Bulletin 79-13: Revision 1, Cracking in Feedwater System Piping
Accession No: 7908220101
SSINS No.: 6830
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
August 30, 1979
IE Bulletin No. 79-13
Revision 1
CRACKING IN FEEDWATER SYSTEM PIPING
Description of Circumstances:
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 or June 5, 1979, that radiographic examination revealed indications
of cracking in feedwater nozzle-to-piping 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.29, 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.
The feedwater nozzle-to-pipe configurations for D.C. Cook and for San Onofre
are shown on the attached figures 1 and 2. A typical feedwater
pipe-to-nozzle weld joint detail showing the principal crack locations for
D.C. Cook and San Onofre are shown on the attached figure 3.
.
IE Bulletin No. 79-13 August 30, 1979
Revision 1 Page 2 of 4
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.
Cracked weldments have been removed from the D. C. Cook Units 1 and 2 and
San Onofre Unit 1 feedwater systems for extensive metallurgical
investigation by Westinghouse. Based on preliminary analysis, Westinghouse
stated the D. C. Cook failure may be "fatigue assisted by corrosion. The San
Onofre cracking was stated to be characteristic of "stress assisted
corrosion."
The cracking experienced at Diablo Canyon, D. C. Cook and San Onofre would
appear to have different cause - effect relationships which are not fully
understood at this time.
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-topping welds is the basis for this Bulletin.
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 this Bulletin.
a. Perform radiographic examination, supplemented by ultrasonic
examination as necessary to evaluate indications, of all feedwater
nozzle-to-piping 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.
.
IE Bulletin No. 79-13 August 30, 1979
Revision 1 Page 3 of 4
b. If cracking is identified during examination of the
nozzle-to-piping weld, all feedwater line welds up to the first
piping support or snubber and high stress points in containment
shall be volumetrically examined in accordance with 1.a. above.
All unacceptable code discontinuities, other than cracking, 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.
a. For steam generator designs having a common nozzle for both main
and auxiliary (emergency) feedwater systems, perform volumetric
examination of all feedwater nozzle-to-pipe weld areas and all
feedwater pipe weld areas inside containment in accordance with
item 1 above.1/ In addition, conduct an examination of welds
connecting auxiliary feedwater piping to the main feedwater line
outside containment. This examination should include an area of at
least one pipe diameter on the main feedwater line downstream of
the connection.
b. For steam generator designs with separate nozzles for main
feedwater and auxiliary feedwater, perform volumetric examination
(in accordance with item 1 above) of all welds inside containment
and upstream of the external ring header or vessel nozzle for each
steam generator. If an external ring header is employed, also
inspect all welds of one inlet riser on each feed ring of each
steam generator.1/
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.
1/ Welds in the feedwater system, (other than the feedwater nozzle-to-pipe
welds) that have been examined since May 1979 need not be reexamined.
.
IE Bulletin No. 79-13 August 30, 1979
Revision 1 Page 4 of 4
5. Provide a written report to the Director of the appropriate NRC
Regional Office within 20 days of the date of this Bulletin 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.
6. A written report of the results of examinations, 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 this Bulletin 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.
Approved by GAO, 8180225 (R0072), clearance expires 7/31/80. Approval was
given under a blanket clearance specifically for identified generic
problems.
Attachments:
Figures 1, 2 and 3
Page Last Reviewed/Updated Tuesday, March 09, 2021