Information Notice 2012-07, Tube-to-Tube Contact Resulting in Wear in Once-Through Steam Generators: Difference between revisions
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| issue date = 07/17/2012 | | issue date = 07/17/2012 | ||
| title = Tube-to-Tube Contact Resulting in Wear in Once-Through Steam Generators | | title = Tube-to-Tube Contact Resulting in Wear in Once-Through Steam Generators | ||
| author name = Dudes L | | author name = Dudes L, Mcginty T | ||
| author affiliation = NRC/NRO, NRC/NRR/DPR | | author affiliation = NRC/NRO, NRC/NRR/DPR | ||
| addressee name = | | addressee name = | ||
Line 13: | Line 13: | ||
| document type = NRC Information Notice | | document type = NRC Information Notice | ||
| page count = 6 | | page count = 6 | ||
}} | }} | ||
{{#Wiki_filter: | {{#Wiki_filter: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== | ==CONTACT== | ||
RESULTING IN WEAR IN ONCE-THROUGH STEAM GENERATORS | |||
===RESULTING IN=== | |||
WEAR IN ONCE-THROUGH STEAM | |||
GENERATORS | |||
==ADDRESSEES== | ==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, | 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== | ==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 | 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 | |||
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 tubes 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-1s 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 3s 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 indications depth on a second 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== | ==BACKGROUND== | ||
NRC IN 2002-21, | 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 NRCs | |||
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 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. | |||
==CONTACT== | ==CONTACT== | ||
This IN requires no specific action or written | This IN requires no specific action or written response. Please direct any questions about this | ||
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 | |||
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:=== | ===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. | |||
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}} | |||
{{Information notice-Nav}} | {{Information notice-Nav}} |
Latest revision as of 07:47, 12 November 2019
ML120740578 | |
Person / Time | |
---|---|
Issue date: | 07/17/2012 |
From: | Laura Dudes, Mcginty T Office of New Reactors, Division of Policy and Rulemaking |
To: | |
References | |
IN-2012-007 | |
Download: ML120740578 (6) | |
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
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 tubes 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-1s 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 3s 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 indications depth on a second 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 NRCs
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 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.
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.
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