JPS58199836A - Zirconium alloy diaphragm with improved anticorrosion - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

DETAILED DESCRIPTION OF THE INVENTION Field of the Invention The present invention relates to improvements in fuel elements for use in the core of nuclear reactors, and more particularly to improvements in fuel elements for use in the core of nuclear reactors.
An improved fuel element comprising a composite cladding vessel comprising a metal selected from iron plus chromium and copper and zirconium and having a metal liner of a lean zirconium alloy bonded to the inner surface of a zirconium alloy cladding substrate.

Background of the invention.

In nuclear reactors currently designed, constructed, and in operation, nuclear fuel is contained in fuel elements of various geometric shapes, such as plates, tubes, or rods. In this case, the nuclear fuel is typically contained in a corrosion-resistant, non-reactive and thermally conductive container or cladding. Fuel assemblies are formed by assembling such fuel elements in a lattice pattern at regular intervals within a channel box for coolant flow, and then a sufficient number of fuel assemblies are combined to produce nuclear fission capable of a sustained fission reaction. A chain reaction system or core is formed. Such a core is housed inside a nuclear reactor vessel, through which coolant is flowed.

The above dressings serve several purposes, but the two main ones are
Here are some examples: First, it prevents contact and chemical reactions between the coolant, the moderator (if present), or both (if both coolant and moderator are present) and the nuclear fuel. useful for. Secondly,
Radioactive nuclear fission, including gaseous ones, Chemical 1. ．． This serves to prevent products from being released from the nuclear fuel into the coolant, moderator, or both (if both are present). Common coating materials include stainless steel, aluminum and its alloys, zirconium and its alloys, niobium (columpinium), and certain magnesium alloys. If cladding failure or loss of seal occurs, the coolant, moderator, and associated systems may become contaminated with long-lived radioactive products, thereby disrupting power plant operations.

The manufacture and operation of fuel elements using certain metals and alloys as claddings presents several problems due to the mechanical or chemical reactions of such claddings under certain circumstances. discovered. Zirconium and its alloys are normally excellent nuclear fuel cladding materials.

That is, they have small neutron absorption cross sections, are tough, ductile, and extremely stable at temperatures below about 750°C (about 398°C), and are commonly used as coolants and moderators in nuclear reactors. Almost no reaction occurs even in the presence of deionized water or steam.

However, when testing the performance of the fuel elements, it was found that nuclear fuel,
It has become clear that brittle rupture occurs as a result of the overall interaction between the cladding material and the fission products produced during the fission reaction. It has also been found that such undesirable performance is exacerbated by the localization of mechanical stress due to differences in thermal expansion between the nuclear fuel and the cladding (stresses in the cladding are concentrated at cracks in the nuclear fuel). ). On the other hand, corrosive fission products released from the nuclear fuel accumulate at the intersections of the nuclear fuel cracks and the inner surface of the cladding. Such nuclear fission products are produced from nuclear fuel by a nuclear fission chain reaction during nuclear reactor operation. As a result of this increased friction between the nuclear fuel and the cladding material, the local stress will further increase.

Inside a sealed fuel element, hydrogen gas may be produced by the slow reaction of the cladding with residual moisture within it. Accumulation of such hydrogen gas can, under certain conditions, reach such levels that it can result in local hydrogenation of the cladding and concomitant local deterioration of the in-machine properties of the cladding. Coatings are also adversely affected by gases such as oxygen, nitrogen, carbon oxides and carbon dioxide over a wide range of temperatures. The zirconium cladding of a fuel element is exposed to one or more of the gases and fission products mentioned above during irradiation within a nuclear reactor. This occurs despite the possible presence of such gases in the reactor coolant and moderator, and even if as much of the gases as possible are excluded from the surrounding atmosphere during the manufacture of the cladding and fuel elements. . That is, refractory sintered ceramic compositions (e.g. uranium dioxide and other compositions) used as nuclear fuels emit significant amounts of these gases when heated (e.g. during the manufacture of fuel elements) and when irradiated. Releases fission products. Furthermore, it is becoming known that refractory particulate ceramic compositions used as nuclear fuels (eg, uranium dioxide powder and other powders) release higher amounts of Be gas upon irradiation. The gases thus released can react with the zirconium cladding containing the nuclear fuel.

As can be seen from the above description, water, water vapor, and other gases (particularly hydrogen) that can react with the cladding from inside the fuel element throughout the period of use of the fuel element in the operation of a nuclear power plant are Therefore, it is desirable to suppress the effect.

