GB2119559A

GB2119559A - Zirconium alloy barrier having improved corrosion resistance

Info

Publication number: GB2119559A
Application number: GB08236441A
Authority: GB
Inventors: Ronald Bert Adamson; Herman Soloman Rosenbaum
Original assignee: General Electric Co
Current assignee: General Electric Co
Priority date: 1982-05-03
Filing date: 1982-12-22
Publication date: 1983-11-16
Also published as: BE895526A; DE3248235A1; ES8506926A1; GB2119559B; SE8300016D0; JPS58199836A; SE459101B; SE8300016L; IT1153911B; CA1209726A; ES518638A0; FR2526213A1; KR910003286B1; KR840003119A; FR2526213B1; IT8224877A1; IT8224877A0

Abstract

A nuclear fuel element for use in the core of a nuclear reactor has a composite cladding container having a substrate and a dilute zirconium alloy liner bonded to the inside surface of the substrate. The dilute zirconium alloy liner forms about 1 to about 20 percent of the thickness of the cladding and is comprised of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper. The dilute zirconium alloy linear shields the substrate from impurities or fission products from the nuclear fuel material and protects the substrate from stress corrosion and stress cracking. The dilute zirconium alloy liner displays greater corrosion resistance, especially to oxidation by hot water or steam than unalloyed zirconium. The substrate material is selected from conventional cladding materials, and preferably is a zirconium alloy.

Description

SPECIFICATION Zirconium alloy barrier having improved corrosion resistance This invention relates broadly to an improvement in nuclear fuel elements for use in the core of nuclear fission reactors and, more particularly, to an improved nuclear fuel element having a composite cladding container having a metal liner of dilute zirconium alloy consisting of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper bonded to the inside surface of the zirconium alloy cladding substrate.

Nuclear reactors are presently being designed, constructed, and operated in which the nuclear fuel is contained in fuel elements which can have various geometric shapes, such as plates, tubes, or rods. The fuel material is usually enclosed in a corrosion-resistant, non-reactive, heat conductive container or cladding. The fuel elements are assembled together in a lattice at fixed distances from each other in a coolant fiow channel or region forming a fuel assembly, and sufficient fuel assemblies are combined to form the nuclear fission chain reacting assembly or reactor core capable of a self-sustained fission reaction. The core, in turn, is enclosed within a reactor vessel through which a coolant is passed.

The cladding serves several purposes and two primary purposes are: first, to prevent contact and chemical reactions between the nuclear fuel and the coolant or the moderator if a moderator is present, or both if both the coolant and the moderator are present; and second, to prevent the radioactive fission products, some of which are gases, from being released from the fuel into the coolant or the moderator or both if both the coolant and the moderator are present. Common cladding materials are stainless steel, aluminum and its alloys, zirconium and its alloys, niobium (columbium), certain magnesium alloys, and others. The failure of the cladding, i.e., a loss of the leak tightness, can contaminate the coolant or moderator and the associated systems with longlived radioactive products to a degree which interferes with plant operation.

Problems have been encountered in the manufacture and in the operation of nuclear fuel elements which employ certain metals and alloys as the cladding material due to mechanical or chemical reactions of these cladding materials under certain circumstances. Zirconium and its alloys, under normal circumstances, make excellent nuclear fuel claddings since they have low neutron absorption cross-sections and at temperatures below about 750cm (about 3980C) are strong, ductile, extremely stable and relatively non-reactive in the presence of demineralized water or steam which are commonly used as reactor coolants and moderators.

However, fuel element performance has revealed a problem with the brittle splitting of the cladding due to the combined interactions between the nuclear fuel, the cladding and the fission products produced during nuclear fission reactions. It has been discovered that this undesirable performance is promoted by localized mechanical stress due to differential expansion of fuel and cladding (stresses in the cladding are concentrated at cracks in the nuclear fuel).

Corrosive fission products are released from the nuclear fuel and are present at the intersection of the fuel cracks with the cladding surface. Such fission products are created in the nuclear fuel during the fission chain reaction during operation of a nuclear reactor. The localized stress is exaggerated by high friction between the fuel and the cladding.

