Information Notice No. 90-68: Stress Corrosion Cracking of Reactor Coolant Pump Bolts
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
October 30, 1990
Information Notice No. 90-68: STRESS CORROSION CRACKING OF REACTOR
COOLANT PUMP BOLTS
Addressees:
All holders of operating licenses or construction permits for pressurized
water reactors (PWRs).
Purpose:
This information notice is intended to alert addressees of a significant
event which occurred at a foreign reactor. The event involves the use of
a material sensitive to intergranular stress corrosion cracking (IGSCC)
in the fabrication of the bolts fastening the turning vanes of the
reactor coolant pumps. 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 do not constitute NRC requirements; therefore,
no specific action or written response is required.
Description of Circumstances:
The U.S. Nuclear Regulatory Commission (NRC) has received information
concerning cracking of the bolts fastening the turning vanes of the
reactor coolant pumps at a foreign reactor plant. There are 23 bolts
fastening the turning vanes in each of three reactor coolant pumps. Five
of these bolts had experienced stress corrosion cracking. The bolts were
made of a stainless steel alloy which is designated by the American
Society for Testing and Materials as A453 grade 660. The material is
commonly described commercially as alloy A-286. The reactor coolant
pumps are similar in design to those manufactured by the Westinghouse
Electric Company.
Considerable information has previously been available regarding the
susceptibility of alloy A-286 to IGSCC. For example, licensees of U.S.
reactors designed by Babcock and Wilcox (B&W) have noted stress corrosion
cracking of bolts fabricated of alloy A-286. Specifically, bolts
fastening the B&W reactor internals, including the core barrel and lower
thermal shield, were fabricated of alloy A-286. Between 1981 and 1984,
bolt cracking and indication of cracking were documented at Oconee,
Rancho Seco, Crystal River and Arkansas Unit 1. In response, the B&W
owners group formed a special task force to study the internal bolts of
the reactor vessel. This task force documented its conclusions in
BAW-1842, "The B&W Owners Group Evaluation of Internals Bolting Concerns
in 177 FA Plants," August 1984. The task force concluded that bolts
fabricated
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IN 90-68
October 30, 1990
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of alloy A-286 are subject to IGSCC at peak stresses above 100 ksi. The
occurrence of IGSCC also appeared to be a function of chromium content,
fabrication practice and environment. As a result of the task force
findings, the design of the reactor internal bolts was modified at B&W
plants to use fastener materials that are less sensitive to IGSCC than
alloy A-286 or to reduce the maximum stress loadings of bolts fabricated
of alloy A-286 to less than 100 ksi.
In addition, the Brookhaven National Laboratory (BNL) performed an
extensive study of bolting which is documented in NUREG/CR-3604, "Bolting
Applications," May 1984. The report discusses the direct relationship
between loading and IGSCC of bolts fabricated of alloy A-286. In this
report, BNL 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 one of the
technical contacts listed below or the appropriate NRR project manager.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contacts: Clifford Sellers, NRR
(301) 492-0703
Walton Jensen, NRR
(301) 492-1157
Attachment: List of Recently Issued NRC Information Notices
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