Patent Publication Number: US-4649023-A

Title: Process for fabricating a zirconium-niobium alloy and articles resulting therefrom

Description:
CROSS-REFERENCE TO RELATED INVENTIONS 
     A process for fabricating products from Zircaloy alloys, and the resultant products having high temperature aqueous environment corrosion resistance, are described in a related application of the present inventors, application Ser. No. 571,122 filed on Jan. 13, 1984 which is a continuation of application Ser. No. 343,787, filed Jan. 29, 1982 assigned to the assignee of the present invention. This related application describes a process where Zircaloy alloy material is beta-treated and subsequently cold worked, with annealing temperatures of about 500° to 600° C. used between cold working steps, to produce a product having a microstructure adjacent a surface of an article having a substantially random distribution of reduced size precipitates. 
     A process for fabricating thin-walled tubing from zirconium alloys containing 1-2.5 percent by weight niobium, and resulting products having corrosion resistance, are described in a related application of the present inventors, application Ser. No. 693,546, filed on even date herewith assigned to the assignee of the present invention. That related application describes a process where zirconium-niobium alloys are treated to produce thin-walled tubing resistant to high temperature aqueous environments, preferably for use in nuclear reactor components. 
     BACKGROUND OF THE INVENTION 
     The present process relates to fabrication of articles, either as intermediate or final products, from a zirconium alloy containing, in addition to zirconium, between about 0.5 to 2.0 percent by weight of niobium, up to about 1.5 percent by weight tin, and a minor amount, up to about 0.25 percent by weight of a third element such as iron, chromium, molybdenum, vanadium, copper, nickel and tungsten. The resultant products, produced according to the present invention, have a particular microstructure that enables the material to resist corrosion in high temperature steam environments. 
     In the development of nuclear reactors, such as pressurized water reactors and boiling water reactors, fuel designs impose significantly increased demands on all of the core components, such as cladding, grids, guide tubes, and the like. Such components are conventionally fabricated from the zirconium-based alloys, Zircaloy-2 and Zircaloy-4. Increased demands on such components will be in the form of longer required residence times and thinner structural members, both of which cause potential corrosion and/or hydriding problems. These increased demands have prompted the development of alloys that have improved corrosion and hydriding resistance, as well as fabricability and mechanical properties that are typical of the conventional Zircaloys. One such class of materials are the zirconium alloys containing zirconium, niobium, tin, and a third element, such as a zirconium alloy containing 1 percent by weight niobium, 1 percent by weight tin, and at least 0.15 percent by weight iron. The only technical limitation which could prevent the utilization of these alloys is that they ordinarily exhibit optimum corrosion and hydriding resistance only after they have been rapidly quenched from high temperatures in the beta-treatment range (˜850°-950°  C.) and then aged for long time periods such as 8-24 hours at about 500°-600° C. This type of treatment cannot readily be applied to any of the required core components, and thus the usefulness of these alloys is severely limited. 
     SUMMARY OF THE INVENTION 
     Intermediate and final products are formed from a zirconium-niobium alloy containing 0.5 to 2.0 percent niobium, up to 1.5 percent tin and up to about 0.25 percent of a third alloying element, such as iron, the article having excellent corrosion resistance to aqueous environments at elevated temperatures, and hydriding resistance. 
     The alloys are beta-quenched and subsequently treated at lower than normal annealing temperatures and fabricating steps. In formation of tubing, for example, the beta-quenched alloy is extruded at a temperature at or below 650° C. and between subsequent cold working steps, the article is subjected to cold working anneals at a temperature at or below 650° C. The resultant article is given a final anneal at a temperature also below 650° C., and preferably around 500° C. The resultant alloy articles have a microstructure that comprises fine precipitates, less than 800 Å, homogeneously dispersed throughout the matrix. 
     These and other aspects of the present invention will become apparent when read in conjunction with the following description and the drawings, in which: 
    
    
     DESCRIPTION OF THE DRAWINGS 
     FIG. 1 is a process flow diagram of an embodiment of the present invention; 
     FIGS. 2A, B, C and D show transmission electron microscopy photomicrographs illustrating the typical precipitate distribution and size observed in a stressrelief-annealed tubing produced according to the present process; and 
     FIGS. 3A, B, C and D show transmission electron microscopy photomicrographs illustrating the typical precipitate distribution and size observed in a fully annealed tubing produced according to the present process. 
    
