Abstract:
Methods and apparatus for improving fretting resistance of zirconium alloy components formed into a shape for use in a nuclear reactor are disclosed in which at least a portion of the outer surface of a component is reacted with material selected from the group consisting of carbon, nitrogen, oxygen and combinations of the foregoing at a temperature below about 700° C. to form a wear resistant layer on the surface of the component.

Description:
FIELD OF THE INVENTION 
     This invention relates to wear resistant nuclear fuel assembly components and, more particularly, to surface hardened, zirconium-based alloy components such as nuclear fuel cladding tubes, spacer elements and channels and to methods of making such components. 
     BACKGROUND OF THE INVENTION 
     The operating environment within a nuclear reactor, including a pressurized water reactor (PWR) and a boiling water reactor (BWR) is particularly hostile. A considerable effort has been expended in the nuclear reactor industry to arrive at materials which are able to withstand the combination of mechanical, thermal, chemical (corrosion) and radiation effects encountered in that environment. At the present time, only a few types of zirconium-based alloys are considered to be acceptable. Those alloys are generally identified as Zircaloy materials. The Zircaloy materials are used for nuclear fuel cladding tubes, spacer elements and channels within the reactor. 
     As a result of experience with long term operation and multiple reloads of nuclear fuel elements, it has been found that certain operating conditions arise which tend to reduce the energy output per unit of fuel (&#34;burn-up&#34;) obtainable and thereby affect operating costs and efficiencies in an undesirable manner. For example, during the operation of nuclear reactors, metal debris which may be present in the reactor can be carried by the cooling water and can impact upon fuel assembly components. The repeated interaction of such debris and the fuel assembly components (such as fuel cladding tubes, channels or spacer elements) can result in fretting (rubbing) damage to the components. 
     While the Zircaloy materials gradually have been optimized with respect to corrosion resistance requirements within a reactor, the fretting wear resistance of Zircaloy, as well as resistance to combined effects of fretting wear and subsequent corrosion have not been optimized. The need to improve fretting wear resistance should not result in any undesirable compromise with respect to corrosion resistance. 
     Zircaloy materials until relatively recently have been treated prior to insertion into a reactor by autoclaving techniques to apply a relatively thin coating (0.5 microns) of oxide material to improve their general operational characteristics. Such an oxide coating has not been found to be resistant to fretting wear or fretting induced corrosion but rather has been found to be subject to being damaged or worn away by the fretting action of the debris. Thereafter, fretting corrosion will occur at the fretting site in the area where the oxide layer has been removed. The corrosion layer which forms is also susceptible to debris fretting wear and will be removed by action of the water and debris. Eventually, after successive cycles of wear and corrosion occur, a hole ultimately can be produced in the base metal itself. In the case of fuel cladding, such a hole will result in the unwanted release of radioactive material and radiation into the cooling water, and if it is in excess of reactor operating limits, will require an untimely shutdown of the reactor for replacement of fuel elements. 
     One approach to avoiding such problems is to improve the wear resistance of the Zircaloy fuel assembly components, especially at their lowermost ends where debris is most often present. 
     Outside of the field of nuclear reactors, it has been proposed that layered structures incorporating whiskers of nitrides, carbides or carbonitrides into Group IVb metals, which include zirconium, will provide a hardened surface condition. (See, e.g., U.S. Pat. Nos. 4,915,734; 4,900,525; and 4,892,792.) Furthermore, dispersions of hard substances in a binder metal, such as zirconium oxide dispersed in iron, cobalt or nickel (see U.S. Pat. No. 4,728,579) have been described as providing improved wear resistance for cutting tools. In addition, in the field of cutting tools, it has been observed (see U.S. Pat. No. 3,955,038) that, before applying an oxide coating such as zirconium oxide to a binder metal, imposition of an intermediate layer such as a carbide or a nitride of a metal in the fourth to sixth subgroups of the periodic system (including zirconium) may impede undesirable diffusion of metal from the substrate into the formed oxide layer. 
     It is understood that the foregoing structures are formed by processes which require temperatures that are incompatible with maintaining the metallurgical state of Zircaloy components to be used in a reactor. That is, such Zircaloy components typically are heat treated to produce particular grain structures and stress relieved conditions in the finished product. 
     In order to preserve the desired metallurgical structure and properties, it is necessary that any additional wear resistant layer be applied utilizing temperatures that are below temperatures at which the desired properties will be changed. 
     For example, in the case of stress relieved cladding of a type used in pressurized water reactors, post annealing processing temperatures should be maintained below about 500° C. In the case of cladding, spacers or channels which have been treated to produce a recrystallized condition (as typically used in boiling water reactors), post annealing processing temperatures should be maintained below about 700° C. and, in some cases, below about 600° C. in order to avoid undesired metallurgical changes in the respective components. 
