Information Notice 2008-07, Cracking Indications in Thermally Treated Alloy 600 Steam Generator Tubes and Tube-To-Tubesheet Welds
ML080040353 | |
Person / Time | |
---|---|
Issue date: | 04/24/2008 |
From: | Michael Case NRC/NRR/ADRO/DPR |
To: | |
References | |
IN-08-007 | |
Download: ML080040353 (7) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, DC 20555-0001 April 24, 2008 NRC INFORMATION NOTICE 2008-07: CRACKING INDICATIONS IN THERMALLY
TREATED ALLOY 600 STEAM GENERATOR
TUBES
ADDRESSEES
All holders of operating licenses or construction permits for pressurized-water reactors except
those that have permanently ceased operations and that have certified that fuel has been
permanently removed from the reactor vessel.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform
addressees of recent operating experience with degradation in steam generator tubes. The
NRC expects that recipients of this IN will review the information for applicability to their facilities
and consider taking actions, as appropriate, to avoid similar problems. However, suggestions
contained in this IN are not NRC requirements; therefore, no specific action or written response
is required.
DESCRIPTION OF CIRCUMSTANCES
Recent operating experience has indicated the potential for cracking to occur in thermally
treated Alloy 600 steam generator tubes at several nuclear stations, as described below.
Operating Experience at Vogtle Electric Generating Plant (Vogtle), Unit 1 Vogtle, Unit 1, has four Westinghouse Model F recirculating steam generators. Each steam
generator has approximately 5600 tubes fabricated from thermally treated Alloy 600. During
fabrication of the steam generators, the lower portions of the tubes were inserted and
hydraulically expanded into a thick plate called a tubesheet. The tubesheet is approximately
533.4 mm (21 inches) thick and has two holes for each tube (one hole on the hot-leg side of the
steam generator and one hole on the cold-leg side). The tubes were hydraulically expanded for
the full depth of the tubesheet. The transition from the expanded portion of the tube within the
tubesheet to the unexpanded portion of the tube at the top of the tubesheet is referred to as the
expansion transition region of the tube. The tubes are welded to the tubesheet at their lower
ends (See Enclosure, Figure 1).
In 2006, Southern Nuclear Operating Company (the licensee) conducted steam generator tube
inspections at Vogtle, Unit 1. At the time of the inspections, Vogtle Unit 1 had operated for
approximately 17.1 effective full-power years. The steam generators have operated at a hot-leg
temperature of approximately 326 °C (618 °F) since commencement of plant operation.
As a result of +PointTM coil inspections on the hot-leg side of the steam generator from 76.2 mm
(3 inches) above to 76.2 mm (3 inches) below the top of the tubesheet, the licensee identified
18 tubes that had indications attributed to outside-diameter stress-corrosion cracking. Of these
18 tubes, 17 tubes contained circumferentially oriented indications and 1 tube contained an
axially oriented indication.
The circumferentially oriented indications were located at the bottom of the expansion transition
at the top of the tubesheet on the hot-leg side of the steam generator. The axially oriented
indication started in the expanded portion of the tube and stopped at the bottom of the
expansion transition, i.e., most of the indication was in the expanded portion of the tube. The 18 indications were spread among the four steam generators, with most of the indications in steam
generator 4. Most of the circumferential indications were located in tubes with low row numbers
and high column numbers (i.e., in one of the corners of the steam generator). Most of the
indications were confirmed to be present using several different eddy current inspection
probes/coils including the Delta probe, the Ghent probe, the 2.032-mm (0.080-inch) diameter
pancake coil, and the 2.921-mm (0.115-inch) diameter pancake coil. Some of the smaller
amplitude signals were not identified with the 2.032-mm (0.080-inch) diameter pancake coil.
The largest amplitude associated with the circumferential indications was 0.55 volts. The
largest percent degraded area for the circumferential indications was estimated to be
approximately 18 percent. The tube with the largest amplitude signal did not have the largest
percent degraded area. All tubes with circumferential indications were stabilized. The axial
indication was estimated to be 11.684 mm (0.46 inches) long and 68 percent through-wall using
phase angle sizing and 92 percent through-wall using amplitude sizing. The amplitude
associated with the axial indication was 1.77 volts.
The tubes with the circumferentially and axially oriented indications were removed from service
by plugging both ends of the tubes. The licensee concluded that all of the tubes with these
indications had adequate structural and leakage integrity.
Operating Experience at Catawba Nuclear Station (Catawba), Unit 2 Catawba, Unit 2, has four Westinghouse Model D5 recirculating steam generators. Each steam
generator has 4570 tubes fabricated from thermally treated Alloy 600. The design of the tube- to-tubesheet joint is similar to that described above for Vogtle Unit 1.
In 2007, Duke Energy (the licensee) conducted steam generator tube inspections at Catawba, Unit 2. At the time of the inspections, Catawba Unit 2 had operated for approximately
17.4 effective full-power years. The steam generators have operated at a hot-leg temperature
between 324 °C (615 °F) and 326 °C (618 °F) since the commencement of plant operation.
During the inspections, the licensee used an array probe to inspect the tubes. During these
inspections, the licensee identified eight tubes with axially oriented indications. One of these
tubes had multiple axial indications. All of the indications initiated from the outside surface of
the tube on the hot-leg side of steam generator B. The indications were located slightly above
the top of the tubesheet in the sludge pile (a region in the steam generator where deposits tend
to accumulate). The array probe data from the tubesheet region were reviewed for all of the
tubes in steam generator B. In the other three steam generators, the array probe data from the
tubesheet region were reviewed for 20 percent of the tubes with no indications identified at the top of the tubesheet. These eight tubes were removed from service by plugging both ends of
the tubes. All of the tubes with these indications had adequate structural and leakage integrity.
