UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001

April 1, 2003

NRC INFORMATION NOTICE 2002-21, SUPPLEMENT 1: AXIAL OUTSIDE-DIAMETER

CRACKING AFFECTING

THERMALLY TREATED ALLOY

600 STEAM GENERATOR

TUBING 

Addressees

All holders of operating licensees for nuclear power reactors, 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 supplement to IN 2002-21 to

inform addressees of the root cause assessment for the axially oriented outside-diameter crack

indications in the thermally treated Alloy 600 steam generator (SG) tubing at Seabrook. It is

expected that recipients will review the information for applicability to their facilities and consider

actions, as appropriate, to avoid similar problems. However, suggestions contained in this

information notice are not NRC requirements; therefore, no specific action or written response

is required.

Background

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

in August of 1990. The unit has operated for approximately 10 effective full-power years

(EFPY).

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

quatrefoil broached tube holes.

During the eighth refueling outage, 42 eddy current indications in 15 “low row” tubes (tubes in

rows 1 through 10) were identified and classified as potential axially oriented outside diameter

stress corrosion cracks (ODSCC). All indications were in one steam generator and all

indications were located in the region where the tube passes through a TSP (i.e., tube-to-tubesupport-plate intersection). Both hot and cold leg tubes were affected. No indications were

observed at the top of the tubesheet. This issue was discussed in NRC IN 2002-21, “Axial

Outside-Diameter Cracking Affecting Thermally Treated Alloy 600 Steam Generator Tubing”,

issued June 25, 2002 (ADAMS Accession No. ML021770094).

ML030900517
IN 2002-21, Sup 1

Page 2 of 4

Description of Circumstances

The licensee completed its root cause evaluation, including destructive examination of two

pulled tubes, confirmed that the indications were axially oriented ODSCC, and also identified

unusually high levels of residual stress in the straight leg sections of both the hot and cold legs. 

Nonoptimal tube processing during SG manufacturing was strongly suspected to be the primary

cause of the high residual stresses and the principal factor increasing the susceptibility of the

affected tubes to stress corrosion cracking. The precise processing steps responsible for the

adverse stress state could not be conclusively determined from a review of the tube processing

records. 

Although an aggressive environment, locally created by concentrating chemistry effects in the

crevice region between the tube and the tube support plate, is a necessary contributing factor

for stress corrosion cracking, evidence of abnormal chemistry was not identified and chemistry

is not believed to have been a significant factor in the early onset of stress corrosion cracking at

Seabrook. Seabrook has maintained secondary chemistry in accordance with EPRI guidelines

throughout plant life and has not experienced any major chemical excursions.

The Alloy 600 material in the pulled tubes complied with established chemical limits and the

microstructure, although not optimal, was considered to be representative of thermally treated

Alloy 600 material. Three material heats were identified as being affected (13 of the 15 cracked

tubes were from one heat). Tubes from the affected heats are used throughout the four steam

generators.

Prior to destructive examination, the pulled tubes were pressure-tested. One pulled tube,

containing the largest flaw, was tested to 7000 pounds per square inch (psi) without signs of

leakage: the tube was not tested to burst pressure in order to save the flaw for fractographic

examination. Other tube portions, with and without flaws, were tested to burst pressures

averaging about 11,000 psi.

During the root cause investigation, the licensee noted that the eddy current signature of the

cracked tubes contained a unique offset or shift on the low-frequency (150 KHz) absolute

channel between the straight leg portion of the tube and the U-bend region. This offset was

attributed to changes in the residual stresses in the tube. No offset in the eddy current data

was expected in the low row tubes (i.e., rows 1 through 10) because the U-bend region is

stress-relieved after bending, resulting in consistently low levels of residual stress throughout

the tube. Since testing of the archived material for the heats of material affected by this

cracking found the expected low levels of stress, the licensee attributed the changes in residual

stress levels and the resultant eddy current offset in these tubes to nonoptimal tube processing.

