Information Notice 2008-07, Cracking Indications in Thermally Treated Alloy 600 Steam Generator Tubes and Tube-To-Tubesheet Welds

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)

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. 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