Within the confines of a sealed fuel element, hydrogen gas can be generated by the slow reaction between the cladding and residual water inside the cladding. This hydrogen gas may build up to levels which, under certain conditions, can result in localized hydriding of the cladding with concurrent local deterioration in the mechanical properties of the cladding. The cladding is also adversely affected by such gases as oxygen, nitrogen, carbon monoxide, and carbon dioxide over a wide range of temperatures.The zirconium cladding of a nuclear fuel element is exposed to one or more of the gases listed above and fission products during irradiation in a nuclear reactor and this occurs in spite of the fact that these gases may not be present in the reactor coolant or moderator, and further may have been excluded as far as possible from the ambient atmosphere during manufacture of the cladding and the fuel element. Sintered refractory and ceramic compositions, such as uranium dioxide and other compositions used as nuclear fuel, release measurable quantities of the aforementioned gases upon heating, such as during fuel element manufacture and further release fission products during irradiation.Particulate refractory and ceramic compositions, such as uranium dioxide powder and other powders used as nuclear fuel, have been known to release even larger quantities of the aforementioned gases during irradiation.

These released gases are capable of reacting with the zirconium cladding containing the nuclear fuel.

Thus, in light of the foregoing, it has been found desirable to minimize attack of the cladding from water, water vapor and other gases, especially hydrogen, which are reactive with the cladding from inside the fuel element throughout the time the fuel element is used in the operation of nuclear power plants. One such approach has been to find materials which will chemically react rapidly with the water, water vapor and other gases to eliminate these from the interior of the cladding. Such materials are called getters.

Another approach has been to coat the nuclear fuel material with any of a variety of materials to prevent moisture coming in contact with the nuclear fuel material. The coating of nuclear fuel material introduces reliability problems in that achieving uniform coatings free of faults is difficult. Further, the deterioration of the coating can introduce problems with the long-lived performance of the nuclear fuel material.

Document GEAP-4555; dated February, 1964, discloses a composite cladding of a zirconium alloy with an inner lining of stainless steel metallurgically bonded to the zirconium alloy, and the composite cladding is fabricated by extrusion of a hollow billet of the zirconium alloy having an inner lining of stainless steel. This cladding has the disadvantage that the stainless steel develops brittle phases, and the stainless steel layer involves a neutron absorption penalty of about ten to fifteen times the penalty for a zirconium alloy of the same thickness.

U.S. Patent No. 3,502,549 discloses a method for protecting zirconium and its alloys by the electrolytic deposition of chromium to provide a composite material useful for nuclear reactors. A method for electrolytic deposition of copper on Zircaloy-2 surfaces and subsequent heat treatment for the purpose of obtaining surface diffusion of the electrolytically deposited metal is presented in Energia Nucleare, Volume 11, No. 9 (September 1964) at pages 505-508. In Stability and Compatibility of Hydrogen Barriers Applied toZirconiumAlloys, by F.Brossa et al (European Atomic Energy Community, Joint Nuclear Research Center, EUR 4098e, 1969), methods of deposition of different coatings and their efficiency as hydrogen diffusion barriers are described along with an Al-Si coating as the most promising barrier against hydrogen diffusion.

Methods for electroplating nickel on zirconium and zirconium-tin alloys and heat treating these alloys to produce alloy-diffusion bonds are disclosed in Electroplating on Zirconium and Zirconium-Tin, by W. C. Schickner et al (BMI-757, Technical Information Service, 1952).

U.S. Patent No. 3,625,821 presents a fuel element for a nuclear reactor having a fuel cladding tube with the inner surface of the tube being coated with a metal of low neutron capture cross-section such as nickel and having finely dispersed particles of a burnable poison disposed therein. Reactor Development Program Progress Report of August, 1973 (ANL-RDP-19) discloses a chemical getter arrangement of a sacrificial layer of chromium on the inner surface of a stainless steel cladding.

Another approach has been to introduce a barrier between the nuclear fuel material and the cladding holding the nuclear fuel material as disclosed in U.S. Patent No. 3,230,150 (copper foil); German Patent Publication DAS 1,238,115 (titanium layer); U.S. Patent No. 3,212,988 (sheath of zirconium, aluminum or beryllium); U.S.

Patent No. 3,018,238 (barrier of crystalline carbon between the UO2 and the zirconium alloy cladding); and U.S. Patent No. 3,088,893 (stainless steel foil). While the barrier concept proves promising, some of the foregoing references involve incompatibie materials with either the nuclear fuel (e.g., carbon can combine with oxygen from the nuclear fuel), or the cladding (e.g., copper and other metals can react with the cladding, altering the properties of the cladding), or the nuclear fission reaction (e.g., by acting as neutron absorbers). None of the listed references disclose solutions to the problem of localized chemical-mechanical interactions between the nuclear fuel and the cladding.