     STATEMENT OF THE INVENTION 
     In accordance with one aspect of the present invention, an improved nuclear fuel element of the type including a zirconium alloy tube, which may or may not be separated from a central core of nuclear fuel material by a barrier layer, has a hard, wear resistant layer produced on at least a portion of the outside surface of the tube by reacting the outside surface of the zirconium alloy with material selected from the group consisting of carbon, nitrogen, oxygen and combinations of the foregoing, the reaction occurring below about 700° C. at a temperature which is sufficiently low to avoid unwanted changes (e.g., annealing which changes desired grain structures) near the outer surface of the tube. 
     In accordance with a further aspect of the present invention, a method of improving fretting resistance of zirconium alloy components for use in a nuclear reactor comprises reacting at least a portion of the surface of the component with material selected from the group consisting of carbon, nitrogen, oxygen and combinations of such materials, the reaction temperature being maintained below about 700° C. at a level to produce a hard, wear resistant layer on the surface without producing undesirable metallurgical changes in the vicinity of the surface. 
     In accordance with yet another aspect of the present invention, an improved structural component formed of a zirconium alloy for use in a nuclear reactor, the improvement comprising a wear resistant layer produced on at least a portion of a surface of the component which contacts cooling fluid in the reactor, the wear resistant layer being produced by reacting the surface with material selected from the group consisting of oxygen, carbon, nitrogen and combinations of such material, the reaction temperature being maintained below about 700° C. so as to maintain a metallurgical state in the vicinity of the surface which existed prior to the formation of the layer. 
    
    
     BRIEF DESCRIPTION OF THE DRAWING 
     The foregoing and other features of the present invention will be more readily apparent from the following detailed description and drawings of illustrative embodiments of the invention in which: 
     FIG. 1 is an elevation view, partially in section, of a typical fuel assembly for a light water nuclear power reactor, the assembly being foreshortened in height and partially broken away for convenience and clarity; and 
     FIG. 2, drawn to a different scale than FIG. 1, is a sectional view of a fuel rod employed in the assembly of FIG. 1 incorporating one version of the present invention. 
    
    
     DETAILED DESCRIPTION 
     Referring to FIG. 1, a typical 14×14 fuel bundle assembly is indicated generally by the reference numeral 10. Fuel assembly 10 includes an upper tie plate 12 and a lower tie plate 14, capturing at opposite ends a plurality of (e.g., 176) fuel rods 13 of tubular shape. A plurality of guide tubes 11 are secured to upper tie plate 12 and to lower tie plate 14. A plurality of grid spacers 15 (e.g., eight several of which are shown) are disposed along the length of fuel rods 13 at locations between tie plates 12 and 14 and form cells, as is well known, through which fuel rods 13 and guide tubes 11 extend. A lowermost one 15&#39; of the grid spacers is illustrated as a debris-resistant spacer of the type shown and described in U.S. Patent No. 4,849,161 of Brown et al. 
     Each of fuel rods 13 encloses a stack of fissionable fuel pellets 16. Pellets 16 in each stack are maintained in close proximity to each other by means of a spring 17 disposed between an upper end of the rod 13 and the uppermost one of pellets 16. A lower end cap 18 of each fuel rod is in close proximity to but spaced away from the upper portion of lower tie plate 14 to take into account the expected linear growth of rods 13 in the operation of the reactor. The total height from the bottom of lower tie plate 14 to the top of the uppermost pellet 16 (i.e., the top of the active fuel) may, for example, be a few inches less than twelve feet. 
     Lower tie plate 14 may be entirely conventional or may comprise a debris resistant design above the lower core support plate 22. 
     Coolant supplied from below the lower tie plate 14 may be expected to carry debris of the type noted above. As the coolant (water) flows upwardly, some debris will be intercepted and can drop down below the plate 22. Some amount of debris, however, can impact upon the exterior surface of fuel rods 13, spacers 15 and, in the case of BWR assemblies, enclosing channel structure, particularly at the lower ends thereof. In the case where a fuel assembly does not include a debris catching device or screen, an even greater amount of debris may be expected to impact upon the exterior surface of fuel rods 13, spacers 15 and other components in the fuel assembly. 
     Fuel rod cladding, spacers or channels may be manufactured in accordance with the present invention to include a method for final hardening treatment of the surface(s) of the components which are exposed to coolant water and accompanying debris. 
