alloys based on zirconium. Until the early 1950’s, zirconium alloys were little studied and found little use. The information available at that time on the alloys was often unreliable as a result of the use of insufficiently pure zirconium and inadequate methods for the preparation of alloys for investigation. All this was altered abruptly in the early 1950’s, when it became possible to obtain purified zirconium, with hafnium impurities removed, and it was discovered that zirconium has a low thermal neutron absorption cross section. This property indicated that zirconium, along with its other favorable properties, could be a very promising material for components of nuclear thermal power reactors. However, initial studies showed that it would be impossible to use unalloyed zirconium for this purpose, primarily because of its low corrosion resistance in hot water. This problem stimulated new intensive research into zirconium alloys, which led to the development of industrial alloys that have found extensive use in the nuclear power industry. Zirconium alloys are used for thermal reactor-core structural elements, such as channels, fuel assemblies, the spacer lattice, and the cladding of the fuel elements.
Zirconium alloys have found the greatest use in water and vapor cooled reactors. Zirconium alloys, in addition to a low thermal neutron absorption cross section, possess high and stable corrosion resistance in water and steam at elevated temperatures and pressures and in other corrosive media, good ductility, and satisfactory strength. The alloying elements of zirconium alloys must meet certain requirements: some of these elements must significantly diminish (repress) the deleterious effect of nitrogen on the corrosion resistance of zirconium (with a permissible nitrogen content in the alloys less than 0.01 percent), while the others must not noticeably increase the thermal neutron absorption cross section, must not reduce the radiation stability, and should increase the strength without markedly diminishing ductility (the alloys must be suitable for the preparation of very thin-walled tubes and sheets and have good weldability). Thus, the selection of alloying additives is limited to relatively few elements; the contents of these elements in zirconium alloys are low. Niobium, tin, iron, chromium, nickel, copper, and molybdenum are used as alloying elements; these metals are introduced in amounts ranging from fractions of a percent to 2–3 percent (total).
Only a few of the large number of zirconium alloys studied have found use. Outside the Soviet Union, the most widely used alloy is the American alloy Zircaloy-2, which contains 1.5 percent tin, 0.1 percent iron, 0.1 percent chromium, 0.05 percent nickel, and not more than 0.01 percent nitrogen. The alloy Zircaloy-4, which differs from Zircaloy-2 by a lower nickel content (0.007 percent), is also used. Zircaloy-2 was specially developed and was initially used for the cladding of the fuel elements of the first American nuclear submarine, Nautilus, and then found use in many nuclear power plants for fuel elements and channels operating in water and steam-and-water mixtures at 250°–300°C.
In the USSR, original zirconium alloys not containing tin have been developed and are now used. These alloys are called ZrlNb and Zr2.5Nb, with 1 and 2.5 percent niobium, respectively. The alloy ZrlNb was first used for fuel elements of the atomic icebreaker Lenin, while the alloy Zr2.5Nb was used for the reactor fuel assemblies of the Novovoronezhskii Atomic Power Plant. In the mid-1970’s, ZrlNb and Zr2.5Nb alloys were used for fuel-element cladding, fuel assemblies, and channels of the reactors of most nuclear power plants in the USSR and the socialist countries. Furthermore, the alloy Zr2.5Nb is used in a series of reactors in Canada. With respect to corrosion resistance, Zr2.5Nb is comparable to Zircaloy-type alloys but has a lower tendency toward hydrogen absorption, is not subject to reduced corrosion resistance upon irradiation, and has greater strength, especially greater resistance to creep.
Despite the high melting point of zirconium (1852°C), the known zirconium alloys do not have high heat resistance and are suitable for operations in steam-and-water media at temperatures not higher than 400°C. At higher temperatures, in addition to loss of strength, zirconium alloys undergo intensive oxidation with the dissolution of oxygen, leading to loss of ductility and to hydrogen absorption, which causes embrittlement because of the formation of hydrides. The strength and ductility of Zircaloy and zirconium-niobium type alloys are about equal upon brief testing (see Table 1) and, like the properties of other metallic materials, depend on the structural state resulting from thermal treatment and deformation.
Zirconium alloys are produced in vacuum arc furnaces with consumable electrodes and in electron-beam furnaces. The zirconium used is reactor-grade zirconium, which is zirconium purified to eliminate hafnium and other impurities with high thermal neutron absorption cross sections. Zirconium-alloy semifinished products are prepared using the same equipment as that used for preparing many nonferrous metals. Annealing is carried out in vacuum furnaces. While zirconium alloys have found extensive use in the nuclear power industry, they have found virtually no use in other areas of technology; in particular, they are inferior to
|Table 1. Mechanical properties of zirconium alloys|
|Alloy||Semifinished product (state)||At 20°C||At 300°C|
|Ultimate strengthsσ0||Relative elongationδ (percent)||Ultimate strengthσ0||Retative elongationρ (percent)|
the stronger, lighter, and less expensive titanium alloys as structural and corrosion-resistant materials.
REFERENCESMetallurgiia tsirkoniia. Moscow, 1959. (Translated from English.)
Trudy vtoroi Mezhdunarodnoi konferentsii po mirnomu ispol’zovaniiu atomnoi energii. Geneva, 1958.
Doklady sovetskikh uchenykh, vol. 3. Moscow, 1959. Page 486.
Rivkin, E. Iu., B. S. Rodchenkov, and V. I. Filatov. Prochnost’ splavov tsirkoniia. Moscow, 1974.
Douglas, D. Metallovedenie tsirkoniia. Moscow, 1975. (Translated from English. Contains references.)
A. A. KISELEV