Information Notice No. 92-57: Radial Cracking of Shroud Support Access Hole Cover Welds

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

                                August 11, 1992

                               COVER WELDS


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 alert addressees to the potential for radial cracking of the welds
of the shroud support access hole covers (AHCs) inside the reactor vessel. 
The radial cracking could propagate into the reactor vessel attachment welds. 
It is expected that recipients will review this 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


Jet pump boiling water reactors (BWRs) are designed with access holes in the
shroud support plate at the bottom of the annulus between the core shroud and
the reactor vessel wall.  Each shroud support plate in the reactor vessel has
two access holes located 180 degrees apart.  These holes are used for access
during construction and are subsequently closed by welding a plate over the
hole.  The AHCs and most shroud support plates are fabricated from Alloy 600. 
The connecting weld material is Alloy 182 or 82.  The high residual stresses
resulting from welding, the crevice geometry of the weld, and the aggressive
water chemistry, could cause intergranular stress corrosion cracking (IGSCC). 

On January 21, 1988, the Philadelphia Electric Company (the licensee) found
significant circumferential cracking in the shroud support AHC welds at the
Peach Bottom Atomic Station, Unit 3, using a remotely operated ultrasonic
testing (UT) fixture specially designed by the General Electric Company (GE)
for this examination.  This was the first time such cracking was reported in a
domestic BWR plant.  The NRC reported the results of a preliminary inspection
of this event in Information Notice No. 88-03, "Cracks in Shroud Support
Access Hole Cover Welds," of February 2, 1988.  In August 1988, the licensee
performed another ultrasonic inspection using several improved examination
techniques.  The licensee found that these cracks, located in the area of the
vertical fusion line on the shroud support ledge side of the welds, appeared
to initiate from the vertical crevices of the AHCs.  


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Licensees used the improved techniques to perform examinations at a number of
other BWRs.  The licensee found similar cracking at the ledge side of the
AHC welds at the Quad Cities Station, Unit 2.  GE provided reactor operating
guidelines for detecting core bypass flow in SIL No. 462, Supplement 2,
Revision 1, of December 19, 1990.  GE also recommended the affected BWR owners
inspect the creviced Alloy 600 access hole cover plates ultrasonically during
the next outage and repeat the examination at intervals of no longer than
3 years.

Description of Circumstances

In SIL No. 462, Supplement 3, of June 6, 1992, GE reported the discovery of
apparent radial cracking in an AHC weld in a domestic BWR/4 plant.  The radial
cracking was discovered during a review of the visual examination data.  Video
examination had been performed to supplement the ultrasonic (UT) examination
to confirm evidence of through thickness cracks.  The UT examination detected
significant circumferential cracks but did not detect the radial cracking. 
Some of the radial cracking appeared to have penetrated into the alloy 182
reactor vessel attachment weld of the shroud support plate.  GE is developing
a specialized UT examination to confirm the extent and the size of radial


GE's preliminary evaluation indicated that the radial cracking would not pose
an immediate safety concern because the attachment weld areas have a large
tolerance for cracks and have large structural margins.  However, GE
recommended that licensees perform safety evaluations at each plant when
cracking is found.  The following issues associated with the radial cracking
in the access hole cover weld were discussed in GE SIL No. 462, Supplement 3:

(1)  Detection of radial cracking.  The current UT procedures recommended by  
     GE for IGSCC inspection of AHCs will not detect the radial cracking. 
     Therefore, special ultrasonic techniques designed for detecting the
     radial cracking are needed to inspect AHCs.

(2)  Repair of radial cracking.  The current repair technique is designed  
     for mitigating circumferential and minor radial cracking of AHCs.  Repair
     methods for extensive radial cracking need to be developed for plants in
     which such cracking is found. 

(3)  Integrity of attachment welds.  GE reported that some radial cracking    
     penetrated into the alloy 182 reactor vessel attachment weld of the
     shroud support plate.  Such cracking could challenge the structural
     margins of the attachment weld depending on the orientation and the
     extent of the crack growth.  Alloy 182 is susceptible to IGSCC which
     could cause further crack growth.  However, the extent of the crack
     growth in the attachment weld would depend on the magnitude and the
     distribution of the welding residual stresses.  .

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(4)  Inspection of access hole covers.  To ensure early detection of cracking
     in AHCs, GE SIL No. 462, Supplement 3, made the following recommendations
     for inspections:
     (a)  BWR owners that have not yet inspected the AHCs should perform the
          UT and visual examinations during the next outage, looking for both
          the circumferential and radial cracking.

     (b)  BWR owners that performed remote visual examination should review
          the data and the video tape records for the presence of radial

     (c)  All BWR owners should examine the AHCs once again for radial
          cracking sooner than the 3-year interval recommended in SIL No. 462,
          Supplement 2, Revision 1. 

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.

                                        ORIGINAL SIGNED BY

                                   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

Attachment:  List of Recently Issued NRC Information Notices

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