Review Texture development and anisotropic deformation of zircaloys K. Linga Murty * , Indrajit Charit North Carolina State University, Raleigh, NC 27695-7909, USA Received 15 June 2005; received in revised form 4 September 2005; accepted 4 September 2005 Abstract This paper is a review of the texture development in zirconium alloys (in the form of thick walled tube reduced extrusion or TREX, thin-walled tubing and sheets) of importance to light and heavy water nuclear reactor technology along with the resultant anisotropic mechanical properties. Quantitative characterization of texture and mechanical anisotropy are emphasized leading to procedures useful to fabricators in optimizing textures for good formability as well as for acceptable in-service performance. A brief history of the development of zirconium alloys is presented followed by texture development and characterization. Mechanical anisotropy is discussed in terms of transverse contractile strain ratios from which the formability (B parameter) is derived. Results on the effect of annealing temperature as well as test temperature on anisotropy parameters are presented. The review concludes with a brief summary of texture effects on creep, stress corrosion cracking and hydride formation. Recent advances in fuel cladding bring out the challenges in characterizing the texture and anisotropy due to Nb additions and microstructural gradients in the new Zircaloyse ,1 such as Zirloe ,2 , Duplexe ,3 and Triclad e ,4 . q 2005 Elsevier Ltd. All rights reserved. Keywords: Zircaloys; Texture; Mechanical anisotropy; Formability; Plasticity; Twinning; CODF 1. Introduction Zirconium alloys are extensively used in various types of fission reactors both light and heavy water types for different applications, examples being thin-walled tubing to clad radioactive fuel, grids, channels in BWRs as well as pressure and calandria tubes in PHWRs. Fig. 1 is an overview of reactor internals depicting the utility of these alloys. The development of zirconium alloys is essentially due to the nuclear industry, where zirconium alloys have been regarded as the proven structural material. Initial studies performed on zirconium suggested, however, that it was not a proper choice for nuclear applications due to its absorption of thermal neutrons essential for power reactor efficiency. That was until researchers at Oak Ridge discovered that naturally occurring zirconium contained approximately 2.5% hafnium which was responsible for the high thermal neutron cross-section. In the 1950s, the decision to use zirconium for nuclear applications was made by Admiral Rickover in the Naval Nuclear Propulsion program for use in their water cooled reactor for submarine Nautilus (Rickover, 1975). Even though zirconium was to Progress in Nuclear Energy 48 (2006) 325–359 www.elsevier.com/locate/pnucene 0149-1970/$ - see front matter q 2005 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2005.09.011 * Corresponding author. Tel.: C1 919 515 3657; fax: C1 919 515 5115. E-mail address: murty@ncsu.edu (K. Linga Murty). 1 Zircaloy is a trade name of Westinghouse company. 2 Zirloe is a proprietary Westinghouse material. 3 Duplexe is a proprietary Siemens material. 4 Triclad is a proprietary General Electric material.