JPH0160797B2 - - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION Field of the Invention The present invention relates generally to improvements in nuclear fuel elements for use in nuclear fission reactor cores, and more particularly, includes a zirconium alloy substrate, an unalloyed zirconium barrier metallurgically bonded to the interior surface of the substrate, and a zirconium An improved nuclear fuel element having a composite cladding vessel comprising a zirconium alloy inner layer metallurgically bonded to a barrier.

BACKGROUND OF THE INVENTION In nuclear reactors currently designed, constructed, and in operation, nuclear fuel is contained in fuel elements having various geometries, such as plates, tubes, or rods. The fuel material is normally contained in a corrosion-resistant, non-reactive, thermally conductive container or container. A plurality of fuel elements are assembled in a grid at regular intervals from each other within a coolant flow region, or channel, to form a fuel assembly, and a sufficient number of fuel assemblies are assembled to form a fission chain reaction assembly, or Construct a reactor core capable of self-sustaining nuclear fission reactions.

The reactor core is further housed within a reactor vessel, with coolant flowing through the reactor vessel.

Kratzding serves several purposes. The two main objectives are:
The second purpose is to prevent contact and chemical reactions between the nuclear fuel and the coolant or moderator, if present, or both the coolant and moderator, if both are present. The goal is to prevent radioactive fission products (some of which are gases) from escaping from the fuel into the coolant or moderator, or both coolant and moderator if both are present. Common cladding materials include stainless steel, aluminum and its alloys, zirconium and its alloys, niobium (columbium), and certain magnesium alloys.
If the cladding fails, ie, loses leak protection, the coolant or moderator and related systems can become contaminated with radioactive long-lived products to the extent that plant operations are disrupted.

In the manufacture and use of nuclear fuel elements using certain metals and alloys as cladding materials,
Several problems arise from the mechanical or chemical reaction of these cladding materials under certain circumstances. Zirconium and its alloys are excellent nuclear fuel cladding materials under normal conditions. Zirconium and its alloys have a small neutron absorption cross section and are strong, ductile, and durable at temperatures below about 398°C (about 750°C), even in the presence of demineralized water or steam, commonly used as reactor coolants and moderators.
Very stable and non-reactive.

However, with the use of fuel elements, the problem of brittle cracking of the cratzding became apparent due to the intertwined interaction between the nuclear fuel, the cratzding, and the fission products produced during the fission reaction. It has been determined that this undesirable phenomenon is promoted by local mechanical stresses due to differential expansion between the fuel and the cladding (stresses during cladding are localized at the fuel pellet interface and sometimes at cracks in the nuclear fuel). This phenomenon is defined by the term "pellet-cracking interaction" (PCI). Corrosive fission products are released from the nuclear fuel and are present at the intersection of the fuel pellet interface and the cladding surface. Such fission products are formed in nuclear fuel during the nuclear fission chain reaction during operation of a nuclear reactor.

Within the sealed fuel element, hydrogen gas is generated by a slow reaction between the cladding and residual moisture within it. When this hydrogen gas accumulates to a certain extent, the cladding is locally hydrogenated under certain conditions, and at the same time, the mechanical properties of the cladding are locally degraded. Cladding is also adversely affected by gases such as oxygen, nitrogen, carbon monoxide and carbon dioxide over a wide temperature range. The zirconium chloride ink of nuclear fuel elements is exposed to one or more of the above gases and fission products during irradiation in a nuclear reactor, which means that these gases are not present in the reactor coolant or moderator. , and occur even though they are excluded as much as possible from the ambient atmosphere during cladding and fuel element manufacturing. Sintered refractory and ceramic compositions, such as uranium dioxide and other compositions used as nuclear fuels, release measurable amounts of the above gases when heated, e.g. during fuel element manufacture, and also release fission products during irradiation. discharge. Particulate refractory and ceramic compositions, such as uranium dioxide powder and other powders used as nuclear fuels, are known to release even greater amounts of these gases during irradiation. These released gases can react with the zirconium cladding containing the nuclear fuel.

