Patent Number: 053496183
Section: description

DETAILED DESCRIPTION According to the principles of this invention, a boiling water nuclear reactor, preferably used for the generation of electricity, includes novel fuel rods and novel assemblies of these fuel rods that improve the reactor performance in comparison to conventional BWRs that are in use for electric power generation. In particular, as described more completely below, hydride fuel pellets are included at selected axial and radial positions within the core so as to improve the neutron moderation while increasing the total length of fuel rods in the core. Herein "hydride fuel" refers to a material which includes hydrogen and at least one fissionable material among its constituents. The fissionable material includes at least one of the fissile isotopes uranium-233, uranium-235, plutonium-239 and plutonium-241. The hydride fuel functions simultaneously as a fuel and as a moderator. The location and concentration of the hydride fuel is adjusted within a fuel rod and within a fuel assembly, as described more completely below, to achieve a more uniform power density. When the hydride fuel is substituted for oxide fuel in undermoderated regions of the core, the fuel itself provides additional moderation. Herein, "oxide fuel" refers to the composition of fuel that is made of the oxide of at least one fissionable material, such as uranium and plutonium. In the presently operating BWRs, the oxide fuel is, typically, uranium oxide (UO.sub.2). Neutrons are moderated better in the core regions where the hydride fuel is substituted for the oxide fuel and the multiplication constant increases in these regions. Therefore, more fissions occur so that enhanced power production is achieved in these regions of the core. Consequently, the moderation provided by the hydride fuel in undermoderated regions flattens the power distribution across the core. This more uniform power distribution permits reduction of the use of burnable poisons and of control rods for power shaping, as used in the prior art BWRs, which in turn improves the fuel utilization, the reactor availability, and the reactor safety. Additional improvement in the fuel utilization, reactor availability and reactor safety is obtained by substituting hydride fuel rods for water rods in the interior of the BWR fuel assemblies, thus increasing the overall length of fuel pellets in the assembly without compromising the moderation capability. According to the principles of this invention, the basic geometry of the BWR core including the fuel rods and fuel assemblies is not changed. Rather, the fuel pellets used in the fuel rods and the axial and radial composition of the fuel within a fuel assembly are modified to provide enhanced performance. While axial and radial dimensions are referred to herein with respect to a BWR core, these dimensional references are illustrative only of the principles of this invention and are not intended to limit the invention to the particular dimensions described. More generally, the axial dimension is a dimension in a first direction and the radial dimension is a dimension in a second direction where the second direction is orthogonal to the first direction. In view of this disclosure, those skilled in the art can implement the invention in a wide variety of geometries. In a preferred embodiment, the BWR uranium oxide fuel pellets of the prior art are utilized within the novel fuel rods of this invention along with uranium-zirconium hydride pellets. The uranium oxide is slightly enriched in the fissile isotope uranium-235 (.sup.235 U). However, as is known to those skilled in the art, a mixed uranium-plutonium oxide may also be utilized in place of uranium oxide. As described above, a novel fuel rod of this invention includes hydride fuel pellets which in the preferred embodiment are made of uranium-zirconium hydride (U-ZrH.sub.x). The number of hydrogen atom per zirconium atom in the uranium-zirconium hydride fuel, denoted above by the x subscript, ranges from zero to about two, and is preferably about 1.6 to 1.7. The weight percent (wt %) of the uranium in the uranium-zirconium hydride fuel ranges from about 30 wt % to about 60 wt %, and is preferably about 45 wt % uranium. The uranium used for the uranium-zirconium hydride fuel is low-enriched uranium (LEU). The enrichment of the uranium ranges from about 2% to about 6%. In the preferred embodiment of this invention, all the hydride fuel pellets used in a given BWR core use the same weight percent uranium and the same ratio of hydrogen to zirconium atoms. In other embodiments of this invention, the uranium weight percent and the hydrogen-to-zirconium atom ratio vary with the location of the hydride fuel within the fuel rod, with the location of the fuel rod within the assembly, as well as with the location of the fuel assembly within