     A process in accordance with this invention, which is applicable to treating either one or more surfaces of zirconium alloy material, whether it is in strip, sheet or tubular form, will produce components having an extended service life. It is preferable, in order to retain the desired metallurgical condition of the component, that hardening operations be conducted without exceeding a temperature of, for example, 500°-700° C., the specific limit temperature being dependent upon the function and/or nature of the component and the environment (PWR or BWR) in which the component is to be employed as noted above. 
     In accordance with the one aspect of the present invention, a method of improving fretting resistance of zirconium alloy components used in nuclear reactors comprises placing the alloy component into a closed vessel, introducing a gaseous atmosphere containing acetylene and raising the temperature within the vessel to a temperature in the range of 450° C. to 500° C., preferably to 475° C .for a period of 8 hours so as to produce a hard, wear resistant carbide layer on the surface of the alloy component. The carbide layer will be of the order of one micron thick (typically slightly less), and will exhibit a hardness acceptable for resisting fretting in a nuclear reactor environment. The carbide layer which is produced in this manner is black in color. 
     Cladding and flat Zircaloy material treated with acetylene as described above to produce a wear resistant layer thereafter was subjected to a standard autoclaving test to determine whether the wear resistant layer was also resistant to waterside corrosion. The results of testing for corrosion resistance alone were that such corrosion resistance was equal to or better than that obtained for components which were the same in all respects except that they were not provided with the wear resistant layer. 
     In accordance with a further aspect of the invention, a method of improving fretting resistance of zirconium alloy components used in nuclear reactors comprises placing at least a portion of an alloy component to be treated in a molten bath containing one or more cyanide salts having an effective melting point less than 500° C. for a period of four to twelve hours to produce a wear resistant layer on the alloy component. In one particular arrangement, a mixture of 60% sodium cyanide and 40% potassium cyanide (weight percent) was used to produce a hard, wear resistant layer two microns thick which was black in color. X-ray fluorescence analysis of the layer so formed indicated the presence of oxide material, although it is normally to be expected that a carbonitride layer is formed by means of a cyanide bath. The resulting surface layer was found to be of increased hardness and wear resistance as compared to the Zircaloy itself. 
     In accordance with a further aspect of the invention, a method of improving fretting resistance of zirconium alloy components used in nuclear reactors comprises placing at least a portion of an alloy component to be treated in a molten bath containing one or more carbonate salts having an effective melting point less than 500° C. for a period of four to twelve hours to produce a wear resistant oxide layer on the alloy component. In one particular arrangement, a mixture of 50% lithium carbonate and 50% potassium carbonate (weight percent) was used to produce a hard, wear resistant layer two microns thick which was black in color. X-ray fluorescence analysis of the layer so formed indicated the presence of oxide material. The resulting surface layer was found to be of increased hardness and wear resistance as compared to the Zircaloy itself. 
     In accordance with a further aspect of the present invention, a method of improving fretting resistance of zirconium alloy components used in nuclear reactors comprises lacing the zirconium alloy component in a closed furnace, introducing an air or oxygen atmosphere into the furnace, and raising the temperature within the furnace to a temperature level between about 400° C. and 500° C. (preferably less than 475° C.) for a period of between about forty and about eighty hours to grow a hard, wear resistant oxide layer on the surface of the alloy component. The oxide layer so formed typically will be in the range of between about 0.7 microns and 1.4 microns thick (that is, of the order of one micron) and will e black in color. 
     The grown oxide layer exhibits a desired resistance to corrosion which might occur in a nuclear reactor environment while, at the same tie, providing an increased fretting resistance. 
     In each of the foregoing examples, the wear resistant layer was produced without compromising the corrosion resistance characteristics of the zirconium alloy component. Particular attention must be paid to the maximum reaction temperature in each case to avoid undesirably changing the metallurgical structure of the components. To that end, temperatures less than about 500° C. were employed in each example for cladding in the stress relieved condition. Appropriate reaction temperatures (e.g., temperatures in the vicinity of about 600° C. to 700° C.) for components other than stress relieved cladding, such as recrystallized cladding, channel and spacers, similarly should be observed to avoid undesired secondary effects on metallurgical characteristics such as grain size or structure. 
     As is shown in FIG. 2, typical fuel cladding 13 constructed in accordance with the present invention includes an outer wear resistant layer 30 over at least a portion of the length thereof (typically at least the lowermost portion). The thickness of layer 30 is exaggerated in FIG. 2 compared to the dimensions of other elements. The inner surface of cladding 13 may be separated from fuel pellets 16 by a barrier layer 32 (as is known). The invention is also useful in connection with non-barrier cladding, spacers and channels as noted above. 
     It should be recognized that various modifications may be made in the apparatus and processes described above without departing from the true scope of the present invention, which is pointed out in the following claims.