Information Notice 2013-11, Official Exhibit - NYS000551-00-BD01 - NRC Information Notice 2013-11: Crack-Like Indication at Dents/Dings and in the Freespan Region of Thermally Treated Alloy 600 Steam Generator Tubes (July 3, 2013)

Official Exhibit - NYS000551-00-BD01 - NRC Information Notice 2013-11: Crack-Like Indication at Dents/Dings and in the Freespan Region of Thermally Treated Alloy 600 Steam Generator Tubes (July 3, 2013)

ML15331A240
Person / Time
Site:	Indian Point
Issue date:	07/03/2013
From:	State of NY, Office of the Attorney General
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 27915, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15331A240 (8)

NYS000551 Submitted: June 9, 2015 UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

OFFICE OF NEW REACTORS

WASHINGTON, DC 20555-0001 July 3, 2013 NRC INFORMATION NOTICE 2013-11: CRACK-LIKE INDICATIONS AT DENTS/DINGS

AND IN THE FREESPAN REGION OF

THERMALLY TREATED ALLOY 600 STEAM

GENERATOR TUBES

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 that have permanently ceased operations

and have certified that fuel has been permanently removed from the reactor vessel.

All holders of and applicants for a power reactor early site permit, combined license, standard

design certification, standard design approval, or manufacturing 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 crack-like indications at dented/dinged locations and in the

freespan region in thermally treated Alloy 600 steam generator tubes. The NRC expects that

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

appropriate, to ensure they meet regulatory requirements. Suggestions contained in this IN are

not NRC requirements; therefore, no specific action or written response is required.

DESCRIPTION OF CIRCUMSTANCES

As discussed below, indications of cracking at dents/dings and in the freespan region were

discovered in thermally treated Alloy 600 steam generator tubes.

Seabrook Station, Unit 1

Seabrook Station, Unit 1 (Seabrook), has four recirculating steam generators, each of which has

approximately 5,600 tubes fabricated from thermally treated Alloy 600. The tubes are supported

in the straight region by a flow distribution baffle and several tube support plates, and in the

U-bend region by anti-vibration bars.

In fall 2012, NextEra Energy Seabrook, LLC (the licensee), conducted steam generator tube

inspections at Seabrook. All tubes were inspected full length with a bobbin coil probe, except

for the U-bend region of the tubes in rows 1 and 2. Various locations of the tube, including the

U-bend region of some of the row 1 and 2 tubes, were inspected using a rotating probe

equipped with a +PointTM coil. At the time of the inspections, Seabrook had operated

approximately 18.95 effective full-power years. The steam generators had operated at a hot-leg

United States Nuclear Regulatory Commission Official Hearing Exhibit

In the Matter of: Entergy Nuclear Operations, Inc.

ML13127A236 (Indian Point Nuclear Generating Units 2 and 3)

ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #: NYS000551-00-BD01 Identified: 11/5/2015 Admitted: 11/5/2015 Withdrawn:

Rejected: Stricken:

Other: temperature of approximately 325.6 degrees Celsius (618 degrees Fahrenheit) since

commencement of commercial operation until implementation of a power uprate in 2005.

Thereafter, the steam generators had operated at 327 degrees Celsius (621 degrees

Fahrenheit).

During the bobbin coil inspections, the licensee detected an indication in the freespan region of

a tube on the hot-leg side of the steam generator between the flow distribution baffle and the

first tube support plate. The licensee used a rotating probe equipped with a +PointTM coil to

further inspect the region where the bobbin coil indication was detected and confirmed that the

indication was axially oriented, crack-like, and had initiated from the outside diameter of the tube

(outside diameter stress corrosion cracking (ODSCC)). The indication had a +PointTM voltage

amplitude of approximately 0.96 volts, a length of 0.52 inches, and a maximum depth of

77 percent of the tube wall thickness.

The rotating probe inspections also identified two additional ODSCC indications, which were not

detected during the bobbin coil inspections. These additional indications were approximately

6 inches above the initially detected indication and were smaller in size. One indication had a

+PointTM voltage of 0.24 volts, a length of 0.3 centimeters (0.15 inches), and a maximum depth

of 45 percent of the tube wall thickness. The other indication had a +PointTM voltage amplitude

of 0.38 volts, a length of 0.46 centimeters (0.18 inches), and a maximum depth of 56 percent of

the tube wall thickness.

Although there was no reportable bobbin signal at these two additional locations, there were

benign signals at these locations since the preservice inspection. These benign signals were

characterized as small dents/dings from the preservice inspection data and had exhibited local

conductivity changes after the first cycle of operation at temperature. The licensee concluded

that the three indications are not components of a single indication since the indications are

separated by ligaments of sound material and are not in the same axial plane.