One way to do this was to find materials that would rapidly chemically react with water, water vapor, and other gases, thereby helping to exclude them from inside the dressing. Such substances are called getters.

Another method has been to coat the nuclear fuel material itself with various materials to prevent it from coming into contact with moisture. Such coatings of the nuclear fuel material itself have reliability problems in that it is difficult to form a uniform coating free of defects.

Furthermore, if coating degradation occurs, problems may arise in terms of the long-term performance of the nuclear fuel material.

In document GFiAP-4555 dated February 1964,
A composite cladding in which a stainless steel lining is metallurgically bonded to a zirconium alloy has been disclosed. Such composite cladding is manufactured by extruding a hollow zirconium alloy billet with a stainless steel lining. Such coatings have the disadvantage that the stainless steel develops a brittle phase and also that the stainless steel layer exhibits approximately 10 to 15 times more neutron absorption impairment than a zirconium alloy of the same thickness.

U.S. Pat. No. 3,502,549 discloses a method for producing composite materials for nuclear reactors by protecting zirconium or its alloys by electrodeposition of chromium. Also, [Energia Nucleale]
ia Nucleare) J Vol. 11 No. 9 (19
September 1964), pages 505-508, Zircaloy-2
A method is described in which copper is electrodeposited on the surface of a metal, followed by surface diffusion of the electrodeposited metal by heat treatment. In addition, a paper by F. F. Florussa et al. [Stability and Compatibility of Hydrogen Barriers Added to Zirconium Alloys]
of Hydrogen BarriersA
pplied to Zirconium A110y
s) J (Eurosono; Atomic Energy Community (Europe)
ean Atomic Energy Community
y), Joint Nuclear Research Center (JointNucl)
, ear Research Center)
, KUR4098e.

(1969) describe methods for installing various coatings and the efficiency of such coatings as hydrogen diffusion barriers. According to this, A as a barrier to hydrogen diffusion.
The T-8i delinquent is said to be the most promising. Furthermore, W, O, 5c
[Electroplating of zirconium and zirconium-tin (Klectroplating)]
ting on Zirconium and Zir
conium-Tin) J [BMI-757,
Technical Information YARVIS
tion 5ervici), 1952] discloses a method of electroplating nickel on zirconium and zirconium-tin alloys, followed by heat treatment to form an alloy diffusion bonding layer.

In the fuel element for a nuclear reactor described in U.S. Pat. In addition, the Zolokram Progress Report on Reactor Development dated August 1973 (Reactor Dev
elopment ProgramProgress
Report ) j (AIJL-RDP-19)
Therein, it is disclosed to place a chromium layer on the inner surface of a stainless steel cladding to be consumed as a chemical getter.

Another method was to introduce a barrier between the nuclear fuel material and the cladding containing the fuel.

Examples of such methods are U.S. Pat.
, U.S. Pat. No. 3,212,988 (zirconium, aluminum or beryllium sheath), U.S. Pat. No. 50,182
No. 38 (Crystalline Carbon Barrier Disposed Between UO3 and Zirconium Alloy Coating Phase) and U.S. Patent No. 308Q8
No. 93 (Stainless Steel Foil). Although the use of septa has proven successful, some of the materials described in the above-mentioned documents have problems with their compatibility. i.e. nuclear fuel (e.g. carbon can combine with oxygen produced from the nuclear fuel), cladding (e.g. copper and other metals can react with the cladding and change its properties) or fission reactions (e.g. copper and other metals can react with the cladding and change its properties). For example, some materials are incompatible with neutron absorbers (7).

Furthermore, none of the above documents discloses a solution to the problem of local chemical-mechanical interaction between nuclear fuel and cladding material.

Other methods for using bulkheads are described in U.S. Pat. No. 396
No. 9186 (molybdenum, tungsten, rhenium, -
U.S. Pat. A method is disclosed in which a liner made of an alloy is disposed between a nuclear fuel and a cladding material, and a coating of a highly slippery material is disposed between the liner and the cladding material.

U.S. Pat. No. 4,045,288 discloses a zirconium alloy substrate comprising a zirconium alloy substrate, a metal partition metallurgically bonded to such substrate, and an inner layer of a zirconium alloy metallurgically bonded to such metal partition. A composite dressing is disclosed. The material for the bulkhead is selected from niobium, anodized minilum, copper-nickel, stainless steel, and iron. These ready-installed metal bulkheads reduce corrosion by fission products and corrosive gases. On the other hand, stress corrosion cracking and liquid metal embrittlement are likely to occur.

