Zirconium alloys
Zirconium alloys are solid solutions of zirconium or other metals, a common subgroup having the trade mark Zircaloy. Zirconium has very low absorption cross-section of thermal neutrons, high hardness, ductility and corrosion resistance. One of the main uses of zirconium alloys is in nuclear technology, as cladding of fuel rods in nuclear reactors, especially water reactors. A typical composition of nuclear-grade zirconium alloys is more than 95 weight percent zirconium and less than 2% of tin, niobium, iron, chromium, nickel and other metals, which are added to improve mechanical properties and corrosion resistance.
The water cooling of reactor zirconium alloys elevates requirement for their resistance to oxidation-related nodular corrosion. Furthermore, oxidative reaction of zirconium with water releases hydrogen gas, which partly diffuses into the alloy and forms zirconium hydrides. The hydrides are less dense and are weaker mechanically than the alloy; their formation results in blistering and cracking of the cladding – a phenomenon known as hydrogen embrittlement.
Production and properties
Commercial non-nuclear grade zirconium typically contains 1–5% of hafnium, whose neutron absorption cross-section is 600 times that of zirconium. Hafnium must therefore be almost entirely removed for reactor applications.Nuclear-grade zirconium alloys contain more than 95% Zr, and therefore most of their properties are similar to those of pure zirconium. The absorption cross section for thermal neutrons is 0.18 barn for zirconium, which is much lower than that for such common metals as iron and nickel. The composition and the main applications of common reactor-grade alloys are summarized below. These alloys contain less than 0.3% of iron and chromium and 0.1–0.14% oxygen.
| Alloy | Sn, % | Nb, % | Vendor | Component | Reactor type |
| Zircaloy 2 | 1.2–1.7 | – | All vendors | Cladding, structural components | BWR, CANDU |
| Zircaloy 4 | 1.2–1.7 | – | All vendors | Cladding, structural components | BWR, PWR, CANDU |
| ZIRLO | 0.7–1 | 1 | Westinghouse | Cladding | BWR, PWR |
| Zr Sponge | – | – | Japan and Russia | Cladding | BWR |
| ZrSn | 0.25 | – | Westinghouse | Cladding | BWR |
| Zr2.5Nb | – | 2.4–2.8 | Fabrica de Aleaciones Especiales | Pressure tube | CANDU |
| E110 | – | 0.9–1.1 | Russia | Cladding | VVER |
| E125 | – | 2.5 | Russia | Pressure tube | RBMK |
| E635 | 0.8–1.3 | 0.8–1 | Russia | Structural components | VVER |
| M5 | – | 0.8–1.2 | Areva | Cladding, structural components | PWR |
*ZIRLO stands for zirconium low oxidation.
Microstructure
At temperatures below 1100 K, zirconium alloys belong to the hexagonal crystal family. Its microstructure, revealed by chemical attack, shows needle-like grains typical of a Widmanstätten pattern. Upon annealing below the phase transition temperature the grains are equiaxed with sizes varying from 3 to 5 μm.Development
Zircaloy 1 was developed after zirconium was selected by Admiral H.G. Rickover as the structural material for high flux zone reactor components and cladding for fuel pellet tube bundles in prototype submarine reactors in the late 1940s. The choice was owing to a combination of strength, low neutron cross section and corrosion resistance. Zircaloy-2 was inadvertently developed, by melting Zircaloy-1 in a crucible previously used for stainless steel. Newer alloys are Ni-free, including Zircaloy-4, ZIRLO and M5.Oxidation of zirconium alloy
Zirconium alloys readily react with oxygen, forming a nanometer-thin passivation layer. The corrosion resistance of the alloys may degrade significantly when some impurities are present. Corrosion resistance of zirconium alloys is enhanced by intentional development of thicker passivation layer of black lustrous zirconium oxide. Nitride coatings might also be used.Whereas there is no consensus on whether zirconium and zirconium alloy have the same oxidation rate, Zircaloys 2 and 4 do behave very similarly in this respect. Oxidation occurs at the same rate in air or in water and proceeds in ambient condition or in high vacuum. A sub-micrometer thin layer of zirconium dioxide is rapidly formed in the surface and stops the further diffusion of oxygen to the bulk and the subsequent oxidation. The dependence of oxidation rate R on temperature and pressure can be expressed as
The oxidation rate R is here expressed in gram/; P is the pressure in atmosphere, that is the factor P1/6 = 1 at ambient pressure; the activation energy is 1.47 eV; kB is the Boltzmann constant and T is the absolute temperature in kelvins.
Thus the oxidation rate R is 10−20 g per 1 m2 area per second at 0 °C, 6 g m−2 s−1 at 300 °C, 5.4 mg m−2 s−1 at 700 °C and 300 mg m−2 s−1 at 1000 °C. Whereas there is no clear threshold of oxidation, it becomes noticeable at macroscopic scales at temperatures of several hundred °C.
