Information Notice 2010-21, Crack-Like Indication in the U-bend Region of a Thermally Treated Alloy 600 Steam Generator Tube

Crack-Like Indication in the U-bend Region of a Thermally Treated Alloy 600 Steam Generator Tube

ML102210244
Person / Time
Issue date:	10/06/2010
From:	Mcginty T Division of Policy and Rulemaking
To:
russell andrea
References
IN-10-021
Download: ML102210244 (5)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, DC 20555-0001 October 6, 2010

NRC INFORMATION NOTICE 2010-21: CRACK-LIKE INDICATION IN THE U-BEND

REGION OF A THERMALLY TREATED

ALLOY 600 STEAM GENERATOR TUBE

ADDRESSEES

All holders of an operating license or construction permit for a nuclear power reactor issued

under Title 10 of the Code of Federal Regulations, Part 50, Domestic Licensing of Production

and Utilization Facilities, 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 information notice (IN) to inform

addressees of recent operating experience with the detection of a crack-like indication in the

U-bend region of a thermally treated Alloy 600 steam generator tube. The NRC expects that

recipients will review the information for applicability to their facilities and consider actions, as

appropriate, to avoid similar problems. Suggestions contained in this IN are not NRC

requirements; therefore, no specific action or written response is required.

DESCRIPTION OF CIRCUMSTANCES

Vogtle Electric Generating Plant (Vogtle), Unit 1, has four Westinghouse Model F steam

generators. Each steam generator has 5,626 tubes fabricated from thermally treated Alloy 600.

The tubes have a nominal outside diameter of 0.688 inch and a nominal wall thickness of

0.040 inch. Following the bending of the tubes, the U-bend region of the tubes installed in the

first 10 rows of the steam generator was thermally stress-relieved at approximately

1,320 degrees Fahrenheit (715 degrees Celsius) for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to relieve the residual stresses

induced by bending. The radius of a Row 1 U-bend is 2.20 inches. The tubes are arranged in a

square pattern, and the centers of the adjacent tubes are 0.98 inch apart.

When Vogtle, Unit 1 was shut down for the 2009 outage, the unit had operated for

approximately 19.8 effective full-power years. The hot leg temperature is approximately

618 degrees Fahrenheit (326 degrees Celsius).

During the cycle before the 2009 refueling outage, no primary-to-secondary leakage was

observed at Vogtle, Unit 1; however, while the unit was being shut down, there were a few

radiation monitor alarms, indicating the presence of activity on the secondary side of the unit.

The leak rate was too small to measure. With the static pressure from the water on the

secondary side of the steam generator acting on the tubes, there was no visible evidence of

leakage coming from any of the tubes.

For this refueling outage, the licensee planned to use an eddy current rotating probe equipped

with a +PointTM coil to inspect the U-bend region of approximately 50 percent of the tubes in

Rows 1 and 2 in each of the four steam generators. During these rotating probe examinations, the licensee found an axially oriented crack-like indication approximately 3 inches above the

uppermost hot-leg tube support plate (i.e., the seventh support plate) in the tube in Row 1, Column 20 (R1C20), in Steam Generator 3. The crack-like indication was close to the apex of

the tube on the extrados and was attributed to primary water stress-corrosion cracking. The

voltage amplitude of the indication, as measured with the midrange +PointTM coil, operated at

300 kilohertz (kHz) was 3.09 volts. The axial length of the indication was 0.54 inch, its

maximum depth was 100 percent through-wall, and it exceeded 80 percent through-wall over

0.40 inch of its length.

As a result of detecting this indication, the rotating probe examinations in the U-bend region

were expanded to include 100 percent of the Row 1 and 2 tubes in each of the four steam

generators and 20 percent of Row 3 tubes in Steam Generator 3. No other crack-like

indications were found in the U-bend region during these examinations. The licensee plugged

the tube in R1C20 in Steam Generator 3.

The U-bend region of the tube in R1C20 in Steam Generator 3 was inspected in prior outages.

The U-bend region was inspected with a bobbin coil in 1986 (baseline inspection), 1991, 1993,

1997, and 2000. A rotating probe examination of the U-bend region of this tube was also

performed in 1997, 2003, and 2009. The practice of inspecting the U-bend region of tubes in

low rows (e.g., Rows 1 and 2) with only a rotating probe (and not with a bobbin probe) is a fairly

common practice today, given the difficulties in performing quality inspections of tight radius

U-bends with a bobbin probe.

