United States Nuclear Regulatory Commission - Protecting People and the Environment

Bulletin 78-12: Atypical Weld Material in Reactor Pressure Vessel Welds

                               UNITED STATES 
                       NUCLEAR REGULATORY COMMISSION 
                    OFFICE OF INSPECTION AND ENFORCEMENT 
                           WASHINGTON, D.C. 20555 
                                     
                            September 26, 1978  

                                                     IE Bulletin No. 78-12 

ATYPICAL WELD MATERIAL IN REACTOR PRESSURE VESSEL WELDS 

Description of Circumstances 

On August 4, 1978, the NRC was informed by the Duke Power Company and the 
Babcock and Wilcox Company (B&W) that the weld wire used in some of the 
reactor vessel welds in Oconee Unit No. 3 may have differed from that 
specified. A chemical analysis of one sample of archive material by B&W 
disclosed that the nickel content was measured to,be 0.1 percent (versus 
0.45 to 0.8 percent nominal specified) and the silicon content was measured 
to be 1.0 percent (versus 0.3 to 0.6 percent nominal specified). The heat of 
weld metal in question was supplied by the Page Company, a Division of the 
American Chain & Cable Co., Bowling Green, Kentucky to B&W, the manufacturer
of the Oconee, Unit 3 vessel. Further checks by B&W of its records have 
identified eleven additional vessels in which the incorrect weld material 
may have been used. Owners of these vessels have been notified. 

The NRC staff has made a determination of the possible effects on reactor 
vessel integrity of the use, or possible use, of the improper weld material.
Weldments containing the atypical material are likely to have higher than 
normal nil-ductility transition temperature characteristics. Therefore to 
maintain reactor vessel safety margins, implementation of new conservative 
pressure/temperature operating limits may be required. 

While the specific problem has been identified as possibly affecting twelve 
vessels manufactured by B&W, it is not possible to conclude in the absence 
of specific information that similar atypical weld material was not also 
supplied to other vessel manufacturers and used in reactor pressure vessel 
fabrication. 

Action To Be Taken By Licensees and Permit Holders: 

For all power reactor facilities with an operating license or a construction
permit, except those already identified as possibly having atypical weld 
material (1): 



(1)  The twelve nuclear units identified as having possible atypical 
     pressure vessel weldments are: Three Mile Island Unit Nos. 1 and 2, 
     Crystal River Unit No. 3, Arkansas Nuclear One Unit No. 1, Oconee Unit 
     No. 3, Rancho Seco Unit No. 1, Midland Unit No. 1, Quad Cities Unit No. 
     2, Browns Ferry Unit No. 1, Turkey Point Unit No. 4 and Zion Unit Nos. 
     1 and 2. 


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IE Bulletin No. 78-12                                  September 26, 1978 

1.   Conduct a record search (2) of all primary reactor pressure vessel 
     weldments (excluding partial penetration welds) and submit the 
     following information (3): 

     a.   The principal vessel manufacturer. If other manufacturers were 
          utilized, identify those companies and the weldments completed by 
          those firms. 

     b.   The type and form of weld materials and the identifying heat and 
          lot numbers used in each weldment. 

     c.   The weld material manufacturer(s) and the types and form of 
          materials supplied. 

     d.   The specified properties of the weld materials and the completed 
          weldments (Chemistry, tensile and impact properties, as 
          appropriate). 

2.   Describe the procedures utilized during fabrication to verify 
     conformance to the specifications. Specifically provide the following: 

     a.   Describe the type, number and dates of tests performed on welding 
          materials to satisfy the material conformance testing requirements
          for each heat, lot or combination of heat and batch, etc., of 
          welding materials used in the construction or repair of the 
          reactor pressure vessel in your facility. Indicate whether each 
          heat, lot or batch subdivision (coil or spool) was tested and the 
          extent of such testing, i.e., were both ends of a coil or spool of 
          wire tested for each sub-arc flux-wire combination or heat-flux 
          batch combination. 

     b.   Describe the type, number and dates of other tests such as 
          procedure qualification, welder performance tests, in-process 
          checks on post-weld tests which were performed. 


(2)  The record search may be performed by the vessel manufacturer and the  
     requested information reviewed as appropriate by each licensee prior to  
     forwarding to the NRC. It is not the intent of this Bulletin to require  
     each licensee to individually examine manufacturing records of a 
     generic nature. Records of nonconforming conditions that may be 
     identified by the manufacturer and are uniquely applicable to a 
     specific vessel should however be carefully examined by the owner. 

(3)  Some of the information requested by this Bulletin may have been 
     previously supplied to the NRC under the Surveillance Program. 
     Information previously submitted may be referenced in lieu of 
     resubmittal. 



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IE Bulletin No. 78-12                                   September 26, 1978 

     c.   For each of the tests described in 2(a) and 2(b) above, describe 
          the parameters of each test and provide the results obtained. 
          Identify the applicability to specific weldments by correlation of
          heat, lot or batch as appropriate. 

3.   Identify those cases of weld filler material which did not meet 
     procurement specifications based on verification tests, i.e. mechanical
     or chemical properties. Describe the disposition action taken or the 
     acceptance basis for utilization in vessel fabrication. In such cases, 
     discuss the effect that the atypical weld composition has on the 
     fracture toughness of the weld metal. 

4.   Provide information on the availability of archive weld materials which
     might be used for verification purposes. 

5.   Please provide your response in writing within 60 days. Reports should 
     be submitted to the Director of the appropriate NRC Regional Office and
     a copy should be forwarded to the U.S. Nuclear Regulatory Commission, 
     Office of Inspection and Enforcement, Division of Reactor Construction 
     Inspection, Washington, D.C. 20555. 

Approved by GAO, B180225 (ROO72); clearance expires 7/31/80. Approval was 
given under a blanket clearance specifically for identified generic 
problems. 



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