SYMPOSIA PAPER Published: 01 January 1981
STP33465S

The Use of Fatigue Crack Growth Technology in Fracture Control Plans for Nuclear Components

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The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code now suggests the use of fatigue crack growth analysis in the evaluation of indications found during in-service inspection of nuclear components. In this paper the role of crack growth analysis in the evaluation process is reviewed in some detail, and the background and philosophy of its implementation is discussed.

Each of the steps in the crack growth analysis process is discussed in order to point out the assumptions possible and their implications. A detailed consideration of crack shape change during growth is presented, as well as a statement of guidelines for choosing a reference crack growth law.

The ASME approach is compared with the approaches taken in other industries and, finally, a number of areas in need of further work are highlighted.

Author Information

Bamford, WH
Westinghouse Electric Corporation, Pittsburgh, Pa.
Jones, DP
Westinghouse Bettis Atomic Power Laboratory, West Mifflin, Pa.
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Details
Developed by Committee: E08
Pages: 281–299
DOI: 10.1520/STP33465S
ISBN-EB: 978-0-8031-4804-8
ISBN-13: 978-0-8031-0717-5