Damage Stucture in Zirconium Alloys Irradiated at 573 to 923°K. [Neutron Fluence 1 X 1025 N. M−2].


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The microstructures of annealed zirconium, Zircaloy-2 and Zr-2.5 wt % Nb alloy and of Zr-2.5 Nb containing .cap alpha.' were studied after neutron irradiation to fluences approximately equal to 1 x 1025 n x m−2 in the temperature range 573 to 923°K. The principal form of damage was dislocation loops which increased in size and decreased in density with increasing temperature and which did not exist above 773°K. The Burgers vector of the loops was consistent with a/3 1120. Half or more of the loops were of vacancy type. No dislocation networks or voids were seen. It is argued that the bias of loops for self-interstitial atoms in .cap alpha.-zirconium is very weak, permitting competitive vacancy and interstitial loops, preventing growth of loops into gross dislocation structure, and depressing the vacancy super-saturation so that voids cannot arise.










Zirconium in the Nuclear Industry


Book Description

Annotation The 41 papers of this proceedings volume were first presented at the 13th symposium on Zirconium in the Nuclear Industry held in Annecy, France in June of 2001. Many of the papers are devoted to material related issues, corrosion and hydriding behavior, in-reactor studies, and the behavior and properties of Zr alloys used in storing spent fuel. Some papers report on studies of second phase particles, irradiation creep and growth, and material performance during loss of coolant and reactivity initiated accidents. Annotation copyrighted by Book News, Inc., Portland, OR.




Zirconium in the Nuclear Industry


Book Description

The proceedings of the Ninth International Symposium on [title], held in Kobe, Japan, November 1990, address current trends in the development, performance, and fabrication of zirconium alloys for nuclear power reactors. the bulk of the most recent work on zirconium alloy behavior has concerned corr




INIS Atomindex


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Electron Microscopy


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Comptes Rendus


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