No. 4,200,492 discloses a composite coating consisting of a zirconium alloy substrate and a cancellous zirconium liner.

Such soft zirconium liners suppress local strain and reduce stress corrosion cracking and liquid metal embrittlement;
On the other hand, it is susceptible to losses due to honing during manufacturing and oxidation. Additionally, zirconium liners exhibit a tendency to rapidly oxidize when cracks form in the cladding, allowing water and/or steam to enter the fuel rod.

Therefore, it remains desirable to develop fuel elements that solve the above-mentioned problems.

SUMMARY OF THE INVENTION In accordance with the present invention, a fuel element that can be used particularly effectively in the core of a nuclear reactor comprises a composite jacketed vessel having a dilute zirconium alloy liner metallurgically bonded to the interior surface of the substrate. It is like containing. Such dilute zirconium alloys consist of zirconium and a metal selected from iron, chromium, iron-decachromium, and copper, and the amount of iron alloyed with the zirconium is from about 0.2 to about 0.03 (by weight). %, the amount of chromium is about 0.05 to about 0.6 (
(weight)%, the total amount of iron chromium is about 0.015% to about 0.3% (by weight), and the weight ratio of iron to chromium is about 1:1 to about 4:1.
, and the amount of copper is about 0.02 to about 02% (by weight).

The base of such a composite cladding vessel is comprised of conventional cladding materials for nuclear reactors;
There is no difference in design and function compared to , and therefore it may be made of a conventional material such as a zirconium alloy. When comprised of a zirconium alloy, the substrate of the composite coated vessel has a higher alloy content than the dilute zirconium alloy liner. Such a dilute zirconium alloy liner forms a continuous barrier between the liquid in the composite cladding vessel and the nuclear fuel material contained therein, and also protects the substrate of zirconium alloy or other material from fission products and gases. Helps cover.

The dilute zirconium alloy liner described above accounts for about 1 to about 20% of the thickness of the composite coated vessel. Such liners remain soft (relative to the substrate) during irradiation to reduce localized stresses within the fuel element, thereby helping to prevent stress corrosion cracking and liquid metal embrittlement of the substrate. The dilute zirconium alloy also serves to protect the substrate from reaction with volatile impurities and fission products present within the fuel element, thereby providing overall protection of the substrate from the effects of volatile impurities and fission products. j9,.

In this way, the present invention provides that the dilute zirconium alloy liner protects the base of the composite coated vessel from contact with fission products, corrosive gases, etc., and also prevents stress corrosion cracking and liquid metal embrittlement of the base. It has this remarkable advantage. However, such liners are unlikely to introduce problems of neutron capture failure, heat transfer failure, or incompatibility between the nuclear fuel and the liner. Moreover, even if cracks occur in the composite coated container, such liners exhibit superior water vapor or hot water acid resistance compared to unalloyed zirconium.

These and other objects of the invention will become apparent to those skilled in the art upon reading the following detailed description in conjunction with the accompanying drawings.

DESCRIPTION OF THE INVENTION Turning first to FIG. 1, a partially cutaway cross-sectional view of a fuel assembly 10 is shown. Such a fuel assembly 10 includes a tubular channel box 11 of approximately square cross section with a lifting handle 12 at the upper end and a nosepiece (not shown since the lower part of the fuel assembly 10 has been omitted) at the lower end. .

The upper end of the channel box 11 is open at the outlet 13, and the lower end of the nosepiece is provided with an opening for coolant flow. A 1W fuel element or fuel rod 14 is housed within the channel box 11 and is supported by an upper tie plate 15 and a lower tie plate (the lower part is omitted and is not shown). usually,
Liquid coolant enters through an opening at the lower end of the nosepiece, passes upwardly around the fuel element 14, and then exits at a high temperature through an outlet 16 at the top. The coolant that flows out in this manner is partially vaporized in a boiling type nuclear reactor, while it is not vaporized in a pressurized type nuclear reactor.

The ends of the fuel rods 14 are hermetically sealed by end plugs 18 welded to the cladding tube 17, and the end plugs 18 are provided with sump struts 19 to facilitate installation of the fuel rods into the fuel assembly. may be established. One end of the fuel element has a
a space or plenum 2 to allow longitudinal expansion of the nuclear fuel material and accumulation of gases released from the nuclear fuel material;
0 is set. Inside the space 20, a nuclear fuel material holding means 24 consisting of a helical member is arranged, in order to suppress the axial movement of the pellet column, especially during handling and transport of the fuel elements.

