Topic
Zirconium alloy
About: Zirconium alloy is a research topic. Over the lifetime, 6548 publications have been published within this topic receiving 78954 citations. The topic is also known as: zircaloy.
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15 Sep 2012TL;DR: In this article, the analytical results of the amount of the radioactive nuclides in the Fukushima-Daiichi nuclear power plant on March 31, 2011 and the following period with the use of the ORIGEN2 code are described.
Abstract: This document describes the analytical results of the amount of the radioactive nuclides in the Fukushima-Daiichi nuclear power plant on March 31, 2011 and the following period with the use of the ORIGEN2 code. The results are given for the irradiated uranium pellet and the activated cladding tube of zirconium alloy in the core and the spent fuel storage pools of the respective reactors. The evaluated values are weight, radioactivity, heat generation, photon generation and neutron generation rate. A CD-ROM is attached as an appendix. (author)
156 citations
01 Feb 1979
TL;DR: In this paper, a handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory.
Abstract: This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.
153 citations
01 Jan 2012
TL;DR: In this article, the twin processes of uniform oxidation and hydrogen embrittlement are described comprehensively in terms of their underlying mechanisms, and the emphasis of this chapter is on uniform oxidisation.
Abstract: When zirconium alloys are used in water-cooled reactors, they are subjected to waterside corrosion. The emphasis of this chapter is on uniform oxidation and hydrogen embrittlement. The twin processes of oxidation and hydriding are described comprehensively in terms of their underlying mechanisms.
153 citations
TL;DR: In this article, the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment were analyzed and the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made.
152 citations
TL;DR: In this article, the diffusion coefficient of hydrogen in zirconium, Zircaloy-2 and Zirca-4 was determined in the temperature range, 275 °C to 700 °C.
152 citations