Further approaches to the barrier concept are disclosed in U.S. Patent No. 3,969,186, (refractory metal such as molybdenum, tungsten, rhenium, niobium and alloys thereof in the form of a tube or foil of single or multiple layers or a coating on the internal surface of the cladding), and U.S. Patent No. 3,925,151 (liner of zirconium, niobium, or alloys thereof between the nuclear fuel and the cladding with a coating of a high lubricity material between liner and the cladding).

U.S. Patent No. 4,045,288 discloses a composite cladding of a zirconium alloy substrate with a metal barrier metallurgically bonded to the substrate and an inner layer of zirconium alloy metallurgically bonded to the metal barrier. The barrier is selected from a group of niobium, aluminum, copper, nickel, stainless steel, and iron.

The buried metal barrier reduces corrosion due to fission products and corrosive gases, but is subject to stress corrosion cracking and liquid metal embrittlement.

U.S. Patent No. 4,200,492 discloses a composite cladding of a zirconium alloy substrate with a sponge zirconium liner. The soft zirconium liner minimizes localized strain, and reduces stress corrosion cracking and liquid metal embrittlement, but is subject to loss due to honing and the like during fabrication and due to oxidation. Furthermore, if a breach in the cladding should occur, allowing water and/or steam to enter the fuel rod, the zirconium liner tends to oxidize rapidly.

Accordingly, it has remained desirable to develop nuclear fuel elements minimizing the problems discussed above.

A particularly effective nuclear fuel element for use in the core of a nuclear reactor has a composite cladding having a metal liner of dilute zirconium alloy metallurgically bonded to the inside surface of the substrate. The dilute zirconium alloy comprises zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium and copper, wherein the amount of iron alloyed with the zirconium is from about 0.2% to about 0.3% by weight; the amount of chromium is from about 0.05% to about 0.3% by weight; the total amount of iron plus chromium is from about 0.15% to about 0.3% by weight and wherein the ratio of the weights of iron to chromium is in the range of from about 1:1 to about 4:1; and wherein the amount of copper is from about 0.02% to about 0.2% by weight.

The substrate of the cladding is completely unchanged in design and function from previous practice for a nuclear reactor and is selected from conventional cladding materials such as zirconium alloys. A zirconium alloy cladding substrate has a higher alloy content than the dilute zirconium alloy liner. The dilute zirconium alloy liner forms a continuous shield between the substrate and the nuclear fuel material held in the cladding, as well as shielding the zirconium alloy or other substrate cladding from fission products and gases.

The dilute zirconium alloy liner forms from about 1 to about 20 percent of the thickness of the cladding. The liner remains soft, relative to the substrate, during irradiation and minimizes localized stress inside the nuclear fuel element, thus serving to protect the cladding from stress corrosion cracking or liquid metal embrittlement.

The dilute zirconium alloy liner shields the substrate from reaction with volatile impurities or fission products present inside the nuclear fuel element and, in this manner, serves to protect the cladding substrate from attack by the volatile impurities or fission products.

This invention has a striking advantage that the substrate of the cladding is protected from stress corrosion cracking and liquid metal embrittlement, in addition to contact with fission products, corrosive gases, etc., by the dilute zirconium alloy liner and the liner does not introduce any appreciable neutron capture penalties, heat transfer penalties, or fuel/liner incompatibility problems. In addition, the liner provides superior resistance to steam or hot water oxidation as compared to unalloyed zirconium in the event of a breach in the cladding.

The present invention will be further described by way of example only, with reference to the accompanying drawings, in which: Fig. 1 is a partial cutaway sectional view of a nuclear fuel assembly containing nuclear fuel elements constructed according to the teaching of this invention; and Fig. 2 is an enlarged transverse cross-sectional view of the nuclear fuel element in Fig. 2 illustrating the teaching of this invention.

