Information Notice No. 94-30: Leaking Shutdown Cooling Isolation Valves at Cooper Nuclear Station


April 12, 1994

                               COOPER NUCLEAR STATION


All holders of operating licenses or construction permits for nuclear power


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to a precursor to an unisolable rupture of shutdown
cooling piping with the potential for core damage and release of radioactive
material outside the containment.  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 are not NRC requirements; therefore, no specific
action or written response is required.

Description of Circumstances

While the Cooper Nuclear Station (Cooper) was operating in May 1992, the
high-pressure alarm on the residual heat removal (RHR) system suction piping
alarmed.  The normal pressure for this piping should not have been above
0.52 MPa [75 psia], the alarm setpoint was 0.79 MPa [115 psia], and the piping
design pressure was 1.14 MPa [165 psia].  The licensee established a path to
vent the low-pressure suction piping into the pressure maintenance system
(a multi-header distribution system that keeps the emergency core cooling
system discharge piping filled with water).  The licensee measured the leakage
through the inboard shutdown cooling suction isolation valve and found it to
be 1.5 lpm [0.38 gpm].  The licensee was not able to measure the leakage
through the outboard valve.

The two shutdown cooling suction isolation valves are Anchor-Darling 20-inch
nominal, double-disk, flex-wedge, gate valves, and perform the pressure
isolation function between the high-pressure primary coolant system and the
low-pressure RHR piping.  These valves also perform a containment isolation
function.  Although the plant technical specifications did not specify a limit
for pressure isolation valve leakage, the licensee concluded that the valves
were operable because the leakage was below the pressure isolation valve
leakage limit of 4 lpm [1 gpm] in the Standard Technical Specifications.  The
licensee did not evaluate the operability of the containment isolation
function of these valves.

9404060186.                                        IN 94-30
                                        April 12, 1994
                                        Page 2 of 3

While the reactor was shut down on March 29, 1993, the licensee performed a
local leak rate test on the inboard valve with pressure applied in the same
direction as the accident pressure.  The valve failed the test.  The leakage
limit was 0.57 m3/h [20 SCFH] and the measured leakage was 1.08 m3/h
[38.1 SCFH].  The integrity of this valve seat and disk is critical because
they perform the pressure and containment isolation functions of the valve.

The licensee disassembled both the inboard and outboard valves and found
cracks in the seating surfaces of both valves.  The majority of the cracks
were in the stellite facing of the normally loaded (outboard) sides of the
valve disks.  The outboard valve had one crack and the inboard valve had five
cracks on the outboard side disk facing.  Cracks were only at the "bottom"
(from about the 4 to 8 o'clock positions) of the disks and reached to the base
metal, except for two cracks in the inboard valve that extended into the base
metal, but not more than 3.2 mm [1/8 inch].  In addition, four "matching"
cracks were found in the inboard valve's loaded side seat ring.  The outboard
valve also had two cracks on the normally unloaded (inboard) side.  Crack
orientation in all cases was radial.  The licensee determined that the
cracking was most likely caused by high residual stresses in the material due
to inadequate stress relief or by fatigue resulting from differential thermal
expansion forces combined with casting voids and small flaws in the disk base
metals and pinholes in the stellite facing.


From May 1992 until March 1993, the licensee for the Cooper Nuclear Station
had operated with a reactor coolant system leak in the RHR shutdown cooling
suction line.  The leakage was initially measured to be 1.5 lpm [0.38 gpm]
through the inboard valve and was sufficient to activate the high-pressure
alarm for the suction line.  Although the cause of the leakage had not been
identified, the licensee established a path to vent the leakage from the RHR
suction piping into the pressure maintenance system, and continued operation.

In March 1993, when the licensee inspected the isolation valves for the
suction line, the licensee found cracks in the valve disks and seats.  A gross
failure of these isolation valves would have created an interfacing-systems
loss-of-coolant accident (ISLOCA), and would have pressurized the suction
piping beyond its design pressure.  A rupture of the suction piping outside
containment, coupled with postulated failure of the two valves, would have
been unisolable and may have led to a release of radioactive material outside
the containment.  Further, because the emergency cooling system could not have
been used in a recirculation mode, core damage accident sequences would have
been more likely.  The licensee failed to promptly identify and correct this
condition and operated for an extended period of time in this condition.

Related Generic Communications

NRC Information Notice 84-74, "Isolation of Reactor Coolant System From
Low-Pressure Systems Outside Containment," September 28, 1984.

.                                                            IN 94-30
                                                            April 12, 1994
                                                            Page 3 of 3

NRC Information Notice 86-40, "Degraded Ability to Isolate the Reactor
Coolant System from Low-Pressure Coolant Systems in BWRs," June 5, 1986.

NRC Generic Letter 87-06, "Periodic Verification of Leak Tight Integrity
of Pressure Isolation Valves," March 13, 1987.

NRC Information Notice 92-36, "Intersystem LOCA Outside Containment,"
May 7, 1992.

This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.


    Brian K. Grimes, Director
    Division of Operating Reactor Support
    Office of Nuclear Reactor Regulation

Technical contacts:  Elmo E. Collins, RIV
   (817) 860-8291

   Neal K. Hunemuller, NRR
   (301) 504-1152

List of Recently Issued NRC Information Notices


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