CA1209727A - Buried zirconium layer - Google Patents

Abstract

BURIED ZIRCONIUM LAYER
ABSTRACT OF THE DISCLOSURE
A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate, an unalloyed zirconium barrier metallurgically bonded to the inside surface of the substrate, and an inner layer metallurgically bonded to the inside surface of the zirconium barrier. In this composite cladding, the inner layer and the zirconium barrier shield the substrate from any impurities or fission products from the nuclear fuel material held within the composite cladding. The zirconium barrier forms about 1 percent to about 30 percent of the thickness of the cladding. The inner layer and the zirconium barrier protect the substrate from stress corrosion and the zirconium barrier restricts the propagation of stress cracking. The substrate and the inner layer of the cladding are selected from conventional cladding material, and preferably are a zirconium alloy.

Description

~2~72~,i BU~IED ZIRCONI~M LAYER
Field of the Inventi'on This invention relates broadly to an improvement in nuclear fuel eIements for use in the core of nuclear fission reactors and, more particularly, to an improved nuclear fuel element having a composite cladding container having a zirconium alloy substrate, a barrier of non-alloyed zirconium metallurgically bonded to the inside surface of the substrate and an inner zirconium alloy layer metallurgically bonded to the zirconium barrier.
Background of the I'nvention Nuclear reactors are presently being designed constructed, and operated in which the nuclear fuel is contained in fuel elements which can have various geometric shapes, such as plates, tubes, or rods. The fuel material is usually enclosed in a corrosion-resistant, non-reactive, heat conductive container or cladding. The fuel elements are assembled together in a lattice at fixed distances from each other in a

2~ coolant flow channeI or region forming a fuel assembly and sufficient fuel assemblies are combined to form the nuclear fission chain reacting assembly or reactor core capable of a self-sustained -fission reaction.
The core, in turn, is enclosed with a reactor ~essel through ~hich a coolant is passed.
The cladding serves several purposes and two primary purposes are: first, to prevent contact and chemical reactions between the nuclear fuel and the -' coolant or the moderator if a moderator is present, or ~ !

12~Z7 24-NT-0~471 both if both the coolant and the moderator are present;
and second, to prevent the radioactive fission products, some of which are gases, from being reIeased from the fuel into the coolant or the moderator, or both if both the coolant and the moderator are present. Common cladding materials are stainless steel, aluminum and its alloys, zirconium and its alloys, niobium (columbium), certain magnesium alloys, and others. The failure of the cladding, i.e, a loss of the leak tightness, can contaminate the coolant or moderator and the associated systems with radioactive long-lived products to a degree which interferes with plant operation.
Problems have been encountered in the manufacture and in the operation of nuclear fuel elements which employ certain metals and alloys as the clad material due to mechanical or chemical reactions of these cladding materials under certain circumstancesO Zirconium and its alloys, under normal circumstances~ are excellent nuclear fuel claddings since they have low neutron absorption cross-sections and at temperatures below about 750F
(about 398C~ are strong, ductile~ extremely stable and non-reactive in the presence of demineralized water or steam which are commonly used as reactor coolants and moderators.
Howe~er, fuel element performance has re~ealed a problem with the brittle splitting of the cladding due to the combined interactions between the nuclear fuel, the cladding and the fission products produced duriny nuclear fission reactions. It has been discovered that -this undesirable performance is promoted by localized mechanical stresses due to fueI cladding differential expansion (stresses in the cladding are localized at fuel pellet interfaces and sometimes at cracks in the nuclear fuel). The phenomenon is defined by the terms pellet-cladding-interaction or PCI. Corrosive fission products are released from the nuclear fuel and are ~ 7Z~ 24-NT-04471 present at the intersection of the fuel pellet interfaces with the cladding surface. Such fission products are created in the nuclear ~uel during the fission chain reaction during operation of a nuclear reactor.
~ithin the confines of a sealed fuel element, hydrogen gas can be generated by the slow reaction between the cladding and residual water inside the cladding. This hydrogen gas may build up to levels which, under certain conditions, can result in localized hydriding of the cladding with concurrent local deterioration in the mechanical properties of the cladding. The cladding is also adversely affected by such gases as oxygen, nitrogen, carbon monoxide, and carbon dioxide over a wide range of temperatures. The zirconium cladding of a nuclear fuel element is exposed to one or more of the gases listed above and fission products during irradiation in a nuclear reactor and this occurs in spite of the fact that these gases may not be present in the reactor coolant or moderator, and further may have been excluded as far as possible from the ambient atmosphere during manufacture of the cladding and the fuel element. Sintered refractory and ceramic compositions, such as uranium dioxide and other compositions used as nuclear fuel, release measurable quantities of the aforementioned gases upon heating, such as during fuel element manufacture, and further release fission products during irradiation. Particulate refractory and ceramic compositions/ such as uranium dioxide powder and other powders used as nuclear fuel, have been known to release even larger quantities of the aforementioned gases during irradiation. These released gases are capable of reacting with the zirconium cladding containing the~nuclear fuel.
Thus, in light of the foregoing, it has been found desirable to minimize attack of the cladding from water, water vapor and other gases, especially hydrogen, 12~7~:~

