ML120740578

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

OFFICE OF NEW REACTORS 

WASHINGTON, DC 20555-0001 

July 17, 2012 

NRC INFORMATION NOTICE 2012-07: TUBE-TO-TUBE CONTACT RESULTING IN 

WEAR IN ONCE-THROUGH STEAM 

GENERATORS 

ADDRESSEES

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

Title 10 of the Code of Federal Regulations (10 CFR) 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. 

All holders of or applicants for an early site permit, standard design certification, standard 

design approval, manufacturing license, or combined license under 10 CFR Part 52, “Licenses, 

Certifications, and Approvals for Nuclear Power Plants.” 

PURPOSE

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

addressees of the detection of wear indications as a result of tube-to-tube contact in 

once-through steam generators and the lessons learned from the discovery of these indications. 

These lessons learned apply to all steam generator types since they address detection of tube 

degradation. It should be noted that as this IN was being written and issued, there is an 

on-going assessment of wear attributed to tube-to-tube contact that occurred at San Onofre 

Nuclear Generating Station, which uses recirculating steam generators. 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

Three Mile Island, Unit 1 (TMI-1), completed the replacement of both of its original once-through 

steam generators in early 2010, after exiting a refueling outage that began in the fall of 2009. 

AREVA (France) fabricated the replacement steam generators. The tubes in the steam 

generator are straight and supported by 15 tube support plates. The center of the tubes is 

located approximately 15 centimeters (6 inches) below the eighth tube support plate (between 

the seventh and eighth tube support plates). 

The first inservice inspection of the TMI-1 replacement steam generators took place in fall 2011. 

During this inspection, the TMI-1 licensee inspected 100-percent of the tubes in both steam 

generators with a bobbin coil probe, which is an eddy current nondestructive examination 

technique. The bobbin coil probe is operated in both an absolute and differential mode. In the 

differential mode, the probe is sensitive to localized variations along the length of the tube, such 

as cracks or pits. However, this mode is not sensitive to gradual variations along the length of 
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the tubing. In the absolute mode, the probe is sensitive to long and gradual changes along the 

length of the tubing. 

During these inspections at TMI-1, the licensee detected several tubes with indications on the 

absolute channel with no discernable signal being observed on the differential channel. A 

comprehensive review of all of the absolute drift indications revealed that most of them were 

near the middle of the tube’s total length (between the eighth and ninth tube support plate), 

were in a radial pattern approximately 76 centimeters to 114 centimeters (30 inches to 45 

inches) from the center of the steam generator, and in adjacent tubes (two or three). In 

addition, the indications in adjacent tubes faced each other, were at the same elevation, and 

had similar lengths and depths. A more detailed investigation led the licensee to conclude that 

these indications are a result of tube wear due to tube-to-tube contact. The length of these 

wear indications ranged from 5 centimeters to 20 centimeters (2 inches to 8 inches) and the 

depths ranged from 1 percent to 21 percent through-wall. 

As a result of these findings, the licensee for TMI-1 informed the licensee for Arkansas Nuclear 

One, Unit 1 (ANO-1), since ANO-1 has similar steam generators. At the time, ANO-1 was shut 

down for a refueling outage. However, the steam generator tube inspections had been 

completed. 

The original once-through steam generators at ANO-1 were replaced in 2005. AREVA (France) 

also fabricated the replacement once-through steam generators at ANO-1, which are similar 

(but not identical) to those used at TMI-1. Based on the information provided by the licensee for 

TMI-1, the ANO-1 licensee reevaluated its previously recorded eddy current data and 

determined it had similar indications of wear as a result of tube-to-tube contact. Some of these 

indications were traceable to the first inservice inspection of ANO-1’s steam generators in 2007. 

The characteristics of the tube-to-tube wear indications at ANO-1, including the depth and 

length, are similar to those at TMI-1. In addition, two neighboring tubes have wear indications in 

two spans rather than one at ANO-1. 

Given that these tube-to-tube wear indications were not originally classified as tube-to-tube 

wear, the licensee for ANO-1 performed an apparent cause evaluation. They determined 

several contributing factors as to why these indications were not identified: (1) the absence of a 

differential channel response to indicate a flaw-like condition, (2) not reporting, 

mischaracterizing, or deleting the absolute indications by the eddy current analysts, and (3) 

distractions to the analysts because of observing bowing of the tie rods used to support and 

connect the tube support plates. 

