The fuel rod components are predominantly made of zirconium alloys. Zirconium is a ductile metal with mechanical properties similar to those of titanium and austenitic stainless steel. Its alloys combine very low neutron absorption with good corrosion resistance in power reactor conditions. Stainless steel was tried for the cladding tube at the beginning of the nuclear era, but was abandoned because of stress corrosion cracking problems and high neutron absorption compared to zirconium.

Zirconium is produced from zirconium ore or sand (zircon, ZrSiO2 ) in two principal ways, namely by chemical reduction of the oxide (sponge zirconium) and by electrolysis of zirconium halides in a salt melt (Neikov et al., 2009). Although the two methods produce rather pure zirconium, the respective products differ in their content of impurities. It has been found that this can lead to differences in the behaviour of the alloys produced from them. Yegorova et al. (2005) studied E110 (the Russian cladding tube alloy traditionally manufactured from electrolytic process zirconium) in loss-of-coolant temperature conditions. They found that material produced from the alternative sponge zirconium showed a significantly reduced oxidation rate and tolerated more oxidation (equivalent cladding reacted, ECR) before the material completely lost its ductility.

Reactor grade zirconium (ASTM B349, 2009) suitable for use in nuclear applications is characterised by its low neutron absorption cross section achieved by removal of hafnium, which occurs in the same mineral in quantities of 1.5 to 4% (the hafnium is used for control rods). The impurities remaining in nuclear grade zirconium after the hafnium extraction process are given in Table 9.3 (Moulin et al., 1984).

For in-core applications, zirconium cannot be used as a pure metal, but has to be alloyed in order to obtain better corrosion resistance. The nuclear era started with alloys called Zircaloy-2 (Zry-2) for BWRs and Zircaloy-4 (Zry-4) for PWRs.

Table 9.2 Typical dimensions of cladding and fuel pellets Feature PWR BWR 14 x 14 15 x 15 16 x 16 17 x 17 18 x 18 9×9 10 x 10 Cladding outer diameter (mm) 10.16-11.18 10.75 9.14-10.75 9.50 9.50 11.00-11.20 9.84-10.28 Wall thickness (mm) 0.570-0.725 0.620-0.725 0.570-0.725 0.570 0.64 0.70-0.71 0.605-0.620 Pellet — cladding diam. gap (mm) 0.170-0.210 0.170-0.190 0.160-0.190 0.156-0.170 0.170 0.180-0.200 0.150-0.170 Pellet height (mm) 10.0-12.6 10.0-12.6 9.8-11.0 9.8-13.5 9.0-9.8 10.0 10.0-10.5

The composition is Zr/1.5% Sn/0.15% Fe/0.1% Cr/0.05% Ni for Zry-2 and Zr/1.5% Sn/0.2% Fe/0.1% Cr for Zry-4 (Schemel, 1977). Later, increasing demands on the fuel cladding due to extended burn-up and in-core residence time led to the development of improved alloys especially for use in PWRs. A variety of products are available from different vendors, e. g. ZIRLO (Westinghouse), M5 (AREVA) and MDA (Mitsubishi). They are characterised by containing 0.5-1.0% niobium, which is not present in the Zircaloys. The Russian E110, which is used in VVER and RBMK reactors, contained 1% Nb from the very beginning. Other vendors developed claddings with an outer layer of corrosion resistant material on a Zircaloy substrate as explained in the next section.