ML080040353 

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. 
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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 tubeto-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 
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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 plant’s 

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-totubesheet 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 
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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. 
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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. 
IN 2008-07 

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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 
IN 2008-07 

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Enclosure