the core. Such variations provide an optimal match between the hydride fuel, the oxide fuel, and the variation in the void fraction in the coolant. The enrichment of the uranium in the hydride fuel typically varies with the location of the hydride fuel in the core. The number of enrichment levels used within one core ranges from one to about ten, and is preferably between 2 and 4. This is similar to the number of enrichment levels of the oxide fuel used in state-of-the-art BWRs. The selection of the optimal location for placement of hydride fuel pellets within the core, of the hydrogen-to-zirconium atom ratio, of the weight percent uranium, and of the uranium enrichment for each hydride fuel pellet can be determined by those skilled in the art by using state-of-the-art computer codes for core design and optimization. (Herein, state-of-the-art computer codes refers to those codes in common use by nuclear engineers for design of the prior art BWR fuel described above.) Accordingly, the following examples are illustrative of the principles of this invention, but the examples are not intended to limit the invention to the specific embodiments disclosed herein. In a first embodiment, a first plurality of fuel rods in a fuel assembly includes both oxide and hydride fuel pellets and a second plurality of fuel rods in the fuel assembly include only hydride fuel pellets. For example, BWR fuel assembly 300 (FIG. 3A) includes a total of 81 fuel rods in a 9.times.9 array. Seventy-two of fuel rods 328, the first plurality, contain both oxide fuel pellets and hydride fuel pellets (mixed hydride-oxide fuel rods) while nine of fuel rods 330, the second plurality, contain only hydride fuel pellets (all-hydride fuel rods). All-hydride fuel rods 330 are arranged within a 3.times.3 array in which a mixed hydride-oxide rod 328 is used to separate all-hydride rods 330. FIG. 3B is another conceptual view of fuel assembly 300. Fuel assembly 300 contains 81 fuel rods in channel 326. The coolant, i.e., water, enters at bottom 320 of channel 326 and flows through space 332 between the fuel rods in axial direction 340 and exits at top 310. As the water flows up through fuel assembly 300, the water is heated and begins to boil at about the location shown by dotted line 350. The steam volume fraction, i.e., the void fraction increases as the water flows upward of dotted line 350 in axial direction 340 and reaches about seventy percent at top 310. Thus, upper region 346 of fuel assembly 300 contains steam, which for neutron moderation acts like a void. In mixed hydride-oxide rods 328, the oxide fuel pellets are, in this embodiment, located below line 360 and the hydride fuel pellets are located above line 360, where line 360 is between onset of boiling line 350 and assembly top 310. The exact preferred location for line 360 can be determined by those skilled in the art by using the state-of-the-art computer codes for core design and optimization. The void fraction in the vicinity of line 360 ranges between 0% to 70%, and is preferably between about 30% to 50%. Thus, fuel rod 328 has a total fueled length in a first direction. A first fueled length is occupied by the oxide fuel pellets. A second fueled length is occupied by the hydride fuel pellets. The second fueled length is smaller than the portion of the fueled length of fuel rod 328 surrounded by a coolant with a non-zero void fraction, i.e, portion 346 of fuel rod 328 above line 350. The compensation in neutron moderation achieved by using the hydride fuel pellets in the region of the core having a void fraction is described more completely below. There are various possibilities for arranging mixed hydride-oxide fuel rods 328 and all-hydride fuel rods 330 within fuel assembly 326. In 9.times.9 fuel assembly 400A (FIG. 4A) the nine innermost fuel rods are all-hydride fuel rods 330, whereas the rest of the fuel rods are mixed hydride-oxide fuel rods 328. Notice that fuel rods 330 are in the interior of fuel assembly 400A and fuel rods 328 surround fuel rods 330. In 9.times.9 fuel assembly 400B (FIG. 4B), only five fuel rods are all-hydride fuel rods 330, whereas the rest of the fuel rods are mixed hydride-oxide fuel rods 328. The innermost fuel rod is an all-hydride fuel rod 330 and in the eight rods surrounding the innermost fuel rod that form a square, the corner fuel rods of the square are also all-hydride fuel rods 330. Notice again that fuel rods 330 are in the interior of fuel assembly 400B and fuel rods 328 surround fuel rods 330. In 8.times.8 BWR fuel assembly 400C (FIG. 4C), eight of the fuel rods are all-hydride fuel rods 330 and the rest of the fuel rods are mixed hydride-oxide fuel rods 328. In a second embodiment, a first plurality of fuel rods in a fuel assembly are mixed hydride-oxide fuel rods, a second plurality of fuel rods in the fuel assembly are all-hydride fuel rods, and