Therefore, in view of the above, during each period during which a fuel element is used in the operation of a nuclear power plant,
Reacts with Kratzding inside the fuel element,
It can be seen that it is desirable to minimize crazing attacks from water, water vapor and other gases, especially hydrogen. One such solution is to find materials that can rapidly chemically react with water, water vapor, and other gases to remove them from the interior of the cladding. Such substances are called getters.

Another solution is coating, as disclosed in US Pat. No. 3,108,936, in which the nuclear fuel material is coated with ceramic to prevent moisture from coming into contact with the nuclear fuel material. In the fuel element described in U.S. Pat. No. 3,085,059, a metal casing contains one or more pellets of fissionable ceramic material, a layer of vitreous material is bonded to the ceramic pellet, and the vitreous layer is bonded to the casing. and the nuclear fuel to maintain uniform and good heat conduction from the pellets to the casing. US Patent No. 2873238
No. 3, a jacketed fissionable uranium slug is placed in a metal case, where the protective jacket or cover for the slug is a zirconium-aluminum bonding layer. The jacketed fissionable body disclosed in U.S. Pat. No. 2,849,387 consists of a plurality of open-ended jacketed body sections of nuclear fuel that have been previously immersed in a molten bath of a bonding material and that are combined with a uranium body section. Forms an efficient thermally conductive bond with the container (or cladding). Coatings are disclosed as metal alloys with good thermal conductivity properties, examples of which include aluminum-silicon and zinc-aluminum alloys. Special Publication No. 47-46559 (1972
Published on November 24, 2017) discloses the consolidation of individual nuclear fuel particles into a carbon-containing matrix fuel composite by coating the individual nuclear fuel particles with a dense, smooth carbon-containing coating. There is. Yet another coating is disclosed in Japanese Patent Publication No. 47-14200, in which one of the two groups of pellets is coated with a layer of silicon carbide and the other group of pellets is coated with a layer of pyrolytic carbon or metal carbide. covered with a layer.

Coating nuclear fuel material presents reliability problems in the sense that it is difficult to obtain a uniform coating free of defects. Additionally, coating degradation is also a problem in connection with the long-life performance of nuclear fuel materials.

U.S. Patent Application No. 330,152, filed February 6, 1973, discloses a method for preventing corrosion of nuclear fuel cladding by adding metals such as niobium to the nuclear fuel. The additive can be in powder form, provided that subsequent fuel processing does not oxidize the metal, or it can also be incorporated into the fuel element as a wire, sheet or other form in, around or between the fuel pellets. can be introduced into

Publication GEAP-4555 (February 1964) discloses a composite cladding of a zirconium alloy with a stainless steel inner lining metallurgically bonded thereto; It is manufactured by extruding a hollow billet of. This cladding has the drawback that the stainless steel develops a brittle phase and the stainless steel layer has a neutron absorption penalty of about 10 to 15 times that of a zirconium alloy layer of the same thickness.

No. 3,502,549 discloses a method of protecting zirconium and its alloys by chroming them to form composite materials useful in nuclear reactors. Energia Nucleare is a method in which the surface of Zircaloy-2 is plated with copper and then heat treated to diffuse the plated metal onto the surface.
Vol. 11, No. 9 (September 1964), pages 505-508. The paper “Stability and Compatibility of Hydrogen Barriers Deposited on Zirconium Alloys” by Brossa et al.
Barriers Applied to Zirconium Alloys)
(Europe Atomic Energy Community, Joint
Nuclear Research Center, EUR4098e, 1969)
describes the method of depositing various coatings and their effectiveness as hydrogen diffusion barriers,
-Si coatings have been shown to be a very promising barrier to hydrogen diffusion. A method for plating zirconium and zirconium-tin alloys with nickel and heat treating these alloys to create alloy-diffusion bonds is described in the paper by WC Schnicker et al.
−Tin)” (BMI−757, Technical Information
Service, 1952). US Patent No.
In the nuclear reactor fuel element having a double cladding tube disclosed in No. 3625821, the inner surface of the cladding tube is made of a metal having a small neutron capture cross section;
For example, it is coated with nickel and embedded with dispersed particles of burnable poison. Reactor Development
Program Process Report (August 1973, ANL-
RDP-19) discloses a chemical getter structure in which a sacrificial layer of chromium is provided on the inner surface of a stainless steel cladding.