Referring now more particularly to Fig. 1, there is shown a partially cutaway sectional view of a nuclear fuel assembly 10. This fuel assembly 10 consists of a tubular flow channel 11 of generally square cross section provided at its upper end with a lifting bail 12 and at its lower end with a nose piece (not shown due to the lower portion of assembly 10 being omitted). The upper end of channel 11 is open at outlet 1 3 and the lower end of the nose piece is provided with coolant flow openings. An array of fuel elements or rods 14 is enclosed in the channel 11 and supported therein by means of an upper end plate 1 5 and a lower end plate (not shown due to the lower portion being omitted).The liquid coolant ordinarily enters through the openings in the lower end of the nose piece, passes upwardly around fuel elements 14, and discharges through the upper outlet 1 3 at an elevated temperature in a partially vaporized condition for boiling reactors or in an unvaporized condition for pressurized reactors.

The nuclear fuel elements or rods 14 are sealed at their ends by means of end plugs 18 welded to the cladding 17, which may include studs 1 9 to facilitate the mounting of the fuel rod in the assembly. A void space or plenum 20 is provided at one end of the element to permit longitudinal expansion of the fuel material and accumulation of gases released from the fuel material. A nuclear fuel material retainer means 24 in the form of a helical member is positioned within space 20 to provide restraint against the axial movement of the pellet column, especially during handling and transportation-of the fuel element.

The fuel element is designed to provide an excellent thermal contact between the cladding and the fuel- material, a minimum of parasitic neutron absorption and resistance to bowing and vibration which is occasionally caused by flow of the coolant at high velocity.

A nuclear fuel element or rod 14 constructed according to the teachings of this invention is shown in a partial section in Fig. 1. The fuel element includes a core or central cylindrical portion of nuclear fuel material 1 6, here shown as a plurality of fuel pellets of fissionable and/or fertile material positioned within a structural cladding or container 1 7. In some cases, the fuel pellets may be of various shapes such as cylindrical pellets or spheres and, in other cases, different fuel forms such as a particulate fuel may be used. The physical form of the fuel is immaterial to this invention. Various nuclear fuel materials may be used including uranium compounds, plutonium compounds, thorium compounds, and mixtures thereof. A preferred fuel is uranium dioxide or a mixture comprising uranium dioxide and plutonium dioxide.

Referring now to Fig. 2, the nuclear fuel material 1 6 forming the central core of the fuel element 14 is surrounded by a cladding 17 which, in this invention, is also referred to as a composite cladding container. The composite cladding container encloses the fissile core so as to leave a gap 23 between the core and the cladding during use in a nuclear reactor. The composite cladding container has an external substrate 21 selected from conventional cladding materials such as a stainless steel and zirconium alloys and, in a preferred embodiment of this invention, the substrate is a zirconium alloy such as Zircaloy-2.

The substrate 21 has metallurgically bonded on the inside circumference thereof a dilute zirconium alloy liner 22 so that the dilute zirconium alloy liner forms a shield of the substrate from the nuclear fuel material 1 6 inside the composite cladding. The dilute zirconium alloy liner preferably forms about 1 to about 20% of the thickness of the cladding. A dilute zirconium alloy liner forming less than about 1% of the thickness of the cladding would be difficult to achieve in commercial production, and a dilute zirconium alloy liner forming more than 20% of the thickness of the cladding provides no additional benefit for the added thickness. Further, a liner more than about 20% of the thickness of the cladding means a concomitant reduction in thickness of the substrate and possible weakening of the cladding.

The dilute zirconium alloy is comprised of zirconium and an alloy addition selected from the group consisting of: iron, chromium, iron plus chromium and copper. As used herein, dilute zirconium alloy means a zirconium alloy with an alloy content sufficiently low to display greater ductility and higher strain rate than does the substrate material under equivalent conditions of stress.

The amount of iron alloyed with zirconium is from about 0.2% to about 0.3% by weight, and preferably from about 0.29/0 to about 0.259/0 by weight.

Chromium is in the range of from about 0.059/0 to about 0.3% by weight and preferably from about 0.1 5% to about 0.25% by weight.

Iron plus chromium may be included so that the total amount of both components is from about 0. 159/0 to about 0.3% by weight and preferably from about 0.2% to about 0.25% by weight and wherein the ratio of the weights of iron to chromium is from about 1:1 to about 4:1 and preferably about 2:1.

Copper is used in the amount of from about 0.02% to about 0.2% by weight and preferably from about 0.05% to about 0.1 5% by weight.

The dilute zirconium alloy liner shields the substrate from gaseous impurities and fission products and protects the substrate portion of the cladding from contact and reaction with such impurities and fission products and prevents the occurrence of localized stress.

