Such alloys are already well-known and comprise a. o. the two alloys Zirkaloy-2 and Zirkaloy-4, both of which have a large number of commercial application areas. For example, different components in nuclear plants often comprise these alloys. Ex- amples of such components include cladding tubes and other components, such as grids and boxes for fuel bundles intended to be used in nuclear reactors, e. g. light-water reactors.

Because the inventive zirconium based alloy is particularly suited for such applications, it will, with the intention of exem- plifying but not to delimit, be described with reference to its use as a component in a nuclear power plant, in which it is subjected to radiation and corrosive media. Thereby, the environment may also be such that it might cause hydration to components lo- cated therein.

At the manufacture of a zirconium based alloy of the type de- fined above, there is obtained a primary phase matrix in which secondary phase particles having a certain size distribution and of a certain area or volume fraction are distributed.

When such a zirconium based material is subjected to oxidation, a barrier layer which protects the material against a continued rapid oxide growth is initially formed. A thick barrier layer with a good protecting effect against corrosion and hydration for a long time is requested.

Zirconium based alloys having the composition corresponding to the specifications of Zirkaloy-2 and Zirkaloy-4 will only solve this problem to a certain extent. A zirconium based alloy having a thicker and better protecting barrier layer than is the case of conventional Zirkaloy-2 and Zirkaloy-4 is therefore requested.

SUMMARY OF THE INVENTION One object of the present invention is to provide a zirconium based alloy which, when being subjected to oxidation, forms a barrier layer that is thick and maintains its protecting effect against corrosion and hydration for long periods of time.

Thereby, said alloy shall have a good resistance against corro- sion and hydration, which are problems that often occur in zir- conium based alloys in corrosive and hydrative environments, e. g. in nuclear power plants. A better corrosion and hydration resistance than the one of conventional Zirkaloy-2 and Zirkaloy- 4 alloys is requested.

This object is obtained by means of a zirconium based alloy of the type initially defined, which is characterised in that the Cr content is in the range of 0.16-0.6 weight-%, and the Fe content is in the range of 0-0.6 weight-%. By conventional Zirkaloy ma- terials, the Cr content is <0.15 weight-%. The invention is based upon the insight that it is the size distribution and the volume fraction of the secondary phase particles present in the zirco- nium based alloy which at least partly determines how thick and protective barrier layers are formed on the surface of the zirco- nium based alloy when the latter is located in corrosive media

and subjected to oxidation. The inventors have noted that, by an optimisation of the size distribution and volume fraction of the secondary phase particles, it is possible to substantially in- crease the thickness of the barrier layer and to improve its pro- tective capacity. Using a higher chrome content than what is specified for Zirkaloy-2 and Zirkaloy-4 is one way of increasing the thickness and improving the protective capacity. It has also been noted that the size distribution and the volume fraction of the secondary phase particles depend on the Fe content in the zirconium based alloy. Both the Fe content and the Cr content are thus important factors when the size distribution and volume fraction of the secondary phase particles is to be optimised in order to obtain a thick and able barrier layer.

According to one embodiment of the inventive alloy, the Fe content is in the range of 0.22-0.6 weight-%. Thereby, the total content of Fe and Cr is >0.37 weight-% and <1.2 weight-%.

Higher contents of Fe + Cr than 1.2 weight-% should be avoided with regard to the requested phase transformation temperatures of the material. Too low phase transformation temperatures should be avoided for security reasons as alloys of this type are used in nuclear power plants and one do not wish to obtain phase transformations in the material upon a possible, large temperature increase in the plant. A higher content will also re- sult in manufacturing problems, e. g. crack formation upon cold rolling. The total, increased content of Fe and Cr in the inventive alloy in relation to prior Zirkaloy materials will however result in a size distribution and a volume fraction of the secondary phase particles that promote a growth of a thicker and more protective barrier layer.

The inventive alloy content results in a larger amount of secon- dary phase particles being formed in the zirconium based alloy upon the manufacture thereof in comparison to a zirconium based alloy which has a lower content of said alloying elements.

Zirconium based alloys having a composition according to prior

art have a tendency of presenting an uncontrolled growth of secondary phase particles in connection to the manufacture of such alloys. Such an uncontrolled growth of small numbers of large secondary phase particles in the alloy will, in its turn, have a devastating effect on the formation of a thick and protective barrier layer of the zirconium based alloy.

The manufacture of zirconium based alloy having a lower total content of alloying elements than what is proposed by the pres- ent invention will thus require a very precise control in order to avoid an unfavourable size distribution of the secondary phase particles. The manufacture of a zirconium based alloy having a composition according to prior art is therefore very delicate and difficult to handle, as the uncontrolled growth of secondary phase particles has to be considered. Thanks to the inventive composition, this problem is reduced or remedied, such that, in the end, a zirconium based alloy having a good ability of forming a thick and a well protecting barrier layer can be obtained while using a substantially simpler and more insensitive manufacturing process.

According to another preferred embodiment, the Cr content is preferably in the range of 0.16-0.40 weight-%.

The Fe content is in the range of 0-0.6 weight-%, and in a pre- ferred embodiment the Fe content is : 0. 22 weight-%, preferably >0.3 weight-%.

In a particular embodiment, the content of Fe is ! 0. 37 weight-%.

According to a preferred embodiment, the inventive alloy con- tains W, the W content being in the range of 0-0.5 weight-%. W results in a higher frequency of secondary phase particles in the alloy, i. e. more but smaller particles. Particularly, this is relevant when the alloying amount is high. According to another pre- ferred embodiment, the W content is in the range of 0-0.2

weight-%. The finer distribution of secondary phase particles thereby results in an improved corrosion resistance.

