JPS58216988A - Buried zirconium layer - 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 This invention relates generally to improvements in nuclear fuel elements for use in the core of nuclear fission reactors, and more particularly to improvements in nuclear fuel elements for use in nuclear fission reactor cores, and more particularly, a zirconium alloy substrate, an unalloyed zirconium barrier metallurgically bonded to the interior surface of the substrate, and a zirconium barrier. An improved nuclear fuel 4-element having a composite cladding vessel comprising a zirconium alloy inner layer metallurgically bonded to a zirconium alloy inner layer.

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 cladding. A plurality of fuel elements are assembled in a lattice pattern with regular spacing from each other within coolant flow regions, or channels, to form fuel assemblies, and a sufficient number of fuel assemblies are assembled to form a fission chain reaction assembly. A reactor core capable of carrying out self-sustaining nuclear fission reactions will be constructed.

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

Cladding serves several purposes. The two main objectives are: The first objective is to improve the relationship between the nuclear fuel and the coolant or moderator, if present, or between the coolant and moderator, if both are present. The second purpose is to prevent contact and chemical reactions, and the second purpose is to remove the radioactive fission products (some of which are gases) from the fuel when cooled or a moderator or when both a coolant and a moderator are present. The key is to prevent it from leaking into both of them. Common cladding materials include stainless steel, aluminum and its alloys, zirconium and its alloys, niobium (columbium), and certain magnesium alloys.

If the cladding fails, i.e., loses leak protection, the coolant or moderator and related systems may 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 reactions 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 and ductile at temperatures below about 398°C (about 750'F), 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 cladding became apparent due to the entangled interaction between the nuclear fuel, the cladding, and the fission products produced during the fission reaction.

We have confirmed that this undesirable phenomenon is promoted by local mechanical stresses due to the differential expansion between the fuel and the cladding (stresses during the cladding are localized at the fuel pellet interface and sometimes at cracks in the nuclear fuel). This phenomenon is described by the term [pellet-clutting interaction (PCI)].
It is defined in J. 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 closed 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, under certain conditions, the cladding is locally hydrogenated, and at the same time, the mechanical properties of the cladding are locally degraded. Ulanding is also adversely affected by gases such as oxygen, nitrogen, carbon oxides and carbon dioxide over a wide temperature range. The zirconium crap 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 This always occurs during the cladding and manufacturing of the fuel element, although it is excluded as much as possible from the ambient atmosphere. Sintered refractory and ceramic compositions, such as uranium dioxide and other compositions used as nuclear fuels, when heated, e.g. during fuel element manufacture, emit a measured amount of the above gases;
It also releases fission products during irradiation. 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.

8- Therefore, in view of the above, during the entire period when the fuel element is used in the operation of a nuclear power plant, there is a possibility that cladding from water, water vapor and other gases, in particular hydrogen, will react with the cladding inside the fuel element. It can be seen that it is desirable to minimize attacks. One such solution is to find materials that can rapidly chemically react with water, water vapor, and 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 a ceramic to prevent moisture from coming into contact with the nuclear fuel material. In the fuel element described in U.S. Pat. A layer is placed between the casing and the nuclear fuel to maintain uniform and good heat transfer from the pellets to the casing.

In US Pat. No. 2,873,238, a fissile uranium slug with jacks 1 is encased in a metal case, where the protective jacket or cover for the slug is a zirconium-aluminum bonding layer. U.S. Patent No. 2,84
The fissionable body with jackets 1 to 1 disclosed in No. 9.387 consists of a plurality of main body parts with open end jackets for nuclear fuel, and these main body parts are immersed in a molten bath of a bonding substance in advance, and the uranium An efficient thermally conductive connection is formed between the main body portion and the container (1ζ is the cladding).

Coatings are disclosed as metal alloys with good thermal conductivity properties, examples of which include aluminum-silicon and zinc-aluminum alloys. Tokuko Showa 4
No. 7-46559 (published November 24, 1972) describes the method of consolidating individual nuclear fuel particles into carbon-containing fn 44 fuel composites by coating the individual nuclear fuel particles with a dense, smooth carbon-containing coating. Another coating is disclosed in Japanese Patent Publication No. 47-1420.
No. 0, in which one group of pellets of two groups of belenes 1-1 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.

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

US 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 nuclear fuel. The additive can be in powder form, provided that subsequent fuel processing does not oxidize the metal;
Alternatively, it can be introduced into the fuel element as a wire, sheet or other form in, around or between the fuel pellets.

