Bulletin 79-13: Revision 2, Cracking in Feedwater System Piping
SSINS: 6830
Accession No.:
7908220135
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
October 16, 1979
IE Bulletin No. 79-13
Revision 2
CRACKING IN FEEDWATER SYSTEM PIPING
Description of Circumstances:
This revision to IE Bulletin No. 79-13 is based on the results of the radio-
graphic examinations and ongoing investigation of the subject problem to
date since the initial Bulletin was issued. The revision reduces in scope
the number and extent of the piping system welds required to be examined.
The requirements for reporting and action time frame remain unchanged.
On May 20, 1979, indiana and Michigan Power Company notified the NRC of
cracking in two feedwater lines at their D. C. Cook Unit 2 facility. The
cracking was discovered following a shutdown on May 19 to investigate
leakage inside containment. Leaking circumferential cracks were identified
in the 16-inch feedwater elbows adjacent to two steam generator nozzle elbow
welds. Subsequent radiographic examination revealed crack indications in all
eight steam generator feedwater lines at this location on both Units 1 and
2.
On May 25, 1979, a letter was sent to all PWR licensees by the Office of
Nuclear Reactor Regulation which informed licensees of the D. C. Cook
failures and requested specific information on feedwater system design,
fabrication, inspection and operating histories. To further explore the
generic nature of the cracking problem, the Office of Inspection and
Enforcement requested licensees of PWR plants in current outages to
immediately conduct volumetric examination of certain feedwater piping
welds.
As a result of these actions, several other licensees with Westinghouse
steam generators reported crack indications. Southern California Edison
reported on June 5, 1979, that radiographic examination revealed indications
of cracking in feedwater nozzle-to-pipe welds on two of three steam
generators of San Onofre Unit 1. On June 15, 1979, Carolina Power and Light
reported that radiography showed crack indications in similar locations at
their H. B. Robinson Unit 2. Duquesne Power and Light confirmed on June 18,
1979, that radiography has shown cracking in their Beaver Valley Unit 1
feedwater piping-to-vessel nozzle weld. Public Service Electric and Gas
Company reported on June 20, 1979 that Salem Unit 1 also has crack
indications. Wisconsin Public Service company decided on June 20, 1979 to
cut out a feedwater nozzle-to-pipe weld which contained questionable
indication, for metallurgical examination. As of June 22, 1979 and since May
25, 1979 seven other PWR facilities have inspected the feedwater
nozzle-to-pipe welds without finding cracking indications.
NOTE: R1 and R2 indicates lines revised or added.
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IE Bulletin No. 79-13 October 16 1979
Revision 2 Page 2 of 5
The feedwater nozzle-to-pipe configurations for D. C. Cook and for San
Onofre are shown on the attached figures l and 2. A typical feedwater
nozzle-to-pipe weld joint detail showing the principal crack locations for
D. C. Cook and San Onofre are shown on the attached figure 3.
On March 17, 1977, during heat-up for hot functional testing of Diablo
Canyon Unit 1, a leak was discovered in the vessel nozzle-to-pipe butt weld
joining the 16-inch diameter feedwater piping to steam generator 1-2.
Subsequent nondestructive examination of all nozzle welds by radiography and
ultrasonics revealed an approximate 6-inch circumferential crack originating
in the weld root heat-affected zone of the leaking nozzle weld. The cause of
this crack-ing was identified as either corrosion fatigue or thermal fatigue
initiating at small cracks probably induced by the welding and postweld heat
treatment cycles. The system was repaired by replacing with a piping
component employing greater controls on the welding including maintaining
preheat temperature until postweld heat treatment.
The potential safety consequences of the cracking is an increased likelihood
of a feedwater line break in the event of a seismic event or water hammer. A
feedwater line break results in a loss of one of the mechanisms of heat
removal from the reactor core and would result in release of stored energy
from the steam generator into containment. Although a feedwater line-break
is an analyzed accident, the identified degradation of these joints in the
absence of a routine inservice inspection requirement of these feedwater
nozzle-to-pipe welds formed the basis of this-Bulletin.
To date the radiographic examinations, supplemented by ultrasonic methods,
have identified cracking in the steam generator nozzle to feedwater piping
weldments at the following W, and C. E. plants.
D. C. Cook Units 1 & 2 Salem Unit 1
Diablo Canyon Surry Unit 1
San Onofre Unit 1 R. E. Ginna
H. B. Robinson Unit 2 Millstone Unit 2
Beaver Valley Unit 1 Palisades **
Kewaunee Yankee Rowe **
Point Beach Unit 2 Maine Yankee
* Found during hot functional testing
** Confirmatory evaluation incomplete
An extensive metallurgical investigation has been conducted by Westinghouse
on a substantial number of cracked weldments removed from the above plants.
Results of the metallurgical analysis lead to the conclusion that a
corrosion fatigue phenomenon is the probable failure-mechanism, except for
the San Onofre, piping which has been characterized as stress assisted
corrosion.
In parallel with the above ongoing analysis, the feedwater piping at D. C.
Cook, H. B. Robinson, R. E. Ginna, Salem and other plants have been
instrumented (Thermocouples, accelerometers, strain gages, and transducers)
to collect data
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IE Bulletin No. 79-13 October 16, 1979
Revision 2 Page 3 of 5
on the potential forcing functions contributing to cracking under steady
state and transient conditions. Preliminary unchecked results of temperature
data has identified cyclic thermal gradients may exist due to stratified
feedwater temperature conditions in the feedpipe weld region during zero and
low power operations. This gradient tends to support the fatigue aspect of
the postulated failure mechanism. No further unexpected operation loading or
forcing functions have been identified by other instrumentation.
