Patent Number: 046876294
Section: description

DETAILED DESCRIPTION OF THE INVENTION In the following description, like reference characters designate like or corresponding parts throughout the several views of the drawings. Also in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like are words of convenience and are not to be construed as limiting terms. IN GENERAL Referring now to the drawings, and particularly to FIG. 1, there is shown a nuclear fuel assembly, generally designated 10 for a boiling water nuclear power reactor (BWR), in which the improvement of the present invention is incorporated. The fuel assembly 10 includes an elongated outer tubular flow channel 12 that extends along substantially the entire length of the fuel assembly 10 and interconnects an upper support fixture or top nozzle 14 with a lower base or bottom nozzle 16. The bottom nozzle 16 which serves as an inlet for coolant flow into the outer channel 12 of the fuel assembly 10 includes a plurality of legs 18 for guiding the bottom nozzle 16 and the fuel assembly 10 into a reactor core support plate (not shown) or into fuel storage racks, for example in a spent fuel pool. The outer flow channel 12 generally of rectangular cross-section is made up of four interconnected vertical walls 20 each being displaced about ninety degrees one from the next. Formed in a spaced apart relationship in, and extending in a vertical row at a central location along, the inner surface of each wall 20 of the outer flow channel 12, is a plurality of structural ribs 22. The outer flow channel 12, and thus the ribs 22 formed therein, are preferably formed from a metal material, such as an alloy of zirconium, commonly referred to as Zircaloy. Above the upper ends of the structural ribs 22, a plurality of upwardly-extending attachment studs 24 fixed on the walls 20 of the outer flow channel 12 are used to interconnect the top nozzle 14 to the channel 12. For improving neutron moderation and economy, a hollow water cross 26 extends axially through the outer channel 12 so as to provide an open cruciform inner channel for subcooled moderator flow through the fuel assembly 10 and to divide the fuel assembly into four, separate, elongated compartments 28. The hollow water cross 26 is mounted to the angularly-displaced walls 20 of the outer channel 12. Preferably, the outer, elongated lateral ends of the water cross 26 are connected such as by welding to the structural ribs 22 along the lengths thereof in order to securely retain the water cross 26 in its desired central position within the fuel assembly 10. Also, the water cross 26 has a lower flow inlet end 30 and an opposite upper flow outlet end 32 which each communicate with the inner channel for providing subcoolant flow therethrough. Disposed within the channel 12 is a bundle of fuel rods 34 which, in the illustrated embodiment, number sixty-four and form an 8.times.8 array. The fuel rod bundle is, in turn, separated into four mini-bundles thereof by the water cross 26. The fuel rods 34 of each mini-bundle, such being sixteen in number in a 4.times.4 array, extend in laterally spaced apart relationship between an upper tie plate 36 and a lower tie plate 38. The fuel rods in each mini-bundle are connected to the upper and lower tie plates 36,38 and together therewith comprise a separate fuel rod subassembly 40 within each of the compartments 28 of the channel 12. A plurality of grids or spacers 42 axially spaced along the fuel rods 34 of each fuel rod subassembly 40 maintain the fuel rods in their laterally spaced relationships. The upper and lower tie plates 36,38 of the respective fuel rod subassemblies 40 have flow openings defined therethrough for allowing the flow of the coolant fluid into and from the separate fuel rod subassemblies. Also, coolant flow paths provide flow communication between the fuel rod subassemblies 40 in the respective separate compartments 28 of the fuel assembly 10 through a plurality of openings 44 formed between each of the structural ribs 22 along the lengths thereof. Coolant flow through the openings 44 serves to equalize the hydraulic pressure between the four separate compartments 28, thereby minimizing the possibility of thermal hydrodynamic instability between the separate fuel rod subassemblies 40. FUEL RODS HAVING ANNULAR FUEL PELLETS WITH SINGLE ENRICHMENT AND DIFFERENT ANNULUS SIZES Turning now to FIGS. 2 through 4, there is shown one of the fuel rods 34 useful in the BWR fuel assembly of FIG. 1 which has been fabricated in accordance with the present invention. The fuel rod 34 includes a hollow cladding tube 46 being sealed at its opposite ends by upper and lower end plugs 48,50 and a plurality of fuel pellets 52 contained in the tube and retained in a stack form therein by a plenum spring 54 being disposed between the upper end plug 48 and the stack of pellets. The pellets 52 are composed of fissile material, such as uranium dioxide, having a single enrichment of U-235, and, in addition, each fuel pellet 52 has an annular configuration. As clearly illustrated in FIG. 3, the central void or annulus 56 of any one of the annular fuel pellets 52 can be of a different diametric size than that of the others in order to vary the enrichment loading from pellet to pellet. For example, in FIG. 3 each annulus 56a-c of each respective pellet 52a-c is of a different size. In such manner, graduation