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

US Pat. No. 3,502,549 discloses a method of protecting zirconium and its alloys by chromium plating to form a composite material useful in nuclear reactors. E 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.
n[! r! Jia N IC1eare, Vo
l, 11. No. 9 (September 1964), 5
It is described on pages 05-508. 13 Rossa et al. [Stability and Compatibility of Hydrogen Barriers Deposited on Zirconium Alloys]
tility of l-11-1ydro 3a
rriers Applied to Zircon
ium' A 1loys > -1 (Europen
Atomic Energy Community
, Joint Nuclear Re5earc
hCenter, EUR 4098e, 1969) describes methods for depositing various coatings and their effectiveness as hydrogen diffusion barriers, with AI-8i coatings being identified as the most promising barrier to hydrogen diffusion. A method for plating zirconium and zirconium-tin alloys with nickel and heat treating these alloys to produce alloy-diffusion bonds has been described by W. C. Schnicker.
[Electroplating on Zi]
rconium and Zirconium-Ti
n) J (BM I-757, Technical
Information3service, 195
2). U.S. Patent No. 3.625,82
In the nuclear reactor fuel element having a double cladding tube disclosed in No. 1, the inner surface of the cladding tube is coated with a metal having a small neutron capture cross section, such as nickel, and dispersed particles of a burnable poison are buried therein. ing. React
or[) development program
Process Report (August 1973)
A chemical getter structure in which a sacrificial layer of chromium is provided on the inner surface of a stainless steel ratting is disclosed in ANL-RDP-19).

Yet another solution is to interpose a barrier between the nuclear fuel material and the cladding that holds it; these methods are described in U.S. Pat. .115 (titanium N), U.S. Pat. No. 3,212°988 (zirconium, aluminum or beryllium sheath), U.S. Pat.
No. 18,238 (crystalline carbon barrier between UO2 and zirconium cladding), and U.S. Pat.
No. 88.893 (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), others with materials that are incompatible with cladding (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) (copper and other metals react with the cladding and change its properties). None of the above patent publications discloses a solution to the recently discovered problem of local chemical-mechanical interaction between nuclear fuel and cladding.

Still other advances on the barrier concept are disclosed in U.S. Pat. in the form of a layer or multilayer tube or foil or as a coating 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, and a coating of a highly lubricious substance is placed between the liner and the cladding. It is located between Klatsding.

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 a low melting point eutectic layer with the zirconium alloy substrate, which is undesirable in the event of a predicted loss of coolant accident.

No. 4,200,492 discloses a composite cladding consisting of a zirconium alloy substrate and an unalloyed zirconium liner. Although soft zirconium alloys minimize local strain and suppress stress corrosion cracking and liquid metal embrittlement, they are susceptible to damage and deterioration during manufacturing due to honing, etc., and are susceptible to corrosion if the cladding ruptures. .

Therefore, it remains desirable to develop nuclear fuel elements that minimize the problems discussed above.

Invention 6 Nuclear fuel elements particularly effective for use in the core of a nuclear reactor are:
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. The substrate of this cladding does not differ in its design and function from that customary in conventional nuclear reactors and is selected from common cladding materials, such as zirconium alloys.

16- The zirconium barrier and inner layer forms a shield between the substrate and the nuclear fuel material held within the cladding and shields the substrate from fission products and gases.

The inner layer also shields the zirconium barrier from fission products released from the fuel and other reactive elements present within the fuel element. This shielding effect may be due to recoil fission products;
or by preventing hardening due to reaction with chemical elements present in the combustible element, allowing the zirconium barrier to maintain maximum purity and ductility.

The zirconium barrier is approximately 1 to 30 times the thickness of the cladding.
%. Zirconium barriers less than about 1% of the cladding thickness are difficult to manufacture industrially, and zirconium barriers greater than about 30% of the cladding thickness provide additional benefits commensurate with the increased thickness. I can't do it. In general, increasing the barrier cladding thickness by more than about 30% reduces the substrate thickness and weakens the composite cladding.

The inner layer is preferably formed to form 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. Due to the purity of the barrier and the shielding effectiveness of the inner layer, the barrier remains soft during irradiation, minimizing local strains inside the nuclear fuel element and thus protecting the substrate from stress corrosion cracking and liquid metal embrittlement. The inner layer and zirconium barrier are the preferred reactive sites for reacting with volatile impurities or fission products present inside the nuclear fuel element! In this way, it serves to protect the barrier and cladding from attack by volatile impurities or fission products.

In addition, the inner layer is effective in preventing loss or damage of the soft barrier during cladding manufacture, thus improving processability. Furthermore, the inner layer protects the barrier from aqueous corrosion in case of fuel element failure.

