Information Notice No. 90-49: Stress Corrosion Cracking in PWR Steam Generator Tubes
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
August 6, 1990
Information Notice No. 90-49: STRESS CORROSION CRACKING IN PWR STEAM
GENERATOR TUBES
Addressees:
All holders of operating licenses or construction permits for
pressurized-water reactors (PWRs).
Purpose:
This information notice is intended to inform licensees of recent problems
involving stress corrosion cracking (SCC) in PWR steam generator (SG) tubes.
In particular, this information notice is intended to alert licensees to
recent findings at Millstone Unit 2 and to recent problems in SCC detection
during inservice inspections. 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 do not constitute NRC requirements; therefore, no
specific action or written response is required.
Description of Circumstances:
1. Circumferential Cracking at Millstone Unit 2
In October 1989, the licensee for Millstone Unit 2 conducted a mid-cycle
inspection of the SG tubing using eddy current testing (ECT).
Circumferential SCC had been observed in previous inspections to be
affecting the outer diameter (OD) surface (that is, the secondary side) of
the tubes at the expansion transition at the top of the tubesheet. The
mid-cycle inspection followed a previous inspection during the February 1989
refueling outage and was performed, in part, out of concern for the
relatively high rate of SCC growth observed during the previous inspection
and to ensure that SCC did not excessively degrade the integrity of the
tubes. Just before the mid-cycle outage, leakage from the primary side to
the secondary side was less than 5 gallons per day (gpd). The plant's
Technical Specifications limit for such leakage is 144 gpd.
The ECT inspections at Millstone 2 were conducted with a rotating pancake
coil (RPC) probe to ensure optimal sensitivity to circumferential cracks.
Tubes found with circumferential crack indications were also inspected by
ultrasonic testing (UT) to obtain additional information regarding the
length and depth of the cracks. In addition, two tubes with circumferential
crack indications were removed and examined.
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The ECT/RPC inspections revealed 104 tubes with circumferential cracks at
the expansion transition. The macrocracks, as defined by ECT/RPC, consisted
of several discontinuous microcracks that were separated by small ligaments
of sound material. The discontinuous nature of the array of microcracks was
confirmed by the UT and examination of the removed tube specimens. As
measured by UT, the macrocracks ranged in circumference from 84 degrees to
329 degrees and ranged in depth up to 100-percent throughwall.
All tubes with crack indications were staked and plugged. In addition, the
licensee evaluated the residual strength of the cracked tubes to assess
their capability to sustain normal operating and postulated accident
loadings before their removal from service. This structural evaluation
considered the profiles for each crack obtained from the UT examination.
This evaluation revealed one cracked tube which failed to meet the ASME
Code, Section III, NB-3225 and Appendix F stress limits for postulated
accident conditions. (Regulatory Guide 1.121, "Bases for Plugging Degraded
PWR Steam Generator Tubes," states that margins should be consistent with
the stress limits in Section III of the code.) Based on these findings, the
staff concludes that the integrity of the subject tube was not ensured under
postulated accident conditions.
The staff has recently identified service induced, circumferential SCC, such
as at Millstone Unit 2, to be a source of significant degradation to tubes
in PWR steam generators. Such cracking is particularly noteworthy because
it is generally not detectable with conventional bobbin probes used
routinely for inservice inspection. Such cracking is generally only
detectable through the use of specialized probes, such as the RPC probe.
Most circumferential cracking has been observed at tube expansion
transitions at or near the top of the tubesheet. In addition to Millstone
Unit 2, circumferential cracking at the expansion transition has recently
been identified at one other Combustion Engineering (CE) plant (Maine
Yankee), at three plants with Westinghouse Model 51 steam generators (North
Anna Unit 1, Trojan Unit 1, and Sequoyah Unit 1), and at one plant with
Westinghouse Model D steam generators (McGuire Unit 1). Tubes in the
affected CE and Westinghouse Model 51 steam generators were explosively
expanded against the tubesheet. Tubes in the McGuire Model D steam
generators were expanded against the tubesheet by mechanical rolling.
In addition to being found at the expansion transition location, widespread
circumferential SCC has been observed at drilled-hole support plate
locations at Palisades (CE steam generators). Isolated instances of
circumferential SCC have been reported at the uppermost support plate of a
pre-replacement Westinghouse Model 44 steam generator of Indian Point Unit 3
and at a row 1 U-bend of a Model 51 steam generator at Zion Unit 1. The
circumferential SCC at Palisades and Indian Point Unit 3 appears to be
associated with significant denting at the support plates.
