Intergranular stress-assisted corrosion cracking of austenitic alloys in water-cooled nuclear reactors-Reference-Cited by-同舟云学术

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Intergranular stress-assisted corrosion cracking of austenitic alloys in water-cooled nuclear reactors

Published:1975-07 Issue:1 Volume:57 Page:11-33
ISSN:0022-3115
Container-title:Journal of Nuclear Materials
language:en
Short-container-title:Journal of Nuclear Materials

Author:

Cheng Craig F.

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering,General Materials Science,Nuclear and High Energy Physics

Reference56 articles.

1. 304 Stainless Steel Steam Generator Tube Experience at N-Reactor;Kratzer;DUN-SA-178,1971

2. Oyster Creek Nuclear Power Plant, Unit No. 1,1968

3. Performance of Tarapur Atomic Power Station;Doyal,1972

4. Versagen von Steuerstahlverlängerungslagern im Atomkraftwerk Kahl;Gutmann;Energ. Tagesfragen,1970

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