reactive with the cladding from inside the fuel element throughout the time the fuel element is used in the operation of nuclear power plants. One such approach has been to find materials which will chemically react rapidly with the water, water vapor and other gases to eliminate these from the interior of the cladding. Such materials are called getters.
Another approach has been to coat the nuclear fuel material with a ceramic to prevent moisture coming in contact with the nuclear fuel material as disclosed in U.S. Patent No, 3~108r936/ issued October 29, 1963 to Gale. U.S. Patent No. 3,085~059, issued April 9, 1963 to Burnham describes a fuel element including a metal casing containing one or more pellets of fissionable ceramic material and a layer of vitreous material bonded to the ceramic pellets so that the layer is between the casing and the nuclear fuel to assure uniformly good heat conduction Erom the pellets to the casing. U.S. Patent No. 2,873~238, issued 20 February 10, 1959 to OhIinger et al, describes jacketed fissionable slugs of uranium canned in a metal case in which the protective jackets or coverings for the slugs are a zinc-aluminum bonding layer. U.S.
Patent No. 2,849,387, issued August 26, 1958 to Brugmann~ discloses a jacketed fissionable body comprising a plurality of open-ended jacketed body sections of nuclear fuel which have been dipped into a molten bath of a bonding material giving an e~fective thermally conductive bond between the uranium 3Q body sections and the container (or cladding). The coating is disclosed as any metal alloy having good thermal conduction properties with examples including aluminum-silicon and zinc-aluminum alloys. Japanese Patent Publication No. SF~O 47-46559 dated November 24, 1972, disclosed consolidating discrete nuclear fuel particles into a carbon-containing matrix fuel composite ~2~ Z~

by coating the fuel particles with a high density, smooth carbon-containing coating around the pellets. Still another coating disclosure is Japanese Patent Publicatlon No.
SHO 14200 in which the coating of one of two groups of pellets is with a layer of silicon carbide and the other group is coated with a layer of pyrocarbon or metal carbide.
The coating of nuclear fuel material introduces reliability problems in that achieving uniform coatings free of faults is difficult. Further, the deterioration of the coating can introduce problems with the long-lived performance of the nuclear fuel material.
One known method for preventing corrosion of nuclear fuel cladding consists of the addition of a metal such as niobium to the fuel. The additive can be in the form of a powder, provided the subsequent fuel processing operation does not oxidize the metal, or the additive can be incorporated into the fuel element as wires, sheets, or other forms in, around or between fuel pellets.
General Electric Atomic Power Document 4555, 20 dated February 1964,at GE NEBO Library 175 Curtner Avenue, San Jose, Calif. 95125, discloses a composite cladding of a zirconium alloy with an inner lining of stainless steel metallurgically bonded to the zirconium alloy, and the composite cladding is fabricated by use of extrusion of a hollow billet of the zirconium alloy having an inner lining of stainless steel. This cladding has the disadvantage that the stainless steel develops brittle phases, and the stainless steel layer involves a neutron absorption penalty of about ten to fifteen times the penalty for a zirconium alloy layer of the same thickness.
U.S. Patent No. 3,502,549, issued March 24, 1970 to Charveriat, discloses a method fvr protecting zirconium and its alloys by the electrolytic deposition of chromium to provide a composite material useful for nuclear reactors. A method for electrolytic deposition of copper on Zircaloy-2 surfaces and subsequent heat ~ 7~7 2~-NT-0~471 treatment for the purpose of obtaining surface diffusion of the elctrolytieally deposited metal is presented in Energia Nucleare, Volume 11, No. 9 (September, 1964) at .
pages 505-508. In Stability and Compatibility of' Hydrogen Barriers Applied to Zirconium All'oys, by ~. ~rossa et al (European Atomic Energy Community, Joint Nuelear Research Center, EUR 4098e~ 1969), methods of deposition of different coatings and their efficiency as hydrogen diffusion barriers are described along with an Al-Si coating as the most promising barrier against hydrogen diffusion. Methods for electroplating nickel on zireonium and zirconium tin alloys and heat treating these alloys to produce alloy-diffusion bonds are disclosed in Electro;plating-on Zirconium and Zirconium-Tin, by W.C.
Schnickner et al (BMI-757, Technical Information Service, 1952). U.S. Patent No. 3,625,821, issued December 7, 1971 to Ricks, presents a fuel element for a nuclear reactor having a duel cladding tube with the inner surface of the tube being coated with a metal of low neutron capture 2Q cross-section such as nickel and having finely-dispersed particles of a burnable poison disposed therein. Reactor Development Program Proc'es's Report of August, 1973 (ANL-RDP-l9) discloses a chemical getter arrangement of a sacrificial layer of chronium on the inner surface of a stainless steel cladding.
Another approach has been to introduce a barrier between the nuclear fuel material and the cladding holding the nuclear fuel material as disclosed in U.S. Patent No.

