Information Notice 1993-20, Thermal Fatigue Cracking of Feedwater Piping to Steam Generators
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C.
20555
March 24, 1993
THERMAL FATIGUE CRACKING OF FEEDWATER
PIPING TO STEAM GENERATORS
Addressees
All holders of operating licenses or construction permits for pressurized
water reactors (PWRs)
supplied by Westinghouse or Combustion Engineering.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to inform addressees of cracks found in the feedwater piping to steam
generators at the Sequoyah Nuclear Power Plant, Units 1 and 2 and the
Diablo Canyon Nuclear Power Plant, Unit 1. Recipients are expected to 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.
Description of Circumstances
In 1992, cracks were found in the feedwater lines at Diablo Canyon Unit 1. In
addition, a through-wall crack and other cracks were found in the feedwater
lines at Sequoyah.
The cracks were attributed to thermal fatigue.
The NRC staff first learned of cracks in feedwater lines to steam generators
which resulted from thermal fatigue in 1979 when the Indiana and Michigan
Electric Company, the licensee for the Donald C. Cook Plant, reported leaks.
In dealing with this problem, the NRC staff issued the following documents:
a letter to PWR licensees pursuant to Paragraph 50.54(f) of Title 10 of
the Code of Federal Regulations, May 25, 1979
Office of Inspection and Enforcement (IE)Bulletin 79-13, "Cracking in
Feedwater System Piping,' June 25, 1979, Revision 1, August 30, 1979, and
Revision 2, October 17, 1979
NUREG/CR-5285, "Closeout of IE Bulletin 79-13,' 1991 In Bulletin 79-13, the NRC requested that licensees perform radiographic and
ultrasonic examinations of feedwater lines.
As a result of these
examinations, cracks were found at 18 of the 54 facilities inspected.
The
staff closed the bulletin on the basis of the results of the one-time
inspection.
The industry had taken the actions recommended in the bulletin
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March 24, 1993 and Instituted the augmented inservice inspection programs, which appeared to
provide for reliable detection of cracks in feedwater piping.
Other technical evaluations were documented in the following reports:
Investigation of Feedwater Line Cracking in PWRs" (Westinghouse, 1980)
NUREG-0691, "Investigation and Evaluation of Cracking Incidents in Piping
In PWRs," (PWR pipe crack study group, 1980)
These studies showed that thermal fatigue was the main cause of the cracks.
Modifications to minimize the effects of thermal stratification In feedwater
lines and augmented licensee inservice inspections were recommended in the
studies.
Recently, some licensees have again reported cracks in feedwater piping.
The
licensee for the Sequoyah units reported an-actual leak despite augmented
inservice inspections. The augmented inspections using ultrasonic techniques
--
showed 'indications that might earlier have revealed the cracks, but the
licensee misinterpreted these as resulting from the geometric configuration of
the pipe. -After finding the leak, the licensee performed radiography on all
feedwater nozzles of both units and found cracks in-five of the eight nozzles.
The licensee for Diablo Canyon Unit 1 reported indications with cracklike
ultrasontc-signal characteristics -in
feedwater-piping to all four steam
generators. The indications varied in length up to 20 cm [7-3/4 inches] in a
circumferential direction, and many were'intermittent.
Some intermittent
indications extended the full circumference with segments up to 5 cm
[2 inches] long.
The licensee tried to verify the indications by radiography
but failed. Later, metallurgical analysis showed the indications to be
cracks. -
'
Only one nozzle at Diablo Canyon had been scheduled for an inspection.
This
inspection was performed in accordance with Section XI of the American Society
of Mechanical Engineers Boiler and Pressure VesselTCode (ASME Code).
However, information on the leak at Sequoyah led the licensee for Diablo Canyon to use
enhanced ultrasonic techniques to inspect all four lines.
These techniques
were considered more appropriate for finding small cracks from thermal fatigue
than techhiques'specified by the ASME Codej which may not be adequate to
detect these types of defects.
Flaw sizing by ultrasonic techniques proved to
be overly conservative at Diablo Canyon, however, presumably because
inclusions in the material led to inaccurate results: 'cracks sized at 8.9 mm
[0.35 inch] deep by ultrasonic inspection were shown by cross sectioning to be
cracks 0.94 mm (0.037 inch] deep.
Discussion
Cracks from thermal fatigue in PWR feedwater lines have proved to be a
recurring problem.
The main cause of crack growth appears to be fatigue
induced by stresses from thermal stratification during cold, low-flow, feedwater injections. Other factors that contribute to crack growth are a
high oxygen content, counterbore weld preparation geometry, and thermal
March 24, 1993 conditions during heatup, hot standby, and low-power operation. A favored
solution has been to replace the degraded piping. However, replacing the
degraded piping in kind without other corrective actions to eliminate or
minimize the factors which cause the cracks can leave the piping susceptible
to the same problem.
Inspection techniques specified in the ASME Code Section XI, do not appear
adequate to find cracks of this type.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
Brian K. Grimes, Director
-Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts:
Lee Banic, NRR
(301) 504-2771
Robert A. Hermann, NRR
(301) 504-2768 Attachment:
List of Recently Issued NRC Information Notices
I I I
I
11 I
Attachment
March 24, 1993
Page I of I
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LIST OF RECENTLi ISSUED
NRC INFORMATION NOTICES
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Notice No.
93-19
93-I8
93-17
93-I6
93-15
93-14
93-13
) 3-12
93-11
93-10
_
.
.
.
