Information Notice No. 90-68: Supplement 1: Stress Corrosion Cracking of Reactor Coolant Pump Bolts

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                            WASHINGTON, D.C.  20555

                                April 14, 1994


NRC INFORMATION NOTICE 90-68, SUPPLEMENT 1:  STRESS CORROSION CRACKING OF
                                             REACTOR COOLANT PUMP BOLTS

Addressees

All holders of operating licenses or construction permits for pressurized-
water reactors (PWRs).

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to the failure of Westinghouse reactor coolant pump
turning vane cap screws at the Millstone Nuclear Generating Station, Unit 3. 
In this event, the failed cap screws were fabricated from material that has
demonstrated susceptibility to intergranular stress corrosion cracking (IGSCC)
as were the ones in the original information notice.  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 September 2, 1993, the licensee for Millstone Unit 3 was inspecting the
reactor lower core support plate before reloading fuel.  The licensee
discovered pieces of a locking cup for the Westinghouse model 93A-1 reactor
coolant pump turning vane cap screws.  The cap screws connect the flanged
interfaces of the turning vane and thermal barrier.  The joint is gasketed to
prevent bypass leakage of reactor coolant beyond the thermal barrier.  

The licensee subsequently removed four turning vane cap screws for inspection. 
A visual and liquid penetrant inspection at the juncture of the head and body
of the cap screws revealed cracks in two cap screws.  One cap screw had no
cracks.  The head of the fourth cap screw was almost completely severed.  The
cap screws are made of alloy A286 stainless steel, designated by the American
Society for Testing and Materials as A453 grade 660.  The cap screw or cap
screw head may deform, loosen, fracture, or fail the locking cup restraints.
Cap screw failures could present a safety hazard because failed parts could
enter the reactor coolant system and cause damage to vital components.   

The licensee has replaced all four Unit 3 reactor coolant pumps.  The
replacement pumps incorporate design improvements recommended by Westinghouse
to reduce the susceptibility of the cap screws to IGSCC.  These improvements
include:



9404080152.                                                           IN 90-68, Supp. 1
                                                           April 14, 1994
                                                           Page 2 of 2   


1.    increased head-to-shank fillet radius between the cap and body of the
      cap screw to reduce stress concentration;

2.    reduced thread engagement to decrease axial and bending stiffness;

3.    reduced bolt shank diameter to lessen bending stresses; and

4.    replacement of the A-286 alloy with cold-worked 316 stainless steel.

Discussion

Information Notice 90-68 described similar reactor coolant pump bolt failures
at a foreign reactor.  The failed bolts at the foreign reactor were also used
to fasten the turning vanes of the reactor coolant pump.  The bolts were of
the same material as that of the Millstone turning vane cap screws.  The
information notice stated that considerable data have been available regarding
the susceptibility of alloy A-286 to IGSCC.  Cap screws fabricated of alloy 
A-286 are subject to IGSCC at peak stresses above 689 MPa [100 ksi].  The 
occurrence of IGSCC also appears to be a function of chromium content,
fabrication practice, and environment.  

The Brookhaven National Laboratory conducted an extensive study of bolting. 
The results of the study are documented in NUREG/CR-3604, "Bolting
Applications," May 1984.  The report discussed the direct relationship between
loading and IGSCC of bolts fabricated of alloy A-286.  This report recommended
that alloy A-286 not be used as a reactor structural material because of its
susceptibility to IGSCC.            

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

                                    /S/'D BY CIGRIMES/FOR


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

Technical contact:  N. Fields, NRR
                    (301) 504-1173