Yet another solution is to interpose a barrier between the nuclear fuel material and the holding material; these methods are described in US Pat.
U.S. Pat. No. 3,212,988 (zirconium, aluminum or beryllium sheath), U.S. Pat.
3018238 (crystalline carbon barrier between UO ₂ and zirconium cladding), and U.S. Pat.
No. 3088893 (stainless steel foil).
Although the idea of creating a barrier has shown promise, some of these precedents involve materials that are incompatible with nuclear fuel (e.g., carbon combines with oxygen from nuclear fuel), and some with materials that are incompatible with crazing (e.g., carbon combines with oxygen from nuclear fuel). For example, some use materials that are incompatible with fission reactions (e.g., act as neutron absorbers). None of the above-mentioned patent publications discloses a solution to the recently discovered problem of local chemical-mechanical interaction between nuclear fuel and crazing.

Still other approaches to barrier thinking are disclosed in U.S. Pat.
Tungsten, rhenium, niobium and their alloys in the form of single or multi-layered tubes or foils or as coatings are provided on the inner surface of the cladding, in the latter case a liner of zirconium, niobium or their alloys is provided between the nuclear fuel and the cladding. , a coating of highly lubricious material is provided between the liner and the cladding.

No. 4,045,288 discloses a composite cladding consisting of a zirconium alloy substrate, a metal barrier metallurgically bonded to the substrate, and an inner layer of zirconium alloy metallurgically bonded to the metal barrier. This barrier is selected from niobium, aluminum, steel, nickel, stainless steel and iron. With the exception of the niobium barrier, all other materials form low melting point eutectic layers with the zirconium alloy substrate, which is undesirable in the event of a potential loss of coolant accident.

No. 4,200,492 discloses a composite cladding consisting of a zirconium alloy substrate and an unalloyed zirconium liner. The soft zirconium liner minimizes local strain and inhibits stress corrosion cracking and liquid metal embrittlement, but is susceptible to corrosion during manufacturing if it is damaged or degraded due to honing, or if the crazing ruptures.

Accordingly, it remains desirable to develop nuclear fuel elements that minimize the problems described above.

SUMMARY OF THE INVENTION A nuclear fuel element particularly useful for use in the core of a nuclear reactor comprises a composite cladding comprising a substrate, an unalloyed zirconium barrier metallurgically bonded to the interior surface of the substrate, and an inner layer metallurgically bonded to the interior surface of the zirconium barrier. equip This crazing substrate is not unlike that used in conventional nuclear reactors in terms of its design and function, and is made of ordinary crazing materials.
For example, it is selected from zirconium alloys.

The zirconium barrier and inner layer form a shield between the substrate and the nuclear fuel material held within the cladding and shield the substrate from fission products and gases. The inner layer also shields the zirconium barrier from other reactive elements present within the fuel element by fission products released from the fuel. This shielding action allows the zirconium barrier to maintain maximum purity and ductility by preventing hardening by recoil fission products or by reaction with chemical elements present in the fuel element. .

The zirconium barrier accounts for approximately 1-30% of the thickness of the cladding. Approximately 1 of the thickness of the Kratzding
Zirconium barriers less than 30% of the thickness of the cladding are difficult to produce industrially, and making the zirconium barrier thicker than about 30% of the thickness of the cladding does not provide additional benefits commensurate with the increased thickness. Additionally, increasing the barrier cladding thickness by more than about 30% reduces the substrate thickness correspondingly and weakens the composite cladding.

The inner layer is preferably formed to have a thickness of 1 to 10% of the total thickness of the cladding. This thickness range was specified in order to obtain the smallest inner layer thickness that could be produced by co-extrusion and co-diameter reduction techniques for tube forming. Based on the purity of the barrier and the shielding effect of the inner layer,
The barrier remains soft during irradiation, minimizing local strain within the nuclear fuel element and thus protecting the substrate from stress corrosion cracking and liquid metal embrittlement. The inner layer and the zirconium barrier provide preferential reaction sites to react with volatile impurities or fission products present inside the nuclear fuel element, thus protecting the barrier and the zirconium barrier from attack by volatile impurities or fission products. act.