3,320,150, issued January 18, 1966 to Martin et al 30 (copper foil), German Patent Publication DAS 1,238,115 (titanium layer), U.S. Patent No. 3,212,988, issued October 19, 1965 to Ringott et al (sheath of zirconium~
aluminum or beryllium), U.S. Patent No. 3~018,238, issued January 23, 1962 to Layer et al(barrier of crystalline earbon between the UO2 and the zirconium cladding), and U.S. Patent No. 3,088,893, issued May 7t 1963 to Spalaris (stainless steel foil). While the barrier concept ~LZ~9~
2~-NT-Q4471 proves promising, some of the foregoing references invol~e incompatible materials with~either the nuclear fuel (e.g.
carbon can combine with oxygen from the nuclear fuel), or the cladding (e.g., copper and other metals can react with the cladding, altering the properties of the cladding), or the nuclear fission reaction ~e.g., by acting as neutron absorbers). None of the listed references disclose solutions -to the recently discovered problem of localized chemical-mechanical interactions between the nuclear fuel and the cladding.
Further approaches to the barrier concept are disclosed in U.S. Pat~nt No. 3,969,186~ issued July 13, 1976 to Thompson et al (refractory metal such as molybdenum, tungsten, rhenium, niobium and alloys thereof in the form of a tube of foil of single or multiple layers or a coating on the internal surface of the cladding), and U.S. Patent No. 3,925,151, issued December 9, 1975 to Klepfer (Iiner of zirconium, niobium, or alloys thereof between the nuclear fuel and the cladding with a coating of a high lubricity material between liner and the cladding).
U.S. Patent No. 4,045,288, issued August 30, 1977 to Armijo discloses a compo~ite cladding of a zirconium alloy substrate with a metal barrier metallurgically bonded to the substrate and an inner layer of zirconium alloy metallurgically bonded to the metal barrier. The barrier is selected from a group of niobium, aluminum, copper7 nickel~ stainless steel, and iron.
With the exception of the niobium barrier, all the other materials will form low melting eutectic phases with the zirconium alloy substrate ! making them undesirable in postulated loss-of-coolant accidents.
U~S. Patent No. 4,200,492, issued April 29, 1980 to Armijo et al discloses a composite cladding of a zirconium alloy substrate with an unalloyed zirconium liner.
The soft zirconium l~ner minimizes loaallzéd strain, and reduces stress corrosion cracking and liquid metal embrittlement, but is subject to damage and losses due to honing and the like during fabrication and to corxosion in the event that the cladding is breached.
Accordingly, it has remained desirable to develop nuclear fuel elements minimizing the problems discussed above.
Summary of the Invention A particularly effectlve nuclear fuel element for use in the core of a nuclear reactor has a composite cladding having a substrate, a non-alloyed zirconium barrier metallurgically bonded to the inside surface of the substrate, and an inner layer metallurgically bonded to the inside surface of the zirconium barrier.
The substrate of the cladding is completely unchanged in design and function from previous practice for a nuclear reactor and is selected from conventional cladding materials such as zirconium alloys.
The zirconium barrier and the inner layer form a shield between the substrate and the nuclear fuel material held in the cladding, as well as shielding the substrate from fission products and gases. The inner layer in turn shields the zirconium barrier from fission products released from the fuel and other reactive elements present in the fuel element. This sheilding allows the zirconium barrier to retain a maximum amount of purity and ductility by preventing hardening by recoiliny fission products or hy reaction with chemical elements present in the Euel element.
The zirconium barrier forms about 1 to about 30 percent of the thickness of the cladding. A zirconium barrier forming less than about 1 percent of the thickness of the cladding would be difficult to achieve in commercial production, and a zirconium barrier forming more than 30 percent of the thickness of the cladding would provide no additional benefit for the added thickness. Further, a 2~-NT-0~471 _g_ barrier more than about 30 percent of the thickness of the cladding would produce a concomitant reduction in thickness of the substrate and weakening of the composite cladding.
The inner layer may be fabricated to constitute from 1 percent to 10 percent of the total cladding thickness. This thickness range has been specified to provide an inner layer of minimum thickness fabricable by tubing co-extrusion and co-reduction techniques. Because of its purity and the shielding effect of the inner layer, the barrier remains soft during irradiation and minimizes localized strain inside the nuclear fuel element, thu~
serving to protect the substrate from stress corrosion cracking or liquid metal embrittlement. The inner la~er and the zirconium barrier provides a preferential reaction site for reaction with volatile impurities or fission products present inside the nuclear fuel element and, in this manner, serve to protect the barrier and the cladding from attack by the volatile impurities or fission products.
In addition, the inner layer is useful during fabrication to prevent losses or damage to the soft barrier and thus improves fabricability. Further, the inner layer protects the barrier from aqueous corrosion in the event of fuel element failure.
This invention has a striking advantage that the substrate of the cladding and the barrier are protected from s-tress corrosion cracking and liquid metal embrittlement, in addition to contact with fission products, corrosive gases, etc., by the inner layer which does not introduce an appreciable neutron capture penalties, heat transfer penalties, or materials incompatibility problems.
Objects of the Invention -It ls an object of this invention to provide a nuclear uel element capable of operating in nuclear reactors for extended periods of time without the oCcurrence of splitting of the cladding/ corrosion of ; the cladding, or other fuel failure problems.