Investigations into the cause of the tube-to-tube contact at ANO-1 and TMI-1 are ongoing. 

In the spring of 2012, the licensee for Oconee, Unit 3 also detected wear attributed to 

tube-to-tube contact in their replacement once-through steam generators. These steam 

generators were designed and fabricated by Babcock and Wilcox in Canada and were installed 

in 2004. The AREVA and Babcock and Wilcox once-through steam generators are similar, but 

not identical. The indications of wear attributed to tube-to-tube contact at Oconee, Unit 3 are 

generally located in the center of the tube bundle, the region of highest compression. The 

length of the wear indications ranged from 2.5 centimeters to 23 centimeters (1 inch to 

9 inches), and the depths ranged up to 20 percent through-wall. All but one of the indications of 

wear attributed to tube-to-tube contact were traceable to the first inservice inspection of the 

Oconee, Unit 3’s steam generators in 2006. The licensee indicated that criterion used by eddy 

current analysts to report a tube-to-tube wear indication in prior outages was the indication had 

to have a voltage greater than 0.5 volts on one channel, and the indication’s depth on a second 
IN 2012-07 

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channel had to measure within 10 percent of the through-wall depth measured on the first 

channel. During the 2012 inspection, this criterion was changed to only require that the 

indication be present on the second channel. The licensee believes the tube-to-tube contact is 

due to compression of the tubes in the region where the indications were observed. 

Discussions with the steam generator manufacturer are on-going. 

The licensees for ANO-1, Oconee, Unit 3, and TMI-1 evaluated the severity of the tube-to-tube 

wear indications in their steam generators. These evaluations concluded that the wear 

indications did not compromise tube integrity (i.e., the tubes could still perform their intended 

function consistent with their original design and licensing basis). The licensees also concluded 

that they could operate until their next scheduled inspection with the wear indications left in 

service without compromising tube integrity. 

BACKGROUND

NRC IN 2002-21, “Axial Outside-Diameter Cracking Affecting Thermally Treated Alloy 600 

Steam Generator Tubing,” dated June 25, 2002, highlighted, in part, the importance of being 

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

long the steam generators have been operating. NRC IN 2002-21 can be found on the NRC’s 

public Web site in the Agencywide Documents Access and Management System (ADAMS) at 

Accession No. ML021770094. 

 

NRC IN 2003-05, “Failure to Detect Freespan Cracks in PWR [Pressurized Water Reactor] 

Steam Generator Tubes,” dated June 5, 2003, highlighted, in part, that the bobbin coil eddy 

current data from the absolute channel can be helpful in detecting long freespan indications and 

observing changes in signals over time (ADAMS Accession No ML031550258). 

NRC IN 2010-21, “Crack-Like Indication in the U-Bend Region of a Thermally Treated Alloy 600 

Steam Generator Tube,” dated October 6, 2010, highlighted, in part, difficulty in detecting new 

or unexpected forms of degradation (ADAMS Accession No. ML102210244). 

DISCUSSION

Technical specifications require steam generator tubes to be inspected. Furthermore, 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 used and at what locations. The objective is to detect flaws of any type that may satisfy the 

applicable tube repair criteria. Tube-to-tube contact and the resultant wearing of the tubes are 

not expected to occur in steam generators. The reevaluation results of the eddy current data at 

ANO-1 and Oconee Unit 3 illustrate the difficulties in identifying new or unexpected forms of 

degradation and the importance of performing robust inspections that will detect both expected 

and potentially new or emerging degradation mechanisms. IN 2010-21 highlighted the difficulty 

of identifying a new form of degradation. IN 2002-21 indicated that the steam generators with 

the most severe operating conditions (e.g., operating time) may not be the first plant at which 

degradation is observed. 