a third plurality of fuel rods in the fuel assembly are all-oxide fuel rods. For example, BWR fuel assembly 500 (FIG. 5) includes a total of 81 fuel rods in a 9.times.9 array. Fifty-two fuel rods 328, the first plurality, contain both oxide and hydride fuel pellets (mixed hydride-oxide fuel rods) while nine fuel rods 330, the second plurality, contain only hydride fuel pellets (all-hydride fuel rods). Twenty fuel rods 574 contain only oxide fuel pellets (all-oxide fuel rods). All-hydride fuel rods 330 are arranged within a 3.times.3 array in which a mixed hydride-oxide fuel rod 328 is used to separate all-hydride fuel rods 330. Many other embodiments are possible using the novel fuel rods of this invention. One embodiment uses all-hydride fuel rods located at the inner region of the fuel assembly with the rest of the fuel rods being all-oxide fuel rods. Such fuel assemblies are identical to those illustrated in FIGS. 4A, 4B, 4C, and 5 except that mixed hydride-oxide fuel rods 328 are replaced by all-oxide fuel rods 574. Another embodiment uses mixed hydride-oxide fuel rods 328 throughout the assembly. Yet another embodiment uses only mixed hydride-oxide fuel rods 328 and all-oxide fuel rods 574. In this embodiment, the hydride-oxide fuel rods are preferably surrounded by the all-oxide fuel rods. Of course, any of the fuel assemblies of this invention may contain one or more water rods or rods containing solid moderators, as proposed for BWRs which do not use hydride fuel. Also, the all-hydride and mixed hydride-oxide fuel rods may contain burnable poisons, such as the burnable poisons used in state-of-the-art BWRs. As is known to those skilled in the art, the onset of boiling in a BWR fuel assembly and the volume fraction occupied by voids at different locations along the assembly first direction vary with the reactor power level, coolant flow rate, coolant inlet temperature, control-rods position, location of burnable poisons, level of fuel burnup, as well as other core design and operating conditions. Typically, multi-dimensional coupled neutronic and thermal hydraulic computer codes are used to determine the optimal composition of a particular fuel rod in a particular fuel assembly during a particular cycle of the reactor. In this embodiment, the fuel rod contains hydride fuel from or above the point of onset of boiling for that particular fuel rod to the top of the fueled region of the fuel rod, and typically from a point where the void fraction is in the range of 30% to 50% to the top of the fueled region. Alternatively, as described above, the entire fuel rod may contain only hydride fuel. Thus, the hydride fuel may occupy from zero to one hundred percent of the fueled length in a fuel rod. An important aspect of the invention is that, unlike prior art designs that traded moderation for fuel volume, the fuel rods of this invention provide improved moderation in regions of the core that are undermoderated and simultaneously maintain the fuel volume. Several alternative embodiments for the axial distribution of the oxide fuel and hydride fuel within a fuel rod are illustrated in FIGS. 6A through 6G. Fuel rod 574 (FIG. 6A) contains only oxide fuel 636 while fuel rod 330 (FIG. 6F) contains only hydride fuel 634. Fuel rods 328A (FIG. 6B), 328B (FIG. 6C), 328C (FIG. 6D) and 328D (FIG. 6E) contain differing amounts of oxide fuel 636 and hydride fuel 634. The point of transition from oxide fuel 636 to hydride fuel 634 is determined for each BWR design and operating plan using coupled neutronic and thermal-hydraulic computer codes in use by those skilled in the art of BWR core design. Note that FIGS. 6A to 6G refer to the fueled region of the fuel rods. As is known to those skilled in the art, a fuel rod includes a plenum above the fuel pellets, and end caps. Fuel rod 328D (FIG. 6E) is referred to as a "predominantly-hydride fuel rod". Fuel rod 328D may be substituted in fuel assembly locations otherwise occupied by all-hydride fuel rods 330 (FIG. 6F). Oxide fuel pellets 636 are used at the lower part of the fuel rod where the neutron density and, therefore, power density tend to decline due to neutron leakage from the bottom of the core. The use of oxide fuel pellets 636 instead of hydride fuel pellets 634 in this relatively low neutron density lower core region will increase the power generated in this region relative to the power generated in the same region when using all-hydride fuel rods. In predominantly-hydride fuel rod 328D, the hydride fuel occupies at least the upper two thirds of the fueled length of the fuel rod. In a typical embodiment, the oxide fuel in fuel rod 328D is confined to the lower one eighth to one sixth of the rod where the neutron density is lower than in the upper core regions and where the steam volume fraction is negligible. Fuel rod 328E (FIG. 6G) is another embodiment of mixed hydride-oxide