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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TL;DR: In this paper, it has been shown that the oxidation of crystal-bar zirconium is limited by electron transport through oxide and across the outer oxide surface between electronic flaws in the oxide and the sites of oxygen reduction.
25 citations
TL;DR: In this article, axial and tangential tensile tests were performed on as-received and machined fuel cladding tube samples of both E110 and E110G Russian zirconium alloys at room temperature to compare their ultimate tensile strengths and the different sample preparation methods.
25 citations
TL;DR: A review of the current understanding of the delayed hydride cracking behavior of zirconium alloy cladding tubes for fuel rods, focusing on the degradation mechanisms in high burnup fuel rods and transient loading scenarios, is presented in this paper.
Abstract: Zirconium alloy cladding tubes used in nuclear fuel rods are susceptible to delayed hydride cracking, which is a time dependent crack growth process resulting from the stress assisted diffusion of hydrogen to the crack tip, followed by the formation of radial hydrides and the subsequent fracture of the hydrides in the crack tip region. This article reviews the current understanding of the delayed hydride cracking behaviour of zirconium alloy cladding tubes for fuel rods, focusing on the degradation mechanisms in high burnup fuel rods and transient loading scenarios, which could potentially lead to substantial changes in the hydride microstructure and cladding failure by delayed hydride cracking following removal from the reactor and during storage and disposal of spent nuclear fuel rods in a waste repository. A brief summary of the general characteristics of delayed hydride cracking in zirconium alloy cladding is presented first. Relevant information on the cladding stresses under various usage conditions is then compiled and categorised into several characteristic stress transients that can be anticipated during reactor operation. Delayed hydride cracking in cladding tubes under stress transients is then examined under various temperatures, cooling rates, burnup levels and loading conditions.
25 citations
TL;DR: In this paper, the superconducting transition temperature and the electronic specific heat for the alloy system scandium-zirconium is investigated, and the results show that the alloys with the largest electronic specific heats are not superconducted down to 0.03 K. This behavior is very similar to the recently reported behavior near the filled end of the transition metals.
Abstract: An investigation of the superconducting transition temperature and the electronic specific heat for the alloy system scandium-zirconium is reported. The results show that the alloys with the largest electronic specific heats are not superconducting down to 0.03\ifmmode^\circ\else\textdegree\fi{}K. When compared with the other early transition metals (those with less than half-filled $d$ shells), the data for this system indicate a sharp decrease in the strength of the net attractive electron-electron interaction as one decreases the number of $d$ electrons (i.e. moves from zirconium to scandium). This behavior, combined with the apparently large exchange enhancement of the spin susceptibility in pure Sc, is very similar to the recently reported behavior near the filled end of the transition metals (i.e., near Pd). It is suggested that the proximity of a $d$-band edge may be causing the anomalous behavior in the superconductivity and the spin susceptibility for both regions of the transition metals.
25 citations
TL;DR: In this paper, the authors conducted autoclave corrosion experiments on a number of zirconium alloys in different heat treatment conditions, including Zircaloy-4, ZIRLO® (ZIRLO is a registered trademark of Westinghouse Electric Company LLC in the USA and may be registered in other countries throughout the world. Unauthorised use is strictly prohibited.
Abstract: Autoclave corrosion experiments were conducted on a number of zirconium alloys in different heat treatment conditions. The alloys tested in the present work were Zircaloy-4, ZIRLO® (ZIRLO is a registered trademark of Westinghouse Electric Company LLC in the USA and may be registered in other countries throughout the world. All rights reserved. Unauthorised use is strictly prohibited.) and two variants of ZIRLO with significantly lower Sn levels, referred to here as A-0·6Sn and A-0·0Sn. Typical corrosion kinetics with a change from pre- to post-initial transition was observed with ZIRLO and Zircaloy-4 displaying the shortest time to the initial transition after 120–140 days of autoclave exposure, followed by A-0·6Sn materials after 140–260 days. A-0·0Sn materials showed no sign of transition even after 360 days although one sample tested to 540 days had gone through transition. Material in the stress relieved condition generally experienced initial transition earlier than the same alloy in the recr...
25 citations