The zirconium based alloy according to the invention comprises Sn, the Sn content being in the range of 0.5-2.5 weight-%. Tin will particularly contribute to an improvement of the mechanical strength of the zirconium based alloy. In certain applications, the Sn content may be substantially higher than in conventional Zirkaloy-2 and Zirkaloy-4, which per definition have a Sn content which does not exceed 1.7 weight-%.

The inventive alloy also comprises Ni, the Ni content being in the range of 0.01-0.2 weight-%. Nickel has the advantage of re- ducing the corrosion speed by zirconium based alloys, but might in certain cases increase the hydrogen absorption of the mate- rial. In order to improve the corrosion resistance of the alloy while simultaneously delimiting its tendency of absorbing hydro- gen, the content of Ni is delimited to the above range.

According to another embodiment, the Fe content is in the range of 0-0.05 weight-%. The total alloy content, but the Cr content, does not then need to be higher than by Zirkaloy materials of prior art. The Fe content is chosen upon the fact that it has been shown that the corrosion speed is lower for a zirconium based alloy which has a low Fe content/Cr content ratio in relation to a corresponding alloy in which the ratio has a higher value. This is particularly relevant for applications in which the zirconium based alloy is arranged in a nuclear power plant and thus being subjected to corrosive media as well as radiation. The difference is due to the fact that secondary phase particles are dissolve when subjected to radiation, and that particles having a rela- tively high Fe content tend to be dissolve faster than particles having a lower Fe content. A zirconium based alloy with this composition will thus have more stable secondary phase parti- cles, and, thereby, an improved corrosion and hydration resis- tance.

The invention also relates to a component in a nuclear power plant, particularly a nuclear fuel component in a light-water re- actor, said component comprising the inventive alloy and being subjected to corrosive media, e. g. reactor water, and an ele- vated radiation, e. g. fast neutrons.

Further features and advantages of the inventive zirconium based alloy will appear in the following description and in the dependent patent claims.

DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS Different embodiments of the inventive, zirconium based alloys will hereinafter be described by way of example but not in a de- limiting purpose.

The contents of certain alloying elements, primarily Fe and Cr, present in the respective alloys are a result of the inventors having realised that the total content of these alloying elements, or at least any of these alloying elements, should be increased in relation to what is the case for Zirkaloy-2 and Zirkaloy-4, such that each respective zirconium based alloy shall obtain such a combination of secondary phase particle size and volume frac- tion that a barrier layer which is thick and has a good protective effect is formed on the alloy when the latter initially oxides in a certain environment.

The zirconium based alloys shown are particularly well suited for forming a constituent in a component arranged in a certain envi- ronment in a nuclear power plant. The component may, for ex- ample, be a cladding tube, a grid or a grid box for fuel bundles intended for use in, for example, a light-water reactor. Thereby, the component is arranged in a corrosive environment and is also subjected to radiation, such as neutron radiation,- radiation or y-radiation.

Table 1 below shows a composition specified for the zirconium based alloys Zirkaloy-2 and Zirkaloy-4, and preferred ranges of the compositions of different embodiments of the inventive alloy.

Table 1: Zr Sn Fe Cr Ni W Alloy Min. Max. Min. Max. Min. Max. Min. Max. Min. Max. Zircaloy-2 Balance 1, 2 1. 7 0,05 0,15 0, 03 0, 15 0,03 0,08 Zircaloy-4 Balance 1, 2 1, 7 0,18 0,24 0, 07 0,13 1 1 Balance 0,6 >0, 15 0, 6 0,01 0,2 2 Balance 0, 5 1, 0 0,05 0,2 3 Balance 2, 0 2, 5 0 0, 6 0, 17 0, 6 0 0,2 4 Balance 0, 5 2, 5 0 0, 05 >0, 15 0, 6 0 0,2 5 Balance 0, 5 2, 5 0 0, 6 >0, 15 0, 6 0 0, 2 0 0, 5

All contents in table 1 as well as in the rest of this application, are given in weight-%. Three particularly preferred alloy compo- sitions are shown in table 2 below: Table 2: Alloy Zr Sn Fe Cr Ni W 1 Balance 1, 3 0,25 0,17 0,06 2 Balance 0,17 0,06 3 Balance 1,3 0,33 0,17 0,06 0, 08 Zirconium based alloys with compositions within the ranges de- fined in table 1 for samples 1-5, and in particular the alloys 1-3 defined in table 2, present a secondary phase particle size dis- tribution and a volume fraction of secondary phase particles which are favorable for the formation of a barrier layer which is thick and keeps its protecting effect against corrosion for a long time. A suitable alloy may have an Fe content in the range of 0.45-0.6 weight-%, and a Cr content in the range of 0.2-0.4 weight-%.

Components made of these alloys are particularly well suited for use in nuclear power plants in which the environment is corro- sive and where they are subjected to an elevated or highly ele- vated radiation. However, they may advantageously be used in

all corrosive environments where there is radiation, also in other environments than the one existing in nuclear power plants.

It should be realised that a number of variants and alternative embodiments of the inventive zirconium based alloy of course will be obvious for a man skilled in the art without thereby de- parting from the scope of the invention such as defined in the annexed patent claims.

Moreover, it should be noted that the impurities which normally exist in zirconium based alloys are specified as below: Tabell 3: Element AI B C Ca Cd Cl Co Max. ppm 75 0,5 270 30 0,5 20 20 Element Cu H Hf Mg Mn Mo N Max. ppm 5025100 2050 50j30 Etement Na Pb Si Ti U Max. ppm 20130 120 503. 5 Si and O are present at contents where Si is 50-120 ppm and O is 900-1600 ppm.