A fuel element such as the one described above achieves excellent thermal contact between the cladding and the nuclear fuel material, minimizes parasitic neutron absorption, and avoids vibrations and curvature occasionally caused by coolant flow at low velocities. Designed to provide resistance to

A fuel element or rod 14 made in accordance with the principles of the present invention is shown in partial cross-section in FIG. Such a fuel element consists of a core or central cylindrical portion of nuclear fuel material (in this case a plurality of fuel pellets of fissionable material and/or fuel parent material) 16 disposed inside a cladding tube or cladding vessel 17. Contains. Depending on the case, fuel pellets of various shapes, such as cylindrical or spherical, may be used, and other forms of fuel may be used, such as granular fuel. Note that the physical form of the fuel is not important to the present invention.

In addition, various types of tSI, including uranium compounds, plutonium compounds, trilam compounds, and mixtures thereof,
8v=1 substance can be used. The preferred fuel is uranium dioxide or a mixture of uranium dioxide and plutonium dioxide.

Next, looking at FIG. 2, the nuclear fuel material 16 constituting the central core of the fuel element 14 is surrounded by a test tube 17, which cladding tube 17 is referred to as a composite cladding vessel in the present invention. . When a fissile core is housed inside the composite coated container, a gap 23 is formed between the core and the container when used in a nuclear reactor. Such a composite coated container consists of a substrate 2 of conventional materials such as stainless steel or zirconium alloys! It has on the outside. In a preferred embodiment of the invention, the substrate is comprised of a zirconium alloy such as Zircaloy-2.

A dilute zirconium alloy liner 22 is provided on the inner surface of the base body 21.
are metallurgically bonded to form a bulkhead that protects the substrate from the nuclear fuel material 16 within the composite cladding vessel. Preferably, the dilute zirconium alloy liner comprises from about 1% to about 20% of the thickness of the composite coated vessel. A dilute zirconium alloy liner that accounts for less than about 1% of the thickness of the composite clad vessel is difficult to achieve in commercial production, and even with a dilute zirconium alloy liner that accounts for more than 20% of the thickness of the composite clad vessel, the thickness However, it is not possible to obtain salt benefits equivalent to an increase in . On the contrary, as the liner exceeds about 20% of the thickness of the composite coated container, the thickness of the substrate decreases, which may reduce the strength of the composite coated container.

The dilute zirconium alloy described above is comprised of zirconium and an alloy additive selected from iron, chromium, iron-chromium, and copper. It should be noted that the term "lean zirconium alloy" as used herein refers to a zirconium alloy that has a sufficiently low alloy content to exhibit greater ductility and strain rate than Substrate 1 under equivalent stress conditions. means.

The amount of iron alloyed with zirconium ranges from about 02 to about 0,000
S% (by weight) preferably from about 0.2 to about 0.25% (by weight).

The amount of chromium used ranges from about 0.05% to about 0.3% (by weight), preferably from about 0.15% to about 0.25% (by weight).

When using iron chromium, the total amount of both components should be approximately 0.1
5 to about 03% (by weight), preferably about 02 to about 025% (electricity storage), and the weight ratio of iron to chromium is about 1 = 1 to
The ratio may be about 4:1, preferably about 2:1.

The amount of copper used is about 1.02% to about 0.2% (by weight), preferably about 0.05% to about 0.15% (by weight).

Such a dilute zirconium alloy liner shields the substrate from gaseous impurities and fission products, protects the substrate portion of the composite envelope from contact and reaction with such impurities and fission products, and protects the substrate from localized stress. Helps prevent occurrence.

The addition of small amounts of metals selected from iron, chromium, iron-chromium, and copper to zirconium improves its corrosion resistance, especially against oxidation by hot water or steam, as long as the amount added is within the specified range for each metal. Helps improve resistance to The lower limit on the amount of each metal alloyed with zirconium is the amount of that metal sufficient to significantly improve corrosion resistance compared to unalloyed zirconium.

The upper limit for the amount of each metal alloyed with zirconium is generally set as the maximum amount of that metal that significantly improves corrosion resistance compared to cancellous zirconium. Addition of metal in amounts exceeding such upper limits not only fails to significantly improve the corrosion resistance of zirconium, but may also be detrimental in reducing the softness and ductility of the liner.

The amounts of each metal to be added to zirconium to achieve the greatest improvement in corrosion resistance are stated as preferred ranges.

