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Information Notice No. 93-20: Thermal Fatigue Cracking of Feedwater Piping to Steam Generators
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 March 24, 1993 NRC INFORMATION NOTICE 93-20: THERMAL FATIGUE CRACKING OF FEEDWATER PIPING TO STEAM GENERATORS Addressees All holders of operating licenses or construction permits for pressurized water reactors (PWRs) supplied by Westinghouse or Combustion Engineering. Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to inform addressees of cracks found in the feedwater piping to steam generators at the Sequoyah Nuclear Power Plant, Units 1 and 2 and the Diablo Canyon Nuclear Power Plant, Unit 1. Recipients are expected to 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 In 1992, cracks were found in the feedwater lines at Diablo Canyon Unit 1. In addition, a through-wall crack and other cracks were found in the feedwater lines at Sequoyah. The cracks were attributed to thermal fatigue. The NRC staff first learned of cracks in feedwater lines to steam generators which resulted from thermal fatigue in 1979 when the Indiana and Michigan Electric Company, the licensee for the Donald C. Cook Plant, reported leaks. In dealing with this problem, the NRC staff issued the following documents: o a letter to PWR licensees pursuant to Paragraph 50.54(f) of Title 10 of the Code of Federal Regulations, May 25, 1979 o Office of Inspection and Enforcement (IE) Bulletin 79-13, "Cracking in Feedwater System Piping," June 25, 1979, Revision 1, August 30, 1979, and Revision 2, October 17, 1979 o NUREG/CR-5285, "Closeout of IE Bulletin 79-13," 1991 In Bulletin 79-13, the NRC requested that licensees perform radiographic and ultrasonic examinations of feedwater lines. As a result of these examinations, cracks were found at 18 of the 54 facilities inspected. The staff closed the bulletin on the basis of the results of the one-time inspection. The industry had taken the actions recommended in the bulletin 9303180065 . IN 93-20 March 24, 1993 Page 2 of 3 and instituted the augmented inservice inspection programs, which appeared to provide for reliable detection of cracks in feedwater piping. Other technical evaluations were documented in the following reports: o "Investigation of Feedwater Line Cracking in PWRs" (Westinghouse, 1980) o NUREG-0691, "Investigation and Evaluation of Cracking Incidents in Piping in PWRs," (PWR pipe crack study group, 1980) These studies showed that thermal fatigue was the main cause of the cracks. Modifications to minimize the effects of thermal stratification in feedwater lines and augmented licensee inservice inspections were recommended in the studies. Recently, some licensees have again reported cracks in feedwater piping. The licensee for the Sequoyah units reported an actual leak despite augmented inservice inspections. The augmented inspections using ultrasonic techniques showed indications that might earlier have revealed the cracks, but the licensee misinterpreted these as resulting from the geometric configuration of the pipe. After finding the leak, the licensee performed radiography on all feedwater nozzles of both units and found cracks in five of the eight nozzles. The licensee for Diablo Canyon Unit 1 reported indications with cracklike ultrasonic signal characteristics in feedwater piping to all four steam generators. The indications varied in length up to 20 cm [7-3/4 inches] in a circumferential direction, and many were intermittent. Some intermittent indications extended the full circumference with segments up to 5 cm [2 inches] long. The licensee tried to verify the indications by radiography but failed. Later, metallurgical analysis showed the indications to be cracks. Only one nozzle at Diablo Canyon had been scheduled for an inspection. This inspection was performed in accordance with Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). However, information on the leak at Sequoyah led the licensee for Diablo Canyon to use enhanced ultrasonic techniques to inspect all four lines. These techniques were considered more appropriate for finding small cracks from thermal fatigue than techniques specified by the ASME Code, which may not be adequate to detect these types of defects. Flaw sizing by ultrasonic techniques proved to be overly conservative at Diablo Canyon, however, presumably because inclusions in the material led to inaccurate results: cracks sized at 8.9 mm [0.35 inch] deep by ultrasonic inspection were shown by cross sectioning to be cracks 0.94 mm [0.037 inch] deep. Discussion Cracks from thermal fatigue in PWR feedwater lines have proved to be a recurring problem. The main cause of crack growth appears to be fatigue induced by stresses from thermal stratification during cold, low-flow, feedwater injections. Other factors that contribute to crack growth are a high oxygen content, counterbore weld preparation geometry, and thermal . IN 93-20 March 24, 1993 Page 3 of 3 conditions during heatup, hot standby, and low-power operation. A favored solution has been to replace the degraded piping. However, replacing the degraded piping in kind without other corrective actions to eliminate or minimize the factors which cause the cracks can leave the piping susceptible to the same problem. Inspection techniques specified in the ASME Code Section XI, do not appear adequate to find cracks of this type. 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 Brian K. Grimes, Director Division of Operating Reactor Support Office of Nuclear Reactor Regulation Technical contacts: Lee Banic, NRR (301) 504-2771 Robert A. Hermann, NRR (301) 504-2768 Attachment: List of Recently Issued NRC Information Notices .
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