In addition, the inner layer is effective in preventing the soft barrier from deteriorating or being damaged during crazing manufacturing.
Therefore, processability is improved. Furthermore, the inner layer protects the barrier from aqueous corrosion in case of fuel element failure.

The present invention provides that the crazing substrate and barrier are not only protected from contact with fission products, corrosive gases, etc. by the inner layer, but are also protected from stress corrosion cracking and liquid metal embrittlement, and that the inner layer itself There are no significant neutron capture penalties, heat transfer penalties, or material incompatibility problems.

OBJECTS OF THE INVENTION It is an object of the present invention to provide a nuclear fuel element that can be used in nuclear reactors for extended periods of time without cladding cracking, cladding corrosion or other fuel failure problems.

Another object of the invention is to provide a composite cladding comprising a substrate, a zirconium barrier metallurgically bonded to the inner surface of the substrate, and an inner layer metallurgically bonded to the inner surface of the zirconium barrier, the metallurgical bond providing a barrier between the substrate and the zirconium barrier. It is also an object to provide a nuclear fuel element having a permanent connection between the zirconium barrier and the inner layer.

The above and other objects of the present invention will be easily understood by those skilled in the art after reading the following description with reference to the accompanying drawings.

DETAILED DESCRIPTION OF THE INVENTION FIG. 1 shows a fuel assembly 10 as a partially cutaway side view. The fuel assembly 10 comprises a tubular flow channel 11 of approximately square cross-section, with a lifting bail 12 at the upper end of the channel and a nosepiece (not shown since the lower part of the fuel assembly 10 has been omitted) at the lower end. ) is provided. The upper end 13 of the channel 11 is open,
Coolant flow openings are drilled in the lower end of the nosepiece. An array of fuel elements or fuel rods 14 is housed within the channel 11 and supported therein by an upper plate 15 and a lower plate (not shown as the lower part has been omitted). Typically, liquid coolant enters through an opening in the lower end of the nosepiece, flows upwardly around the fuel element 14, and exits through the upper outlet 13 at a high temperature and, in the case of a boiling reactor, in a partially vaporized state. , and in the case of pressure reactors, it leaves the reactor unevaporated.

Nuclear fuel elements or rods 14 have end plugs 1 welded to cladding 17 at both ends thereof.
8 and the end plug 18 is provided with a stud 19 to facilitate assembly of the fuel rods into an assembly. A space or plenum 20 is provided at one end of the fuel element 14 to permit longitudinal expansion of the fuel material and to store gases released from the fuel material. The fuel material retention means 24 in the form of a helical member are arranged within the space 20 to restrain the pellet column against axial movement, particularly during handling and transportation of the fuel elements.

The fuel element is designed to have good thermal contact between the cladding and the fuel material, minimal parasitic neutron absorption, and to withstand warping and vibrations occasionally caused by the high velocity flow of coolant.

Nuclear fuel element or rod constructed according to the invention 1
4 is shown in cross section in FIG. The fuel element comprises a core or central cylindrical portion of fuel material 16, shown here as a plurality of fuel pellets of fissile material and/or parent material, disposed within a structural cladding or vessel 17. In some cases, the fuel pellets can be of various shapes, such as cylindrical or spherical pellets, and in other cases, different fuel forms can be used, such as particulate fuel. The physical form of the fuel is not important to the invention. A variety of nuclear fuel materials can be used including uranium compounds, plutonium compounds, thorium compounds and mixtures thereof. The preferred fuel is uranium dioxide or a mixture of uranium dioxide and plutonium dioxide.

Referring to FIG. 2, the nuclear fuel material 16 forming the central core of the fuel element 14 is surrounded by a cladding 17. Cladding 17 is also referred to as composite cladding in the present invention.
A composite cladding vessel surrounds the fissile core to leave a gap 26 between the core and the cladding vessel during operation in a nuclear reactor. The composite cladding has an outer substrate 21 selected for conventional fuel cladding materials, and in the preferred embodiment of the invention the substrate is a zirconium alloy, such as Zircaloy-2 or Zircaloy-4.

A non-alloyed zirconium barrier 22 is metallurgically bonded to the inner circumferential surface of the substrate 21, so that the zirconium barrier 2
2 forms a shield that shields the base body 21 from the nuclear fuel material 16 inside the composite cladding. Preferably, the zirconium barrier 22 comprises about 1-30% of the thickness of the composite cladding.

