UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001

June 25, 2002

NRC INFORMATION NOTICE 2002-21: AXIAL OUTSIDE-DIAMETER CRACKING

AFFECTING THERMALLY TREATED ALLOY 600

STEAM GENERATOR TUBING

Addressees

All holders of operating licenses for pressurized-water reactors (PWRs), except those who have

permanently ceased operations and have certified that fuel has been permanently removed

from the reactor.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to inform

addressees of preliminary indications of axial outside-diameter (OD) cracking of thermally

treated Alloy 600 steam generator (SG) tubing at Seabrook. The NRC anticipates that

recipients will review the information for applicability to their facilities. No specific action or

written response is required.

Description of Circumstances

Seabrook is a four-loop Westinghouse 1198 MWe PWR unit. Commercial operation started in

August of 1990. The unit has operated for approximately 9.67 effective full-power years

(EFPY). 

Seabrook has four Westinghouse Model-F recirculating steam generators (A, B, C, D). Each

steam generator contains 5626 thermally treated Inconel 600 tubes, which are nominally 0.688

inch in diameter with a wall thickness of 0.040 inch. Prior to installation, the tubes in rows 1

through 10 were stress relieved to relieve the stresses from bending the tubes. Each steam

generator contains eight stainless steel tube support plates and six anti-vibration bars in the

U-bend region. The first tube support plate is a partial plate, consisting of only a plate ring with

drilled tube holes. The remaining seven plates contain quatrefoil broached tube holes. Plates

1 through 4 are 0.75 inches thick and plates 5 through 8 are 1.12 inches thick.

Seabrook started its eighth refueling outage in May 2002. During the outage, the licensee

performed steam generator tube inspections as required by the plant technical specifications. 

The initial scope of the inspections included two of the four steam generators, SGs A and D. 

While performing bobbin probe inspections of tubes in steam generator D, the licensee 

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detected indications at a number of tube to tube support plate intersections. Subsequent pluspoint probe inspection confirmed the indications as axially oriented linear indications, initiating

on the OD tube surface. The licensee also performed ultrasonic testing (UT), which further

confirmed the findings. At that stage, the licensee classified these indications as potential axial

OD cracking. 

It was noted that all indications were confined to the intersections of tube and tube support

plate, at the location within the thickness of the tube support plates, opposite the broached tube

hole lands. As of May 23, 2002, the licensee found that 42 of these intersections in 15 tubes

between Row 2 and Row 9 of SG D contain axial OD indications. The indications are located at 

tube support plates 2 through 6 on the hot-leg side and 3 through 5 on the cold-leg side. The

maximum depth of the indications was estimated to be 62 percent through-wall using available

techniques. The lengths ranged from 0.3 to 0.75 inch. 

The licensee completed inspections of SG A and analyzed all the data collected. No similar

degradation was noted in SG A. The inspection scope was expanded to steam generators B

and C. Results of these inspections indicated no similar degradation. Based on the data

analysis, the licensee has concluded that there was no similar degradation detectable in the

other steam generators.

The licensee has pulled two tubes for metallurgical analysis to characterize the apparent

degradation and identify the root cause. These activities are ongoing. The pulled tubes have

four locations with indications, including the one with the maximum measured depth. All

affected tubes were plugged.

Discussion

Most steam generators placed into service before the early 1980's used tubing fabricated from

mill-annealed Alloy 600. This tubing was found to be susceptible to degradation, including

stress corrosion cracking. This degradation became extensive and has led to steam generator

replacement at more than 30 units to date in the United States. The safety significance is that if

tube degradation is not monitored, repaired, and managed appropriately, it may lead to primaryto-secondary leakage and/or a tube rupture. 

To reduce susceptibility to stress corrosion cracking, many steam generators placed into

service during the 1980s were made of thermally treated Alloy 600 tubing. Operating

experience has confirmed the superior corrosion resistance of thermally treated Alloy 600

tubing. Until the recent findings at Seabrook, no known or likely instances of stress corrosion

cracking affecting thermally treated Alloy 600 tubing had been reported in the United States.1 2 

__________________________

1. Stress corrosion cracking in thermally treated Alloy 600 tubing has been reported at foreign units.

2. In a few prior cases indications found in thermally treated Alloy 600 tubing at U.S. plants could not be

ruled out as stress corrosion cracks; however, the general industry consensus is that these

indications do not represent stress corrosion cracking (i.e., they were inconclusive, fabrication

anomalies, or false calls).
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Currently, 17 U.S. plants have steam generators with thermally treated Alloy 600 tubing. 

Another plant has thermally treated Alloy 600 tubes in the first 10 rows of tubes (the remaining

tubes are mill-annealed Alloy 600). Steam generators with thermally treated Alloy 600 tubes

have had as much as 20 calendar years and over 15 EFPYs of total service time. Seabrook

has operated for only about 10 EFPYs. The Seabrook findings underscore the importance of

being alert during inspections to evidence of possible stress corrosion cracking, regardless of

how long the steam generators have been operating.

If you have any questions about the information in this notice, please contact one of the

technical contacts listed below or the appropriate project manager in the NRC’s Office of

Nuclear Reactor Regulation (NRR).

/RA/

William D. Beckner, Program Director

Operating Reactor Improvements Program

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Technical contacts: Z. Bart Fu, NRR Emmett L. Murphy, NRR

301-415-2467 301-415-2710

Email: zbf@nrc.gov Email: elm@nrc.gov

Attachment: List of Recently Issued NRC Information Notices