United States Nuclear Regulatory Commission - Protecting People and the Environment

Bulletin 82-03: Revision 1: Stress Corrosion Cracking in Thick-Wall, Large-Diameter, Stainless Steel, Recirculation System Piping at BWR Plants

                                                           SSINS NO.: 6820 
                                                           IEB 82-03 Rev. 1 

                               UNITED STATES 
                        NUCLEAR REGULATORY COMMISSION
                    OFFICE OF INSPECTION AND ENFORCEMENT
                           WASHINGTON, D.C. 20555 
                                     
                              October 28, 1982

IE BULLETIN NO. 82-03, REVISION 1: STRESS CORROSION CRACKING IN THICK-WALL, 
                                   LARGE-DIAMETER, STAINLESS STEEL, 
                                   RECIRCULATION SYSTEM PIPING AT BWR PLANTS

Addressees:  

Those licensees of operating boiling water reactors (BWRs) identified in 
Table 1 for action. All other licensees and holders of construction permits 
(CPs) for information only. 

Purpose: 

This bulletin is to notify all licensees and CP holders about a matter that 
may have a high degree of safety significance, and to require specific 
actions as set forth below for those licensees listed in Table 1. 
Specifically, this matter involves the degradation in the recirculation 
system piping in the reactor coolant pressure boundary . (RCPB) that was 
found at the Nine Mile Point Unit 1 Nuclear Generating Station. This 
information was described in considerable detail in Information Notice 
82-39, dated September 21, 1982. Action by the affected licensees identified 
in Table 1 is required to (1) provide a reasonable level of assurance that 
inspections which are currently being performed or scheduled are sufficient 
to detect cracking in BWR thick-wall recirculation piping welds* and (2)  R1
to assist the NRC in determining the generic significance of the piping 
degradation found at Nine Mile Point.  The affected licensees are those 
owners whose plants are currently in or scheduled to be in a refueling mode 
or extended outage through January 31, 1983. 

This bulletin is provided to all other licensees and holders of construction
permits for information only at this time. Licensees not listed in Table 1 
will be notified by January 15, 1983 as to the scope and extent of any 
required actions.  

Description of Circumstances: 

During a primary system hydrotest in March 1982 at Nine Mile Point Unit 1 
(NMP-1), leakage was visually detected at two of the ten  

*Large bore piping that is not designated as "Service-Sensitive" in       R1
 accordance with NUREG-0313, Rev. 1. It should be noted that NUREG-0313,  R1
 Rev. 1 designates the recirculation riser lines as "Service-Sensitive.   R1



8208190240 
.                                                         IEB 82-03, Rev. 1 
                                                         October 28, 1982  
                                                         Page 2 of 5       

                                  Table 1 

                Plants Currently in or Scheduled to Be in 
       a Refueling Mode or Extended Outage Through January 31, 1983 

LICENSEE                                PLANT 

Northern States Power Company           Monticello Nuclear Generating 
Station 

Tennessee Valley Authority              Browns Ferry Unit 2 Nuclear  
                                        Generating  Station 

Commonwealth Edison Company             Quad Cities Unit 1 Nuclear 
                                        Generating Station 
                                        Dresden Unit 2 Nuclear Generating 
                                          Station 

Northeast Utilities                     Millstone Unit 1 Nuclear Generating 
                                          Station 

Georgia Power Company                   Hatch Unit 1 Nuclear Generating 
                                          Station 

Carolina Power & Light Company          Brunswick Unit 1 Nuclear Generating 
                                          Station* 

Jersey Central Power & Light Company    Oyster Creek Nuclear Generating 
                                          Station 

Iowa Electric Light & Power Company     Duane Arnold Nuclear Generating 
                                          Station 

*To be performed during the November 1982 refueling outage, not the current 
 outage. 
.

                                                         IEB 82-03, Rev. 1 
                                                         October 28, 1982  
                                                         Page 3 of 5       

furnace-sensitized, recirculation system safe-ends. Further visual 
inspection revealed three pinhole indications and a single 1/2-inch-long 
axial indication, all of which were located in the heat-affected zone of the 
welds where the safe-end joined the pipe. About nine months before the leak, 
these safe-ends were ultrasonically (UT) inspected; at that time, the 
inspection did not disclose any reportable indications. Subsequent to the 
leak, the UT procedure was modified; UT examination of the two affected 
safe-ends and one other safe-end confirmed the presence of indications of 
intermittent cracking around the pipe's inside diameter (ID). Additional 
examinations revealed cracking in heat-affected zones of recirculation pump 
discharge welds. Dye penetrant examination confirmed these crack 
indications. The UT examinations werg extended to other welds in the five 
loops of the recirculation system. The results of these examinations 
disclosed ID cracking in a large number of the welds examined. 

Two boat samples removed from the area of the through-wall cracks in one 
safe-end were sent for evaluation--one to General Electric Co. and the other
to Battelle Laboratories. In addition, a boat sample from the crack region 
of the elbow weld was evaluated by Sylvester Associates, consultants to the 
licensee. The results of these metallurgical evaluations concluded that the 
degradation resulted from intergranular stress corrosion cracking (IGSCC) in
the sensitized region of the weld's heat-affected zones. 

Based on the fact that NMP-1 has furnace-sensitized safe-ends, the licensee 
decided to replace all 10 recirculation system safe-ends without further 
investigation beyond that described above. Based on recirculation system 
findings, the licensee decided to also replace all recirculation system 
piping while the facility was shut down for safe-end replacement. 

