Flaw Assessment Procedure for High-Temperature Reactor Components

[+] Author and Article Information
R. A. Ainsworth

Nuclear Electric plc, Berkeley Nuclear Laboratories, Berkeley, Gloucestershire, UK

M. B. Ruggles

Oak Ridge National Laboratory, Oak Ridge, TN 37831-8051

Y. Takahashi

Central Research Institute of Electric Power Industry, Komae Research Laboratory, Tokyo, Japan

J. Pressure Vessel Technol 114(2), 166-170 (May 01, 1992) (5 pages) doi:10.1115/1.2929024 History: Received August 06, 1991; Revised November 05, 1991; Online June 17, 2008


An interim high-temperature flaw assessment procedure is described. This is a result of a collaborative effort between Electric Power Research Institute in the US, Central Research Institute of Electric Power Industry in Japan, and Nuclear Electric plc in the UK. The procedure addresses pre-existing defects subject to creep-fatigue loading conditions. Laws employed to calculate the crack growth per cycle are defined in terms of fracture mechanics parameters and constants related to the component material. The crack growth laws may be integrated to calculate the remaining life of a component or to predict the amount of crack extension in a given period. Fatigue and creep crack growth per cycle are calculated separately, and the total crack extension is taken as the simple sum of the two contributions. An interaction between the two propagation modes is accounted for in the material properties in the separate calculations. In producing the procedure, limitations of the approach have been identified. Some of these limitations are to be addressed in an extension of the current collaborative program.

Copyright © 1992 by The American Society of Mechanical Engineers
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