fuel rod. It uses oxide pellets 636 at its far top, in addition to oxide pellets 636 at its lower part. Typically, the volume of oxide pellets 636 at the top of fuel rod 328E is small, so that the use of oxide fuel at this highly voided core region will not significantly reduce the average multiplication constant of the upper part of the core. Solid moderator can be placed in the reflector right above the upper oxide fuel pellets to compensate for the lack of hydrogen in the oxide fuel. The purpose of placing the oxide fuel at the top is to increase the power density in the upper core region where the neutron density declines due to leakage. The fuel in all the fuel rods of this invention is contained within a cladding 638 (FIGS. 6A to 6G) made of, for example, the zirconium alloy "zircaloy". The mechanical design of the fuel rods for the innovative mixed hydride-oxide fuel rods 328 and all-hydride fuel rods 330 and the fuel assemblies containing these fuel rods are similar to the mechanical design of the all-oxide fuel rods and fuel assemblies used for typical prior art BWRs. Examples of such typical nuclear fuel assemblies were depicted and described in the above-identified patents of Venier et al., Lass, and Fritz et al. The fuel rods of this invention differ from the typical BWR fuel rods in their fuel composition and their cladding design. In any of the fuel rods of this invention containing a hydride fuel, gaseous hydrogen fills the small gap between the pellets and the fuel rod cladding, as well as the volume of the plenum above the pellets. The free hydrogen gas may hydrogenize the cladding material, or diffuse out through it. Hydrogenization of the cladding material may impair its mechanical integrity and is preferably prevented. If hydrogen gas permeates through the cladding, the hydrogen gas pressure inside the cladding drops, and part of the hydrogen dissociates from the hydride fuel pellets and become free hydrogen. Thus, if the hydrogen permeation rate is large, the dissociation of the hydride fuel may impair its neutron moderation ability. To avoid impairment of the mechanical integrity of the fuel rod cladding and lose of an unacceptably large fraction of the hydrogen of the hydride fuel, the hydride fuel pellets are surrounded by a sealed hydrogen permeation barrier. Several different embodiments are available for designing this hydrogen permeation barrier. In the preferred embodiment, the hydrogen permeation barrier design is the design proposed by Weitzberg for zirconium hydride moderator containing fuel rods in his above identified patent. FIGS. 7A, 7B, 7C, 7D and 7E illustrate a partial elevation view cross-section of fuel rods for a number of embodiments. In the embodiment illustrated in FIG. 7A, oxide fuel pellets 736, which are indicated by the capital letter "O", fill the lower part of fuel rod 328 while hydride fuel pellets 734, which are indicated by the capital letter "H", fill the upper part of fuel rod 328. There is no barrier between oxide fuel pellets 736 and hydride fuel pellets 734. Between fuel rod cladding 738 and pellets 734, 736 there is a barrier 744 to inhibit the permeation of hydrogen through cladding 738 to the water which surrounds this cladding. Barrier 744 is either a coating on internal side 738A of cladding 738 or a cylindrical sleeve with an inner surface adjacent pellets 734, 736 and an outer surface adjacent internal side 738A of cladding 738. The coating or sleeve is made of a material having (i) no hydrogenization, (ii) a high resistance to hydrogen permeation, and (iii) preferably a relatively low probability for neutron absorption. As used herein, "a high resistance to hydrogen permeation" is a resistance that prevents the loss of more than 50% and, preferably, of more than 10% of the hydrogen content of the hydride pellets during the residence time of these pellets in the core. Also as used herein, "a relatively low probability for neutron absorption" is measured relative to the probability for a neutron absorption in the fuel. As is known to those skilled in the art, the fission neutrons are preferably moderated and absorbed in the fuel. Different materials and material thicknesses are possible for hydrogen permeation barrier 744. One preferred embodiment is a layer of stainless steel, typically 0.05 to 0.1 mm in thickness. Such a barrier is proposed by Gylfe in his above-identified patent. Preferably, the stainless steel is oxidized to further improve its hydrogen retention capability. In another embodiment, hydrogen permeation barrier 744 is an oxidation layer on internal side 738A of zircaloy cladding 738. In still another embodiment, internal side 738A of zircaloy cladding 738 is a glass-enamel coating, as suggested in the above-identified Weitzberg patent. According to Weitzberg, glass-enamel coating metal cladding, about 0.08 mm thick, has been