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

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.

It is an object of the present invention 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 bond 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.

DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 shows a partially cutaway side view of a nuclear fuel assembly 10. The fuel assembly 10 has a 9 square 19-section tubular flow channel 11 with a lifting pail 12 at its upper end and a nosepiece (not shown since the lower part of the fuel assembly 10 is omitted) at its lower end. ) is provided. The upper end 13 of the channel 11 is open and the lower end of the nosepiece is provided with coolant flow openings. 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 1- (not shown as the lower part has been omitted). Usually liquid coolant enters through an opening in the lower end of the nosepiece;
It flows upwardly around the fuel element 14 and through the upper outlet 13 at a high temperature and in a partially evaporated state in the case of boiling reactors or in a non-vaporized state in the case of pressure reactors. Go out.

Nuclear fuel element or rod 14 is cladded 1 at both ends thereof.
The end plug 18 is sealed by an end plug 18 welded to the fuel rod 7 and has a single short shaft (stud) 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. A nuclear fuel material retention means 24 in the form of a helical member is 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 occasional warping and vibration caused by high velocity coolant flow.

A nuclear fuel element or rod 14 constructed in accordance with the present invention is
The figure is shown in cross section. The fuel element comprises a core or central cylindrical portion of fuel material 16, shown here as a number 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, nuclear fuel material 16 forming the central core of fuel element 14 is surrounded by 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 use 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.

An unalloyed zirconium barrier 22 is metallurgically bonded to the inner peripheral surface of the substrate 21, such that the zirconium barrier 22 forms a shield that shields the substrate 21 from the nuclear fuel material 16 within 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. Preferably, the inner layer 23 comprises about 1-10% of the thickness of the cladding;
Formed from a conventional cladding TTA material, in the preferred embodiment of the invention the inner layer is a zirconium alloy, such as Zircaloy-2 or Silka[1-4]. The zirconium barrier acts as a reaction site that reacts with gaseous impurities and fission products that have passed through cracks and defects in the inner layer 23,
It protects the base portion of the cladding from contact with and reactions with impurities and fission products, and minimizes the occurrence of local stress and cladding due to pellet-cladding 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 the nuclear fuel element of the present invention 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 from 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.

The unalloyed zirconium that forms the barrier in composite cladding is highly resistant to radiation hardening, and this property allows the zirconium barrier to maintain desirable structural properties, such as yield strength and hardness, even after long-term irradiation, to a greater extent than conventional zirconium alloys. can be maintained at a low level. In fact, the zirconium barrier does not harden as significantly during irradiation as conventional zirconium alloys, and this, combined with its initially low yield strength, causes the zirconium barrier to plastically deform during power transients and to reduce pellet-induced stresses within the fuel element. can be absorbed. burn it
The pellet-induced stress within the element is determined by e.g. the reactor operating temperature (
300-350℃), nuclear fuel pellets undergo thermal expansion and/or
or swelling, thus resulting from contact of berene 1~ with the cladding.

Roughly, the thickness of the crutching is preferably about 5 to 15
% of the zirconium barrier, particularly preferably 10% of the thickness of the cladding, a zirconium barrier bonded to the outer substrate of the zirconium alloy achieves sufficient barrier effect and stress relief to prevent failure of the composite cladding. Ru.

A preferred embodiment according to the principles of the invention uses "hypoxic sponge" grade zirconium as the buried barrier layer, but higher purity "crystal rod zirconium" grade and lower purity "quaternary sponge" zirconium can also be used. Residual impurities in sponge zirconium serve to impart special properties to the zirconium barrier. Generally, sponge zirconium contains about i
oo.

Impurities of greater than or equal to about 5000 ppm, preferably less than /I'200+11)In, are included. Preferably, the oxygen is maintained within the range of about 200-1200111) m. Other typical impurity levels are shown below. Aluminum 751) l1m or less, boron 0.4
ppm or less, cadmium 0.4. f'll 1 m or less, carbon 27 o ppm or less, rim [lm 2001) l) m or less, cobal 1 to 2 o ppm or less, copper 501] pm or less,
Hafnium 1100 ppm or less, hydrogen 25111) m or less, iron 1500 ppm or less, magnesium 20 ppm or less, manganese 50 ppm or less, molybdenum 50 DDm
Below, Nicke Schiff 0IIDm or less, Niobium 1ooppm
Below, nitrogen 80 pl) m or less, silicon 1201) 11 m
Below, tin 50 ppm or less, tungsten 100111
1 m or less, titanium 50 pl'1 m or less, and uranium 3.5 pl' m or less.