An inner layer 23 is metallurgically bonded to the inner peripheral surface of the zirconium barrier 22, so that the inner layer 23 constitutes the closest portion of the composite cladding to the nuclear fuel material 16. The inner layer 23 is approximately 1 to 10 times thicker than the Kratzding.
% and is preferably made of a conventional cladding material; in a preferred embodiment of the invention, the inner layer is a zirconium alloy, such as Zircaloy-2 or Zircaloy-4. The zirconium barrier acts as a reaction site to react with gaseous impurities and fission products that have passed through cracks and defects in the inner layer 23, and protects the base portion of the cladding from contact with and reactions with such impurities and fission products. It also minimizes the occurrence of local stress and cracking damage due to pellet-cludding interaction (PCI).

In an embodiment of the invention, the zirconium barrier layer is about 3 mils thick and the Zircaloy-2 inner layer is about 1 mil thick. Both the inner layer and the barrier layer must be continuous layers. That is, the layer must be free of holes and seams.

The composite cladding of nuclear fuel elements of the present invention comprises:
It has an unalloyed zirconium barrier metallurgically bonded to the substrate and an inner layer metallurgically bonded to the zirconium barrier. Metallographic examination shows that the cross-diffusion occurring between the substrate and the zirconium barrier and between the zirconium barrier and the inner layer is sufficient to form a metallurgical bond, but not so much as to significantly reduce the purity of the zirconium barrier itself. I confirmed it. As can also be seen in FIG. 2, the zirconium barrier can be referred to as a "buried" zirconium barrier. This is because the zirconium barrier is sandwiched between the substrate and the inner layer.

Unalloyed zirconium, which forms the barrier in composite cladding, is highly resistant to radiation hardening;
This property allows zirconium barriers to maintain desirable structural properties, such as yield strength and hardness, at levels significantly lower than conventional zirconium alloys even after long-term irradiation. In fact, the zirconium barrier does not harden as significantly as normal zirconium alloys during irradiation, and this, combined with its initially low yield strength, causes the zirconium barrier to plastically deform during power transients and absorb pellet-induced stresses within the fuel element. be able to. Pellet-induced stresses within the fuel element are caused, for example, by thermal expansion and/or swelling of the nuclear fuel pellets through reactor operating temperatures (300-350 DEG C.) and thus contact of the pellets with the cladding.

Furthermore, failure of the composite cladding is prevented by a zirconium barrier bonded to the zirconium alloy external substrate, preferably on the order of about 5-15% of the thickness of the cladding, particularly preferably 10% of the thickness of the cladding. A sufficient barrier effect and stress relief is achieved.

A preferred example of the principle of the present invention is "hypoxic sponge"
grade zirconium is used as the buried barrier layer, although higher purity "crystal rod zirconium" grade and lower purity "reactor grade sponge" zirconium can also be used. Residual impurities in the sponge zirconium serve to impart special properties to the zirconium barrier. Generally, about 1000ppm for sponge zirconium
More than about 5000 ppm of impurities, preferably
Contains less than 4200ppm of impurities. Oxygen about 200
Preferably it is maintained within the range of ~1200 ppm. Other typical impurity levels are shown below. Aluminum 75ppm or less, Boron 0.4ppm or less, Cadmium 0.4ppm or less, Carbon 270ppm or less, Chromium
200ppm or less, cobalt 20ppm or less, copper 50ppm or less, hafnium 100ppm or less, hydrogen 25ppm or less,
Iron 1500ppm or less, magnesium 20ppm or less, manganese 50ppm or less, molybdenum 50ppm or less, nickel 70ppm or less, niobium 100ppm or less, nitrogen 80ppm
Below, 120ppm or less of silicon, 50ppm or less of tin, 100ppm or less of tungsten, 50ppm or less of titanium, and 3.5ppm or less of uranium.

Sponge zirconium is typically produced by reduction with elemental magnesium at elevated temperatures and atmospheric pressure. The reaction is carried out in an inert atmosphere, for example helium or argon.

