Patent Number: 042960746
Section: summary

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to the removal of stainless steel, zirconium or zirconium alloy cladding materials from a metallic element selected from the group consisting of uranium, thorium and mixtures thereof. The present invention is particularly applicable to the selective destructive removal of cladding materials from nuclear fuel elements containing fissile or fertile fuels such as uranium, thorium and mixtures thereof. 2. Prior Art There are numerous types of nuclear power fuel elements. The present invention is particularly applicable to those nuclear fuel elements of the solid type which comprise a body or core of thermal neutron fissionable uranium, thorium or mixtures thereof which may be present in an elemental state or alloyed with zirconium, niobium or other low cross section materials which are clad in a low cross section corrosion resistant material such as stainless steel, zirconium, or zirconium alloys. Nuclear power fuel elements generally contain two types of nuclear fuel material, both of which are valuable. It is essential that the fuel element contain a fissionable nuclear fuel material such as uranium isotopes U 233 or U 235. Fuel elements also contain nuclear fuel materials that are not originally fissionable, but which can be converted to fissionable material and are, therefore, said to be fertile or potential nuclear fuel materials. For example, U 238 is a fertile material often present in fuel elements in considerable amounts. In some instances as much as 99.3% of the uranium content may be present in the form of U 238 in the case of an unenriched element. During the course of the use of the element in a power reactor, the fissionable material such as U 233 and U 235 releases neutrons. Some of the neutrons are trapped by the fertile but unfissionable U 238 present in the element and the U 238 eventually becomes Pu 239 which is fissionable. In the same way, thorium which is a fertile but unfissionable material, absorbs neutrons to become U 233 which is fissionable and useful as nuclear fuel material. Fuel elements of the solid type, with which the present invention is particularly applicable, deteriorate due to radiation damage long before the useful content of the fissionable material is used. At the same time, radioactive fission products accumulate in the fuel element. Some are gases and others are solid; however, each is objectionable in reducing the efficiency of the reactor as a whole and each exert some part in the destruction or disintegration of the fuel element. More particularly, many of the fission products have a high neutron capture cross section thus reducing the total amounts of neutrons available for production of thermal energy. In addition, the gaseous fission products build up pressure within the cladding material which can result in permanent structural damage to the elements and possibly to the reactor. Since these deleterious effects occur at a time when only a small fraction of the fissile values have been burned by the fission process and since the unburned fuel is too valuable to be wasted, it advantageously is reprocessed to render it fit for reuse. None of the heretofore known methods for recovering fuel and fertile uranium or thorium from such elements has been completely satisfactory. One method, for recovering unburned fissile and fertile fuel values from solid neutron irradiated fuel elements, involves dissolution of the cladding and the fuel followed by a liquid-liquid solvent extraction process in which an aqueous nitrate feed solution containing said values is selectively extracted by contact with an organic aqueous immiscible extractant. An example of a solvent extraction process for recovering uranium values, for example, is found in U.S. Pat. No. 2,848,300. A major disadvantage of aqueous dissolution of cladding, however, is that large aqueous feed volumes containing dissolved metals must be carried through the solvent extraction process. This in turn leads to a large radioactive waste volume requiring expensive waste storage and handling. In addition, the solutions generally are highly corrosive and have a high chloride content. Removal of chloride from the aqueous feed must be accomplished prior to solvent extraction for recovery of the thorium or uranium. In an attempt to reduce the volume of high level radioactive waste pollution, various other methods have been proposed, such as separately dissolving the cladding material in concentrated sulfuric acid thus making the fuel core available for ready dissolution in a nitric acid solution. However, a cladding material such as stainless steel is relatively passive in sulfuric acid and even when it does react, there is a high probability that cross contamination between the decladding solution and the core solution will result, thus further complicating the problem of recovering the fuel. U.S. Pat. No. 2,827,405 suggests a method of desheathing fuel rods of uranium metal bars by puncturing the sheath to expose the uranium core at a plurality of points. The rod then is reacted with steam at an elevated temperature to oxidize the uranium and break the bond between the sheath and the uranium. The fuel is recovered as an oxide requiring expensive processing to convert it back to a metal. Another method suggested in U.S. Pat. No. 2,962,371 comprises reacting the element at an elevated temperature with essentially pure anhydrous hydrogen for a time sufficient to hydride the cladding so that it falls from the core. This invention however, is concerned with zirconium-clad fuel elements although it is suggested that it is also applicable to elements that are clad in alloys of zirconium. Another process for recovering the core of a zirconium-clad fuel element is disclosed in U.S. Pat. No. 3,007,769. The process comprises immersing the clad element in a substantially neutral solution of ammonium fluoride to effect the dissolution of the zirconium and separate the neutron fissionable material values from the solution. U.S. Pat. No. 3,089,751 suggests a process for the selective separation of uranium from ferritic stainless steels. In accordance with the process disclosed therein, a nuclear fuel element consisting of a core of uranium clad in a ferritic stainless steel is heated to a temperature in the range of 850.degree. C. to 1050.degree. C. for a period of time sufficient to render the cladding susceptible to intergranular corrosion. The heated element is then cooled rapidly to a temperature range of 850.degree. C. to 615.degree. C. and then to about room temperature. The cooled element then is contacted with an aqueous nitrate solution to selectively and quantitatively dissolve the uranium from the core. Gas phase processes for effecting the dissolution of fuel or the cladding material are disclosed in U.S. Pat. Nos. 3,149,909; 3,156,526 and 3,343,924. The problem of handling and containing gaseous fuel, however, is even greater than that for liquid phase processes. U.S. Pat. No. 3,929,961 suggests a method of treating a nuclear fuel element enclosed in a stainless steel metal sheath which comprises disposing the fuel element with a portion thereof in an induction coil, subjecting the induction coil to a radio frequency magnetic field to induce local induction heating of the metal sheath sufficient to raise the temperature of the portion of the sheath within the coil to its melting temperature and effect local melting therein. The fuel element is moved axially relative to the induction coil with continued heating to rupture the metal sheath. The fuel values are subsequently recovered by dissolution. Thus it is seen that the prior art processes either convert the fuel to an oxide or at some point require liquid or gas phase processing with all of the problems associated therewith. SUMMARY OF THE INVENTION The present invention provides a method of treating an assembly comprising an element selected from the group consisting of uranium, thorium and mixtures thereof encased in a cladding of stainless steel or a zirconium alloy to separate the selected element from the cladding. In accordance with the present method, the assembly is subjected to a scoring or perforating step to expose the selected element. Thereafter, the assembly is exposed to hydrogen at a pressure of from about 0.5 to 2.0 atmospheres (360 to 1400 torr) and a temperature of 450.degree. C. to 680.degree. C. to form a hydride of the element. The hydride, having a greater volume than the elemental metal, expands, rupturing the cladding material. Thereafter, the temperature is further increased to a range of from about 700.degree. C. to 900.degree. C. to decompose the hydride back to the element. The dehydriding results in the element being in the form of friable particulates such that after at least one and preferably after about three successive hydriding-dehydriding steps, the selected element is readily recoverable from the cladding material, utilizing conventional mechanical separation techniques such as sieving or the like. In a particularly preferred embodiment of the invention, during the hydriding step, the temperature is cycled between about 500.degree. C. and 650.degree. C. to enhance the completeness of the hydriding and maximize the removal or evolution of any volatile compounds contained within the assembly. The present invention is particularly applicable to the treatment of irradiated fuel elements for the recovery of fissionable and fertile values therefrom.