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 Attachment: List of Recently Issued IE Information Notices .