2.07.4.1 Alloying Elements and Phase Diagrams

Like any metal, pure Zr exhibits rather poor engi­neering properties. To improve the properties of a given metal, the metallurgical engineering proce­dures are always the same: It consists in finding addi­tions, any species of the periodic table could be considered, with significant solubility, or heat treat­ments producing new phases that could improve the properties. The relative solubility ofthe various alloy­ing elements in the a — and р-phases is therefore one basis for the choice of additions, as well as for devel­oping the heat treatments, for microstructure control.

For the nuclear applications, neutron physics requirements restrict the possibilities, by rejection
of the isotopes having high interaction cross-sections, or isotopes that would transmute to isotopes of high capture cross-section or having high irradiation impact (Co). Elements such as Hf, Cd, W, and Co have therefore not been considered for alloy devel­opments. With low nuclear impact, O, Sn, and Nb have been selected (Al and Si having also low nuclear impacts were not retained because of degradation in corrosion resistance), while other TMs (Fe, Cr, Ni, etc.) can be accepted up to limited concentrations (below 0.5% total).

The additions have to improve the engineering properties. The main properties to be improved are the corrosion behavior in hot water and the mechan­ical strength (yield stress, ductility, and creep). As described below, Sn and Nb are added for corrosion resistance, and elements forming secondary phases (Nb and Fe, Cr, and Ni) or solid solutions are also used for increasing the mechanical properties.

Last, the microstructure obtained after the ther­momechanical processing should not change without control under irradiation. Therefore, hardening obtained by precipitation or strain hardening can be considered only if the irradiation-induced evolution of the initial microstructure will be compensated by the development of irradiation-induced microstruc­tural defects. In this respect, the evolution of preci­pitates in Zircaloys is of high importance for corrosion behavior and geometrical integrity. These points are discussed in Chapter 5.03, Corrosion of Zirconium Alloys and Chapter 4.01, Radiation Effects in Zirconium Alloys.

Most of the binary phase diagrams with Zr are already known and many ternary or higher-level
diagrams of industrial interest are now known.17 The need for a better control of the processing of the current alloys and the aim of finding new alloys and structures without too much experimental work have been a driving force for the modern trend in numeri­cal simulation for material science. It is now also possible to extrapolate the binary data to multicom­ponent systems. In that respect, a thermodynamic database for Zr alloys, called ZIRCOBASE, has been developed under the Calphad methodology.18 This database contains 15 elements and is frequently updated. The most complex ternary or quaternary phase diagrams available are optimized or computed using this database, and, in the case of missing basic thermodynamic data, with the contribution of ‘ab — initio’ computations.19 The phase diagrams presented in this review were obtained according to this procedure.

Oxygen is highly soluble in the a-phase, and stabi­lizes at high temperature (Figure 4). Oxygen has to be considered as an alloying element. This use of oxygen for strengthening is rare in metallurgy, com­pared to the use of nitrogen. However, the use of nitrogen for strengthening would severely deteriorate the corrosion resistance, and nitrogen is removed as much as possible. The purpose of oxygen additions is to increase the yield strength by solution strength­ening, without degradation of the corrosion resis­tance. The O content is not specified in the ASTM standards, but usually it is added to concentrations in the range of 600-1200 ppm, and this has to be agreed between producer and consumers. High O concen­trations (O > 2000 ppm) reduce the ductility of the alloys; therefore, O additions above 1500 ppm are not recommended. In addition, O atoms interact with the dislocations at moderate temperatures,
leading to age-strengthening phenomena in temper­ature ranges depending on strain rate.20 The oxygen in solid solution in a-zirconium is an interstitial in the octahedral sites. In the Zr-O system, the only available stable oxide is ZrO2. A monoclinic phase is stable at temperatures up to about 1200 °C, above which it transforms to a tetragonal structure. The impact on corrosion of the different phases of ZrO2, according to temperature and pressure is dis­cussed in Chapter 5.03, Corrosion of Zirconium Alloys.