of enrichment loading can be provided between the stacked annular fuel pellets 52 in the axial direction along the fuel rod 34. Such fuel rod configuration is particularly useful in BWR fuel assemblies. Since the core axial power distribution in BWRs is bottom-peaked, the fuel in the top half of the core is under-utilized. Thus, it would be advantageous to place the annular pellets in the top portion of each fuel rod. This increases the energy utilization of the fuel since the kw/kg of uranium is increased. The flattening of the core axial power shape produced by placing the annular pellet fuel in the top portion of each assembly improves the operating margins to the fuel thermal limits by reducing the axial power peaking and maximum hot spot power peaking. Referring now to FIGS. 5 through 7, a comparison can be made between the fuel bundle of a prior art BWR fuel assembly 10a, being partially represented schematically in FIG. 5, and the fuel bundle of a BWR fuel assembly 10b fabricated in accordance with the present invention, being partially represented schematically in FIG. 6. The fuel bundle of the prior art fuel assembly 10a includes fuel rods 34a which each contain fuel pellets that all have a solid configuration. To obtain the desired enrichment loading across the fuel assembly 10a, there are five different enrichments used in the solid fuel pellets of the prior fuel rods 34a. These fuel rods 34a which have the different enrichments are identified by letters A-E within respective circles and the values of the respective enrichments are listed in FIG. 7. In contrast to the prior fuel assembly 10a of FIG. 5, the fuel bundle of the BWR fuel assembly 10b of the present invention in FIG. 6 includes fuel rods 34b which each contain fuel pellets composed of fissile uranium having a single U-235 enrichment. The single enrichment chosen for the fuel rods 34b in FIG. 6 is the predominant enrichment of the fuel rods 34a of FIG. 5, that being 3.236 w/o U-235. All of the fuel rods 34b in FIG. 6 having a uranium loading intended to match the loading of the predominant enrichment of the fuel rods 34a in FIG. 6 contain solid fuel pellets, as symbolized by an empty circle in FIGS. 6 and 7. All of the other fuel rods 34b, with the exception of one, contain annular fuel pellets. There are three categories of pellets with different respective void sizes, being specified in FIG. 7 and symbolized by respective circles containing a dot, s smaller circle and an "x" in a smaller circle in FIGS. 6 and 7. The one exception is the fuel rod symbolized by the circle containing the letter A. This rod has not been replaced with one containing annular pellets because the void percent or fraction would be excessive at approximately forty-five percent. Instead, a PWR enrichment could be used, such as a standard 2.6 w/o U-235 which would result in an approximately thirty-one percent void fraction which is a practical annulus size. Therefore, with the exception of one, all solid pellet-containing fuel rods 34a in fuel assembly 10a having enrichments below the predominant enrichment are replaced by annular pellet-containing fuel rods 34b having a single enrichment and different annular or void percent sizes. In such manner, the desired uranium loading of the annular pellet-containing fuel rods can be provided by properly sizing the annulus of their pellets. In the illustrated example, as explained in FIG. 7, the predominant enrichment is 3.236 w/o U-235. In the complete prior art BWR fuel assembly 10a of FIG. 5, there are forty-one such fuel rods. These remain as solid pellet-containing fuel rods in the new fuel assembly 10b of FIG. 6. Thus, in view that there are sixty-four fuel rods 34b in the assembly 10b, a majority of them will contain fuel pellets having a solid configuration. Consequently, the remainder of the fuel pellets 34b, a minority of the total number of fuel rods, have an annular configuration. As seen in FIG. 7, there are three different void or annulus sizes, which provides the desired graduation of enrichment loading between the solid and annular fuel pellets of the respective majority and minority of fuel rods. The value of the single U-235 enrichment is at the level of the maximum or predominant enrichment loading of the prior art fuel assembly 10a of FIG. 5. Consistent PHOENIX 2-D lattice calculations were made comparing the solid pellet fuel (FIG. 5) and annular pellet fuel (FIG. 6) designs from a neutronics behavior standpoint. Comparisons were made for radial pin-wise power distribution within the bundle, bundle reactivity, reactivity coefficients and defects. The results of these calculations indicated the following: (a) very comparable radial pin-wise power distributions; (b) a void coefficient benefit for the annular pellet design; (c) a shutdown margin benefit for the annular pellet design; and (d) essentially identical Doppler coefficient and defect. In sum, the neutronics behavior of the annular pellet design is expected to be acceptable in all areas. It also has comparable fuel cycle cost to the solid pellet design. While the present invention has been described with reference to fuel rods for a BWR fuel assembly, it is equally applicable to fuel rods intended for use in other types of nuclear reactors. It is thought that the invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely a preferred or exemplary embodiment thereof.