, as would be expected, but in the region where
the tubes pass through the tube support plates. Historically, cracking has been observed first
at the top of the tubesheet due to increased levels of stress in the expansion transition and the
buildup of contaminants that collect at the top of the tubesheet. The cracking was also identified
in both the cold and hot legs, which is unexpected because the lower temperatures in the cold
leg typically result in less degradation. Cracking was identified in three material heats, but the
degradation mechanism does not appear to be heat dependent as these heats are used
throughout the steam generators. The licensee has indicated that according to vendor records, these three heats have been used for steam generator tubes in other PWRs as well.
A unique eddy current signal offset was identified in the cracked tubes. It was reported to result
from high residual stresses caused by nonoptimal tube processing. The high stresses are
principally responsible for creating conditions fostering ODSCC. All tubes were screened for
the signal offset; however, since the magnitude of the eddy current signal is relative, it may be
difficult to adequately screen for susceptibility to ODSCC based on observing an eddy current
offset. That is, tubes with consistently high residual stresses throughout their length may not
display the eddy current offset, and yet these tubes may be susceptible to stress corrosion
cracking.
Heat treatment and tube processing is a special process requiring in-process controls to
provide reasonable assurance of end product quality. Although nonoptimal tube processing is
unexpected with strict in-process controls, problems in manufacturing can occur and could
generically affect mill-annealed Alloy 600, thermally treated Alloy 600, or thermally treated Alloy
690 steam generator tubes.
The unexpected nature of the Seabrook cracking, the potential applicability to other tube
materials, and the ability to screen tubes which may be more susceptible to ODSCC using the
eddy current offset technique illustrates the need for thorough inspections and strong inservice
inspection programs which remain vigilant to the potential for stress corrosion cracking
regardless of the material, location, or steam generator history. This example of unanticipated
cracking should also be considered in determining appropriate frequencies for inspecting the
reactor coolant pressure boundary to ensure that its integrity is maintained consistent with the
plants design and licensing basis.
IN 2002-21, Sup 1 This information notice requires no specific action or written response. If you have any
questions about this notice, contact one of the persons listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
/RA/
William D. Beckner, Program Director
Operating Reactor Improvements Program
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical Contact:
Martin Murphy, NRR
301-415-3138
301-415-1243 E-mail: mcm2@nrc.gov
E-Mail: rkc1@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
ML030900517 DOCUMENT NAME: G:\\RORP\\OES\\Staff\\Caldwell\\IN 2002-21 Sup 1 Outside Diameter cracking\\IN 2002-21 Sup 1 Tech Ed 1 31 03.wpd
OFFICE
OES:RORP:DRIP
Tech Editor
DE:EMCB
DE:EMCB
DE:EMCB
NAME
RKCaldwell
PKleene
MCMurphy
KJKarwoski
LALund
DATE
03/03/2003
01/30/2003
03/06/2003
03/06/2003
03/14/2003 OFFICE
BC:DE:EMCB
LPD1:DLPM
SC:OES:RORP:DRIP
PD:RORP:DRIP
NAME
WHBateman
JWClifford
TReis
WDBeckner
DATE
03/28/2003
03/28/2003
03/31/2003
04/01/2003
______________________________________________________________________________________
OL = Operating License
CP = Construction Permit
Attachment
IN 2002-21, Sup 1 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
_____________________________________________________________________________________
Information
Date of
Notice No.
Subject
Issuance
Issued to
_____________________________________________________________________________________
2003-04 Summary of Fitness-For-Duty
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All holders of operating licensees
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All holders of operating licenses
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2002-36
Incomplete or Inaccurate
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Contractor or Subcontractor
Employee
12/27/2002
All materials and fuel cycle
licensees and certificate holders.
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