Iron, chromium and copper are leached into zirconium. 1
By heat-treating a dilute zirconium alloy containing at least one such metal, it is possible to obtain a material in which fine metal particles are dispersed inside due to the inertness of the zirconium matrix. Since the alloying components are dissolvable, solid solution strengthening of zirconium almost never occurs. This reinforcement effect is sufficient to maintain the softness required for the lean zirconium alloy liner to prevent or reduce fuel element failure due to pellet-to-cladding interaction.

The diluted zirconium alloy liner in the composite coated container resists radiation hardening compared to Zircaloy and other conventional zirconium alloys. Therefore, even after long-term irradiation, dilute zirconium alloy liners are able to maintain desirable structural properties such as yield strength and hardness at significantly lower levels than with conventional zirconium alloys. In fact, when exposed to irradiation, dilute zirconium alloy litchi does not harden as much as regular zirconium alloys. As a result, the diluted zirconium alloy liner, which originally has a low yield strength, exhibits plastic deformation, which can be applied at, for example, the operating temperature of a nuclear reactor (300-5
Since the nuclear fuel pellet expands and comes into contact with the target material at a temperature of 50° C., it is possible to alleviate pellet-induced stress that may occur in the fuel element.

It is made of zirconium and a metal selected from iron, chromium, iron-decachromium, and copper, and is bonded to a common zirconium alloy substrate, and preferably about 5 to 15% of the thickness of the composite coated container. A dilute zirconium alloy liner, such as one that accounts for 10% of the zirconium alloy, provides sufficient stress relief to prevent or reduce failure of the composite coated vessel.

The purity of the zirconium metal that is alloyed with iron, chromium, iron-chromium, or copper is important and contributes to imparting special properties to the dilute zirconium alloy liner.

Generally, less than 00 ppm of impurities are present in zirconium metal.

Among them, oxygen should be as small as possible, but about 1
000 ppm is acceptable.

The composite cladding vessel of the fuel element of the present invention includes a dilute zirconium alloy liner metallurgically bonded to a substrate. Metallographic examination indicates that there is sufficient interdiffusion between the substrate and the diluted zirconium alloy liner to produce a nationwide bond, but not enough interdiffusion to result in significant alloying with the diluted zirconium alloy liner itself. It can be seen that there is no diffusion.

Among the common zirconium alloys that may be used as suitable substrates are Zircaloy-2 and Zircaloy-4. Zircaloy-2 is approximately 1.5% (by weight) of tin, 012
(by weight)% iron, 009 (1 acid)% chromium, and 0
It contains 0.005% (heavy loading) nickel and is widely used in water-cooled nuclear reactors. Zircaloy-4 has a lower nickel content than Zircaloy-2, but a slightly higher iron content. The composite coated container used in the fuel element of the present invention can be manufactured by any of the following methods.

According to one method, a tube of dilute zirconium alloy litchi material is inserted into a hollow billet of the material selected for the substrate. The tube is then joined to the viresol by subjecting such assembly to an explosive joining operation. The composite thus obtained is subjected to a conventional tube extrusion operation at an elevated temperature of about 1000-1400° C. (about 538-760° C.). The extruded composite is then formed into a cladding tube of the desired size by a process that includes conventional diameter reduction operations. In this case, the relative wall thicknesses of the hollow billet and the dilute zirconium alloy liner tube are selected to provide the desired thickness ratio in the finished cladding tube.

According to another method, a tube of dilute zirconium alloy liner material is inserted into a hollow billet of material 1 selected for the substrate. The assembly is then subjected to compressive stress (e.g. 750°C) resulting in good metal-to-metal contact.
Diffusion bonding of the tube and billet is achieved by applying a heating operation (for 8 hours). The composite obtained by such diffusion bonding is subjected to a conventional tube extrusion operation as described above. The extruded composite is then formed into a cladding tube of the desired size by a process that includes conventional diameter reduction operations.

According to yet another method, a tube of dilute zirconium alloy liner material is inserted into a hollow billet of the material selected for the substrate. The assembly is then subjected to conventional pre-extrusion operations as described above. The extruded composite is then formed into a cladding tube in the desired manner by a process that includes conventional diameter reduction operations.

The method of manufacturing the composite cladding of the present invention is significantly more economical than other methods used to manufacture cladding (eg, electroplating or vapor deposition).

In manufacturing the fuel element, first a composite cladding container with one end open is created. Such a composite coated vessel consists of a base body and an inner dilute zirconium alloy liner, the latter consisting of zirconium and a metal selected from iron, chromium, ferrochromium, and copper, and metallurgically bonded. The composite cladding container is then filled with nuclear fuel material, leaving a space at the open end. After inserting the nuclear fuel material retaining means into the space, an end plug is installed at the open end of the composite cladding vessel, leaving the space in communication with the nuclear fuel material.

