The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking of a reactor pressure vessel (RPV) due to a pressurized thermal shock (PTS) event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.” The §50.61a rule, which is optional, requires licensees to analyze the results from periodic volumetric examinations required by the American Society of Mechanical Engineers (ASME) Code. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory has been working on a program to assess the ability of current inservice inspection ultrasonic testing (UT) techniques, as qualified through the ASME Code to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, “Spirit of Appendix VIII reactor vessel examination,” a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.

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