The addition to zirconium of small amounts of a metal selected from the group of iron, chromium, iron plus chromium and copper improves corrosion resistance, especially resistance to oxidation by hot water or steam if the addition is within the stated range for that metal. The lower limit of the amount of each metal alloyed with zirconium provides sufficient quantity of that metal to significantly improve the corrosion resistance as compared to unalloyed zirconium.

The upper limit of the amount of each metal alloyed with zirconium is generally set at the maximum amount of the metal which significantly improves the corrosion resistance as compared with sponge zirconium. Additions of the metal exceeding the upper limit fail to significantly enhance the corrosion resistance properties of zirconium and may have a detrimental effect in reducing the softness and ductility of the liner.

The additions of each metal to zirconium that impart the greatest improvement in corrosion resistance are stated as the preferred ranges.

Iron, chromium and copper are sparingly soluble in zirconium. Diiute zirconium alloys involving one or more of these metals can be heat treated to provide a material with a fine dispersion of intermetallic particles which are noble with respect to the zirconium matrix.

Because the alloy constituents are sparingly soluble, little solid solution strengthening of the zirconium occurs. The strengthening effect is sufficiently low to maintain the softness required for the dilute zirconium alloy liner to prevent or mitigate fuel failure by pellet-cladding interaction.

The dilute zirconium alloy liner in the composite cladding resists irradiation hardening relative to Zircaloy or other conventional zirconium alloys, and this enables the dilute zirconium alloy liner, after prolonged irradiation, to maintain desirable structural properties such as yield strength and hardness at levels considerably lower than those of conventional zirconium alloys.

In effect, the dilute zirconium alloy liner does not harden as much as conventional zirconium alloys when subjected to irradiation and this, together with its initially low yield strangth, enables the dilute zirconium alloy liner to deform plastically and relieve pellet-induced stresses in the fuel element that can be brought about, for example, by swelling of the pellets of nuclear fuel at reactor operating temperatures (3000C to 3500C) so that the pellet comes into contact with the cladding.

A dilute zirconium alloy liner comprising zirconium and a metal selected from the group comprising of iron, chromium, iron plus chromium, and copper and preferably about 5 to 1 5 percent of the thickness of the cladding bonded to a conventional zirconium alloy substrate provides stress reduction sufficient to prevent or mitigate failures in the composite cladding.

The purity of the zirconium metal that is alloyed with iron, chromium, iron plus chromium or copper is important and serves to impart special properties to the dilute zirconium alloy liner. Generally, there is less than 5000 ppm impurities in the zirconium metal. Of these oxygen should be as low as practical but may vary up to about 1000 ppm.

The composite cladding of the nuclear fuel element of this invention has a dilute zirconium alloy liner metallurgically bonded to the substrate.

Metallographic examination shows that there is sufficient cross-diffusion between the substrate and the dilute zirconium alloy liner to form metallurgical bonds, but insuffient cross-diffusion to alloy significantly with the dilute zirconium alloy liner itself.

Among the conventional zirconium alloys serving as suitable substrates are Zircaloy-2 and Zircaloy-4. Zircaloy-2 has on a weight basis about 1.5 percent tin; 0.12 per cent iron; 0.09 percent chromium and 0.005 percent nickel and is extensively employed in water-cooled reactors.

Zircaloy-4 has less nickel than Zircaloy-2, but contains slightly more iron than Zircaloy-2. The composite cladding used in the nuclear fuel elements of this invention can be fabricated by any of the following methods.

In one method, a tube of the dilute zirconium alloy liner material is inserted into a hollow billet of the material selected to be the substrate, and then the assembly is subjected to explosive bonding of the tube to the billet. The composite is extruded using conventional tube shell extrusion at elevated temperatures of about 10000F to 1 40O0F (about 5380C to 76O0C). Then the extruded composite is subjected to a process involving conventional tube reduction until the desired size of cladding is achieved. The relative wall thickness of the hollow billet and the dilute zirconium alloy liner tube are selected to give the desired thickness ratios in the finished cladding tube.

In another method, a tube of the dilute zirconium alloy liner material is inserted into a hollow billet of the material selected to be the substrate, and then the assembly is subjected to a heating step (such as at 7500C for 8 hours) under compressive stress to assure good metal-to-metal contact and diffusion bonding between the tube and the billet. The diffusion bonded composite is extruded using conventional tube shell extrusion such as described above in the immediately preceding paragraph. Then the extruded composite is subjected to a process involving conventional tube reduction until the desired size of cladding is achieved.