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2. Axial Cracking at Support Plates
Licensees have reported secondary side-initiated, axial SCC at several
plants with Westinghouse Model 51 and Model D steam generators at support
plate intersections exhibiting little or no denting. Recent difficulties
experienced in the detection of such cracks were the subject of Westinghouse
Customer Information Letter GEN-LTR-90-006, "Steam Generator Tube Outer
Diameter Stress Corrosion Cracking at Tube Support Elevation-Eddy Current
Detection Issue," which was issued on or about February 8, 1990, to all
utilities with Westinghouse steam generators. Westinghouse reported that
metallographic examinations of tubes removed from the field have revealed
the presence of OD-initiated SCC at tube support plate (TSP) intersections
that were not reported by personnel using a bobbin probe to perform field
eddy current tests. For example, these examinations revealed one tube
containing axial cracks within two 30-degree-wide bands on opposite sides of
the tube, with the deepest crack penetrating to 62-percent throughwall. The
field eddy current interpretation of the signal for this TSP location was
"no detectable degradation" (NDD) using the plant voltage threshold
criteria.
The EPRI "Steam Generator Examination Guidelines, Revision 2" contains
Figure C-55 showing the qualitative relationship between bobbin probe signal
amplitude and crack depth determined metallographically. The staff believes
that the amplitude threshold criteria used at the plant in the
above-mentioned example were taken from the actual data used to develop
Figure C-55. The recent evidence cited by Westinghouse suggests that this
data is not conservative for all plants.
Industry meetings attended by representatives from a number of vendors
providing SG inspection services, EPRI, and the Westinghouse Owners Group
have been conducted to examine various proposals to detect and measure SCC
at TSP locations. The minutes of the EPRI Guidelines Revision Committee
meeting on February 13, 1990, note that general principles still apply,
pending development of updated guidance, for the detection and measurement
of SCC at support plates. Section 4.6.1 of the EPRI guidelines states that
as a general rule, an "extremely conservative position" should be adopted
for the resolution of distorted indications or undefined signals not covered
by existing analysis guidelines. Specifically, tubes with these types of
indications should be recommended for plugging unless other supporting data
exists (tube pulls or NDE diagnostic data) that justifies their retention as
active tubes.
Discussion:
The reliable detection and sizing of SCC during inservice inspections pose a
significant challenge to current ECT technology and practice in view of the
low signal-to-noise ratios frequently exhibited by such cracks. Experience
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indicates that SCC is frequently not detected until it has penetrated beyond
40-percent throughwall. Fortunately, the vast majority of SCC flaws consist
of short axial or circumferential crack segments. The staff believes that
such flaws can be detected before they grow sufficiently large to degrade
the structural margins of the tube to below the Regulatory Guide 1.121
criteria.
Tubes are generally inspected once per refueling cycle. Depending on the
rate of crack growth and the number of tubes involved, this frequency may or
may not be sufficient to ensure that all cracks are detected before they
become sufficiently large to degrade structural margins to less than the
Regulatory Guide 1.121 criteria. A structural assessment of the crack
geometries found during an inspection, such as performed at Millstone Unit
2, provides a means for assessing whether the inspection frequency is
sufficient to ensure adequate structural margins for all tubes between
inspections.
The staff believes that the effectiveness of eddy current testing for
detecting and sizing SCC can be enhanced through improved criteria for the
qualification and performance demonstration of the eddy current data
acquisition equipment (including probes), data analysis procedures, and data
analysts. The staff and the industry, including the EPRI Steam Generator
Reliability Project, are evaluating this issue. In the meantime, field
experience indicates that careful attention to the potential for SCC at the
tubesheet region, at tube expansion transitions, at support plates, and at
tube U-bends is important. The recommendations in Section 4.6.1 of the EPRI
steam generator examination guidelines indicate the importance of being
alert to distorted and undefined signals at these locations and employing
diagnostic measures as appropriate to establish the cause of these signals
and to validate the plant data analysis procedures and criteria.
Finally, the findings at Millstone Unit 2 illustrate that cracks will not
necessarily cause leakage approaching the Technical Specifications limit
before the structural margin in the affected tube drops below the Regulatory
Guide 1.121 criteria for postulated accident conditions. This point is
further illustrated by the steam generator tube rupture (SGTR) event that
occurred at McGuire Unit 1 on March 7, 1989, as a result of axially oriented
SCC. Leakage before the SGTR event was about 15 gpd, which was small
compared to the plant's Technical Specifications limit of 500 gpd. The
McGuire event occurred under normal operating conditions. Thus, the McGuire
event was preceded by a period during which the subject tube was vulnerable
to rupture if challenged by a postulated accident.
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This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
the technical contact listed below or the appropriate NRR project manager.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contact: E. Murphy, NRR
(301) 492-0710
Attachment: List of Recently Issued NRC Information Notices
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