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. Although crystal bar zirconium is more expensive, it has fewer impurities and is more resistant to radiation damage 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 molybdenum from the inside of the tube. By embedding the relatively expensive unalloyed zirconium in the tube wall, the cheaper zirconium alloy is lost during manufacturing, so that the unalloyed zirconium is completely
100%) 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. This is because zirconium alloys are easier to machine, such as T1 honing, than softer unalloyed zirconium.

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 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, 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-2J. The composite cladding used in the nuclear fuel element of the present invention can be manufactured by the following method.

In the first 1j method, an unalloyed zirconium barrier tube is inserted into a hollow billet of the selected material as the substrate, a tube of the selected material as the inner layer is inserted into the zirconium barrier tube, and then this assembly is Explosion 28- Process and join the tube to the billet. The composite tube is heated to approximately 538-760°C (approximately 10
Extrusion at high temperature (00-1400'F).

The extruded composite tube is then processed, including conventional tube forming, to obtain the desired cladding dimensions.

The relative wall thicknesses of the hollow billet 1, the zirconium barrier tube, and the inner layer tube are selected to demonstrate the desired wall thickness ratio for the finished tartarding tube.

In the second method, a tube of unalloyed zirconium barrier material is selected as the substrate and inserted into a hollow billet at t=U;
A tube of the material chosen as the inner layer is inserted into the zirconium barrier tube and the assembly is then subjected to a heating step under compressive pressure (e.g. 750° C. for 8 hours) to ensure good metal-to-metal contact and diffusion between the tube and the billet. Achieve union. The diffusion-bonded composite tube is processed by conventional tube shell extrusion, e.g.
Extrude as described in the method. The extruded composite tube is then processed, including conventional tube forming, to obtain the desired cladding dimensions.

In yet another method, unalloyed zirconium alloy 1￥31
The tubing is inserted into the hollow billet of the selected tubing as the substrate, the tubing of the selected tubing as the inner layer is inserted into the zirconium barrier tube, and the assembly is extruded by conventional tube shell extrusion as described above. The extruded composite tube is then processed, including conventional tube forming, to obtain the desired cladding dimensions.

All of the above methods of manufacturing the composite cladding of the present invention are more economical than other methods used to manufacture cladding, such as plating and vapor deposition.

A method of manufacturing a nuclear fuel element of the present invention includes forming an open-ended composite cladding vessel comprising 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. death,
This composite cladding container is filled with nuclear fuel material, a space is left at the open end side, a nuclear fuel material holding means is inserted into the space, a sealing member is applied to the space communicating with the nuclear fuel, and then the end of the cladding container is filled with nuclear fuel material. and a sealing member to form a hermetic seal therebetween.

The present invention reduces the chemical interactions of cluttering,
This minimizes local stress on the zirconium alloy base portion of the cladding, minimizes stress corrosion on the zirconium alloy base portion of the cladding, and minimizes the stress corrosion of the zirconium alloy base portion of the cladding.
It has many benefits that help extend the operational life of nuclear fuel elements, including reducing the potential for crack failure in zirconium alloy substrates as a result of cladding interactions (PCI). The present invention further prevents direct contact between fission products and the zirconium alloy substrate, and
Prevents local stress from occurring on the zirconium alloy substrate.

Thus, the present invention prevents the occurrence or transmission 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, the 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 in the conductor. Unalloyed Silcony 31-ram is also less susceptible to iodine stress corrosion and also tends to resist crack propagation.

An important feature of the composite cladding of the present invention is that the improvements described above are achieved without any additional neutron penalty. Such rudding can be readily applied to nuclear reactors since it does not form eutectics in the event of a loss of cooling house accident or a reactivity insertion accident, including a control rod drop.

Additionally, composite claddings do not have a heat transfer penalty in the sense that they do 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 the present invention. The invention is to be construed in the broadest sense within the scope of the claims.