In regard to B&W plants a total of 95 welds in the main and separate
auxiliary feedwater piping, risers and, steam generator nozzles regions have
been examined at Crystal River Unit 3 and Davis Besse. No indications of a
cracking problem was found.
In view of the findings to date, the revised inspections outlined below is
considered acceptable to meet this intent of IE Bulletin No. 79-13.
Actions to be Taken by Licensees
For all pressurized water reactor facilities with an operating license:
1. Facilities which have steam generators fabricated by Westinghouse or
Combustion Engineering that have not conducted volumetric examination
of feedwater nozzles since May 1979 shall complete the inspection
program described below at the earliest practical time but no later
than 90 days after the date of Bulletin No. 79-13.
a. Perform radiographic examination, supplemented by ultrasonic
examination as necessary to evaluate indications, of all feedwater
nozzle-to-pipe welds and of adjacent pipe and nozzle areas (a
distance equal to at least two wall thicknesses). Evaluation shall
be in accordance with ASME Section III, Subsection NC, Article
NC-5000. Radiography shall be performed to the 2T penetrameter
sensitivity level, in lieu of Table NC-5111-1, with systems void
of water.
b. In the event cracking is identified during examination of the
nozzle-to-pipe weld, all feedwater line welds up to the first
piping support or snubber outboard of the nozzle shall be
volumetrically examined in accordance with 1.a above. All
unacceptable code discontinuities shall be subject to repair
unless justification for continued operation is provided.
c. Perform a visual inspection of feedwater system piping supports
and snubbers in containment to verify operability and conformance
to design.
2. All pressurized water reactor facilities shall perform the inspection
program described below at the next outage of sufficient duration or at
the next refueling outage after the inspection required by item 1.
.
IE Bulletin No. 79-13 October 16, 1979
Revision 2 Page 4 of 5
a. For steam generator designs with a common nozzle for both main and
auxiliary feedwater systems, perform volumetric examination of the
feedwater nozzle-to-pipe welds, the feedwater piping welds to the
first support, and the feedwater line-to-containment penetration
welds in accordance with Item 1 above. In addition, examine an
area of at least one pipe diameter of the main feedwater line
downstream at the auxiliary feedwater to main feedwater
connection.
b. For steam generator designs utilizing auxiliary feedwater systems
connected by means of welded nozzle connections, perform
volumetric examination of all auxiliary feedwater nozzle to piping
welds and the first adjacent outboard pipe-to-pipe welds (risers)
in accordance with item 1 above.
For designs utilizing auxiliary feedwater systems connected to the
steam generator by means of bolted flange connections, perform
volumetric examination of the flanged nozzle to piping and first
outboard pipe-to-pipe welds (risers) in accordance with item 1
above.
The examinations specified in 2.b above are not required provided
that during startup, hot standby or cold shutdown operations, the
feedwater level within the steam generator is maintained
essentially constant and no intermittent cold auxiliary feedwater
injection is utilized; i.e., auxiliary feedwater injection where
used, is preheated during the forementioned operating modes.
c. Perform a visual inspection of all feedwater system piping
supports and snubbers in containment to verify operability and
conformance to design.
3. Identification of cracking indications in feedwater nozzle or piping
weld areas in one unit of a multi-unit facility shall require shutdown
and inspection of other similar units which have not been inspected
since May 1979, unless justification for continued operation is
provided.
4. Any cracking or other unacceptable code discontinuities identified
shall be reported to the Director of the appropriate NRC Regional
Office within 24 hours of identification.
5. Provide a written report to the Director of the appropriate NRC
Regional Office within 20 days of the date of the original Bulletin
(June 25, 1979) addressing the following:
a. Your schedule for inspection if required by item 1.
b. The adequacy of your operating and emergency procedures to
recognize and respond to a feedwater line break accident.
c. The methods and sensitivity of detection of feedwater leaks in
containment.
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IE Bulletin No. 79-13 October 16, 1979
Revision 2 Page 5 of 5
6. A written report of the results of examination, in accordance with
requests by Regional Offices preceding this Bulletin and with Bulletin
item 1 and 2 including any corrective measures taken, shall be
submitted within 30 days of the date of the original Bulletin No. 79-13
(June 25, 1979) or within 30 days of completion of the examination,
whichever is later, to the Director of the appropriate NRC Regional
Office with a copy to the NRC Office of Inspection and Enforcement,
Division of Reactor Operations Inspection, Washington, D. C. 20555.
Actions to be Taken by Designated Applicants for Operating Licenses:
1. On completion of the hot functional testing program and prior to fuel
loading, perform the inspections described in item 1 above.
2. During the first refueling outage, perform the inspections described in
item 2 above.
3. Submit reports as described in Items 4, 5. and 6 above based on the
date of Revision 1 to Bulletin No. 79-13 (August 30. 1979)
Approved by GAO, B180225 (R0072), clearance expires 7/31/80. Approval was
given under a blanket clearance specifically for identified generic
problems.
Attachments:
Figures 1, 2, and 3
.
DESIGNATED APPLICANTS FOR OPERATING
LICENSES
Salem 2
North Anna 2
Diablo Canyon 1 & 2
Sequoyah 1
McGuire 1
San Onofre 2
Summer
Watts Bar 1 & 2
Page Last Reviewed/Updated Tuesday, March 09, 2021
Page Last Reviewed/Updated Tuesday, March 09, 2021