In still another method, a tube of the dilute zirconium alloy liner material is inserted into a hollow billet of the material selected to be the substrate, and the assembly is extruded using conventional tube shell extrusion as described above. Then the extruded composite is subjected to a process involving conventional tube reduction until the desired size of cladding is achieved.

The foregoing processes of fabricating the composite cladding of this invention give economies over other proceses used in fabricating cladding such as electroplating or vapor deposition. A nuclear fuel element can be forged by making a composite cladding container which is open at one end, the cladding container having a substrate and an inner dilute zirconium alloy liner consisting of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper metallurgically bonded to the inside surface of the substrate, filling the composite cladding container with nuclear fuel material and leaving a cavity at the open end, inserting a nuclear fuel material retaining means into the cavity, applying an enclosure to the open end of the container leaving the cavity in communication with the nuclear fuel, and then bonding the end of the cladding container to said enclosure to form a tight seal therebetween.

The present invention offers several advantages promoting a long operating life for a nuclear fuel element, including the reduction of chemical interaction of the cladding, the minimization of localized stress on the zirconium alloy substrate portion of the cladding, the minimization of stress corrosion on the zirconium alloy substrate portion of the cladding, and the reduction of the probability of a splitting failure occurring in the zirconium alloy substrate.

In addition to minimizing stress and stress corrosion on the substrate, the dilute zirconium alloy liner is resistant to oxidation by steam and hot water in the event that the cladding is breached, whereas unalloyed zirconium oxidizes rapidly under these conditions. The dilute zirconium alloy exhibits plasticity similar to unalloyed zirconium and provides the benefits thereof while also providing increased resistance to corrosion, especially to oxidation by hot water and steam.

An important property of the composite cladding of this invention is that the foregoing improvements are achieved with no substantial additional neutron penalty. Such a cladding is readily accepted in nuclear reactors since the cladding would have no eutectic formation during a loss-of-coolant accident or an accident involving the dropping of a nuclear control rod. Further, the composite cladding has a very small heat transfer penalty in that there is no thermal barrier to transfer of heat such as results in the situation where a separate foil or liner is inserted in a fuel element. Also, the composite cladding of this invention is inspectable by conventional nondestructive testing methods during various stage of fabrication and operation.

As will be apparent to those skilled in the art, various modifications and changes may be made in the invention described herein. It is accordingly the intention that the invention be construed in the broadest manner within the spirit and scope as set forth in the accompanying claims.

Claims

1. A nuclear fuel element comprising: a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium, and mixtures thereof; and an elongated composite cladding container enclosing said core including an outer portion forming a substrate, and a dilute zirconium alloy liner formed of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper metallurgically bonded on the inside surface of the substrate, said dilute zirconium alloy liner comprising from about 1 to about 20 percent of the thickness of the composite cladding container.

2. The nuclear fuel element of claim 1 in which the dilute zirconium alloy liner comprises from about 0.2% to about 0.3% by weight iron, the balance being zirconium.

3. The nuclear fuel element of claim 1 in which the dilute zirconium alloy liner comprises from about 0.2% to about 0.25% by weight iron, the balance being zirconium.

4. A nuclear fuel element of claim 1 in which the dilute zirconium alloy liner comprises from about 0.05% to about 0.3% by weight chromium, the balance being zirconium.

5. A nuclear fuel element of claim 1 in which the dilute zirconium alloy liner comprises from about 0.15% to about 0.25% by weight chromium, the balance being zirconium.

6. A nuclear fuel element of claim 1 in which the dilute zirconium alloy liner comprises iron and chromium wherein the total amount of iron plus chromium is from about 0.15% to about 0.3% by weight, the balance being zirconium and wherein the weight ratio of iron to chromium is from about 1:1 to about 4:1.

7. A nuclear fuel element of claim 6 wherein the weight ratio of iron to chromium is about 2:1.

8. A nuclear fuel element of claim I in which the dilute zirconium alloy liner comprises iron and chromium wherein the total amount of iron plus chromium is from about 0.2% to about 0.25% by weight, the balance being zirconium and wherein the weight ratio of iron to chromium is from about 1:1 to about 4:1.

9. A nuclear fuel element of claim 8 wherein the weight ratio of iron to chromium is about 2:1.

10. The nuclear fuel element of claim 1 in which the dilute zirconium alloy liner comprises about 0.02% to about 0.2% by weight copper, the balance being zirconium.