[Brief explanation of the drawing]

FIG. 1 is a partially cutaway side view of a nuclear fuel assembly including 32 nuclear fuel elements constructed in accordance with the present invention, and FIG. 2 is an enlarged cross-sectional view of the nuclear fuel element of the present invention. 10... Fuel assembly, 11... Channel, 14... Fuel element, 16... Nuclear fuel material, 17... Cladding, 21 ...Base body, 22 ... Zirconium barrier, 23 ... Inner layer. Patent Applicant General Electric Company Representative (7
630) Raw Numa Toku 2Fig, 1 11 Ft'g-2

Claims (1)

Claims: 1. (a> a central core of a body of nuclear fuel material selected from the group consisting of uranium, plu-1-nium, 1-lium compounds, and mixtures thereof; and (11) an elongated body surrounding said core. a composite cladding vessel, the composite cladding vessel having an outer substrate, a continuous zirconium barrier of unalloyed zirconium metallurgically bonded to an interior surface of the substrate and comprising about 1 to 30% of the thickness of the cladding vessel; and the zirconium barrier. 2. A nuclear fuel element comprising a continuous inner layer metallurgically bonded to the inner surface of the cladding vessel and comprising about 1 to 10% of the thickness of the cladding vessel. 2. The outer substrate is formed of a zirconium alloy. The nuclear fuel element of claim 1. 3. The nuclear fuel element of claim 1, wherein the inner layer is formed of a zirconium alloy. 4. The nuclear fuel element of claim 1, wherein the zirconium barrier is about 5-15% of the thickness of the cladding vessel. The nuclear fuel element according to claim 1. 5. The nuclear fuel element according to claim 1, wherein the unalloyed zirconium barrier is sponge zirconium. 6. The nuclear fuel element according to claim 1, wherein the unalloyed zirconium barrier is crystal rod zirconium. 7. The nuclear fuel element according to claim 1. 7. The nuclear fuel element according to claim 1, wherein the nuclear fuel material is selected from the group consisting of uranium compounds, plu-1-nium compounds, and mixtures thereof. 8. 9. The nuclear fuel element according to claim 1, wherein the nuclear fuel material is uranium dioxide. 9. The nuclear fuel element according to claim 1, wherein the nuclear fuel material is a mixture of uranium dioxide and plutonium dioxide. 10. (a) a central core of a body of nuclear fuel material selected from the group consisting of uranium, plutonium, thorium compounds, and mixtures thereof; and (b) an elongate composite cladding vessel surrounding said core, the composite cladding vessel comprising zirconium and thorium compounds. a continuous zirconium outer portion formed of a material selected from zirconium alloys and forming a base body, and a continuous zirconium formed of unalloyed zirconium metallurgically bonded to the inner surface of the base body and comprising about 1 to 30% of the thickness of the cladding vessel; a barrier metallurgically bonded to the inner surface of the zirconium barrier and having a thickness of about 1 to 10 times the thickness of the cladding vessel.
A nuclear fuel element characterized in that it consists of a continuous inner layer formed of zirconium or a zirconium alloy comprising %. 11. A zirconium alloy outer portion forming a substrate, a continuous zirconium barrier formed of unalloyed zirconium metallurgically bonded to the inner surface of the substrate and comprising about 5 to 15% of the thickness of the cladding vessel; and the metal barrier. A composite cladding vessel for a nuclear reactor, comprising a continuous inner layer of zirconium alloy metallurgically bonded to the inner surface of the cladding vessel and comprising about 1 to 10% of the thickness of the cladding vessel. 12. Claim 11, wherein said unalloyed zirconium barrier comprises about 5-15% of the thickness of the cladding vessel.
Composite cladding container as described in Section 1. 13. The composite cladding container of claim 11, wherein said unalloyed zirconium barrier is formed from cancellous zirconium. 14. The composite cladding container of claim 11, wherein said unalloyed zirconium barrier is formed of crystal rod zirconium. 15. An outer portion formed of a material selected from zirconium and zirconium alloys and forming a base body, and formed of unalloyed zirconium metallurgically bonded to the inner surface of the base body and comprising about 5 to 15% of the thickness of the cladding vessel. an elongated composite cladding vessel consisting of a continuous zirconium barrier made of metal and a continuous inner layer of zirconium or zirconium alloy metallurgically bonded to the inner surface of the metal barrier and comprising about 1 to 10% of the thickness of the cladding vessel: uranium, plutonium, thorium. A center consisting of a nuclear fuel material selected from compounds and mixtures thereof, disposed within the container and partially filling the container leaving an internal void within the container; a sealing member secured and sealed; and nuclear fuel material retaining means disposed in the cavity, such that the cladding vessel leaves a gap between the core and the cladding during use of the core in a nuclear reactor. A nuclear fuel element characterized by enclosing. 16. A hollow composite cladding vessel for nuclear fuel suitable for use in a nuclear reactor having an outer substrate of a zirconium alloy and an inner liner of a zirconium alloy, wherein unalloyed zirconium is metallurgically bonded between the outer substrate and the inner liner. A composite cladding container characterized by having a barrier layer.