Information Notice No. 88-01: Safety Injection Pipe Failure

                                  UNITED STATES
                          NUCLEAR REGULATORY COMMISSION
                             WASHINGTON, D.C.  20555

                                January 27, 1988

Information Notice No. 88-01:  SAFETY INJECTION PIPE FAILURE 


All holders of operating licenses or construction permits for nuclear power 


This information notice is to alert addressees to a potentially generic 
problem concerning the reliability of piping in safety-related systems due to 
valve leakage which results in thermal cycling of the piping.  Recipients are 
expected to review the information for applicability to their facilities and 
consider actions, if appropriate, to preclude similar problems from occurring 
at their facilities.  However, suggestions contained in this information 
notice do not constitute NRC requirements; therefore, no specific action or 
written response is required. 

Description of Circumstances: 

On December 9, 1987, while restarting Farley Unit 2 after a refueling outage, 
the licensee noted increased moisture and radioactivity within containment.  
The unidentified leak rate for the RCS was determined to be 0.7 gpm.  After 
entering containment to identify the location of the leak, licensee personnel 
determined that the leak could not be isolated.  The reactor, which was at 33 
percent power, was shut down to facilitate repair. 

By ultrasonic testing, the licensee found an indication of a crack on the 
interior surface of the 6-inch ECCS piping connected to the cold leg of RCS 
Loop B.  The indication was located at a weld connecting an elbow and a hori-
zontal spool, as shown in Attachment 1.  Further, the indication was on the 
underside of the pipe and extended circumferentially 60 degrees in both direc-
tions from the bottom of the pipe.  The crack extended through the wall for 
approximately 1 inch at the center of the indication.  Visual and metallo-
graphic examinations showed that the weld had failed as a result of fatigue 
after roughly one million stress cycles.  The licensee examined the operating 
records and determined that the number of stress cycles imposed by starting up 
and shutting down and by safety injections was significantly less than the 
relevant design criteria. 

.                                                            IN 88-01 
                                                            January 27, 1988 
                                                            Page 2 of 3 

On the basis of this information, the licensee postulated that the stress 
loads were (1) thermal and created by valve leakage or convective flow cells 
or (2) mechanical and created by flow-induced vibrations.  To test these 
postulations, the licensee replaced the failed piping and installed sensors 
for temperature and acceleration near the location of the failed weld and at a 
location 25 to 30 inches upstream from the failed weld, that is, on the other 
side of the check valve.  The licensee also installed sensors at similar loca-
tions on the ECCS pipe connected to Loop C.   At each location the sensors 
were distributed circumferentially around the pipe. 

Data from the sensors demonstrated that there was an adverse temperature 
distribution in the Loop B ECCS piping as shown in Attachment 1.  The 
circumferential temperature difference at the location of the failed weld was 
215� F.  Further, the temperature at the bottom of the pipe fluctuated as much 
as 30� F in 30 seconds.  This spatial and temporal distribution was caused by 
failure of the valve in the bypass pipe around the boron injection tank (BIT) 
to seat properly.  The valve, which is shown in Attachment 2, is believed to 
be the cause of failure of the weld.  Leakage through the valve apparently 
caused the check valves in the Loop B ECCS pipe to partially open, or chatter, 
admitting relatively cold coolant to the unisolable portion of the pipe 
between the nozzle and the first check valve.  Temporarily redirecting the 
valve leakage away from the ECCS manifold changed the temperature 
distribution, as shown in Attachment 1.  It should be noted that there may be 
other safety-related piping in both PWRs and BWRs which could experience 
similar fatigue due to thermal cycling. 

Data from the temperature sensors for Loop C indicated that the check valves 
in that pipe were not chattering and that the temperature distribution was 
normal.  Further, none of the accelerometers indicated adverse mechanical 
stress cycling. 

Examination of the analysis of record for the small-break, loss-of-coolant 
accident indicated that double-ended failure of the unisolable ECCS pipe may 
not have been enveloped. 


A generic safety question may exist for those plants having dual purpose pumps 
that are used for charging the RCS with coolant during normal operation and 
injecting emergency core coolant at high pressure following an accident.  
During normal operation, with one of the pumps providing charging flow to the 
RCS via the normal charging piping and with a leaking valve allowing coolant 
to flow to the ECCS manifold, pressure in the manifold will exceed RCS 
pressure and check valves in the ECCS piping will open admitting relatively 
cold coolant to the RCS.  The flow rate via this additional path or paths is 
determined by the throttling that occurs in the leaking valve.  If the check 
valves in more than one ECCS pipe open, then more than one unisolable ECCS 
failure may occur.  Subjecting the flawed piping to excessive stresses induced 
by a seismic event, water hammer, or some other cause conceivably could result 
in simultaneous double-ended failure of more than one ECCS pipe.   
.                                                            IN 88-01 
                                                            January 27, 1988 
                                                            Page 3 of 3 

Corrective action for this common-mode failure would include redesigning the 
piping, instrumenting unisolable and adjacent portions of the piping to detect 
cyclic or abnormal thermal stresses, instrumenting the ECCS manifold to detect 
pressure resulting from valve leakage, or providing additional surveillance. 

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 

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

Technical Contact:  Roger Woodruff, NRR
                    (301) 492-7096

1.  Farley 2 Temperature Data
2.  Farley 2 ECCS
3.  List of Recently Issued NRC Information Notices
.                                                            Attachment 3
                                                            IN 88-01 
                                                            January 27, 1988 
                                                            Page 1 of 1

                             LIST OF RECENTLY ISSUED
                            NRC INFORMATION NOTICES 
Information                                  Date of 
Notice No._____Subject_______________________Issuance_______Issued to________

86-81,         Broken External Closure       1/11/88        All holders of OLs
Supp. 1        Springs on Atwood & Morrill                  or CPs for nuclear
               Main Steam Isolation Valves                  power reactors. 

87-67          Lessons Learned from          12/31/87       All holders of OLs
               Regional Inspections of                      or CPs for nuclear
               Licensee Actions in Response                 power reactors. 
               to IE Bulletin 80-11 

87-66          Inappropriate Application     12/31/87       All holders of OLs
               of Commercial-Grade                          or CPs for nuclear
               Components                                   power reactors. 

87-28,         Air Systems Problems at       12/28/87       All holders of OLs
Supp. 1        U.S. Light Water Reactors                    or CPs for nuclear
                                                            power reactors. 

87-65          Plant Operation Beyond        12/23/87       All holders of OLs
               Analyzed Conditions                          or CPs for nuclear
                                                            power reactors. 

87-64          Conviction for Falsification  12/22/87       All nuclear power 
               of Security Training Records                 reactor facilities
                                                            holding an OL or 
                                                            CP and all major 
                                                            fuel facility 

87-35,         Reactor Trip Breaker          12/16/87       All holders of OLs
Supp. 1        Westinghouse Model DS-416,                   or CPs for nuclear
               Failed to Open on Manual                     power reactors. 
               Initiation From the Control 

87-63          Inadequate Net Positive       12/9/87        All holders of OLs
               Suction Head in Low Pressure                 or CPs for nuclear
               Safety Systems                               power reactors. 

87-62          Mechanical Failure of         12/8/87        All holders of OLs
               Indicating-Type Fuses                        or CPs for nuclear
                                                            power reactors. 
OL = Operating License
CP = Construction Permit 

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