    
     DETAILED DESCRIPTION 
     The fabrication of intermediate and final products from a zirconium-niobium-tin alloy containing a minor amount of a third alloying element is effected according to the present invention with the production of resultant articles that exhibit excellent corrosion resistance and resistance to hydride formations. 
     The alloys that are the subject of the present process are zirconium alloys which contain, in percentages by weight, 0.5 to 2.0 percent niobium, up to 1.5 percent tin, and up to about 0.25 percent of a third alloying element such as iron, chromium, molybdenum, vanadium, copper, nickel and tungsten. While tin and the third alloying element are present in an amount up to the percentages by weight listed, the minimum amount present would be that sufficient to give the desired corrosion resistance in the articles produced therefrom. Preferably the alloy contains about 1 percent by weight niobium and about 1 percent by weight tin. It is also preferred that the level of the third alloying element be about 0.1 weight percent. A particularly useful alloy has been found to be a zirconium alloy containing 1 percent by weight niobium, 1 percent by weight tin, and 0.1 percent by weight iron. 
     The present invention produces articles of the zirconium alloy wherein the alloying elements are homogeneously dispersed throughout the zirconium in a finely divided state. The particles, homogeneously dispersed, are of an average particle size below 800 angstroms (Å), and preferably the particle size is below about 500 angstroms. 
     The alloys of the present invention are first subjected to a beta-treatment by heating the alloy to about 950°-1000° C. and water-quenching the same to a temperature below the alpha+beta to alpha transus temperature. In formation of tubing, to which the following description is directed, the billet is then prepared for extrusion by drilling an axial hole along the center line of the billet, machining the outside diameter to desired dimensions, and applying a lubricant to the surfaces of the billet. The billet diameter is then reduced by extrusion at a lower than conventional temperature, below about 700° C., through a frustoconical die and over a mandrel. A beta-anneal of the extruded tube shell may then be effected, depending upon the alloy, by heating to about 850°-1050° C., followed by rapid cooling. The billet may then be cold worked by pilgering, at a source of primary fabrication, to reduce the wall thickness and outside diameter. This intermediate production is called a TREX (Tube Reduced Extrusion), which may then be sent to a tube mill for fabrication by cold working, intermediate low temperature annealing, and a final anneal under the fabricating steps of the present invention to produce tubing. 
     In accordance with the present process, a zirconium alloy was processed into thin-walled tubular cladding, as illustrated in the process outline flow diagram of FIG. 1, from a zirconium alloy containing, by weight, 1 percent niobium, 1 percent tin, and 0.1 percent iron, with the balance being zirconium. A zirconium alloy ingot, of the composition given in Table IV, was broken down by conventional treatment and beta-forged from a twelve inch diameter to about six inch diameter billet (Step 1). The six inch billet was beta-treated (Step 2) by holding the ingot in a furnace at about 968°-996° C. (1775°-1825° F.) for a period of fifteen minutes and then water-quenching the billet. At this point, the beta-treated billet was machined, bore-holed, and inspected in preparation for extrusion. The hollow zirconium alloy billet was then heated to about 649° C. (1200° F.) and extruded (Step 3) to a hollow tube having an outside diameter of 2.5 inches and a wall thickness of 0.43 inch. 
     The extruded hollow tube was beta annealed (Step 4) at 954° C. (1750° F.) for a period of fifteen minutes in preparation for the first cold working step. 
     The beta-annealed extrusion was pilgered in Step 5 to a TREX having an outside diameter of 1.75 inches and a wall thickness of 0.3 inch. The TREX was then subjected to a cold working anneal (Step 6) for an 8 hour period at 600° C. (1112° F.). Following the cold working anneal of the TREX, the same was cold pilgered (Step 7) to a tube shell having an outside diameter of 1.25 inch and a wall thickness of 0.2 inch. The tube shell was subjected to another cold working anneal (Step 8) at about 580° C. (1076° F.) for 8 hours. The annealed tube shell was again cold pilgered (Step 9) to a tube shell having an outside diameter of 0.70 inch and a wall thickness of 0.07 inch. A further cold working anneal (Step 10) was effected on the tube shell for 8 hours at about 580° C. (1076° F.). A third pilgering of the tube shell was effected, (Step 11) to produce a tube having an outside diameter of 0.423 inch and a wall thickness of 0.025 inch. This tube was then subjected to a final anneal (Step 12) for a period of 7.5 hours at about 480° C. (896° F.). 
     Nuclear fuel cladding produced according to the present process was subjected to out-reactor and in-reactor performance tests and illustrate corrosion resistance and hydriding resistance significantly superior to that of Zircaloy-4 nuclear fuel cladding. 
     Stress relieved sections of tubing prepared according to the present process were corrosion tested in a static autoclave in 316° C. (10.6 MPa) water, 360° C. (18.6 MPa) water and 427° C. (10.3 MPa) steam, with corrosion rate and hydriding data compared with Zircaloy-4. The results of the tests are listed in Table I (corrosion) and Table II (hydriding): 
     
