Information Notice No. 88-01: Safety Injection Pipe Failure
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
January 27, 1988
Information Notice No. 88-01: SAFETY INJECTION PIPE FAILURE
Addressees:
All holders of operating licenses or construction permits for nuclear power
reactors.
Purpose:
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.
8801210097
. 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.
Discussion:
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
office.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contact: Roger Woodruff, NRR
(301) 492-7096
Attachments:
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
licensees.
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
Room
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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