United States Nuclear Regulatory Commission - Protecting People and the Environment

Bulletin 79-13: Revision 2, Cracking in Feedwater System Piping

                                                       SSINS: 6830  
                                                       Accession No.: 

                               UNITED STATES  
                           WASHINGTON, D.C. 20555 

                             October 16, 1979  

                                                  IE Bulletin No. 79-13  
                                                  Revision 2 


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 

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 

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. 

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 

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 

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 

     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 

          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 

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 

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 

Figures 1, 2, and 3


Salem 2

North Anna 2

Diablo Canyon 1 & 2  

Sequoyah 1

McGuire 1 

San Onofre 2 


Watts Bar 1 & 2
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