In addition to these three indications of axially oriented ODSCC in one tube, another axially

oriented ODSCC indication was detected in another tube. This latter indication was associated

with a dented/dinged region of the tube at the top tube support plate on the hot-leg side of the

steam generator. There were two dents/dings in this tube at the uppermost tube support plate:

one at the bottom edge of the tube support plate had a bobbin voltage amplitude of 11.35 volts, and one at the upper edge of the tube support plate had a bobbin voltage amplitude of

8.96 volts. The crack-like indication was associated with the dent/ding at the lower edge of the

tube support plate and was detected during the rotating probe inspections of dents/dings. A

rotating probe is typically used to inspect dents/dings that have bobbin voltage amplitudes

greater than 5 volts since the bobbin coil is not qualified to detect crack-like indications in such

dents/dings.

The crack-like indication had a +PointTM coil voltage amplitude of 0.89 volts, a length of

0.56 centimeters (0.22 inches), and a maximum depth of 76 percent of the tube wall thickness.

Since the original scope of inspections only included a 50 percent sample of the hot-leg and

U-bend dents/dings that had bobbin voltage amplitudes greater than 5 volts, the inspection

scope in the steam generator in which the crack-like indication was detected was expanded to

include: 100 percent of the hot-leg and U-bend dents/dings, with bobbin voltage amplitudes

greater than 5 volts, 100 percent of the cold-leg dents/dings with bobbin voltage amplitudes

greater than 5 volts at the uppermost support plate, a 20 percent sample of the hot-leg and

U-bend dents/dings that had bobbin voltage amplitudes greater than 2 volts and less than or

equal to 5 volts, and a 20 percent sample of the dents/dings that had bobbin voltage amplitudes greater than 2 volts and less than or equal to 5 volts at the uppermost cold-leg tube support

plate. The scope of the inspections in the other three steam generators was not expanded. No

additional crack-like indications associated with dents/dings were detected in any of the four

steam generators during the inspection.

The tubes with the axially oriented ODSCC were removed from service by plugging both ends of

the tubes. Both of the tubes with these indications had adequate structural and leakage

integrity. Neither of the tubes had any evidence of high residual stress as a result of

non-optimal tube processing as discussed in NRC IN 2002-21, Supplement 1, Axial

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

Additional information is available in Seabrook StationSteam Generator Tube Inspection

Report, dated December 31, 2012 (Agencywide Documents Access and Management System

(ADAMS) Accession No. ML13008A160).

Braidwood Station, Unit 2

Braidwood Station, Unit 2 (Braidwood Unit 2), has four recirculating steam generators, each of

which has 4,570 tubes fabricated from thermally treated Alloy 600.

In the fall of 2012, Exelon Generation Company, LLC (the licensee), conducted steam generator

tube inspections at Braidwood Unit 2. All tubes were inspected full length with a bobbin coil

probe, except for the U-bend region of the tubes in rows 1 and 2. The licensee inspected

various locations of the tube using a rotating probe equipped with a +PointTM coil. At the time of

the inspections, Braidwood Unit 2 had operated approximately 21.3 effective full-power years.

The steam generators have generally operated at a hot-leg temperature of approximately

322 degrees Celsius (611 degrees Fahrenheit).

As a result of the operating experience at Seabrook described above, the licensee for

Braidwood Unit 2 confirmed that its automatic eddy current data analysis system would identify

the types of flaws observed at Seabrook. In addition, the licensee trained its eddy current data

analysts to identify these types of indications.

During the 2012 inspections, one tube at Braidwood, Unit 2, was identified to have three axially

oriented indications that were attributed to ODSCC. Two of these indications were located at

tube support plate elevations (at the 3H and 5H tube support plate) and one was in the freespan

region of the tube between the 3H and 5H tube support plate elevations (approximately

86 centimeters (34 inches) above the 3H tube support). The indications were not aligned axially

along the length of the tube, as evidenced from the +PointTM data, which was acquired from

8 centimeters (3 inches) below the 3H support plate to 8 centimeters (3 inches) above the 5H

support plate.

The indication at 3H had a +PointTM voltage amplitude of 0.64 volts, a length of 1.4 centimeters

(0.56 inches), and a maximum depth of 69.6 percent of the tube wall thickness. The indication

at 5H had a +PointTM voltage amplitude of 0.25 volts, a length of 1.2 centimeters (0.48 inches),

and a maximum depth of 50 percent of the tube wall thickness. The freespan indication had a

+PointTM voltage amplitude of 0.34 volts, a length of 0.48 centimeters (0.19 inches), and a

maximum depth of 56.4 percent of the tube wall thickness. There was no evidence of a scratch

along the length of the tube. The freespan indication was associated with a ding with a bobbin

voltage amplitude of approximately 1 volt. The affected tube was identified as potentially having

elevated residual stresses caused by non-optimal tube processing since the tube had no U-bend offset signal (typically referred to as a 2-sigma tube). This tube was removed from

service by plugging both ends of the tube.