~'~7 ~
~ u ~ ~ ~4-NT-04471 It is another object o~ this invention to provide a nuclear fuel element with a composite cladding having a substrate, a zirconium barrier metallurgically bonded to the inside surface of the substrate, and an inner layer metallurgically bonded to the inside surface of the zirconium barrier so that the metallurgical bonds provide a long-lived connection between the substrate and the zirconium barrier and between the zirconium barrier and the inner layer.
The foregoing and other objects of this invention will become apparent to persons skilled in the art from reading the following specification and the drawings described immediately hereinafter.
Description of the Drawing Figure 1 is a partial cutaway sectional view of a nuclear fuel assembly containing nuclear fuel elements constructed according to the teaching of this invention and Figure 2 is an enlarged trans~erse cross-sectional view of the nuclear fuel element in Figure 2illustrating the teaching of this invention.
Description of the Invention Referring now more particularly to Figure 1, there is shown a partially cutaway sectional view of a nuclear fuel assembly 10. This fuel assembly 10 consists of a tubular flow channel 11 of generally-square cross-section provided at its upper end with a lifting bail 12 and at its lower end with a nose piece ~not shown due to the lower portion of the assembly 10 being omitted). The upper end of channel 11 is open at 13 and the lower end of the nose piece is provided with coolant flow openings.
An array of fuel elements or rods 14 is enclosed in the channeI 11 and supported therein by means of an upper end plate 15 and a lower end plate (not shown due to the lower portion being omitted). The liquid coolant ordinarily enters through the openings in the lower end of the nose piece, passes upwardly around fuel elements t 14, and discharges through the upper outlet 13 at an elevated temperature in a partially vaporized condition for boiling reactors in an unvaporized condition for pressurized reac-tors.
The nuclear fuel elements or rods 14 are sealed at their ends by means of end plugs 18 welded to the cladding 17, which may include studs 19 to facilitate the mounting of the fuel rod in the assemhly. A void space or plenum 20 is provided at one end of the element to permit longitudinal expansion of the fuel material and accumulation of gaes released from the fuel material. A nuclear fuel material retainer means 24 in the form of a helical member is positioned within space 20 to provide restraint against the axial movement f the pellet column, especially during handling and transportation of the fuel element.
The fuel element is designed to provide an excellent thermal contact between the cladding and the fuel material, a minimum of parasitic neutron absorption and resistance to bowing and vibration which is occasionally caused by flow of the coolant at high velocity.
A nuclear fuel element or rod 14 constructed according to the teachings of this invention is shown in a partial section in Figure 1. The fuel element includes a core or central cylindrical portion of nuclear fuel material 16, here shown as a plurality of fuel pellets of fissionable and/or fertile material positioned within a structural cladding or container 17. In some cases, the fuel pellets may be of various shapes such as cylindrical pellets or spheres and, in other cases, different fuel forms such as a particulate fuel may be used. The physical form of the fuel is immaterial to this invention. Various nuclear fuel materials may be used including uranium compounds, plutonium compounds, thorium compounds, and mixtures thereof. A preferred fuel is uranium dioxide or a mixture comprising uranium dioxide and plutonium dioxide.

