United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 87-19: Perforation and Cracking of Rod Cluster Control Assemblies

                                                     SSINS No.: 6835 
                                                        IN 87-19

                                UNITED STATES
                        NUCLEAR REGULATORY COMMISSION
                    OFFICE OF INSPECTION AND ENFORCEMENT
                           WASHINGTON, D.C. 20555

                                April 9, 1987

Information Notice No. 87-19:   PERFORATION AND CRACKING OF ROD CLUSTER 
                                   CONTROL ASSEMBLIES 

Addressees: 

All Westinghouse nuclear power pressurized-water reactor (PWR) facilities 
holding an operating license or a construction permit. 

Purpose: 

This notice is provided to inform recipients of a potentially significant 
safety problem that could result from the perforation and cracking of the 
rod cluster control assemblies (RCCAs) in Westinghouse PWRs. It is expected 
that recipients will review the information for applicability and consider 
action, as appropriate to preclude a similar problem from occurring at their 
facilities. However, suggestions contained in this information notice do not 
constitute NRC requirements; therefore, no specific action or written 
response is required. 

Description of Circumstances: 

An estimate that was intended to be conservative indicated that the RCCAs 
would last for at least 15 years before the absorber cladding, a thin tube, 
would show excessive thinning as a result of sliding wear. These components 
were inspected at Point Beach Nuclear Plant, Unit 2, in 1983 after 13 years 
of operation. As a result of this inspection Point Beach reported on August 
18, 1983 that sliding wear was minor, but one control rod had a 2-in. crack 
near the tip of the rod and severe fretting wear had occurred on several 
tubes. Subsequent inspections at the Kewaunee and Haddam Neck nuclear power 
plants, which have been in operation for more than 12 years, confirmed the 
fretting wear. In addition, Haddam Neck reported tube cracking in 32 of 47 
RCCAs. 

In the event of a breach of the tubing resulting from wall thinning, 
perforation, or cracking, the immediate effect is the introduction of 
activation products from the neutron absorber material into the reactor 
coolant. Although there are large margins, another concern is the potential 
reduction in shutdown margin and negative reactivity worth. 

8704080095
.

                                                            IN 87-19     
                                                            April 9, 1987 
                                                            Page 2 of 3  

Discussion: 

Each RCCA contains 16 rods. The rods at Point Beach, Kewaunee, and Haddam 
Neck were constructed with an outer tube of 0.019-in.-thick 304 stainless 
steel that retains the absorber material (80% silver, 15% indium, 5% 
cadmium). Some newer plants use hafnium as the absorber material, while 
others use boron carbide surrounded by a 0.038-in.-thick tube. The control 
RCCAs are inserted or withdrawn to compensate for various reactivity changes 
during operation of the reactor and can trip to provide shutdown capability. 
The shutdown RCCAs are fully withdrawn from the core when the reactor is 
critical. 

At Kewaunee, marks of fretting wear about 1 inch in length, were found 
adjacent to the guide blocks that position the rods when the RCCAs are in 
their withdrawn position. The 1-in.-thick stainless steel blocks are spaced 
on 12-in. centers and each rod in the cluster passes through all eight of 
the blocks. At Point Beach the tubing wore in two modes: fretting and 
sliding of the rods over the guide blocks during rod motion. Five RCCAs at 
Haddam Neck had wall thinning resulting from fretting and four of these were 
actually wearing into the absorber material. All of the others had fretting 
wear, but to a lesser extent. 

The fretting resulted from flow-induced vibratory contact between the rods 
and the guide blocks during long periods of steady-state power operation. 
Vibration is hydraulically induced by flow of the reactor coolant; therefore 
it is a continuous process when the reactor coolant pumps are in operation. 
According to Westinghouse Electric Corp. fretting wear encompassed one-third 
of the circumference of the rod and the depth varied, with the amount of 
time the RCCAs were in the withdrawn position. 

At Point Beach significant number of short hairline cracks at the lower 
extremity of the tubing were observed near the end plug region of the rod. 
The cracks extended axially for 4 in. and penetrated the stainless tubing, 
exposing the absorber material. No circumferential cracks were found. 
Examination of the cracks showed that irradiation-induced swelling of the 
absorber was the principal cause of tensile stress in the cladding, which 
resulted in cracking after substantial irradiation. 

Where excessively worn rods were found, they have been replaced. While the 
issue is being studied by NRC and the industry, several licensees have been 
given approval to slightly change the position of the fully withdrawn RCCA 
in order to distribute the wear among different locations on the tubing. 
Westinghouse Electric Corp. reported that an increase in the amount of the 
silver isotope, Ag-110m in the reactor water is a reliable indication of 
exposure of absorber material due to cracking or fretting wear.
.

                                                            IN 87-19
                                                            April 9, 1987 
                                                            Page 3 of 3  

The NRC is continuing review of the safety significance of this information 
to determine whether further NRC action is warranted. No specific action or 
written response is required by this information notice. If you have any 
questions about this matter, please contact the Regional Administrator of 
the appropriate regional office or this office. 




                         Edward L. Jordan Director
                         Division of Emergency Preparedness
                           and Engineering Response
                         Office of Inspection and Enforcement 

Technical Contact:  Paul Cortland, IE
                    (301) 492-4175

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