Subject
Slab Hopper Bulging
Portable Moisture-Density
Cauge User Responsibilittes
during Field Operations
Safety Systems Response
to Loss of Coolant and
Loss of OffsIte Power
Failures of Nut-Locking
Devices in Check Valves
Failure to Verify the
Continuity of Shunt Trip
Attachment Contacts In
Manual Safety Injection
and Reactor Trip Switches
Clarification of
10 CFR 40.22, Small
Quantities of Source
Material
Undetected Modification
of Flow Characteristics
in the High Pressure
Safety Injection System
Off-Gassing in Auxiliary
Feedwater System Raw
Water Sources
Sigle Failure Vulner- ability of Engineered
Safety Features
Actuation Systems
Dose Calibrator Quality
Control
Issuance
03/17/92
03/10/93
03/08/93
02/19/93
02/18/93 Issued to
All nuclear fuel cycle
licensees.
All U.S. Nuclear Regulatory
Commission licensees that
possess moisture-density
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
.
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02/18/93
All licensees who possess
source material.
02/16/93
All holders of OLs or CPs
for nuclear power reactors.
02/11/93
All holders of OLs or CPs
for nuclear power reactors.
02/04/93
All holders of OLs or CPs
for nuclear power reactors.
02/02/93 All Nuclear Regulatory Com- mission medical licensees.
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March 24, 1993 conditions during heatup, hot standby, and low-power operation. A favored
solution has been to replace the degraded piping. However, replacing the
degraded piping in kind without other corrective actions to eliminate or
minimize the factors which cause the cracks can leave the piping susceptible
to the same problem.
Inspection techniques specified in the ASME Code Section XI, do not appear
adequate to find cracks of this type.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
Original igned by
Brian K. Grimes
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts:
Lee Banic, NRR
(301) 504-2771
Robert A. Hermann, NRR
(301)
504-2768 Attachment:
List of Recently
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Issued NRC Information Notices
Document Name: 93-20.IN
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IN 93-XX
March xx, 1993 conditions during heatup, hot standby, and low-power operation. A favored
solution has been to replace the degraded piping.
However, replacing the
degraded piping in kind without other corrective actions to eliminate or
minimize the factors which cause the cracks would leave the piping susceptible
to the same problem.
Inspection techniques specified in the ASME Code Section XI, do not appear
adequate to finding cracks of this type.
This information notice requires no specific action or written response.
If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts:
Lee Banic, NRR
(301) 504-2771
Robert A. Hermann, NRR
(301) 504-2768 Attachment:
List of Recently Issued NRC Information Notices
Document Name: DIANINF.IN
- SEE PREVIOUS CONCURRENCES
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IN 93-XX
March xx, 1993 replace the degraded piping. However, replacing the degraded piping in kind
without other corrective actions to eliminate or minimize stratification will
leave it susceptible to cracking again.
Inspection techniques specified in the ASME Code,Section XI, appear to be
inadequate for finding the cracking.
This information notice requires no specific action or written response.
If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts:
Lee Banic, NRR
(301) 504-2771
Robert A. Hermann, NRR
(301) 504-2768
List of Recently Issued NRC Information Notices
Attachment:
Document Name: DIANINF.IN
- SEE PREVIOUS CONCURRENCES
D/DORS:NRR
C/OGCB:DORS:NRR*RPB:ADM
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IN 93-XX
February xx, 1993 in the ASME Code,Section XI, appear to be inadequate for finding the
cracking. A favored solution has been to replace the degraded piping.
However, replacing the degraded piping in kind without other corrective
actions to eliminate or minimize stratification will leave it susceptible to
cracking again.
This information notice requires no specific action or written response.
If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts:
Lee Banic, NRR
(301) 504-2771
Robert A. Hermann, NRR
(301) 504-2768
List of Recently Issued NRC Information Notices
Attachment:
Document Name: DIANINF.IN
- SEE PREVIOUS CONCURRENCES
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IN 93-XX
January xx, 1993 in the ASME Code,Section XI, appear to be inadequate for finding the
cracking. A favored solution has been to replace the degraded piping.
However, replacing the degraded piping in kind without other corrective
actions to eliminate or minimize stratification will leave it susceptible to
cracking.
This information notice requires no specific action or written response.
If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts:
Lee Banic, NRR
(301) 504-2771
Robert A. Hermann, NRR
(301) 504-2768 Attachment:
List of Recently Issued NRC Information Notices
Document Name: DIANINF.IN
- SEE PREVIOUS CONCURRENCES
OGCB:DORS:NRR
CVHodge*
01/25/93 EMCB:DE:NRR
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January xx, 1993 Discussion
Cracking from thermal fatigue is a persistent problem. The main cause of
crack growth is fatigue induced by stresses from thermal stratification during
cold, low-flow, feedwater injections. Other factors are a high oxygen
content, counterbore weld preparation geometry, and thermal conditions during
heatup, hot standby, and low-power operation.
Inspection techniques specified
in the ASME Code,Section XI, appear to be inadequate for finding the
cracking. A favored solution has been to replace the degraded piping.
However, replacing the degraded piping in kind without other corrective
actions to eliminate or minimize stratification will leave it susceptible to
cracking.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts:
Lee Banic, NRR
(301) 504-2771
Robert A. Hermann, NRR
(301) 504-2768 Attachment:
List of Recently Issued NRC Information Notices
Document Name: DIANINF.IN
- SEE PREVIOUS CONCURRENCES
D/DORS:NRR
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one of the technical ontacts listed below or the appro riate Office of
Nuclear Reactor Regulation (NRR) project -manager.
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical
contacts:
Lee Banic, NRR
(301) 504-2771
Robert A. Hermann, NRR
(301) 504-2768 Attachment:
List of Recently Issued NRC Information Notices
Document Name:
DIANINF.IN
OGCB:DORS:NRR
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01/ /93 D/DORS:NRR
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