Another preferred embodiment uses a buried barrier layer formed of crystal rod zirconium. Zirconium crystal rods are produced by gas phase decomposition of zirconium tetraiodide. Crystal rod zirconium is more expensive, but
It has fewer impurities and has greater radiation damage resistance than sponge zirconium.

The use of a buried layer of zirconium also provides desirable manufacturing advantages. The process from tube forming to finishing tends to remove a small amount of material from the inside of the tube. By embedding the relatively expensive unalloyed zirconium in the tube wall, what is lost during manufacturing is the cheaper zirconium alloy.
The result is completely (100%) unalloyed zirconium
Available. Furthermore, manufacturing defects on the inside of the tube occur in relatively unimportant inner layers, ensuring continuity of the zirconium barrier, which is typically only a few mils thick. Additionally, alloyed zirconium inner layers are better than containers with unalloyed zirconium inner layers. The reason is that zirconium alloys are easier to machine than the softer unalloyed zirconium.
This is because honing and the like are easy.

However, if it is desired to provide a buried layer on the inner surface of the cladding vessel, the inner layer of zirconium alloy can be removed by etching after the tube has been finished to its final dimensions.

Zirconium alloys that serve as suitable alloy substrates include Zircaloy-2 and Zircaloy-4. Zircaloy-2 is approximately 1.5% tin by weight,
It contains 0.12% iron, 0.09% chromium and 0.005% nickel and is widely used in water-cooled nuclear reactors. Zircaloy-4 contains less nickel than Zircaloy-2, but slightly more iron than Zircaloy-2. The composite cladding used in the nuclear fuel element of the present invention can be manufactured by the following method.

In the first method, a tube of unalloyed zirconium barrier material is inserted into a hollow billet of the material selected as the substrate, a tube of the material selected as the inner layer is inserted into the zirconium barrier tube, and the assembly is then subjected to explosive processing. Connect the tube to the billet. The composite tube is heated to approximately 538 to 760℃ (approx.
Extrude at a high temperature of 1000~1400〓). The extruded composite tube is then processed, including conventional tube forming, to obtain cladding of the desired dimensions. hollow billet,
The relative wall thicknesses of the zirconium barrier tube and inner layer tube are selected to achieve the desired wall thickness ratio for the finished cladding tube.

In the second method, a tube of unalloyed zirconium barrier material is inserted into a hollow billet of the material selected as the substrate, a tube of the material selected as the inner layer is inserted into the zirconium barrier tube, and the assembly is then placed under compressive pressure. (e.g. 750° C. for 8 hours) to achieve good metal-to-metal contact and diffusion bonding between the tube and billet. The diffusion bonded composite tube is extruded by conventional tube shell extrusion, eg, as described in the first method. The extruded composite tube is then processed, including conventional tube forming, to obtain cladding of the desired dimensions.

Yet another method involves inserting a tube of unalloyed zirconium barrier material into a hollow billet of the material selected as the substrate, inserting a tube of the material selected as the inner layer into the zirconium barrier tube, and completing the assembly as described above. Extruded by conventional tube shell extrusion. The extruded composite tube is then processed, including conventional tube forming, to obtain cladding of the desired dimensions.

All of the above methods of making the composite cladding of the present invention are more economical than other methods used to make cladding, such as plating and vapor deposition.
In the method of manufacturing a nuclear fuel element of the present invention,
A composite cladding vessel with an open end is formed of a substrate, an unalloyed zirconium barrier metallurgically bonded to the inner surface of the substrate, and an inner layer metallurgically bonded to the inner surface of the zirconium barrier, and the composite cladding vessel is filled with nuclear fuel material. , leaving a space on the open end side, inserting the nuclear fuel material holding means into the space, applying a sealing member to the space communicating with the nuclear fuel, and then joining the end of the crudding container to the sealing member to seal both forming an airtight seal between them.

The present invention reduces the chemical interactions of the cladding, minimizes local stress on the zirconium alloy substrate portion of the cladding, minimizes stress corrosion of the zirconium alloy substrate portion of the cladding, and minimizes the pellet-cladding interaction ( It has many benefits that help extend the operational life of nuclear fuel elements, including reducing the potential for crack failure that occurs in zirconium alloy substrates as a result of PCI. The present invention further prevents direct contact between the fission products and the zirconium alloy substrate, and prevents the generation of local stress on the zirconium alloy substrate. Accordingly, the present invention prevents the occurrence or propagation of stress corrosion cracking in alloy substrates.