7~

Referring now to Figure 2, the nuelear ~uel material 16 forming the central eore of the fuel element 14 is surrounded by a cladding 17 which, in this invention, is also referred to as a composite cladding.
The composite cladding container encloses the fissile core so as to leave a gap 24 between the eore and the eladding container during use in a nuclear reactor. The composite eladding has an external substrate 21 selected for conventional fuel eladding materials and, in a preferred embodiment of this invention, the substrate is a zireonium alloy sueh as Zirealoy-2 or Zirealoy-4.
The substrate 21 has metallurgieally bonded on the inside eircumference thereof an unalloyed æirconium barrier 22 so that the zireonium barrier forms a shield o~ the substrate from the nuelear fuel material 16 inside the composite cladding. The zirconium barrier preferably forms about 1 percent of about 30 percent of the thiekness of the eomposite eladding.
The zireonium barrier 22 has metallurgieally bonded on the inside eireumference thereof an inner layer 23 so that the inner layer is the portion of the composite cladding cosest to the nuelear fuel material 16. The inner layer preferably forms about 1 percent to about 10 percent of the thickness of the cladding and is comprised of conventional cladding materials, and in a preferred embodiment of this invention, the substrate is a zirconium alloy such as Zircaloy-2 or Zircaloy-4.
The zirconium barrier serves as a reaetion site for gaseous impurities and fission products whieh have penetrated through craeks or defects in the inner layer 23 and protects the substrate portion oF the cladding from contact and reaction with such impurities and fission products, and minimizes the oeeurrenee of localized stress and cladding failure by pellet-eladding-interaetion.

~2~ 7Z~ 24-NT-04~71 In an exemplary embodiment, the zirconium barrier layer is about three mils thick and the inner layer of Zircaloy-2 is about one mil thick. Each of the inner layers and barrier layers should be continuous, that is be free of perforations or seams.
The composite cladding o~ the nuclear fuel element of this invention has an unalloyed zirconium barrier metallurgically bonded to the substrate and an inner layer metallurgically bonded to the zirconium barrier. ~etallographic examination shows that there is sufficient cross-diffusion between the substrate and the zirconium barrier and between the zirconium barrier and the inner layer to form metallurgical bonds, but insufficient cross diffusion to significantly reduce the purity of the zirconium barrier itself. Also, from Figure 2, it is apparent that the zirconium barrier could be termed a "buried" zirconium barrier since it is sandwiched between the substrate and inner layer.
Non alloyed zirconium forming the barrier in the composite cladding is highIy resistant to radiation hardening and this enables the zirconium barrier, after prolonged irradiationl to maintain desirable structural properties such as yield strength and hardness at levels considerably lower than those of conventional zirconium alloys. In effect, the zirconium barrier does not harden as much as conventional zirconium alloys when subjected to irradiation and this, together with its initially low yield strength, enables the zirconium barrier to deform plastically and relieve pellet-induced stresses in the fuel elemeIlt duxing power transients. Pellet-induced stresses in the fuel element can be brought about, for example, by thermal expansion and/or swelling of the pellets of nuclear fuel at re~ctor operating temperatures (300 C to 350 C) so that the pellet comes into contact with the cladding.