The indication at the 3H tube support plate was in-situ pressure tested, with no leakage

observed at any test pressure, including the test pressure associated with three times the

normal operating differential pressure. Only the indication at 3H was tested since it exceeded

the threshold for performing in-situ pressure testing.

During the original production analysis of the bobbin coil eddy current data, only the indication at

the 3H tube support plate was identified. The other two indications in this tube (at 5H and in the

freespan) were not identified by either the primary or secondary analysis of the data, but rather

by the independent qualified data analyst. The primary analysis (of the bobbin coil data) was

performed using an automated data analysis system operated in the interactive mode, and the

secondary analysis was performed using human analysts. An investigation into why the

freespan indication was not identified by the automated analysis system revealed that the

freespan indication had a phase angle of 151 degrees, whereas the flaw identification algorithm

was set to only identify indications that were less than 150 degrees. As a result of these

findings, the licensee increased its criterion to 151 degrees. The criterion was not increased

above 151 degrees because of concerns that many nonflaw-like signals would be identified.

The automated data analysis system missed the indication at the 5H tube support plate

because the flaw identification algorithm was not applied at this location. In order for the

automated flaw identification algorithm to apply at a tube support plate, the entire tube support

plate must be contained within a data evaluation window size of 27. Since the entire 5H tube

support plate was not within this window size, the automated system did not apply the flaw

identification algorithm at this location. The licensee increased the window size to 31 to ensure

the flaw identification algorithm would be applied to all tube support plates. The licensee also

reduced the voltage threshold for identifying the tube support plate region from 1 volt to

0.8 volts.

As a result of these findings, all bobbin coil data was re-analyzed with the automated data

analysis system operated in the interactive mode with the revised criteria. The re-analysis

identified no additional crack-like indications.

The licensee reviewed the prior inspection data for the three indications attributed to ODSCC.

This review indicated that there was a 20 degree change in the phase angle of the freespan

indication (which appeared ding-like) from 1990 to the present. For the indications at the tube

supports, there were no indications present in the 2009 data at either support and there was no

indication present in the 2011 data for the 5H tube support plate. However, with hindsight, some evidence of a signal could be seen in the 2011 data for the signal at the 3H tube support

plate (but the signal would not have been reportable).

Additional information is available in Braidwood Station, Unit 2 Steam Generator Tube

Inspection Report for Refueling Outage 16, dated February 5, 2013 (ADAMS Accession

No. ML13039A042).

BACKGROUND

Related NRC Generic Communications

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 (ADAMS Accession No. ML102210244). This

IN alerted addressees to the detection of a crack-like indication in the U-bend region of a

thermally treated Alloy 600 steam generator tube.

NRC IN 2010-05, Management of Steam Generator Loose Parts and Automated Eddy Current

Data Analysis, dated February 3, 2010 (ADAMS Accession No. ML093640691). This IN

alerted addressees to loose parts (foreign objects) in steam generators and the use of

automatic steam generator eddy current data analysis systems.

NRC IN 2008-07, Cracking Indications in Thermally Treated Alloy 600 Steam Generator

Tubes, dated April 24, 2008 (ADAMS Accession No. ML080040353). This IN alerted

addressees to degradation in steam generator tubes.

NRC IN 2005-09, Indications in Thermally Treated Alloy 600 Steam Generator Tubes and

Tube-to-Tubesheet Welds, dated April 7, 2005 (ADAMS Accession No. ML050530400). This

IN alerted addressees to degradation in steam generator tubes and tube-to-tubesheet welds.

NRC IN 2004-17, Loose Part Detection and Computerized Eddy Current Data Analysis in

Steam Generators, dated August 25, 2004 (ADAMS Accession No. ML042180094). This IN

alerted addressees to (1) challenges associated with detection of loose parts and related tube

damage in steam generators, and (2) computerized data screening algorithms used in the

evaluation of steam generator tube eddy current data.

NRC IN 2002-21, Supplement 1, Axial Outside-Diameter Cracking Affecting Thermally Treated

Alloy 600 Steam Generator Tubing, dated April 1, 2003 (ADAMS Accession

No. ML030900517). This IN alerted addressees to the root cause assessment for the axially

oriented outside-diameter crack indications in the thermally treated Alloy 600 steam generator

tubing at Seabrook.