On September 16, 1982, a meeting was held between General Electric, BWR 
licensees, and NRC staff to review past IGSCC experiences and the general 
implications of NMP-1 IGSCC degradation in main recirculation piping welds. 
The staff had the benefit of the metallurgical evaluation of the NMP-1 event
and an update of the general IGSCC experiences relative to all operating BWR
plants. 

On September 27, 1982, a meeting was held between BWR licensees and the NRC 
staff to discuss the extent and results of examining welds in the 
recirculation system for all BWR licensees with plants currently in or 
scheduled to be in a refueling mode or extended outage through January 31, 
1983. As a result of this meeting, the NRC staff has determined that 
additional information is needed to assess the effectiveness of the UT 
methods employed or planned to be used and to determine whether such piping 
should be designated "service- sensitive" in accordance with NUREG-0313, 
Rev. 1, issued by NRC letter dated February 26, 1981. 

To provide a reasonable level of assurance that inspections which are 
currently being performed or scheduled are sufficient to detect cracking in 
thick-wall, recirculation system piping welds and to assist the NRC in 
further evaluating this issue, the affected licensees (identified in Table 
1) are requested to take the following actions. 
.

                                                         IEB 82- 03, Rev. 1 
                                                         October 28, 1982  
                                                         Page 4 of 5       

Actions To Be Taken by Licensees of BWR Facilities Identified in Table 1: 

1.   Before resuming power operations following the current refueling or 
     extended outage, the licensee is to demonstrate the effectiveness of 
     the detection capability of the ultrasonic methodology used or planned 
     to be used to examine welds in recirculation system piping. This 
     demonstration shall be made on representative service-induced cracked 
     pipe samples. Arrangements should be made to allow NRC to witness this 
     demonstration. This demonstration shall employ those procedures and 
     standards, the same type of equipment (same transducer size, 
     frequencies and calibration-standards), and representative UT personnel 
     from the inservice inspection (ISI) organization utilized or to be 
     utilized in the examinations at the plant site.* 

2.   Before resuming power operations following the current refueling or 
     extended outage, the licensee is to provide a listing of results of 
     recirculation system piping inspections. 

 3.  Before resuming power operations following the current refueling or 
     extended outage, the licensee (if the inspections indicate the presence
     of cracks) is to describe the corrective actions taken and report these
     in accordance with the appropriate regulations. 

4.   To assist NRC's further evaluation of this issue, the following shall 
     be submitted by December 1, 1982: 

     a.   A description of the sampling plan used or to be used during this 
          outage for UT examinations of recirculation system piping welds 
          and the bases for the plan. The description should: 

          (1)  Provide an isometric drawing of the recirculation system 
               piping showing all the welds, and the number of welds and 
               their location that have been examined or will be examined. 

          (2)  Identify criteria for weld sample selection (e.g., stress 
               rule index, carbon content, high stress location, and their 
               values for each weld examined). 

          (3)  Describe piping material(s), including material type, 
               diameter, and wall thickness. 

          (4)  Estimate the occupational radiation exposure incurred or 
               expected and briefly summarize measures taken to maintain 
               individual and collective exposures as low as reasonably 
               achievable. 

*We understand that Electric Power Research Institute (EPRI) has arranged to
 have samples from the Nine Mile Point Unit 1 plant available for industry 
 demonstrations of UT methodology. The samples have been taken to Battelle 
 Memorial Institute in Columbus, Ohio for characterization and subsequent 
 use. 
.

                                                         IEB 82-03, Rev. 1 
                                                         October 28, 1982  
                                                         Page 5 of 5       

     b.   A summary description of the UT procedures and calibration 
          standards used or to be employed in the examination at the 
          licensee's plant site. This description should include the 
          scanning sensitivity, the evaluation sensitivity and the recording 
          criteria. 

     c.   A summary of the results of any previous inspection of the 
          recirculation system piping welds which used the validated 
          examination methodology as discussed in Action Item 1 above. 

     d.   An evaluation of the crack-detection capability of ultrasonic 
          methodology used or planned to be used to examine recirculation 
          system piping welds. This evaluation should result from conducting
          the demonstration required in Action Item 1 above, and should 
          include a comparison of the service-induced pipe crack sample to 
          those welds actually examined in the licensee's plant in terms of 
          pipe wall thickness and diameter, weld geometry, and materials. 

5.   The written reports required by Items 2, 3, and 4 shall be submitted to
     the appropriate Regional Administrator under oath or affirmation under 
     provisions of Section 182a, Atomic Energy Act of 1954, as amended. The 
     original copy of the cover letters and a copy of the reports shall be 
     transmitted to the U.S. Nuclear Regulatory Commission, Document Control
     Desk, Washington, D.C. 20555 for reproduction and distribution. 

This request for information does not require Office of Management and 
Budget approval since the number of plants asked to provide the information 
is limited to nine reactor plants. 

Although no specific request or requirement is intended, the following 
information would help the NRC evaluate the cost of implementing this 
bulletin: 

     o    Staff time to perform requested demonstration 

     o    Staff time to prepare written responses 

If you have any questions regarding this matter, please contact the Regional
Administrator of the NRC Regional Office or one of the technical contacts 
listed below. 


                                   Richard C. DeYoung, Director 
                                   Office of Inspection and Enforcement 

Technical Contact:  William J. Collins, IE
                    492-7275 

                    Warren Hazelton, NRR
                    492-8075

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