The prior inspections of this tube indicated the presence of a manufacturing indication, referred

to as a Blairsville bump (because the bump was most likely introduced during bending of the

tube at a facility in Blairsville, PA). This bump is located at the start of the bent region of the

tube (i.e., the start of the U-bend region). During the review of the bobbin coil data obtained in

2000, one of the analysts reviewing the data (typically two analysts review all eddy current data)

identified a nonquantifiable indication at the location where the crack-like indication was

eventually discovered. This indication was eventually dismissed by the resolution analyst (an

analyst who oversees the review of the primary and secondary data analysts) because the 1997 rotating probe examination indicated that no flaws were present at this location, the bobbin coil

data indicated that the signal had not changed since the 1986 inspection, and there was a

general absence of any cracking in tubes fabricated from thermally treated Alloy 600 tubing at

the time of the inspection. During the review of the 2003 rotating probe data, an axial indication

was reported by one of the analysts at the location where the crack-like indication was

eventually discovered. This indication was also dismissed by the resolution analyst because the

indication from the rotating probe did not change appreciably from 1997 (1.75 volts as

measured from the 300-kHz channel) to 2003 (1.83 volts as measured from the 300-kHz

channel).

The tube in R1C20 in Steam Generator 3 was pressure-tested in situ in 2009 to confirm its

integrity. During the test, no leakage was observed under simulated normal operating conditions, approximately 0.002 gallons per minute was observed under simulated steamline

break conditions, and approximately 0.09 gallons per minute was observed at three times the

normal operating differential pressure. The tube did not burst at three times the normal

operating differential pressure, demonstrating that it retained adequate structural integrity. The

leak rate from this tube under accident conditions, when combined with all other sources of

accident-induced leakage, was within acceptance limits, demonstrating that the steam generator

had satisfied the accident-induced leakage performance criteria.

BACKGROUND

Previous related generic communications include the following:

• NRC IN 2003-13, Steam Generator Tube Degradation at Diablo Canyon, dated

August 28, 2003 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML032410215)

• NRC IN 97-26, Degradation in Small-Radius U-Bend Regions of Steam Generator

Tubes, dated May 19, 1997 (ADAMS Accession No. ML031060007)

DISCUSSION

Cracking in the U-bend region of steam generator tubes has been observed for years. Many of

these cracks have been located in the U-bends of tubes in Rows 1 and 2 as discussed in NRC

IN 97-26; however, there have been some cracks in the U-bends of higher row tubes as

discussed in NRC IN 2003-13. All of these cracks have been in tubes made with mill-annealed

Alloy 600 tubing. (Note: Although NRC IN 97-26 indicates that an axial indication was

discovered in the U-bend region of a thermally treated Alloy 600 tube at Braidwood Station, Unit 2, subsequent evaluation indicated that this indication was not a crack).

The findings at Vogtle, Unit 1 are particularly noteworthy because this is the first confirmed

instance of cracking in the U-bend region of a thermally treated Alloy 600 tube. Although there

was a manufacturing indication near the location where the crack developed, it is not known

whether this condition is necessary to lead to the initiation of a crack in this region of the tubing.

In addition, although the crack at Vogtle, Unit 1 occurred in the U-bend region of a Row 1 tube, it is not clear that this phenomenon would be primarily limited to low-row (i.e., Rows 1 and 2)

tubes as it primarily was for units with mill-annealed Alloy 600 steam generator tubing. In the

units with mill-annealed Alloy 600 tubing, the U-bends were typically not stress relieved after

bending. As a result, the residual stresses; and therefore, the susceptibility to cracking, would

normally be expected to be highest in the lowest row tube (i.e., the one with the shortest bend

radius), provided there were no significant changes in the bending processes from one tube row

to another. However, in the case of the units with thermally treated Alloy 600 tubing, the

U-bends of the low-row tubes (i.e., Rows 1 through 8, 9, or 10, depending on the model) were

stress relieved after bending. Assuming the stresses were fully relieved as a result of this

process, the resistance to cracking should generally be the same if all other factors (such as

material microstructure) are the same. In 2003, the licensee concluded that there was no flaw at the location where the crack was

eventually observed in 2009 in the tube in R1C20 in Steam Generator 3. Its conclusion was

based on observing only a minor change in the eddy current signal (bobbin and rotating probe)

between 1997, 2000, and 2003 and the general lack of cracking in units with thermally treated

