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                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                          WASHINGTON, DC  20555-0001

                                April 18, 1997


NRC INFORMATION NOTICE 97-19:  SAFETY INJECTION SYSTEM WELD FLAW AT
                               SEQUOYAH NUCLEAR POWER PLANT, UNIT 2


Addressees

All holders of operating licenses or construction permits for nuclear power
reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to potential cracking of a safety injection system
line detected in a Westinghouse plant.  The crack occurred in the upstream
weld of the check valve adjacent to the unisolable section of this line, and
was determined to have been caused by intergranular stress-corrosion cracking
(IGSCC).  However, it is also possible that similar cracks could be caused in
these lines by thermal fatigue cycling from possible reactor coolant leakage
through the valve seat.  The concern therefore involves potential cracking of
piping in portions of safety-related systems due to either IGSCC or thermal
stresses induced by intermittent leakage of hot water through isolation
valves, impacting welds with considerable residual stresses due to extensive
reworking. Because both abnormal conditions could lead to similar results,
this information notice addresses the potential for both failure mechanisms.
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

On May 2, 1996, a routine American Society of Mechanical Engineers (ASME)
Section XI ultrasonic test (UT) inservice inspection (ISI) at Sequoyah Unit 2
revealed a pipe crack indication in a low-pressure safety injection system
(SIS) line.  The UT indicated a 7-inch (17.8 cm)-long, 75-percent through-wall
7-inch circumferential crack on the inside surface of the (25.4 cm) 10-inch
Schedule 140 stainless steel line.  The crack was located at the top dead
center of the base metal adjacent to the upstream weld of the check valve
closest to the Reactor Coolant System (RCS) cold leg (see figure).  This check
valve isolates the RCS from the SIS.  Although the crack was on the upstream
side of the check valve, the circumstances that created this crack initially
appeared to also reflect  those described in NRC Bulletin 88-08, "Thermal
Stresses in Piping Connected to Reactor Coolant Systems," and a review of the
plant records revealed that this valve had a history of leakage. In addition,
the records also showed that this particular weld had been reworked four times
during plant construction before it was finally accepted.


9704140153.                                                            
                                                            IN 97-19
                                                            April 18, 1997
                                                            Page 2 of 3


Discussion

Cracks in piping attached to the reactor coolant loop have been found in other
plants.  NRC Bulletin 88-08 and associated supplements describe leaking cracks
found in unisolable sections of safety injection lines and residual heat
removal lines at Farley, Unit 2.  A similar crack was also found at Nine Mile
Point, Unit 1.  These cracks are thought to have resulted from thermal cycling
induced by intermittent leakage through isolation valves.

The initiation and propagation of this particular crack was initially
attributed to intermittent leakage of hot reactor coolant through the check
valve seat into the stagnant cold fluid side of the check valve (a cause
similar to that described in the earlier bulletin).  The weld and adjacent
base metal were replaced, and a subsequent metallurgical examination
established that the cracking began at the inside diameter surface of the
counterbore region of the pipe-to-valve weld and progressed radially outward
through the sensitized region of the heat- affected zone (HAZ).  The
metallurgical evaluation concluded that the cracking in the pipe was caused by
intergranular stress-corrosion cracking which, in turn, was caused by exposure
of the sensitized region of the HAZ to oxygenated boric acid water in the SIS. 
This situation was exacerbated by additional sensitization due to the large
residual stresses caused by the four cycles of repair to this weld zone.

Identification of potential crack locations of this type may be facilitated by
considering the history of the weld zones and the leak-tightness of the check
valves in low-pressure SISs connected to RCS piping when performing required
ISI testing .  

Related Generic Communications

Problems relating to cracks in safety-related lines connected to the reactor
coolant loop have been discussed in the following NRC communications:

      NRC Bulletin 88-08:  "Thermal Stresses in Piping Connected to Reactor
      Coolant Systems," dated June 22, 1988.

      NRC Bulletin 88-08, Supplement 1:  "Thermal Stresses in Piping Connected
      to Reactor Coolant Systems," dated June 24, 1988.

      NRC Bulletin 88-08, Supplement 2:  "Thermal Stresses in Piping Connected
      to Reactor Coolant Systems," dated August 4, 1988.

      NRC Bulletin 88-08, Supplement 3:  "Thermal Stresses in Piping Connected
      to Reactor Coolant Systems," dated April 11, 1989.

.                                                           IN 97-19
                                                            April 18, 1997
                                                            Page 3 of 3


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. 


                                                signed by M.M. Slosson for

                                    Thomas T. Martin, Director
                                    Division of Reactor Program Management
                                    Office of Nuclear Reactor Regulation

Technical contacts:  Mark Hartzman, NRR         
                     (301) 415-2755                   
                     E-mail: [email protected]

                     William Burton, NRR
                     (301) 415-2853
                     E-mail: [email protected]

Attachments:  
1.  Figure - Weld Flaw Location 
(see WordPerfect file IN97019.ZIP for text and figure)
```