                       TABLE I                                                     
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Linear Out-of-Pile Post-Transition Corrosion Rates Exhibited              
by the Zr--1.0 Nb--1.0 Sn--0.1 Fe Alloy and Reference                     
Zircaloy-4 Fuel Claddings                                                 
               Temperature                                                
                          Corrosion Rate                                  
Alloy          (°C.)                                               
                          (mg/dm.sup.2 /day)                              
______________________________________                                    
Zr--Nb--Sn--Fe 316        0.09                                            
&#34;              360        0.33                                            
&#34;              427        2.48                                            
Zircaloy-4     316        0.10                                            
&#34;              360        0.57                                            
&#34;              427        6.05                                            
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                       TABLE II                                                    
______________________________________                                    
Out-of-Pile Hydrogen Pickup Rates Exhibited by the Zr--1.0                
Nb--1.0 Sn--0.1 Fe Alloy and Reference Zircaloy-4                         
Fuel Claddings                                                            
           Temperature                                                    
                      % of       Hydriding Rate                           
Alloy      (°C.)                                                   
                      Theoretical*                                        
                                 (mg/dm.sup.2 /day)                       
______________________________________                                    
Zr--Nb--Sn--Fe                                                            
           316        4.4 ± 2.2                                        
                                 0.6                                      
&#34;          360        6.2 ± 2.2                                        
                                 2.3                                      
&#34;          427        27.7 ± 5.8                                       
                                 79.5                                     
Zircaloy-4 316        12.2 ± 9.0                                       
                                 2.2                                      
&#34;          360        12.7 ± 4.4                                       
                                 8.6                                      
&#34;          427        51.6 ± 13.1                                      
                                 122.4                                    
______________________________________                                    
 *-x±-                                                                 
 
    
     Inspection of the above tables reveal that the Zr-Nb-Sn-Fe fuel cladding of the present invention exhibits lower thermal post-transition corrosion rates at all three temperatures, and hydrogen pickup rates that are a factor of 3-4 lower than that of Zircaloy-4. 
     Fuel rods using cladding fabricated according to the present invention were irradiated in the BR-3 pressurized water reactor located in Mol, Belgium. Post-irradiation examinations (PIE) performed after one cycle exposure indicate in-reactor corrosion performance superior to Zircaloy-4, consistent with the out-of-pile results. A comparison of the results is listed in Table III: 
     