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

Steam Generator Tubing, dated June 25, 2002 (ADAMS Accession No. ML021770094). This

IN alerted addressees to preliminary indications of axial outside-diameter cracking of thermally

treated Alloy 600 steam generator tubing at Seabrook.

DISCUSSION

There are 17 units in the United States with thermally treated Alloy 600 steam generator tubes.

The steam generators at these units have been in service, on average, for approximately

25 calendar years. In 2002, the first incidence of corrosion-related cracking was reported in

units with thermally treated Alloy 600 steam generator tubing. This cracking was attributed to

non-optimal tube processing (refer to NRC IN 2002-21). Since then, several other units with

thermally treated Alloy 600 tube material observed crack-like indications in their steam

generators. These crack-like indications occurred in the United States at several different

locations along the length of the tube, including in the tubesheet region, at the top of the

tubesheet, at tube support plate elevations, and in the U-bend. 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.

The recent instances of cracking at Seabrook and Braidwood, Unit 2, are the first reported

instances of cracking in the freespan region of the tube and at dented/dinged regions in units

with thermally treated Alloy 600 tubing. The crack-like indications not initially detected with the

bobbin coil at Seabrook and Braidwood Unit 2 illustrate the challenges in identifying crack-like

indications and the need to be diligent in reviewing inspection data. The role the small dings

played in initiating the freespan crack-like indications, if any, is not known since the tubes were

not removed for destructive examination.

As discussed below, the finding of the one crack-like indication at a dented/dinged location at

Seabrook potentially illustrates two other limitations related to (1) relying on the temperature

dependence of cracking to focus inspections and (2) using a sampling strategy when the

number of flaws that potentially exist is low.

Stress corrosion cracking is a temperature-dependent phenomenon and typically results in

cracks being more prevalent at units operating at higher temperatures and finding a larger

number of cracks in hotter regions of the tube. Assuming dents/dings with similar severities are

located in hotter regions of the tube at Seabrook (i.e., at lower locations on the hot-leg side of

the tube), the finding of a crack-like indication at the uppermost hot-leg tube support plate, rather than at one of these postulated dents/dings in the hotter regions of the tube, indicates a

potential weakness of a sampling strategy that focuses the dent/ding inspections at the hotter

tube locations. This is because the potential for cracking depends not only on the temperature, but also on the tube material, the stresses in the tube, and other operating parameters

(e.g., water chemistry). In some instances, it is difficult to quantify all of these parameters such

that a simple sampling strategy can be developed.

Given the relatively low number of crack-like indications being found in units with thermally

treated Alloy 600 tubing (e.g., one crack-like indication in a dent/ding), a sampling strategy in

lieu of inspecting all susceptible locations may result in missing crack-like indications.

The operating experience described above illustrates the importance of inspecting locations

susceptible to degradation with probes capable of detecting that degradation and the challenges

and limitations of implementing a sampling strategy (e.g., based solely on the temperature

dependence of cracking or when only a limited number of flaws may be present in a steam

generator). In addition, the operating experience indicates that crack-like indications may be

missed during inspections. These items should be considered in establishing the appropriate

operating interval between inspections.

The findings at Braidwood Unit 2, regarding the initial setup of the computerized data analysis

system indicate the importance of properly establishing the parameters for computerized data

analysis algorithms. A rigorous technical basis should exist for these parameters to provide

assurance that the inspections are performed with the objective of detecting flaws that may

satisfy the applicable tube plugging or repair criteria.

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 NRC project manager.

/RA Sher Bahadur Acting for/ /RA/

Lawrence E. Kokajko, Director Laura A. Dudes, Director

Division of Policy and Rulemaking Division of Construction Inspection

Office of Nuclear Reactor Regulation and Operational Programs

Office of New Reactors

Technical Contacts: Kenneth J. Karwoski, NRR

Telephone: 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 the NRC Library

ML13127A236 *via e-mail TAC MF1394 NRR/DE/ESGB/ NRR/DPR/PGC

OFFICE NRR/DE* Tech Editor* NRR/DE/D*

BC* B/PM

NAME KKarwoski JDougherty GKulesa PHiland ARussell

DATE 06/03/13 05/14/13 06/04/13 06/13/13 06/14/13 NRR/DPR/ NRR/DPR/PGCB

OFFICE NRO/DCIP/D NRR/DPR/DD NRR/DPR/D

PGCB/LA* /BC

NAME CHawes TMensah (A) LDudes SBahadur LKokajko

DATE 06/18/13 06/18/13 6/19/13 6/28/13 7/3/13