Information Notice No. 82-39: Service Degradation of Thick Wall Stainless Steel Recirculation System Piping at a BWR Plant

                                                            SSINS No.: 6835 
                                                            IN 82-39  

                               UNITED STATES 
                          WASHINGTON, D. C. 20555 

                             September 21, 1982 

Information Notice No. 82-39:   SERVICE DEGRADATION OF THICK WALL 
                                   STAINLESS STEEL RECIRCULATION SYSTEM 
                                   PIPING AT A BWR PLANT 


All boiling water reactor facilities holding an operating license (OL) or 
construction permit (CP). 


This notice is to provide licensees and construction permit holders 
available information about the degradation of the primary pressure boundary 
at Nine Mile Point Unit 1 due to intergranular stress corrosion cracking. 
Recipients should review this information relative to their facilities. If 
NRC evaluation so indicates, further licensee action may be requested. In 
the interim, we expect licensees to review the relevance of this information 
for applicability to their facilities. 

Description of Circumstances: 

The Nine Mile Point Nuclear Station Unit 1 (NMP Unit 1) was shut down in 
order to replace recirculation pump seals. On March 23, 1982, leakage was 
visually detected at two of the ten recirculation loop safe ends during a 
primary system hydrotest at 900 psig to test the seals. Further visual 
inspection identified three pin-hole 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. 

On March 26, 1982, an ultrasonic examination of the two affected safe ends 
and one other safe end confirmed the presence of intermittent cracking 
indications around the pipe's inside diameter. Further ultrasonic 
examination of the welds joining the pump discharge casting to the riser 
elbow also revealed cracking in weld heat affected zones on the inside 
diameter (ID) of the elbows. This was later confirmed by dye penetrant 

Because the cracks were confirmed at the welds of the safe ends and riser 
elbows, the ultrasonic examination was extended to all of the remaining 
welds in the five loops of the primary system, wherever radiation levels 
permitted. The results of this examination show ID cracking at 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 to General Electric and 
Battelle Laboratories, respectively, for evaluation. A boat sample from the 
crack region of the elbow weld was also evaluated by Sylvester Associates, 
consultants to the licensee. The results 


                                                      IN 82-39  
                                                      September 21, 1982  
                                                      Page 2 of 3 

of these metallurgical evaluations concluded the degradation was due to 
intergranular stress corrosion cracking (IGSCC) in the sensitized region of 
the welds' heat affected zones. Further metallurgical investigation is being
pursued to determine, as far as possible, the probable cause(s) of the 

Based on the results of the examinations and investigations to date, the 
licensee will replace the safe ends and 28-inch recirculation piping in all 
five loops of the system. Replacement of the branch piping out to the first 
isolation valve is also being considered; however, no final decision in this
regard has been made at this time. 

All replacement material will be stainless steel type 316 nuclear grade 
consistent with NUREG-0313, Revision 1 requirements. The actual replacement 
will be accomplished in accordance with ASME Boiler and Pressure Vessel 
Code, Section XI, 1977 Edition and Addenda through summer 1978. Welding will 
be performed in accordance with Section IX, 1978. Fitup requirements will be 
in accordance with ANSI Pressure Piping Code B31.1-1977 and Addenda through 
winter 1979. The replaced system configuration will duplicate the original 

All ten recirculation system safe ends at NMP Unit 1 had been previously 
examined volumetrically by ultrasonic techniques at each refueling outage 
under an augmented inservice inspection program. This was in addition to the
ASME code required inservice inspection program applied to other system 
welds. The augmented program was required because of IGSCC problems 
experienced with furnace-sensitized safe ends at this and other BWR plants. 

It is important to note that the programs conducted under the normal and 
augmented programs did not indicate a pending problem. Examinations were 
performed during 1979 and 1981. The procedure employed during the 1981 
augmented program for the safe ends was based on ultrasonic test (UT) using 
the EPRI transducer with a flat calibration block which was stated to be 
capable of detecting IGSCC at the code required gain or sensitivity level. 
The procedure differed from the GE recommended procedures in specifying less
gain, and differed significantly in the calibration standards and data 
recording requirements, thus resulting in reduced sensitivity compared to 
the GE recommended procedures. 

After leakage was visually observed on March 23, 1982, a UT examination of 
the safe ends was performed using the same method employed in the 1981 
augmented program. Many safe ends exhibited code "reportable," but not 
rejectable indications. However, when an ultrasonic sensitivity of 10 
decibels above code calibration sensitivity was employed, greater 
reliability was realized in detecting the presence and full extent of the 
IGSCC problems with the thick wall piping welds, both at the safe ends and 
at other locations in the reactor coolant system. The generic implications 
of the above variances is under further review by the NRC staff. 

This Information Notice No. is to advise licensees of further occurrences of 
the prevailing IGSCC problem that is under continuing review by the NRC 

                                                      IN 82-39  
                                                      September 21, 1982  
                                                      Page 3 of 3 

If you have any questions regarding this matter, please contact the Regional
Administrator of the appropriate Regional Office, or this Office. 

                                   Edward L. Jordan, Director  
                                   Division of Engineering and  
                                     Quality Assurance  
                                   Office of Inspection and Enforcement 

Technical Contact:  W. J. Collins 

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