Tin tends to extend the a-domain, and has a maximal solubility in the hcp Zr of 9 wt% at 940 °C (Figure 5). It was originally added at concentrations of 1.2-1.7% to increase the corrosion resistance, especially by mitigating the deleterious effects of nitrogen. The amount of Sn needed to compensate the effect of 300 ppm of N is about 1% of Sn. How­ever, in N-free Zr, Sn has been observed to deterio­rate the corrosion resistance. Therefore, the modern trend is to reduce it, but only slightly, in order to maintain good creep properties.21

Iron, chromium, and nickel, at their usual concentra­tions, are fully soluble in the р-phase (Figure 6). However, in the a-phase, their solubilities are very low: in the region of 120 ppm for Fe and 200 ppm for Cr at the maximum solubility temperature.22 In the pure binary systems, various phases are obtained: ZrFe2 and ZrCr2 are Laves phases with cubic or hexagonal structure, while Zr2Ni is a Zintl phase with a body-centered tetragonal C16 structure. These precipitates are called the Second Phase Par­ticles (SPPs).

In the Zircaloys, the Fe substitutes for the corresponding TM and the intermetallic compounds found in Zircaloy are Zr2(Ni, Fe) and Zr(Cr, Fe)2.

The formation of these precipitates, and more complex ones in industrial alloys, is analyzed in detail for the control of the corrosion behavior of the Zircaloys. Indeed, a strong correlation has been observed between precipitate size distributions and corrosion kinetics, the behavior being opposite for BWRs and PWRs. A better uniform corrosion resis­tance is obtained for Zircaloys used in PWRs if they contain large precipitates, while better resistance to the localized forms of corrosion is seen in BWRs in materials that have finely distributed small precipi­tates 23,24 With an increase in the particle diameters from 0.05 to 0.1 pm or higher, the in-pile corrosion of Zircaloy cladding diminishes appreciably. However, nodular corrosion may occur in BWR cladding with a further increase in the particle diameters above about 0.15 pm25 (Figure 7).

Due to the low solubility of the transition metals (Fe, Cr and Ni) in the Zr matrix, coarsening of the precipitates, after the last p-quench, occurs at very
low rates, during the intermediate annealing heat treatments, following each step of the rolling process. Therefore, the precipitate growth integrates the ther­mal activation times of each recovery, and their tem­peratures and durations can be used to control the size of SPPs. This integrated coarsening activation time is referred as the ‘A ’ or ‘SA ’ parameter.

The A-parameter calculates the integral of the activation processes for the different anneal durations and temperatures. The annealing parameter is defined as A = S, (t, exp(—Q/RT)), where t, is the time (in hours) of the ith annealing step, at tempera­ture T (in K); Q/T is the activation temperature of the process involved. The activation energy for the process should have been taken as the one controlling the coarsening, that is, the diffusion. However, as the early studies were undertaken with the aim of improving the corrosion resistance, an unfortunate practice has been induced to take 40 000 K as the value of Q/T. A more correct value would be

о

Figure 7 Effect of precipitate size on the corrosion kinetics of Zircaloys. Reproduced from Garzarolli, F.; Stehle, H. Behavior of structural materials for fuel and control elements in light water cooled power reactors, IAEA STI/PUB/721; International Atomic Energy Agency: Vienna, 1987; p 387.

32 000 K, which fits very well with the recrystalliza­tion kinetics. The influence of the A-parameter on the corrosion of Zircaloy is discussed in more detail in Chapter 5.03, Corrosion of Zirconium Alloys. High resistance to uniform corrosion in PWR is obtained for the A-parameter close to (1.5—6.0) x 10~19h. In BWR, the A-parameter value for the Zircaloy-2 cladding in BWR has to be in the range (0.5-1.5) x 10~18h (Figure 7).25 This corresponds to precipitates larger than 0.18 pm. The SA approach has been developed for the Zircaloys and is clearly not applicable for other alloys, such as the Zr-Nb alloys.

Niobium (columbium) is a p-stabilizer that can extend the bcc domain to a complete solid solution between pure Zr and pure Nb at high temperatures (Figure 8). A monotectoid transformation occurs at about 620 °C and around 18.5 at.% Nb. The solubil­ity of Nb in the a-phase is maximal at the monotectic temperature, and reaches 0.65%.