Based upon the above findings, the licensee reviewed the eddy current data from the prior

outage to determine the number of tubes that may have high residual stresses (i.e., exhibit the

offset). This review included not only low row tubes, where the residual stresses are expected

to be consistent throughout the tube, but also the higher row tubes (i.e., those not receiving the

local U-bend stress relief), where the residual stresses are expected to be higher in the U-bend

region (when compared to the straight portion of the tube). Review of the eddy current data

from the tubes in all four steam generators identified 21 tubes, including the 15 tubes with

cracks, which exhibited the eddy current offset. The 15 degraded tubes (including the two

tubes pulled for destructive examination) have been plugged. The six additional tubes identified
IN 2002-21, Sup 1

Page 3 of 4

as having the offset showed no signs of degradation and were also located in the low row tubes

(rows 1 through 10). The licensee indicated that the six tubes would be plugged during the next

outage. The 21 tubes identified with the offset were all located in SG D.

A summary of the licensee’s root cause analysis presentation to the staff and the root cause

analysis report may be found under ADAMS Accession Nos.: ML023300457 and 

ML023240524.

Discussion

The indications of axially oriented ODSCC in thermally treated Alloy 600 tubing at Seabrook,

reported in IN 2002-21, have been confirmed through destructive examination.

Tube cracking at Seabrook was both unexpected and unusual. Thermally treated Alloy 600

material has been successfully used for over 20 years with no prior reports of ODSCC in the

United States. Seabrook has significantly less operating history, roughly 10 effective full-power

years, than other plants with Model F steam generators. The first signs of cracking were 

observed not in the top of the tube sheet region, as would be expected, but in the region where 

the tubes pass through the tube support plates. Historically, cracking has been observed first

at the top of the tubesheet due to increased levels of stress in the expansion transition and the

buildup of contaminants that collect at the top of the tubesheet. The cracking was also identified 

in both the cold and hot legs, which is unexpected because the lower temperatures in the cold

leg typically result in less degradation. Cracking was identified in three material heats, but the

degradation mechanism does not appear to be heat dependent as these heats are used

throughout the steam generators. The licensee has indicated that according to vendor records,

these three heats have been used for steam generator tubes in other PWRs as well.

A unique eddy current signal offset was identified in the cracked tubes. It was reported to result

from high residual stresses caused by nonoptimal tube processing. The high stresses are

principally responsible for creating conditions fostering ODSCC. All tubes were screened for

the signal offset; however, since the magnitude of the eddy current signal is relative, it may be

difficult to adequately screen for susceptibility to ODSCC based on observing an eddy current

offset. That is, tubes with consistently high residual stresses throughout their length may not

display the eddy current offset, and yet these tubes may be susceptible to stress corrosion

cracking.

Heat treatment and tube processing is a special process requiring in-process controls to

provide reasonable assurance of end product quality. Although nonoptimal tube processing is

unexpected with strict in-process controls, problems in manufacturing can occur and could

generically affect mill-annealed Alloy 600, thermally treated Alloy 600, or thermally treated Alloy

690 steam generator tubes.

The unexpected nature of the Seabrook cracking, the potential applicability to other tube

materials, and the ability to screen tubes which may be more susceptible to ODSCC using the

eddy current offset technique illustrates the need for thorough inspections and strong inservice

inspection programs which remain vigilant to the potential for stress corrosion cracking

regardless of the material, location, or steam generator history. This example of unanticipated

cracking should also be considered in determining appropriate frequencies for inspecting the

reactor coolant pressure boundary to ensure that its integrity is maintained consistent with the

plant’s design and licensing basis.
IN 2002-21, Sup 1

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This information notice requires no specific action or written response. If you have any

questions about this notice, contact one of the persons listed below or the appropriate Office of

Nuclear Reactor Regulation (NRR) project manager.

/RA/

William D. Beckner, Program Director

Operating Reactor Improvements Program

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Technical Contact: Martin Murphy, NRR Bob Caldwell, NRR

 301-415-3138 301-415-1243

 E-mail: mcm2@nrc.gov E-Mail: rkc1@nrc.gov

Attachment: List of Recently Issued NRC Information Notices