~z~7~ ~ 24~NT-04471 It has ~urther been discovered -that a zirconium barrier of the order of preferably about 5 percent to 15 percent of the thickness of the cladding and a particularly preferred thickness of 10 percent of the cladding bonded to the external substrate of zirconium alloy provides stress reduction and a barrier effect sufficient to prevent failures in the composite cladding.
A preferred embodiment according to the principles of this invention comprises "low oxygen sponge"
grade zirconium as the buried barrier layer, although higher purity "crystal bar zirconiuml' grade and lower purity "reactor grade sponge" zirconium are also possible.
The residual impurity content of the sponge zirconium serves to impart special properties to the zirconium barrier. Generally, there are at least about 1000 parts ppm impurities in sponge zirconium and preferably less than 4200 ppm. Oxygen is preferably kept within the range of about 200 to about 1200 ppm. Other typical impurity levels are listed as follows: aluminum -75 ppm or less; boron - 0.4 ppm or less; cadmium -0.4 ppm or less; carbon- 270 ppm or less; chromium -200 ppm or less; cobalt - 20 ppm or less; copper -50 ppm or less; hafnium - 100 ppm or less; hydrogen -25 ppm or less; iron - 1500 ppm or less; magnesium -20 ppm or less; manganese - 50 ppm or less; molybdenum -50 ppm or less; nickel ~ 70 ppm or less; niobium -100 ppm or less; nitrogen - 80 ppm or less; silicon -120 ppm or less; tin 50 ppm or less; tungsten -100 ppm or less; titanium - 50 ppm or less; and uranium -3.5 ppm or less.
Sponge zirconium is typically prepared by reduction with elemental magnesium at elevated temperatures at atmospheric pressure. The reaction takes place in an inert atmosphere such as helium or argon.
~nother preferred embodiment comprises a buried barrier layer -formed of crystal bar zirconlum.

.

~ 24-NT-04471 Crystal bar zirconium is produced by the vapor-phase decomposition of zirconium tetraiodide~ Crystal bar zirconium is more expensive, but has fewer impurities and displays a greater resistance to radiation damage than sponge zirconium.
The use of a buried layer of zirconium also results in desirable fabricati~n benefits. Tube reduction to finishing tends to remove some material from the inside of the tube. sy burying the more expensive unalloyed zirconium in the wall or the tube, the manufacturing losses are of the less-costly zirconium alloy, resulting in 100 percent utilization of the unalloyed zirconiumr Further any manufacturing defects on the inside of the tube are in the less-critical inner layer, assuring continuity of thezirconium barrier which is typically only a few mils thick. Further, an alloyed zirconium inner layer is better than a container with an inner layer of unalloyed zirconium, since zirconium alloys are more readily machined, honed, etc., than the softer unalloyed zirconium.
However, if it is desired to have the buried layer at the inside surface of the cladding container, the inner layer of zirconium alloy can be removed by etching after the tube is finished to its final dimensions.
Among the zirconium alloys serving as suitable alloy substrates are Zircaloy-2 and Zircaloy-4.
Zircaloy-2 has on a weight basis about 1.5 percent tin;
0.12 percent iron; 0.09 percent chromium and 0.005 percent nic~sel and is extensively employed in water-cooled reactors. Zircaloy-4 has less nickel than Zircaloy-2, but contains slightly more iron than Zircaloy-2. The composite cladding used in the nuclear fuel elements of this invention can be fabricated by any of the following methods.

,, ~2~ 2~
24~NT-04471 ~ n one method, a tube of unalloyed zirconium barrier material is inserted into a hollow billet of the material selected to be the substrate, a tube of the material selected to be the inner layer is inserted into the zirconium barrier tube, and then the assembly is subjected to explosive bonding of the tubes to the billet. The composite is extruded using conventional tube shell extrusion at elevated temperatures of about 1000F to 1400F (about 538C to 760 C). Then the extruded composite is subjected to a process involving conventional tube reduction until the desired size of cladding is achieved. The relative wall thickness o~ the hollow billet, the zirconium barrier tube and the inner layer tube are selected to give the desirable thickness ratios in the finished cladding tube.
In another method, a tube of unalloyed zirconium barrier material is inserted into a hollow billet of the material selected to be the substrate, a tube of the material selected to be the inner layer is inserted into the tube of the zirconium barrier, and then the assembly is subjected to a heating step ~such as at 750 C for 8 hours) under compressive stress to assure good metal-to-metal contact and diffusion bonding between the tubes and the billet. The diffusion bonded composite is extruded using conventional tube shell extrusion such as described above in the immediately preceding paragraph.
Then the extruded composite is subjected to a process involving conventional tube reduction until the desired size of cladding is achieved.
The foregoing process of fabricating the composite cladding of this invention gives economies over other processes used in fabricating cladding such as electroplating or vapor deposition. The invention includes a method of producing a nuclear fuel element comprising making a composite cladding container which is open at one end, the cladding container having a substrate, an unalloyed zirconium barrier metallurgically bonded to the inside 37~ ~