Alloy 600 tubing. Following the discovery of the crack in 2009, the NRC staff reviewed the 2003 and 2009 rotating probe eddy current data for the tube in R1C20. Although the NRC staff did

not have all of the information available to the licensee, the NRC staffs review of the 2003 rotating probe data indicated the presence of a flaw-like signal. These results highlight the

limitation of confirming flaw signals based on signals exhibiting change from one inspection to

the next and the difficulties in detecting new or unexpected forms of degradation. Detecting

primary water stress-corrosion cracking (i.e., cracking that initiates from the inside diameter of

the tubing) may be enhanced by using smaller probes operated at high frequencies. Although

the noise levels in the data obtained at higher and lower frequencies may be comparable, the

signal level associated with a flaw on the inside diameter of the tube may be significantly greater

at the higher frequencies. This will increase the signal-to-noise level for this type of flaw and

improve the likelihood that it would be detected.

As discussed above, at least one of the eddy current data analysts during the 2000 and 2003 inspections identified a signal at the location where the crack was eventually observed in 2009.

These signals were dismissed by a resolution analyst. To limit the potential for human error and

to improve detection of flaws, some licensees require two independent resolution analysts to

review the signals identified by the primary and secondary data analysts rather than assigning

this review to a single resolution analyst. In addition, when dismissing indications based on a

historical review of the data, two independent analysts reviewing the data may improve the

detection of flaws or changing conditions.

CONTACT

This IN requires no specific action or written response. Please direct any questions about this

matter to the technical contact listed below or to the appropriate Office of Nuclear Reactor

Regulation (NRR) project manager.

/RA/ by TQuay for

Timothy J. McGinty, Director

Division of Policy and Rulemaking

Office of Nuclear Reactor Regulation

Kenneth J. Karwoski, NRR

301-415-2752 E-mail: kenneth.karwoski@nrc.gov

Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections. In 2003, the licensee concluded that there was no flaw at the location where the crack was

eventually observed in 2009 in the tube in R1C20 in Steam Generator 3. Its conclusion was

based on observing only a minor change in the eddy current signal (bobbin and rotating probe)

between 1997, 2000, and 2003 and the general lack of cracking in units with thermally treated

Alloy 600 tubing. Following the discovery of the crack in 2009, the NRC staff reviewed the 2003 and 2009 rotating probe eddy current data for the tube in R1C20. Although the NRC staff did

not have all of the information available to the licensee, the NRC staffs review of the 2003 rotating probe data indicated the presence of a flaw-like signal. These results highlight the

limitation of confirming flaw signals based on signals exhibiting change from one inspection to

the next and the difficulties in detecting new or unexpected forms of degradation. Detecting

primary water stress-corrosion cracking (i.e., cracking that initiates from the inside diameter of

the tubing) may be enhanced by using smaller probes operated at high frequencies. Although

the noise levels in the data obtained at higher and lower frequencies may be comparable, the

signal level associated with a flaw on the inside diameter of the tube may be significantly greater

at the higher frequencies. This will increase the signal-to-noise level for this type of flaw and

improve the likelihood that it would be detected.

As discussed above, at least one of the eddy current data analysts during the 2000 and 2003 inspections identified a signal at the location where the crack was eventually observed in 2009.

These signals were dismissed by a resolution analyst. To limit the potential for human error and

to improve detection of flaws, some licensees require two independent resolution analysts to

review the signals identified by the primary and secondary data analysts rather than assigning

this review to a single resolution analyst. In addition, when dismissing indications based on a

historical review of the data, two independent analysts reviewing the data may improve the

detection of flaws or changing conditions.

CONTACT

This IN requires no specific action or written response. Please direct any questions about this

matter to the technical contact listed below or to the appropriate Office of Nuclear Reactor

Regulation (NRR) project manager.

/RA/ by TQuay for

Timothy J. McGinty, Director

Division of Policy and Rulemaking

Office of Nuclear Reactor Regulation

Kenneth J. Karwoski, NRR

301-415-2752 E-mail: kenneth.karwoski@nrc.gov

Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.

DISTIBUTION:

DLR/RARB Allen Hiser

ADAMS Accession Number: ML102210244 TAC ME4181 OFFICE NRR:DCI TECH EDITOR D:DCI

NAME KKarwoski (AJohnson for) KAzariah-Kribbs MEvans

DATE 09/29/10 08/16/10 e-mail 10/04/10

OFFICE PM:PGCB:DPR LA:PGCB:DPR BC:PGCB:DPR D:DPR

NAME ARussell CHawes SRosenberg TMcGinty

DATE 10/04/10 10/4/10 10/5/10 10/6/10

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