The successful identification of wear because of tube-to-tube contact at TMI-1 indicates that a 

comprehensive review of the locations and characteristics of all detected eddy current 

indications may be an effective diagnostic tool for evaluating inspection data. This 

comprehensive review may include plotting all locations where indications have been detected 

both radially within the steam generator tube bundle and axially along the tube length. It may 

involve reviewing all indications together or looking at subsets of various types of indications. It 
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may also include determining the directions the indications face and plotting changes in signal 

amplitudes to determine if the indications are changing with time. Performance of the 

comprehensive review may, for example, reveal a clustering of eddy current indications or 

another pattern that may warrant additional attention (e.g., migration of a loose part left in a 

steam generator or the tube-to-tube wear phenomenon). 

The ANO-1 and TMI-1 inspections highlight the importance of reviewing data from the absolute 

channels in addition to data from the differential channels. Slowly varying flaws such as 

long-tapered wear scars and cracks may not be detectable on the differential channels. 

However, they may create clearly discernible signals on the absolute channel. The usefulness 

of reviewing data from the absolute channel was discussed in IN 2003-05. 

The indications at ANO-1 may have been characterized earlier as wear flaws if a more 

comprehensive comparison was performed of the data obtained during the preservice and first 

inservice inspection of the tubes. Since none of the tube-to-tube wear indications were present 

in the preservice inspection, there was clearly a change in the eddy current data. The difficulty 

of attributing this clear change in eddy current data to tube degradation, however, is that some 

of this change also could be the result of operating the tubes at temperature for a cycle (i.e., the 

first heating of the tubes) and normal test repeatability. This highlights the importance of 

understanding the magnitude of the change in eddy current signals that typically occur as a 

result of the first heat cycle and test repeatability so that any higher-than-normal changes can 

be further investigated as possible indications of tube degradation. 

The findings at Oconee Unit 3 demonstrate the importance of properly establishing the reporting 

criteria used by eddy current analysts to ensure flaws are identified. In addition, the findings at 

Oconee Unit 3 demonstrate the importance of using operating experience since the results from 

TMI-1 and ANO-1 were used to target the inspections at Oconee Unit 3 to detect tube-to-tube 

wear indications. The ANO-1 findings also highlight the importance of staying attentive to all 

inspection results and not only focusing on specific issues. 

From a broader perspective, the ANO-1 and TMI-1 findings highlight the importance of 

performing comprehensive inspections of new and replacement equipment to ensure that it is 

performing as expected. 

 
IN 2012-07 

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CONTACT

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

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

Regulation (NRR) project manager. 

/RA/ /RA/ 

Laura A. Dudes, Director Timothy J. McGinty, Director 

Division of Construction Inspection Division of Policy and Rulemaking 

 and Operational Programs Office of Nuclear Reactor Regulation 

Office of New Reactors 

Technical Contact: Kenneth J. Karwoski, NRR 

 301-415-2752 

E-mail: kenneth.karwoski@nrc.gov

Note: NRC generic communications can be found on the NRC public Web site, 

http://www.nrc.gov, under NRC Library. 
IN 2012-07 

Page 5 of 5 

CONTACT

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

the technical contact listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project 

manager. 

/RA/ /RA/ 

Laura A. Dudes, Director Timothy J. McGinty, Director 

Division of Construction Inspection Division of Policy and Rulemaking 

 and Operational Programs Office of Nuclear Reactor Regulation 

Office of New Reactors 

Technical Contact: Kenneth J. Karwoski, NRR 

 301-415-2752 

E-mail: kenneth.karwoski@nrc.gov

Note: NRC generic communications can be found on the NRC public Web site, http://www.nrc.gov, 

under NRC Library. 

ADAMS Accession No.: ML120740578 TAC ME7921 

OFFICE DE Tech Editor BC:NRR/DE/ESGB D:NRR/DE 

NAME KKarwoski J Dougherty GKulesa PHiland 

DATE 4/5/12 email 04/2/12 e-mail 5/14/12 5/21/12 

OFFICE BC:NRO/DE/CIB PM:NRR/PGCB LA:NRR/PGCB 

NAME DTerao DBeaulieu CHawes 

DATE 5/14/12 5/21/12 4/10/12 

OFFICE BC:NRR/PGCB LA:NRR/PGCB D:NRO/DCIP D:NRR/DPR 

NAME KMorganButler CHawes LDudes TMcGinty 

OFFICE 4/16/12 4/16/12 via e-mail 7/13/12 7/16/12 

OFFICIAL RECORD COPY