Information Notice 1997-26, Degradation in Small-Radius U-Bend Regions of Steam Generator Tubes
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
May 19, 1997
NRC INFORMATION NOTICE 97-26: DEGRADATION IN SMALL-RADIUS U-BEND
REGIONS OF STEAM GENERATOR TUBES.
Addressees
All holders of operating licensees or construction permits for pressurized-water reactors
(PWRs).
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to
disseminate information about recent degradation affecting small-radius (rows I and 2)
U-bend regions of tubes in recirculating steam generators (SGs), in order to alert utilities to
potential problems in ensuring the integrity of the small-radius U-bends, and to provide
information about action taken by certain licensees to ensure adequate integrity. It is
expected that recipients will review the information for applicability to their facilities and
consider this information, as appropriate, in their SG inspection programs. However, suggestions contained in this information notice are not NRC requirements; therefore, no
specific action or written response is required.
DescriDtion of Circumstances
Licensees that use Westinghouse-designed recirculating SGs have for many years identified
indications in the U-bend regions of tubes with small radii. During the late 1970s and early
1980s, many units, as a preventative measure, plugged small-radius U-bend tubes to avoid
potential forced outages due to leakage. However, some licensees subsequently unplugged
these tubes and performed in situ stress relief to reduce the susceptibility for degradation.
Also SG designs evolved over time and a number of different material conditions are
represented in currently operating PWRs. These include mill-annealed alloy 600, mill- annealed alloy 600 in situ stress relieved, thermally treated alloy 600, and thermally treated
alloy 690. The following discussion of experience at four plants represents recent operating
experience regarding U-bend degradation that involved various tube material conditions.
During a 1996 inspection, Commonwealth Edison Company (ComEd) identified a total of
64 axially oriented and 2 circumferentially oriented indications in the U-bends of the row 1 SG
tubes at Zion Unit 2. ComEd characterized the indications as primary water stress-corrosion
cracking. The tubes at Zion Unit 2 were fabricated with mill-annealed alloy 600 material, and
the U-bends had not been heat treated. As a result of the inspection findings, ComEd
preventively plugged all the row I tubes at Zion Unit 2.
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IN 97-26 May 19, 1997 During a 1996 inspection, the Tennessee Valley Authority (TVA) identified axial indications in
17 small-radius U-bends of SG tubes at Sequoyah Unit 2 and characterized the degradation
as primary water stress-corrosion cracking. The tubes were fabricated with mill-annealed
alloy 600 material and the U-bends werein situ heat treated during the cycle 6 outage in
1994. TVA plugged the 17 row I tubes that contained the U-bend indications.
During 1992 and 1995 inspections, Pacific Gas & Electric Company (PG&E) identified
circumferential indications having relatively small arc angles in the small-radius U-bends
of SG tubes at Diablo Canyon Unit 1. The tubes were fabricated with mill-annealed alloy
600 material and the small-radius U-bends were in situ heat treated after the second
refueling outage in 1988. PG&E plugged the degraded tubes.
During a 1996 inspection, ComEd identified a single axial indication in the U-bend of one of
the SG tubes at Braidwood Unit 2. The Braidwood Unit 2 tubes were fabricated with
thermally treated alloy 600 tubes and the U-bends in the first seven rows received additional
thermal stress relief after bending during the manufacturing process. ComEd plugged the
degraded tuve.
A small number of axial indications originating on the outside diameter of the tubes have
been reported in the small-radius U-bend regions of the SGs at Palo Verde 1, 2, and 3 and
St. Lucie 1. These SGs were designed by Combustion Engineering.
Discussion
U-bend degradation has occurred in mill-annealed alloy 600 tubes irrespective of whether
they have been heat treated. Tubes with thermally treated alloy 600 material are less
susceptible to degradation than mill-annealed alloy 600 tubes. However, thermally treated
alloy 600 tubes have also begun to experience U-bend degradation. None of the degraded
thermally treated alloy 600 tubes have been removed from SGs for confirmation of the
degradation mechanism. Reports of U-bend degradation have been based on eddy current
inspection results. The susceptibility to cracking in small-radius U-bends and the findings of
recent field inspections have emphasized the importance of inspection of this area of SGs
with techniques capable of accurately detecting U-bend degradation.
U-bend degradation can potentially impair tube integrity if not effectively managed. Concerns
in this regard stem from limitations of eddy current testing to detect and size U-bend cracks, the potential for some U-bend cracks to have relatively long lengths, and the potential for
high crack growth rates for some of these cracks. The industry standard bobbin coil has
proven unreliable for detecting U-bend cracks and, in addition, is not qualified for this
application under the Electric Power Research Institute (EPRI) technique qualification
protocol. The industry has developed special probes for these inspections. The industry has
qualified a rotating pancake coil and a Plus Point coil for detecting indications in small-radius
U-bends, in accordance with enhanced qualification criteria developed by EPRI.
