Japanese Activities Concerning Nuclear Codes and Standards—Part II

[+] Author and Article Information
Yasuhide Asada

 Thermal and Nuclear Power Engineering Society, Terayama Pacific Bldg., 1-23-11 Toranomon, Minato-ku, Tokyo, 105-0001 Japanasada@tenpes.or.jp

J. Pressure Vessel Technol 128(1), 64-70 (Oct 19, 2005) (7 pages) doi:10.1115/1.2138063 History: Received September 01, 2005; Revised October 19, 2005

This series of papers has been issued to give general views on recent Japanese activities related to nuclear codes and standards. Part II of the series describes component aging aspects and future trends. The component aging aspects include evaluation methods for vessels based on elastic-plastic fracture mechanics, environmental fatigue evaluation guidelines and inspection and evaluation guidelines for reactor internals. With respect to future trends, the development of the International Thermonuclear Experimental Reactor code and the System-Based Code are introduced.

Copyright © 2006 by American Society of Mechanical Engineers
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Figure 1

Ductile crack initiation behavior of a small test vessel

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Figure 2

Ductile crack growth behavior of a small test vessel

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Figure 3

Flow chart of the development of simplified EPFM

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Figure 4

Effect of the whole circumferential crack depth of the core shroud H7a weld joint (load-displacement relation)

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Figure 5

Relationship between crack depth of the core shroud H7a weld joint and operating time

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Figure 6

Margins in SBC: from accumulation to distribution





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