Oxidation of zirconium by steam
One disadvantage of metallic zirconium is in the case of a loss-of-coolant accident in a nuclear reactor. Zirconium cladding rapidly reacts with water steam above.Oxidation of zirconium by water is accompanied by release of hydrogen gas. This oxidation is accelerated at high temperatures, e.g. inside a reactor core if the fuel assemblies are no longer completely covered by liquid water and insufficiently cooled. Metallic zirconium is then oxidized by the protons of water to form hydrogen gas according to the following redox reaction:
Zirconium cladding in the presence of D2O deuterium oxide frequently used as the moderator and coolant in next gen pressurized heavy water reactors that CANDU designed nuclear reactors use would express the same oxidation on exposure to deuterium oxide steam as follows:
This exothermic reaction, although only occurring at high temperature, is similar to that of alkali metals with water. It also closely resembles the anaerobic oxidation of iron by water.
This reaction was responsible for a small hydrogen explosion accident first observed inside the reactor building of Three Mile Island Nuclear Generating Station in 1979 that did not damage the containment building. This same reaction occurred in boiling water reactors 1, 2 and 3 of the Fukushima Daiichi Nuclear Power Plant after reactor cooling was interrupted by related earthquake and tsunami events during the disaster of March 11, 2011, leading to the Fukushima Daiichi nuclear accident. Hydrogen gas was vented into the reactor maintenance halls and the resulting explosive mixture of hydrogen with air oxygen detonated. The explosions severely damaged external buildings and at least one containment building. The reaction also occurred during the Chernobyl Accident, when the steam from the reactor began to escape. Many water cooled reactor containment buildings have catalyst-based passive autocatalytic recombiner units installed to rapidly convert hydrogen and oxygen into water at room temperature before the explosive limit is reached.
Formation of hydrides and hydrogen embrittlement
In the above oxidation scenario, 5–20% of the released hydrogen diffuses into the zirconium alloy cladding forming zirconium hydrides. The hydrogen production process also mechanically weakens the rods cladding because the hydrides have lower ductility and density than zirconium or its alloys, and thus blisters and cracks form upon hydrogen accumulation. This process is also known as hydrogen embrittlement. It has been reported that the concentration of hydrogen within hydrides is also dependent on the nucleation site of the precipitates.In case of loss-of-coolant accident in a damaged nuclear reactor, hydrogen embrittlement accelerates the degradation of the zirconium alloy cladding of the fuel rods exposed to high temperature steam.
Deformation
Zirconium alloys are used in the nuclear industry as fuel rod cladding due to zirconium's high strength and low neutron absorption cross-section. It can be subject to high strain rate loading conditions during forming and in the case of a reactor accident. In this context, the relationship between strain rate-dependent mechanical properties, crystallographic texture and deformation modes, such as slip and deformation twinning.Slip
has a hexagonal close-packed crystal structure at room temperature, where 〈?〉prismatic slip has the lowest critical resolved shear stress. 〈?〉 slip is orthogonal to the unit cell 〈?〉 axis and, therefore, cannot accommodate deformation along〈?〉. To make up the five independent slip modes and allow arbitrary deformation in a polycrystal, secondary deformation systems such as twinning along pyramidal planes and 〈? + ?〉slip on either 1st order or 2nd order pyramidal planes play an important role in Zr polycrystal deformation. Therefore, the relative activity of deformation slip and twinning modes as a function of texture and strain rate is critical in understanding deformation behaviour. Anisotropic deformation during processing affects the texture of the final Zr part; understanding the relative predominance of deformation twinning and slip is important for texture control in processing and predicting likely failure modes in-service.The known deformation systems in Zr are shown in Figure 1. The preferred room temperature slip system with the lowest critical resolved shear stress in dilute Zr alloys is 〈?〉 prismatic slip. The CRSS of 〈?〉prismatic slip increases with interstitial content, notably oxygen, carbon and nitrogen, and decreases with increasing temperature. 〈?〉basal slip in high purity single crystal Zr deformed at a low strain rate of 10−4 s−1 was only seen at temperatures above 550 °C. At room temperature, basal slip is seen to occur in small amounts as a secondary slip system to 〈?〉 prismatic slip, and is promoted during high strain rate loading. In-room temperature deformation studies of Zr, 〈?〉 basal slip is sometimes ignored and has been shown not to affect macroscopic stress-strain response at room temperature. However, single crystal room temperature microcantilever tests in commercial purity Zr show that 〈?〉 basal slip has only 1.3 times higher CRSS than 〈?〉 prismatic slip, which would imply significant activation in polycrystal deformation given a favourable stress state. 1st order 〈? + ?〉 pyramidal slip has a 3.5 times higher CRSS than 〈?〉 prismatic slip. Slip on 2nd-order pyramidal planes are rarely seen in Zr alloys, but 〈? + ?〉 1st-order pyramidal slip is commonly observed. Jensen and Backofen observed localised shear bands with 〈? + ?〉 dislocations on planes during 〈?〉 axis loading, which led to ductile fracture at room temperature, but this is not the slip plane as 〈? + ?〉 vectors do not lie in planes.