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**UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF NUCLEAR REACTOR REGULATIONWASHINGTON, D.C. 20555-0001**

July 24, 1998

**Addressees**

All holders of operating licenses for pressurized-water reactors (PWRs) except those who have permanently ceasedoperation and have certified that fuel has been permanently removed from the reactor vessel.

**Purpose**

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to inform licensees of instances ofsteam generator tube-end cracking. 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 in required.

**Description of Circumstances**

Entergy, the licensee for Arkansas Nuclear One, Unit 1 (ANO-1), conducted leak testing on each once-through steamgenerator (OTSG) during the spring 1998 refueling outage to identify the source of primary-to-secondary leakage measuredduring the previous operating cycle. The leak testing revealed a small leak in one tube of each OTSG. The leak in each tubewas in flaws in the tube hardroll just below the upper tubesheet seal weld. Subsequent eddy current inspections identifiedprimary-water stress-corrosion cracking (PWSCC) in each of the two tubes. The through-wall flaws in the tubes wereoriented both axially and circumferentially in the tubes. Examinations of all upper tubesheet seal weld areas in both OTSGsrevealed possible flaws in 1896 additional tubes. For these tubes, the licensee installed an additional hard roll in the uppertubesheet, thus redefining the reactor coolant pressure boundary.

Eddy current inspections completed at Davis-Besse Nuclear Power Station in April 1998 identified five tubes with "tubeend anomalies" believed to be related to the flaws identified at ANO-1. Duke Power Co. recently completed an assessmentto determine if its Babcock and Wilcox-designed, operating units, Oconee Nuclear Station, Units 1 and 3, were affected bytube-end cracking. The licensee concluded that 372 indications in the Unit 1 OTSGs and 61 indications in the Unit 3OTSGs exceeded the repair criteria in the Oconee Technical Specifications. Tube-end cracking has also been reported forthe steam generators at Prairie Island Nuclear Generating Plant.

Steam generator tubes are secured into tubesheets using both mechanical expansion and welding. In general, each tubeprotrudes slightly beyond the tubesheet. A seal weld provides a leak-tight configuration. The mechanical expansion andseal welding processes introduce high residual stresses in the ends of each tube. In addition, the high temperaturesassociated with welding alter the metallurgical characteristics of the tube material in the heat-affected zone. Although thetubes in the ANO-1 OTSGs were heat treated prior to operation (which should have improved the material's resistance tostress-corrosion cracking), the licensee's preliminary conclusion is that the rolling and welding processes contributed to thedevelopment of PWSCC in the tubes at the seal weld. The licensee is currently conducting a root-cause evaluation todetermine any future corrective measures necessary to prevent recurrence or to mitigate the consequences of this mode ofdegradation.

**Discussion**

One of the safety functions of steam generators is to maintain a barrier against fission product release. Tube-to-tubesheetseal welds in steam generators minimize the potential for primary-to-secondary leakage under normal-operating andpostulated-accident conditions. The development of PWSCC tube cracking in the vicinity of these welds may degrade theleak-tight barrier between the primary and secondary coolant systems. Although the primary-to-secondary leakage duringnormal operation from a single tube-end flaw is likely to be small relative to the limit specified in plant technicalspecifications, the combined leakage from tube-end flaws could become significant if a large number of tubes havethrough-wall cracks.

PWR technical specifications require tubes containing defects with depths that exceed the repair criteria (typically 40percent of nominal wall thickness) to be removed from service. The repair criteria often require licensees to consider theentire tube in determining whether tubes are degraded or defective. Defects located in any area of a tube are evaluatedagainst the repair criteria unless the area is specifically excluded from the surveillance requirements. Upon confirming thepresence of tube-end flaws at ANO-1, the licensee utilized an industry-approved repair method to install an additional hardroll in the upper tubesheet, thus redefining the reactor coolant pressure boundary. The tubes in most recirculating steamgenerator designs are fully expanded within the tubesheet and do not offer the option of this type of repair. While fully-expanded designs inherently provide considerable resistance to bypass leakage, licensee technical specifications may stillrequire licensees to consider flaws near tube-end seal welds in dispositioning tubes as acceptable for continued service.

Steam generator tube-end cracking may not be limited to OTSGs. All PWR steam generators contain similar welds thatmay be equally susceptible to the development of PWSCC in the heat-affected zone of the tube. Due to the location offlaws associated with tube seal welds, the use of traditional inspection techniques and analysis procedures may not beentirely effective in detecting and characterizing this mode of degradation. The licensee for the Oconee units found itnecessary to perform supplemental eddy current inspections on mockup test samples to adequately address tube-endcracking.

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.

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