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1
IN 97-26 May 19, 1997 There continues to be an absence of pulled tube information to confirm that the detection
threshold for these cracks is better than 40 or 50-percent through wall. In addition, available
inspection techniques are not capable of reliably sizing crack depths and, for this reason, it
has been industry's practice to "plug on detection" U-bend indications that are found.
Information available on crack growth rates bting experienced in the field is very limited by
virtue of the inability to perform reliable crack depth measurements and the resulting need to
"plug on detection." However, U-bend cracks have led to leakage as early as the first cycle
of operation and, thus, crack growth rates may potentially be high for some cracks. Given
the relatively high detection thresholds, the relatively long operating cycles, and the
potentially high growth rates, the depth of cracks may be in excess of 50-percent through
wall when they are first detected.
In view of these concems, effective management of the degradation of SG tubes is
important to ensure that adequate tube integrity is being maintained in accordance with
10 CFR Part 50, Appendices A and B. One such approach being Implemented by a number
of licensees involves the use of tube integrity assessments to ensure that inspection
sensitivity to U-bend cracks and the frequency and scope of inspection are sufficient to
ensure that U-bend flaws are being detected and removed from service before tube integrity
is impaired.
For example, ComEd performed in situ pressure tests at Zion Unit 2 on four tubes having
the longest axial U-bend indications and on two tubes with circumferential U-bend indica- tions using pressure loading consistent with the margins recommended in Regulatory Guide
(RG) 1.121, "Bases for Plugging Degraded PWJR Steam Generator Tubes." The two tubes
having circumferential indications satisfied RG 1.121 margins without leaking. Three of the
four tubes having axial indications leaked at a pressure of main steamline break conditions
but did not burst under a pressure loading of three-times-normal operating pressure.
Because of the limitations of the test equipment, the pressure in the fourth tube did
not reach the three-times-normal operating pressure criterion of RG 1.121. For this tube, ComEd performed analyses to show that the tube would not burst under a pressure loading
of three-times-normal operating pressure. These analyses are based on eddy current
test measurements. Since these measurements may have large uncertainties, ComEd
conservatively assumed that the cracks were 100-percent through wall. On the basis
of the leakage measurements at main steamline break pressures, CoinEd was able to
demonstrate that accident leakage would satisfy the requirements of 10 CFR Part 100.
For U-bend indications at Sequoyah Unit 2, TVA did not perform in situ pressure testing;
instead, it performed bounding analyses to show that the three tubes having the largest
U-bend cracking satisfied RG 1.121 criteria. However, it should be noted that in situ pressure
testing provides more definitive assurance of structural and leakage integrity than analyses.
For axial indications in the small-radius U-bend regions of the SGs at Palo Verde 1, 2, and 3 and St. Lucie 1, the licensees plugged the tubes.
IN 97-26 May 19, 1997 As shown by the examples discussed above, the integrity of the small-radius U-bend regions
can be more fully ensured by efforts that include performing inspections of rows 1 and 2 U-bends using qualified eddy current techniques; performing in situ pressure testing, as
necessary, to assess the condition of defective tubes; taking appropriate corrective actions, including plugging defective tubes; and assessing the appropriate operating intervals until the
next SG tube inspection.
This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact the technical contacts list below
or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
Marylee M. Slosson, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contact:
John C. Tsao, NRR
(301) 415-2702 E-mail: jct@nrc.gov
Eric J. Benner, NRR
(301) 415-1171 E-mail: ejbl @nrc.gov
Attachment: List of Recently Issued 1RCInformatiog Notices
Attachment
IN 26
May 19, 1997 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information
Date of
Notice No.