Finally, the ends of the composite coated container are sealed by joining them to the end plugs.

According to the present invention, several advantages are achieved that result in an extension of the service life of the fuel element. These benefits include reduced chemical interactions in the composite coated vessel, relaxation of localized stresses on the zirconium alloy base portion of the composite coated vessel;
and a reduced likelihood of failure of the zirconium alloy substrate.

The dilute zirconium alloy liner of the present invention not only alleviates stress and stress corrosion on the surrounding body, but also resists steam and hot water oxidation even if the substrate is cracked. In contrast, unalloyed zirconium oxidizes rapidly under these conditions. On the other hand, the dilute zirconium alloy liner of the present invention exhibits plasticity comparable to unalloyed zirconium. These advantages are obtained along with increased corrosion resistance, especially resistance to oxidation by hot water or steam.

One of the important properties of the composite coated vessel of the present invention is that the various improvements described above are achieved without introducing substantial neutron capture obstacles. Such a composite coated vessel does not produce eutectic alloys even in the event of a loss of coolant accident or a control rod fall accident, and therefore can be easily accommodated in a nuclear reactor. Moreover, such composite coated containers present significantly less heat transfer impairment. This is because there is no barrier to heat transfer, unlike when a separate foil or liner is inserted into the fuel element. The composite coated containers of the present invention can also be tested by conventional non-destructive testing methods at various stages of manufacture and operation.

As will be apparent to those skilled in the art, various changes and modifications may be made to the embodiments of the invention described herein. Therefore, it goes without saying that the scope of the present invention should be interpreted in the broadest sense according to the scope of the claims.

4. BRIEF DESCRIPTION OF THE DRAWINGS FIG. 1 is a partially cut away sectional view of a fuel assembly containing fuel elements made in accordance with the principles of the invention; FIG. 2 is a detailed description of the invention. FIG. 2 is an enlarged cross-sectional view of the fuel element of FIG. 1 serving as a fuel element;

In the figure, 10 is a fuel assembly, 11 is a channel box, 1
2 is a handle, 14 is a fuel element or fuel rod, 15 is an upper tie plate, 16 is a core body, 17 is a composite jacket container, 18 is an end plug, 19 is a strut, 20 is a plenum, 21 is a base body, 22
represents the dilute zirconium alloy liner and 23 represents the air gap.

Fig, 1 Fig, ;)

Claims (9)

[Claims] (1) (a) an outer part made of a zirconium alloy forming a substrate; and (b) a metallurgical component consisting of zirconium and a metal selected from among iron, chromium, ferrochromium, and copper, and metallurgical to the inner surface of said substrate. a diluted zirconium alloy liner and a diluted zirconium alloy liner that are bonded together, the diluted zirconium alloy liner accounting for about 5 to about 15% of the thickness of the resulting composite cladding vessel. Composite coated container. 2. The composite coated container of claim 1, wherein said dilute zirconium alloy liner comprises about 0.2 to about 0.3 percent (by weight) iron and the balance zirconium. (3) The dilute zirconium alloy liner has a thickness of about 0.2
A composite coated container according to claim 1, comprising ~0.2% (by weight) iron and the balance zirconium. 4. The composite coated container of claim 1, wherein said dilute zirconium alloy liner comprises about 0.005% to about 3% (by weight) chromium and the balance zirconium. (5) The dilute zirconium alloy liner is about 0.015
A composite coated container according to claim 1, comprising from about (125% (by weight)) chromium and the balance zirconium. (6) The dilute zirconium alloy liner comprises a total of about α15 to about 0.3% (by weight) of ten chromium iron and the balance zirconium, and the weight ratio of the iron to the chromium is about 1 to about 1. Claim 1 where approximately 4=1
Composite coated container as described in section. (7) The dilute zirconium alloy liner is approximately α in total.
2 to about 0.25% (by weight) of iron chromium and the balance zirconium, wherein the weight ratio of said iron to said chromium is from about 1:1 to about 4=1. Composite coated container. 8. The composite coated container of claim 1, wherein said dilute zirconium alloy liner comprises from about 0.02% to about 0.2% (by weight) copper and the balance zirconium. (9) The dilute zirconium alloy lychee is about 0.005
A composite coated container according to claim 1, comprising ~0.1 s (by weight) percent copper and the balance zirconium.