of the substrate, and an inner layer metallurgically bonded to the inside surface of the zirconium barrier, filling the composite cladding container with nuclear fuel material and leaving a cavity at the open endl inserting a nuclear fuel material retaining means into the cavity, applying the cavity in communication with the nuclear fuel, and then bonding the end of the cladding container to said enclosure to form a tight seal therebetween.
The present invention offers several advantages promoting a lony operating life for a nuclear fuel element, including the reduction of chemical interaction of the cladding, the minimization of localized stress on the zirconium alloy substrate portion of the cladding, the minimization of stress corrosion on the zirconium alloy substrate portion of the cladding, and the reduction of the probability of a splitting failure occurring in the zirconium alloy substrate as a result of pellet-cladding-interaction. The invention further prevents direct contact between the fission products and the zirconium alloy substrate and the occurrence of localized stress on the zirconium alloy substrate. The invention thus prevents the initiation or propagation of stress corrosion cracks in the alloy substrateO
There are particular advantages to using unalloyed zirconium as a buried barrier. The unalloyed zirconium is quite ductile and in the event stress corrosion cracks initiate in the inner layer, -their propagation can be effectively stopped in the zirconium barrier. It is believed that the radius of curvature at the end of a crack at the unalloyed zirconium is much larger than in zirconium alloys; thereby re~uiring much higher stress levels for propagation. The unalloyed zirconium is also less susceptible to iodine stress corrosion~ further ~! ~ tending to inhibit crack propagation.
., ~Z~ 7 24-NT-04~71 An importan-t property of the composite cladding of this invention is that the foregoing improvements are achieved with no additional neutron penalty. Such a ~ladding is readily accepted in nuclear reactors since the cladding would have no eutectic formation during a loss-of-coolant accident or a reactivity insertion accident involving the dropping of a nuclear control rod.
Further, the composite cladding has no heat transfer penalty in -that there is no thermal barrier to transfer of heat such as results in the situation where a separate foil or liner is inserted in a fuel element. Also, the composite cladding of this invention is inspectable by conventional, non-destructive testing methods during various stages of fabrication and operation.
As will be apparent to those skilled in the art, various modifications and changes may be made in the invention described herein. It is accordingly the intention that the invention be construed in the broadest manner within the spirit and scope as set forth in the accompanying claims.

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Claims (31)

The embodiments of the invention in which an exclusive property or privilege is claimed are defined as follows:

1. A nuclear fuel element comprising:
(a) a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium, and mixtures thereof; and (b) an elongated composite cladding container enclosing said core comprising an exterior substrate, a continuous zirconium barrier formed of unalloyed zirconium metallurgically bonded on the inside surface of the substrate, said zirconium barrier comprising from about 1 percent to about 30 percent of the thickness of the cladding container, and a continuous inner layer metallurgically bonded on the inside surface of the zirconium barrier, said inner layer comprising from about 1 percent to about 10 percent of the thickness of the cladding container.

2. The nuclear fuel element of Claim 1 in which the exterior substrate is formed of a zirconium alloy.

3. The nuclear element of Claim 1 in which the inner layer is formed of a zirconium alloy.

4. The nuclear element of Claim 1 in which the zirconium barrier comprises from about 5 percent to about 15 percent of the thickness of the cladding container.

5. The nuclear fuel element of Claim 1 in which the unalloyed zirconium barrier is sponge zirconium.

6. The nuclear fuel element of Claim 1 in which the unalloyed zirconium barrier is crystal bar zirconium.

7. The nuclear fuel element of Claim 1 in which the nuclear fuel material is selected from a group consisting of uranium compounds, plutonium compounds, and mixtures thereof.

8. The nuclear fuel element of Claim 1 in which the nuclear fuel material is comprised of uranium dioxide.

9. The nuclear fuel element of Claim 1 in which the nuclear fuel material is a mixture comprised of uranium dioxide and plutonium dioxide.