Subject
Issuance
Issued to
87-10,
Sup. 1
Potential for Water
Hammer During Restart of
Residual Heat Removal
Pumps
05/15/97
All holders of OLs or CPs
for boiling-water reactors
97-25 Dynamic Range Uncertain- ties in the Reactor Vessel
Level Instrumentation
05/09/97
All holders of OLs or CPs
for Westinghouse pressurized- water reactors
97-24
97-23
97-22
97-21
97-20
Failure of Packing Nuts
on One-inch Uranium
Hexafluoride Cylinder
Valves
Evaluation and Reporting
of Fires and Unplanned
Chemical Reactor Events
at Fuel Cycle Facilities
Failure of Welded-Steel
Moment-Resisting Frames
During the Northridge
Availability of Alternate
AC Power Source Designed
for Station Blackout Event
Identification of
Certain Uranium
Hexafluoride Cylinders
that do not comply
with ANSI N14.1 Fabrication
Standards
05/08/97
05/07/97
04/25/97
04/18/97
04/17/97
All U.S. Nuclear Regulatory
Commission licensees and
certificatees authorized
to handle uranium hexa- fluoride in 30- and 48-inch
diameter cylinders
All fuel cycle conversion, enrichment, and fabrication
facilities
All holders of OLs or
CPs for nuclear power
reactors
All holders of OLs
for nuclear power
reactors
All holders of OLs
for nuclear power
OL = Operating License
CP = Construction Permit
_J
IN 97-26 May 19, 1997 As shown by the examples discussed above, the integrity of the small-radius U-bend regions
can be more fully ensured by efforts that include performing inspections of rows 1 and 2 U-bends using qualified eddy current techniques; performing in situ pressure testing, as
necessary, to assess the condition of defective tubes; taking appropriate corrective actions, including plugging defective tubes; and assessing the appropriate operating intervals until the
next SG tube inspection.
This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact the technical contacts list below
or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
original signed by S.H. Weiss for
Marylee M. Slosson, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contact:
John C. Tsao, NRR
(301) 415-2702 E-mail: jct@nrc.gov
Eric J. Benner, NRR
(301) 415-1171 E-mail: ejbl@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
Tech Editor has reviewed and concurred on 04/07/97
- SEE PREVIOUS CONCURRENCES
DOCUMENT NAME: G:\\EJB1\\UBEND.IN
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Contacts
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BC/PECB:DRPM
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NAME
JTsao*
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JStrosnider*
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MS
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EBenner
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/
DATE
04/08/97
04/16/97
05/08/97
05'397
04/08/97
[OFFICIAL RECORD COPY
IN 97-XX
May XX, 1997 As shown by the examples discussed above, the integrity of the small-radius U-bend regions
can be more fully ensured by efforts that include performing inspections of rows I and 2 U-
bends using qualified eddy current techniques; performing in situ pressure testing, as
necessary, to assess the condition of defective tubes; taking appropriate corrective actions, including plugging defective tubes; and assessing the appropriate operating intervals until the
next SG tube inspection.
This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact the technical contact list below
or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
Marylee M. Slosson, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contact:
John C. Tsao, NRR
(301) 415-2702 E-mail: jct~nrc.gov
Eric J. Benner
(301) 415-1171 E-mail: ejblnrc.gov
Attachment: List of Recently Issued NRC Information Notices
- SEE PREVIOUS CONCURRENCES
DOCUMENT NAME: G:XEJBIUBEND.IN
OFC
Contacts
BC/EMCB
BCIPECB:DRPM
D/DRPM
NAME
JTsao*
JStrosnider
AChaffee
MSlosson
EBenner
__
__
_
_
_
_
DATE
04/08197
04/16/97 T/r/97 I,197
04/08/97
[OFFICIAL REC.RD COPY
IN 97-XX
May XX, 1997 This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact the technical contact list below
or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
Marylee M. Slosson, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contact:
John C. Tsao, NRR
(301) 415-2702 E-mail: jct@nrc.gov
Eric J. Benner
(301) 415-1171 E-mail: ejbl@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
- SEE PREVIOUS CONCURRENCES
DOCUMENT NAME: G:NEJBIUBEND.IN
OFC
Contacts
BC/EMCB
BC/PECB:DRPM
D/DRPM
NAME
JTsao*
JStrosnider*
AChaffee
tuY
MSlosson
EBenner*
DATE
04/08197
04/16/97 QJi97 I /97
04,08/97
________
__COPY
[FFICIALRECORD COPY
.
I
IN 97-XX
May XX, 1997 This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact the technical contact list below
or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
Marylee M. Slosson, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contact:
John C. Tsao, NRR
(301) 415-2702 E-mail: jct@nrc.gov
Eric J. Benner
(301) 415-1171 E-mail: ejblnrc.gov
Attachment: List of Recently Issued NRC Information Notices
- SEE PREVIOUS CONCURRENCES
DOCUMENT NAME: G:\\EJB1UBEND.IN
OFC
Contacts
BCIEMCB
BCIPECB:DRPM
DIDRPM
NAME
JTsao*
JStrosnider*
AChaffee
MSlosson
EBenner*
l
DATE
04/08/97
04116/97
1/97 I /97
04/08/97 I
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IN 97-xx
May XX, 1997 This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact the technical contact list below
or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
Thomas T. Martin, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contact:
John C. Tsao, NRR
(301) 415-2702 E-mail: jct@nrc.gov
Eric J. Benner
(301) 415-1171 E-mail: ejbl@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
DOCUMENT NAME: G:XEJB1%UBEND.IN
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