United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 92-50: Cracking of Valves in the Condensate Return Lines of a BWR Emergency Condenser System

                               UNITED STATES
                          WASHINGTON, D.C.  20555

                               July 2, 1992

                               LINES OF A BWR EMERGENCY CONDENSER SYSTEM


All holders of operating licenses or construction permits for boiling water
reactors (BWRs).


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to inform addressees of cracking found in valves in the condensate
return lines of the emergency condenser system at the Nine Mile Point Nuclear
Station, Unit 1.  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

Following a reactor trip at Nine Mile Point, Unit 1, on May 1, 1992, the
licensee (Niagara Mohawk Power Corporation) inspected the drywell to
investigate the cause of a recent increase of unidentified leakage in the
reactor coolant system.  The licensee found a 0.5 gpm leak coming from a
manual gate valve at a 1-inch drain line connection.  The leaking gate valve,
designated valve 39-02, is located in the condensate return line for the
loop 12 emergency condenser system.  

The emergency condenser system has two independent loops (loops 11 and 12). 
Figure 1 shows the configuration of the condensate return line in loops 11 
and 12.  In the condensate return line, a manual gate valve is connected
downstream of a tilting disc check valve.  At each of those two valves, two
1-inch drain lines are connected to the bottom part of the valve body with
one drain line at the upstream side and the other one at the downstream side
of the valve.  The valve bodies are made of CF8M cast stainless steel. 

While investigating the leakage at the manual gate valve 39-02, the licensee
removed the internal components of the adjacent check valve to perform a
visual test (VT), a radiographic test (RT), and an ultrasonic test (UT).  The
licensee visually observed cracks on the inside surfaces at both valves in
loop 12.  At gate valve 39-02, the licensee found cracks near each of the two
drain holes.  At check valve 39-04, the licensee found cracks near a
downstream drain hole and found evidence of cracking in the threads of the

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upstream drain.  These cracks were further examined radiographically and
ultrasonically.  The licensee found four cracks including a throughwall crack
near the drain hole upstream of gate valve 39-02.  The licensee reported the
throughwall crack to be about 3.5 inches long and oriented radially outward
from the hole.  The other three cracks were all reported to be within 0.15
to 0.35 inch of passing through the wall (1.25 inch wall thickness).  The
licensee found two cracks in the drain hole area downstream of the gate valve
(39-02), which is the unisolable side of the valve body.  The licensee
reported the deepest crack to be about 1 inch long and within 0.15 inch of
passing through the wall.  The licensee found four cracks near the drain hole
downstream of check valve 39-04 with the deepest reported to be within
0.1 inch of passing through the wall.  The licensee visually observed one
small indication on the seat ring in manual gate valve 39-02.  The licensee
also examined valves 39-01 and 39-03 in the condensate return line for loop
11 of the emergency condenser system and found two cracks near the drain hole
upstream of manual gate valve 39-01.  The largest crack was reported to be
about 1.25 inch long and 1 inch deep.  The licensee reported the cracking of
the valve body in loop 11 to be less severe than that in loop 12.  The
licensee observed cracking indications on the inside surface of a butt weld
that joins the gate valve to the check valve but did not confirm these
indications by the radiographic examination.  The licensee ultrasonically
examined selected piping welds inboard of the condensate return isolation
valves and found no indications.  

The licensee removed a boat sample containing a 0.5-inch long crack from 
manual gate valve 39-02 in loop 12.  The licensee examined the boat sample
metallographically and fractographically (using a scanning electron
microscope) and found that the crack had propagated transgranularly with very
little secondary cracking.  These features are typical of fatigue crack
propagation.  The licensee noted possible fatigue striations that were not
well-developed.  The licensee measured the delta ferrite content of the boat
sample to be about 15 percent.


The emergency condensate system at Nine Mile Point, Unit 1, which is
connected directly to the reactor coolant system, operates by natural
circulation and acts as a backup for the main condenser to remove the reactor
decay heat following a reactor isolation.  The emergency condenser system at
Nine Mile Point, Unit 1, as shown in Figure 2, has two loops (loop 11 and
loop 12) with two condensers in each loop.  During normal plant operation,
the condensate return isolation valves (39-05 and 39-06) are closed, and the
steam isolation valves (39-07, 08, 09, and 10) are open in each loop.  As
shown in Figure 1, two valves, a manual gate valve and a tilting disc check
valve, are located in  horizontal sections of the condensate return line. 
The horizontal sections are connected to the suction side of the
recirculation piping system.  The manual gate valves are maintenance valves
and are open during normal operation.

The licensee postulated thermal fatigue as the root cause of the cracking in
the valve bodies, upon considering the straight and transgranular cracking
morphology, the location of the cracks on the bottom surface near.                                                            IN 92-50
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discontinuities, and the orientation of the cracks.  However, the licensee
did not find the direct causes of the apparent thermal stratification and
cycling at the affected valves.  The licensee speculated that the observed
cracking may have been caused by the leaking of the cold water from the
condensate isolation valves (39-05 and 06) and the periodic opening of the
tilting disc in the check valve.  The licensee provided a limited history of
the time and temperature as evidence of thermal cycling in the loop 12
condensate return line valve 39-06.  Although the licensee also observed
cracking in loop 11, it did not observe such thermal cycling on the
condensate return line during a 1-week test.  The sections of the emergency
condenser condensate return lines that showed evidence of cracking are
classified as American Society of Mechanical Engineers (ASME) Code Class 1. 
The licensee extended its current outage to complete acceptable code repairs
because of the extent of the cracks in the reactor coolant pressure boundary
and, in particular, the cracks found at the downstream drain line hole for
valve 39-02.  

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.

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

Technical contacts:  William H. Koo, NRR
                     (301) 504-2706

                     Robert A. Hermann, NRR
                     (301) 504-2768

1.  Figure 1, NMP-1 Emergency Condenser System Condensate Return
      Line Configuration Inside Drywell
2.  Figure 2, Nine Mile Point Unit One Emergency Condenser System
      Simplified Diagram
3.  List of Recently Issued NRC Information Notices
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