                       TABLE III                                                   
______________________________________                                    
Oxide Thicknesses Measured on BR-3 Fuel Rods                              
After One Irradiation Cycle (7.5 months exposure)                         
                        Oxide Thickness                                   
                                     Mean                                 
Alloy      Elevation (mm)                                                 
                        (μm)      (μm)                              
______________________________________                                    
Zr--Nb--Sn--Fe                                                            
            81          2.56                                              
&#34;          287          3.55                                              
&#34;          467          3.97         3.468                                
&#34;          666          3.96                                              
&#34;          848          3.30                                              
Zircaloy-4  81          2.63                                              
&#34;          272          4.00                                              
&#34;          416          4.23                                              
&#34;          675          4.22         3.656                                
&#34;          869          3.20                                              
______________________________________                                    
 
    
     The present process provides uniform distribution of very fine precipitate particles in the microstructure of the alloy. The microstructure of a zirconium alloy tubing containing 1.0 percent niobium, 1.0 percent tin and 0.1 percent iron that has been stress-relief-annealed with a final anneal of 7.5 hours at 480° C. (896° F.) is illustrated in FIGS. 2A, B, C and D. FIGS. 3A, B, C and D show the microstructure of a tubing, of the same alloy of FIG. 2, that has been fully annealed at about 590° C. (1094° F.) for an 8 hour period. 
     In fully annealed tubing produced according to the present invention, formed from a zirconium alloy containing 1.0 percent niobium, 1.0 percent tin and 0.1 percent iron, which was fully annealed for 8 hours at 600° C. (1112° F.), the average particle size of the precipitates was 330 Å in average diameter, with a number density of 4×10 14  /cm 3 . This precipitate size represents a relatively fine dispersion, especially when compared to an average diameter of 3000 Å for particle sizes in a conventionally processed Zircaloy-4 alloy. 
     While the advantages of the present invention have been described in connection with a zirconium alloy containing 1 percent niobium, 1 percent tin, and 0.1 percent iron, in fabricating a thin-walled tubing, having a wall thickness less than about 0.04 inch, for use as a fuel cladding, it is believed that the present invention is also applicable to other alloys as hereinbefore defined and to fabrication of other articles such as sheet material or plates. The present invention is thus believed to be applicable to zirconium alloys containing niobium, tin and iron in other percentages by weight as aforedescribed and to zirconium alloys containing niobium, tin and other third alloying elements, such as chromium, molybdenum vanadium, copper, nickel and tungsten, in substitution for or in addition to iron, with total third alloying element being in an amount of less than about 0.25 percent by weight. 
     In the formation of sheet material or plates according to the present process, the rolling temperature is reduced to about 650° C. or less, with a number of hot rolling passes made below that temperature as needed, with a final anneal also below that temperature to provide articles with the homogeneously dispersed precipitates of less than about 800 angstroms throughout the material. 
     
                       TABLE IV                                                    
______________________________________                                    
Ingot Chemistry of Zirconium Alloy Containing 1 Weight                    
Percent Niobium, 1 Weight Percent Tin, and 0.1 Weight                     
Percent Iron, Processed in Accordance with the Invention                  
Spec.           Top of Ingot                                              
                           Bottom of Ingot                                
______________________________________                                    
Nb    0.9-1.1       1.01       0.96                                       
Sn    0.9-1.1       0.97       0.94                                       
Fe    0.09-0.11     0.10       0.09                                       
O     1000-1600 ppm 1370       1490                                       
Zr    BALANCE                                                             
Al    75            &lt;35        &lt;35                                        
B     0.5           &lt;0.2       &lt;0.2                                       
Cd    0.5           &lt;0.2       &lt;0.2                                       
C     120           60         70                                         
Cl    20            &lt;5         &lt;5                                         
Co    20            &lt;10        &lt;10                                        
Cu    50            &lt;25        &lt;25                                        
Cr    50            &lt;50        &lt;50                                        
H     20            5          6                                          
Hf    75            40         39                                         
Pb    50            &lt;50        &lt;50                                        
Mn    50            &lt;25        &lt;25                                        
N     50            29         28                                         
Ni    50            &lt;35        &lt;35                                        
Si    80            &lt;50        &lt;50                                        
Ti    50            &lt;40        &lt;40                                        
W     100           &lt;25        &lt;25                                        
U.sub.T                                                                   
      3.5           0.5                                                   
U-235 0.025         0.0045                                                
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