11. The nuclear fuel element of claim 1 in which the dilute zirconium alloy liner comprises about 0.05% to about 0.1 5% by weight copper, the balance being zirconium.

12. The nuclear fuel element of claim 1 in which the dilute zirconium alloy liner comprises from about 5 to about 1 5 percent of the thickness of the composite cladding container.

1 3. A composite cladding container for nuclear reactors comprising a zirconium alloy outer portion forming a substrate and a dilute zirconium alloy liner formed of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper metallurgically bonded on the inside surface of the substrate, said zirconium alloy liner comprising from about 5 to about 1 5 percent of thickness of the composite cladding container.

14. A composite cladding container as claimed in claim 13 in which the dilute zirconium alloy liner comprises from about 0.02% to about 0.3% by weight iron, the balance being zirconium.

1 5. A composite cladding container as claimed in claim 13 in which the dilute zirconium alloy liner comprises from about 0.2% to about 0.25% by weight iron, the balance being zirconium.

1 6. A composite cladding container as claimed in claim 13 in which the dilute zirconium alloy liner comprises from about 0.05% to about 0.3% by weight chromium, the balance being zirconium.

1 7. A composite cladding container as claimed in claim 13 in which the dilute zirconium alloy liner comprises from about 0.1 5% to about 0.25% by weight chromium, the balance being zirconium.

1 8. A composite cladding container as claimed in claim 13 in which the dilute zirconium alloy liner comprises iron and chromium wherein the total amount of iron plus chromium is from about 0.15% to about 0.3% by weight, the balance being zirconium and wherein the weight ratio of iron to chromium is from about 1:1 to about 4:1.

1 9. A composite cladding container as claimed in claim 13 in which the dilute zirconium alloy liner comprises iron and chromium wherein the total amount of iron plus chromium is from about 0.2% to about 0.25 by weight, the balance being zirconium and wherein the weight ratio of iron to chromium is from about 1:1 to about 4:1.

20. A composite cladding container as claimed in claim 13 in which the dilute zirconium alloy liner comprises about 0.02% to about 0.2% by weight copper, the balance being zirconium.

21. A composite cladding container as claimed in claim 13 in which the dilute zirconium alloy liner comprises about 0.05% to about 0.15% by weight copper, the balance being zirconium.

22. A hollow composite cladding container comprising an outer portion formed of a zirconium alloy forming a substrate and a continuous dilute zirconium alloy liner formed of zirconium and a metal selected from the group consisting of iron, in the range of from about 0.2% to about 0.3% by weight; chromium in the range of from about 0.05% to about 0.3% by weight; iron plus chromium in the range of from about 0.15% to about 0.3% by weight wherein the weight ratio of iron to chromium is from about 1:1 to about 4:1; and copper in the range of from about 0.02% to about 0.2% by weight.

23. A composite cladding container as claimed in claim 22 in which the dilute zirconium alloy liner comprises from about 0.2% to about 0.25% by weight iron, the balance being zirconium.

24. A composite cladding container as claimed in claim 22 in which the dilute zirconium alloy liner comprises from about 0.15% to about 0.25% by weight chromium, the balance being zirconium.

25. A composite cladding container as claimed in claim 22 in which the dilute zirconium alloy liner comprises iron and chromium wherein the total amount of iron plus chromium is from about 0.2% to about 0.25% by weight, the balance being zirconium.

26. A composite cladding container as claimed in claim 22 in which the dilute zirconium alloy liner comprises about 0.05% to about 0.15% by weight copper, the balance being zirconium.

27. A composite cladding container as claimed in claim 22 in which the dilute zirconium alloy liner comprises from about 1 to about 20 percent of the thickness of the composite cladding container.

28. A composite cladding container as claimed in claim 22 in which the dilute zirconium alloy liner comprises from about 5 to 1 5 percent of the thickness of the composite cladding container.

29. In a hollow composite cladding container for nuclear fuel for use in a nuclear reactor comprising an outer substrate of zirconium alloy and an inner liner metallurgically bonded to the zirconium alloy substrate, the improvement comprising forming the liner of a dilute zirconium alloy consisting essentially of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper.

30. A composite cladding container as claimed in claim 29 in which the dilute zirconium alloy liner comprises from about 0.2% to about 0.3% by weight iron, the balance being zirconium.