In addition to the indications detected in the tubes near the top of the tubesheet, the licensee
also detected indications near the tube end in both the hot leg and the cold leg. Indications near
the tube end in the hot-leg side of the steam generator had been observed during prior
inspections. There were 10 tubes identified with 15 circumferential indications on the cold-leg
side of steam generators A and D. Nine of these 10 tubes were in steam generator D. These
tubes were left in service since the licensee had been granted an amendment to the plants
technical specifications to leave tubes with flaws near the tube end in service.
Following the 2007 inspection outage at Catawba, Unit 2, eddy current analysts from several
utilities and vendors reviewed the indications identified near the cold-leg tube end at Catawba, Unit 2. These analysts also reviewed data from a mockup of the tube-to-tubesheet joint that
had axial and circumferential notches near the tube end with at least one circumferential notch
in the tube-to-tubesheet weld. Based on this review, the consensus of the industry analysts was
that the cold-leg tube end indications at Catawba, Unit 2 most likely exist in the tube; however, some of the indications extend close enough to the tube end that the possibility that the
indications extend into the tube-to-tubesheet weld could not be ruled out.
BACKGROUND
Before 2006, crack-like indications had been detected in thermally treated Alloy 600 steam
generator tubes at U.S. plants in the tubesheet region and at the tube support plate elevations.
The crack-like indications in the tubesheet region initiated from the inside surface of the tubes
and were either axially or circumferentially oriented. They were located in bulges within the
tubesheet (circumferential) and near the tube end and possibly extending into the tube-to- tubesheet weld (axial and circumferential). Additional information concerning these types of
indications appears in IN 2005-09, Indications in Thermally Treated Alloy 600 Steam Generator
Tubes and Tube-to-Tubesheet Welds, dated April 7, 2005. The crack-like indications at the
tube support plate elevations were axially oriented, were initiated from the outside surface of the
tube, and were associated with nonoptimal tube processing. Additional information concerning
these types of indications appears in IN 2002-21, Supplement 1, Axial Outside-Diameter
Cracking Affecting Thermally Treated Alloy 600 Steam Generator Tubing, dated April 1, 2003.
DISCUSSION
In the United States, 17 units have thermally treated Alloy 600. The steam generators at these
units have been in service for approximately 20 years, on average. In 2002, the first incidence
of corrosion-related cracking was reported in units with thermally treated Alloy 600. This
cracking was attributed to nonoptimal tube processing (refer to IN 2002-21). Since then, a few
other units with thermally treated Alloy 600 tube material have observed crack-like indications in
their steam generators. These crack-like indications have occurred at several different
locations.
Crack-like indications that initiate from the inside surface of the tube have been observed near
the tube end and possibly extending into the tube-to-tubesheet weld and within bulges inside
the tubesheet region. The crack-like indications within the bulges have only been observed on the hot-leg side of the steam generators; however, the crack-like indications near the tube-end
and possibly extending into the tube-to-tubesheet weld have been observed both on the hot-leg
and cold-leg sides of the steam generators.
Crack-like indications that initiate from the outside surface of the tube have been observed in
the expansion transition region, in the portion of tube slightly above the expansion transition, and at the tube support plate elevations (in tubes with nonoptimal tube processing). All of the
indications that initiated from the outside surface of the tube near the top of the tubesheet were
on the hot-leg side of the steam generator. The crack-like indications at the tube support plate
elevations were observed both on the hot- and cold-leg side of the steam generator.
The number of tubes identified with corrosion-related cracking is small in comparison to the
approximately 275,000 thermally treated Alloy 600 tubes in service. Although only a small
number of tubes have been identified with crack-like indications, these findings indicate that the
tubes are potentially susceptible to cracking at a variety of locations. In addition, the rate at
which new cracks are found is expected to increase with time. These findings also illustrate the
importance of carefully evaluating the potential for cracking to occur at other locations. The
potential for cracking depends not only on the tube material, but also on the stresses in the tube
and the operating environment (e.g., water chemistry, temperature). In some instances, it is
difficult to quantify all of these parameters, making it important to have a conservative approach
for inspecting tubes that are susceptible to corrosion-related degradation.
Technical specifications require the steam generator tubes to be inspected, and they require
licensees to perform an assessment to determine the types and locations of flaws to which the
tubes may be susceptible, as well as to determine which inspection methods need to be
employed and at what locations. The objective is to detect flaws of any type that may satisfy the
applicable tube repair criteria. Given the findings discussed above, additional locations along
the length of the tube may need to be inspected for cracking.
CONTACT
This IN does not require any specific action or written response. Please direct any questions
about this matter to the technical contact listed.
/RA by TQuay for/
Michael J. Case, Director
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
Technical Contact:
Kenneth J. Karwoski, NRR
301-415-2752 email: kjk1@nrc.gov
Enclosure: Tube Installed in the Tubesheet
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
CONTACT
This IN does not require any specific action or written response. Please direct any questions
about this matter to the technical contact listed.
/RA by TQuay for/
Michael J. Case, Director
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
Technical Contact:
Kenneth J. Karwoski, NRR
301-415-2752 email: kjk1@nrc.gov
Enclosure: Tube Installed in the Tubesheet
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
Distribution:
IN Reading File
ADAMS Accession Number: ML080040353 OFFICE DCI BC:CSGB:DCI TECH EDITOR D:DCI
NAME KKarwoski AHiser CHsu MEvans
DATE 4/1/08 4/16/08 1/11/08 4/22/08 OFFICE LA:PGCB:DPR PGCB:DPR BC:PGCB:DPR D:DPR
NAME CHawes DBeaulieu MMurphy TQuay for MCase
DATE 4/24/08 4/24/08 04/24/08 04/24/08 OFFICIAL RECORD COPY Enclosure