10. A nuclear fuel element comprising:
(a) a central core of a body of nuclear fuel material selected from the group consisting of compounds or uranium, plutonium, thorium, and mixtures thereof; and (b) an elongated composite cladding container enclosing said core and including an outer portion formed of a materialselected from the group of zirconium and zirconium alloys for forming a substrate, a continuous zirconium barrier formed of unalloyed zirconium metallurgically bonded on the inside surface of the substrate, said zirconium barrier comprising from about 1 percent to about 30 percent of the thickness of the cladding container, and a continuous inner layer formed of zirconium or zirconium alloys metallurgically bonded on the inside surface of the zirconium barrier, said inner layer comprising from about 1 percent to about 10 percent of the thickness of the cladding container.

11. The nuclear fuel element of Claim 10 in which the zirconium barrier comprises from about 5 percent to about 15 percent of the thickness of the cladding container.

12. The nuclear fuel element of Claim 10 in which the unalloyed zirconium barrier is sponge zirconium.

13. The nuclear fuel element of Claim 10 in which the unalloyed zirconium barrier is crystal bar zirconium.

14. The nuclear fuel element of Claim 10 in which the nuclear fuel material is selected from the group consisting of uranium compounds, plutonium compounds, and mixtures thereof.

15. The nuclear fuel element of Claim 10 in which the nuclear fuel material is comprised of uranium dioxide.

16. The nuclear fuel element of claim 10 in which the nuclear fuel material is a mixture comprising uranium dioxide and plutonium dioxide.

17. A composite cladding container for fuel material for service in nuclear reactors comprising a zirconium alloy outer portion forming a substrate, a continuous zirconium barrier formed of unalloyed zirconium metallurgically bonded on the inside surface of the substrate, said zirconium barrier comprising from about 5 percent to about 15 percent of the thickness of the cladding container and a continuous inner layer formed of zirconium alloy metallurgically bonded on the inside surface of the metal barrier, said inner layer comprising from about 1 percent to about 10 percent of the thickness of the cladding container.

18. A composite cladding container according to claim 17 in which the unalloyed zirconium barrier comprises from about 5 percent to about 15 percent of the thickness of the cladding container.

19. A composite cladding container according to claim 17 in which the unalloyed zirconium barrier is sponge zirconium.

20. A composite cladding container according to claim 17 in which the unalloyed zirconium barrier is crystal bar zirconium.

21. A nuclear fuel element which comprises an elongated composite cladding container including an outer portion formed of a material selected from he group of zirconium and zirconium alloys forming a substrate, a continuous zirconium barrier formed of unalloyed zirconium metallurgically bonded on the inner surface of the substrate, said zirconium barrier comprising from about 5 percent to about 15 percent of the thickness of the cladding container, and a continuous inner layer formed of zirconium metallurgically bonded on the inside surface of the metal barrier, said inner layer comprising from about 1 percent to about 10 percent of the thickness of the cladding container, a central core of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium, and mixtures thereof, disposed in and partially filling said container and leaving an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container, and a nuclear fuel material retaining means positioned in the cavity, said cladding container enclosing said core so as to leave a gap between said core and said cladding during use in a nuclear reactor.

22. The nuclear fuel element of Claim 21 wherein the unalloyed zirconium barrier is sponge zirconium.

23. The nuclear fuel element of Claim 21 wherein the unalloyed zirconium barrier is crystal bar zirconium.

24. In a hollow composite cladding container for nuclear fuel for use in a nuclear reactor comprising an outer substrate of zirconium alloy and an inner liner of zirconium alloy, the improvement comprising a barrier layer of unalloyed zirconium metallurgically bonded between the outer substrate and the inner liner.

25. A container as recited in Claim 24 wherein the unalloyed zirconium barrier layer has a thickness in the range from about 1 percent to about 30 percent of the thickness of the cladding container.

26. A container as recited in Claim 24 wherein the unalloyed zirconium barrier layer has a thickness in the range from about 5 percent to about 15 percent of the thickness of the cladding container.

27. A container as recited in Claim 24 wherein the unalloyed zirconium barrier layer is formed of sponge zirconium.

28. A container as recited in Claim 24 wherein the unalloyed zirconium barrier layer is formed of crystal bar zirconium.

29. A container as recited in claim 24 wherein the inner liner of zirconium alloy has been removed by chemical etching so that the composite cladding container comprises an inside surface of unalloyed zirconium.

30. A container as recited in claim 29 wherein the unalloyed zirconium is sponge zirconium.

31. A container as recited in claim 29 wherein the unalloyed zirconium is crystal bar zirconium.