31. A composite cladding container as claimed in claim 29 in which the dilute zirconium alloy liner comprises from about 0.2% to about 0.25% by weight iron, the balance being zirconium.

32. A composite cladding container as claimed in claim 29 in which the dilute zirconium alloy liner comprises from about 0.05% to about 0.3% by weight chromium, the balance being zirconium.

33. A composite cladding container as claimed in claim 29 in which the dilute zirconium alloy liner comprises from about 0.1 5% to about 0.25% by weight chromium, the balance being zirconium.

34. A composite cladding container as claimed in claim 29 wherein the dilute zirconium alloy liner comprises iron and chromium wherein the total amount of iron plus chromium is from about 0.1 5% to about 0.3% by weight, the balance being zirconium and wherein the weight ratio of iron to chromium is from about 1:1 to about 4:1.

35. A composite cladding container as claimed in claim 29 wherein the dilute zirconium alloy liner comprises iron and chromium wherein the total amount of iron plus chromium is from about 0.2% to about 0.25% by weight, the balance being zirconium and wherein the weight ratio of iron to chromium is from about 1:1 to about 4:1.

36. A composite cladding container as claimed in claim 29 in which the dilute zirconium alloy liner comprises about 0.02% to about 0.2% by weight copper, the balance being zirconium.

37. A composite cladding container as claimed in claim 29 in which the dilute zirconium alloy liner comprises about 0.05% to about 0.15% by weight copper, the balance being zirconium.

38. A composite cladding container as recited in claim 29 wherein the dilute zirconium alloy liner has a thickness in the range of from about 1 to about 20 percent of the thickness of the composite cladding container.

39. A composite cladding container as claimed in claim 29 wherein the dilute zirconium alloy liner has a thickness in the range of from about 5 to about 1 5 percent of the thickness of the composite cladding container.

40. A nuclear fuel element comprising: a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium, and mixtures thereof; and an elongated composite cladding container enclosing said core including an outer portion of zirconium alloy forming a substrate, and a zirconium alloy liner having a lower alloy content than the substrate formed of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper metallurgically bonded on the inside surface of the substrate, said zirconium alloy liner comprising from about 1 to about 20 percent of the thickness of the composite cladding container.

41. The nuclear fuel element of claim 40 in which the zirconium alloy liner comprises from about 0.2% to about 0.3% by weight iron, the balance being zirconium.

42. The nuclear fuel element of claim 40 in which the zirconium alloy liner comprises from about 0.2% to about 0.25% by weight iron, the balance being zirconium.

43. A nuclear fuel element of claim 40 in which the zirconium alloy liner comprises from about 0.05% to about 0.3% by weight chromium, the balance being zirconium.

44. A nuclear fuel element of claim 40 in which the zirconium alloy liner comprises from about 0.15% to about 0.25% by weight chromium, the balance being zirconium.

45. A nuclear fuel element of claim 40 in which the zirconium alloy liner comprises iron and chromium wherein the total amount of iron plus chromium is from about 0.1 5% to about 0.3% by weight, the balance being zirconium and wherein the weight ratio of iron to chromium is from about 1:1 to about 4:1.

46. A nuclear fuel element of claim 45 wherein the weight ratio of iron to chromium is about 2:1.

47. A nuclear fuel element of claim 1 in which the zirconium alloy liner comprises iron and chromium wherein the total amount of iron plus chromium is from about 0.2% to about 0.2t% by weight, the balance being zirconium and wherein the weight ratio of iron to chromium is from about 1:1 to about 4:1.

48. A nuclear fuel element of claim 47 wherein the weight ratio of iron to chromium is about 2:1.

49. The nuclear fuel element of claim 40 in which the zirconium alloy liner comprises about 0.02% to about 0.2% by weight copper, the balance being zirconium.

50. The nuclear fuel element of claim 40 in which the zirconium alloy liner comprises about 0.05% to about 0.1 5% by weight copper, the balance being zirconium.

51. The nuclear fuel element of claim 40 in which the zirconium alloy liner comprises from about 5 to about 1 5 percent of the thickness of the composite cladding container.

52. A nuclear fuel element as claimed in claim 1, substantially as hereinbefore described with reference to and as illustrated in the accompanying drawing.

53. A composite cladding container as claimed in claim 13, 22, 29 or 40 substantially as hereinbefore described with reference to and as illustrated in the accompanying drawing.