United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 86-106, Supplement 3: Feedwater Line Break

                                UNITED STATES
                           WASHINGTON, D.C. 20555

                              November 10, 1988

Information Notice No. 86-106, SUPPLEMENT 3:  FEEDWATER LINE BREAK 


All holders of operating licenses or construction permits for nuclear power 


This supplement to Information Notice (IN) 86-106 is intended to provide 
addressees with additional information about a potential problem that 
resulted in thinning of secondary system piping at both units of an 
operating nuclear power station and the catastrophic failure of a main 
feedwater suction pipe. It is expected that recipients will review the 
information for applicability to their facilities and consider actions, as 
appropriate, to avoid similar problems. However, suggestions contained in 
this information notice do not constitute NRC requirements; therefore, no 
specific action or written response is required. 


IN 86-106, Supplement 1, was issued on February 13, 1987, to provide 
additional information regarding the catastrophic feedwater line break at 
Surry Power Station, Unit 2. Results of the licensee's failure analysis and 
the NRC technical panel conclusions concerning the pipe break failure 
mechanism were discussed in Supplement 1. Supplement 2 addressed system 

During the September 1988 outage, the Surry licensee discovered that pipe 
wall thinning had occurred more rapidly than expected. On the suction side 
of one of the main feedwater pumps, an elbow that was installed during the 
1987 refueling outage lost 20 percent of its 0.500 inch wall in 1.2 years. 
In addition, wall thinning is continuing in safety-related main feedwater 
piping and in other non-safety-related condensate piping. 

On the basis of partial inspection results, the licensee indicated that the 
broad area thinning rate for the replacement piping, installed during the 
last refueling outage, is roughly 60 mils/year. The maximum localized 
thinning rate is 90 mils/year. These rates were higher than the 20 to 30 
mils/ year rate estimated previously. The estimated rate of 20 to 30 
mils/year was based on a single measurement and an assumption that wall 
thinning had been 


                                                  IN 86-106, Supplement 3 
                                                  November 10, 1988 
                                                  Page 2 of 2 

progressing linearly since initial full-power operation was achieved. This 
new rate of wall thinning, which is based on a second data point, indicates 
that significant wall thinning may have coincided with a reduction in 
feedwater dissolved-oxygen concentration subsequent to steam generator 
replacement. The lower rate of wall thinning associated with a higher 
feedwater dissolved-oxygen concentration is consistent with, the low rates 
of erosion/ corrosion reported in IN 88-17, "Summary of Responses to NRCB 
87-01, 'Thinning of Pipe Walls in Nuclear Power Plants'," for boiling water 
reactors (BWRs), which typically operate at a feedwater dissolved-oxygen 
concentration of approximately 30 ppb. The licensee is continuing its 
failure analysis to determine the cause(s) of the increase in the estimated 
pipe wall thinning rate. Because the measured rate of pipe wall thinning is 
in excess of the previously estimated rate, the scheduled frequency of 
future inspections may need to be reexamined to ensure that code-allowable 
wall thickness is maintained. 

Additional information pertaining to erosion/corrosion in feedwater-conden-
sate system piping can be found in Information Notice No. Nos. 86-106; 
86-106, Supplement 1; 87-36, "Significant Unexpected Erosion of Feedwater 
Lines"; 88-17; and NRC Bulletin 87-01, "Thinning of Pipe Walls in Nuclear 
Power Plants." No specific action or written response is required by this 
information notice. If you have any questions about this matter, please 
contact the technical contact listed below or the Regional Administrator of 
the appropriate regional office. 

                              Charles E. Rossi, Director
                              Division of Operational Events Assessment 
                              Office of Nuclear Reactor Regulation 

Technical Contact:  Paul Wu, NRR 

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