There are particular advantages to using unalloyed zirconium as a buried barrier. Unalloyed zirconium is very ductile, and when stress corrosion cracking occurs in the inner layer, propagation of the crack is effectively stopped by the zirconium barrier. This is believed to be because the radius of curvature at the end of the crack in unalloyed zirconium is significantly larger than in zirconium alloy, which requires significantly higher stress levels for propagation. Unalloyed zirconium is also less susceptible to iodine stress corrosion and also tends to inhibit crack propagation.

The important features of the composite cladding of the present invention are:
The improvements described above are achieved without any additional neutron penalty. Such cladding is readily applicable to nuclear reactors because it does not form eutectics during loss of coolant accidents or reactivity insertion accidents, including control rod drops. Furthermore, composite cladding does not have a heat transfer penalty in the sense that it does not create a thermal barrier to heat transfer as would occur in the situation of inserting a separate foil or liner within the fuel element. The composite cladding 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 can be made to this invention. The invention is to be interpreted in the broadest sense within the scope of the claims.

[Brief explanation of drawings]

FIG. 1 is a partially cutaway side view of a nuclear fuel assembly including nuclear fuel elements constructed according to the present invention;
FIG. 2 is an enlarged cross-sectional view of a nuclear fuel element of the present invention. 10... Fuel assembly, 11... Channel, 1
4... Fuel element, 16... Nuclear fuel material, 17...
Cladding, 21...Substrate, 22...Zirconium barrier, 23...Inner layer.

Claims (1)

Claims: 1 (a) a central core 16 of a nuclear fuel body material selected from the group consisting of compounds of uranium, plutonium, and thorium, and mixtures thereof; and (b) an elongated composite vessel 17 surrounding the core. In the nuclear fuel element 14, the composite container has a zirconium alloy substrate 21 on the outside and non-alloyed zirconium metallurgically bonded to the inner surface of the substrate, and the composite container has a thickness of about 1 to 30% of the vessel wall thickness of the composite container. % and a continuous inner layer 2 of zirconium alloy metallurgically bonded to the inner surface of said zirconium barrier and comprising about 1 to 10% of the thickness of the vessel wall of said composite container.
A nuclear fuel element characterized by consisting of 3. 2. The nuclear fuel element of claim 1, wherein the zirconium continuous barrier 22 accounts for about 5-15% of the thickness of the wall of the composite vessel 17. 3. The nuclear fuel element of claim 1, wherein said zirconium continuous barrier is comprised of sponge zirconium. 4. The nuclear fuel element of claim 1, wherein the zirconium continuous barrier is comprised of crystal bar zirconium. 5. The nuclear fuel element of claim 1, wherein during use in a nuclear reactor, the composite vessel surrounds the core such that a space 25 remains between the core and the composite vessel. 6. The nuclear fuel element of claim 1, wherein the nuclear fuel is selected from the group consisting of uranium compounds, plutonium compounds and mixtures thereof. 7. The nuclear fuel element according to claim 1, wherein the nuclear fuel comprises uranium dioxide. 8. The nuclear fuel element according to claim 1, wherein the nuclear fuel is a mixture of uranium dioxide and plutonium dioxide. 9. An outer part made of a zirconium alloy forms the base, and a continuous barrier of zirconium formed by metallurgically bonding unalloyed zirconium to the inner surface of such base accounts for about 1 to 30% of the thickness of the cladding vessel; A composite cladding vessel for a nuclear reactor, in which a continuous inner layer formed by metallurgically bonding a zirconium alloy to the inner surface of the barrier occupies approximately 1 to 10% of the thickness of the cladding vessel. 10. The composite cladding container of claim 9, wherein the continuous barrier comprises about 5-15% of the thickness of the cladding container. 11. The composite cladding container of claim 9, wherein the continuous barrier is formed of sponge zirconium. 12. The composite cladding container of claim 9, wherein the continuous barrier is formed of crystal rod zirconium.