Water p-quenching of small pieces leads to the precipitation of a’ martensite supersaturated in Nb. Tempering at intermediate temperature results in p-Nb precipitation within the a’ needles and subsequent transformation of a’ into a. When quenching is performed from an a + p region, a uniform distribution of a — and p-grains is obtained, and the Nb-rich p-phase does not trans­form. By aging at temperatures in the range of 500 °C, the metastable Nb-rich p-phase can be decomposed into an hcp ю-phase. This gives a sharp increase in mechanical strength because of the fine micro­structure obtained by the p-ю transformation.26 In the usual form of the Zr-2.5% Nb, the cold work condition after a + p extrusion and air-cooling, the microstructure consists of Zr grains with layers of p-Nb rich phase (close to eutectoid composition). Owing to the affinity of Fe for the p-phase, most of this element is found in the minor p-grains. These p-grains are metastable and decompose, upon aging, to a mixture of a-Zr and pure p-Nb. The Nb dis­solved in the a-hcp Zr phase is itself metastable and the irradiation-induced precipitation of the supersat­urated Nb solid solution is believed to be the origin of the improvement in corrosion resistance under irradiation of these alloys.27

In the case of Zr-1% Nb used for VVER and RBMK, or M5® in PWRs, the concentration of Nb in the Zr matrix after processing corresponds to the maximum solubility near the monotectoid tempera­ture, which is higher than the solubility at the service temperature. Owing to the slow diffusion of Nb, the equilibrium microstructure cannot be obtained ther­mally. However, the irradiation-enhanced diffusion allows precipitation of fine p-Nb needles in the grains after a few years in reactors.28

Sulfur has recently been observed to be extremely efficient in improving the creep resistance, even at concentration as low as 30-50 ppm. This chemical species, formerly not considered as important, is now deliberately added during processing to reduce the scatter in behavior and to improve the high tempera­ture mechanical properties.29 The efficiency of such low concentrations on the creep properties has been explained by the segregation of the S atoms in the core of the dislocations, changing their core configurations. It does not affect the corrosion properties.30

In the case of complex alloys, other thermody­namical interactions are expected and intermetallic compounds including three or four chemical ele­ments are observed. The chemistry and the crystal­lography of these phases may be rather complex.

200 nm

Two examples will be given of the complex structure

and behavior of these intermetallics.

• For the Zr-Cr and Zr-Ni binary alloys, the stable forms of the second phase are Zr2Ni or ZrCr2. These phases are effectively the ones observed in the Zircaloys, with Fe substituting for the corresponding TM. Therefore, the general formu­lae ofthe intermetallic compounds in Zircaloys are Zr2(Ni, Fe) and Zr(Cr, Fe)2. The crystal structure of the Zr(Cr, Fe)2 precipitates is either fcc (C15) or hcp (C14), depending on composition and heat treatment. Both structures are Laves phases, with characteristic stacking faults as seen in Figure 9. The equilibrium crystallographic structure is dependent upon the Fe/Cr ratio, cubic below 0.1 and above 0.9, and hexagonal in the middle. Under irradiation, these precipitates transform to amor­phous state and release their Fe in the matrix, with strong impact on corrosion behavior under irradiation.31

• In the Zr-Nb-Fe ternary, other intermetallic com­pounds can be observed (Figure 10): the hexago­nal Zr(Nb, Fe)2 phase and the cubic (Zr, Nb)4Fe2.32 Although of apparent similar composition, the two phases are indeed different: Nb can substitute Fe in the hexagonal phase, while it will substitute Zr in the cubic phase. In these alloys, due to the slow diffusion of Nb, metastable phases are often present and the equilibrium microstructure after industrial heat treatments may be far from the stable one. Therefore, the final microstructure is strongly dependent on the exact thermomechanical history.

In addition, the low solubility of these elements at operating temperatures drastically reduces the dif­fusion kinetics and requires more than a year to reach equilibrium at 450 ° C, in the absence of irradiation.3

Other minor constituents are often found in the form of precipitates. Among them are the carbide fcc — ZrC and silicides or phosphides of various stoichio­metries (Zr3Si, ZrSi2, ZrP, and Zr3P) that act as nucleation sites for the p! a-phase transformation during quenching and, therefore contribute to con­trol the a-platelets thickness and density.