successfully utilized in SNAP reactors at temperatures up to 700.degree. C. Of course, cladding 738 may be made from a material, such as stainless steel, Which does not interact with hydrogen and which has a high resistance to hydrogen permeation. Such a cladding is used in TRIGA reactors. Although the simplest to implement, the latter approach is the most wasteful on neutrons, as the stainless steel has a higher neutron absorption probability than zirconium. In the embodiment illustrated in FIG. 7B, oxide fuel pellets 736 again fill the lower part of the fuel rod while hydride fuel pellets 734 are in the upper part. Hydrogen permeation cladding barrier 742 is a special material layer which encloses, i.e., clads, all hydride pellets 734 and is contained within fuel rod cladding 738. Hydrogen permeation barrier cladding 742 for hydride fuel pellets 734 extends, in one embodiment, above the upper most hydride fuel pellet to provide a plenum for accumulating gaseous fission products that are emitted from hydride fuel pellets 734. Hydrogen permeation barrier cladding 742 is one of stainless steel, a glass-enamel coated metal cladding, and any other material having high resistance to hydrogen permeation and low neutron absorption probability. In the embodiment of FIG. 7C, oxide fuel pellets 736 and hydride fuel pellets 734 are oriented as described above in the two previous embodiments. Each hydride fuel pellet 734 is surrounded by a hydrogen permeation barrier 746 in the form of a cladding or coating and the cladded (coated) hydride fuel pellet is contained within fuel rod cladding 738. In the embodiment of FIG. 7D, there is not a single boundary between hydride fuel pellets 734 and oxide fuel pellets 736. The transition from the completely oxide fuel region to completely hydride fuel region within the fuel rod is done gradually. Hydrogen permeation barrier 744, as described above for FIG. 7A, is utilized and that description is incorporated herein by reference. Alternatively, each hydride fuel pellet 734 is individually encased as illustrated in FIG. 7C. Of course, another alternative is to individually encase each hydride fuel pellet 734 in any transition regions, and each group of hydride fuel pellets 734 in fuel rod 728 are to use hydrogen permeation barrier 742 (FIG. 7B). This embodiment is a combination of embodiments 746 and 742. In fuel rod 330 (FIG. 7E) with all-hydride fuel pellets 734, any one of the embodiments 744, 742 and 746 may be used. In FIG. 7E, hydrogen permeation barrier 744 is illustrated. The density of hydrogen atoms per unit volume of U-ZrH.sub.1.6, the preferred fuel, is approximately 4.7.times.10.sup.22 hydrogen atoms per cubic centimeter of fuel. This is very close to the hydrogen density of 4.8.times.10.sup.22 in liquid water at the BWR operating temperature of about 280.degree. C. As in a typical BWR the volume of fuel in a fuel rod is nearly 62% of the volume of water which surrounds the fuel rod, the hydride fuel significantly increases the hydrogen content in the fuel assembly. Table 1 compares the relative amounts of hydrogen in selected elevations in the fuel assembly, when the preferred hydride fuel is used instead of the regular oxide fuel. The elevations considered feature the steam volume fractions given in Table 1. In the fuel assembly regions in which the void fraction is less than 60%, the use of the hydride fuel increases the hydrogen content to above its value in the reference BWR assembly lower section which does not experience boiling. At the outlet from the fuel assembly, where the void fraction is nearly 70%, the replacement of oxide fuel by the hydride fuel increases the hydrogen content by nearly three fold, bringing the hydrogen content to 90% of its value at the bottom of the reference BWR core. The effect of the improved moderation provided by the hydride fuel on the axial power profile is described more completely below. TABLE 1 ______________________________________ Assembly Hydrogen Content With The Hydride Fuel Void fraction (%) 0 40 50 60 70 ______________________________________ Relative to liquid water 1.6 1.2 1.1 1.0 0.9 Relative to oxide fuel 1.6 2.0 2.2 2.5 3.0 ______________________________________ When the preferred hydride fuel rod is used to replace the water rods or water rod segments within a prior art BWR fuel assembly, the hydrogen density in these rods and rod segments changes insignificantly. Thus, the hydride fuel provides practically the same improvement in moderation as provided by the special water rods, while adding nuclear fuel to these rods. The fuel addition converts these rods and rod segments to power producing, thus making a better utilization of the fuel assembly volume. In the above embodiment, a hydride fuel has been used to offset the problems that result from undermoderation associated with boiling or with uneven moderation resulting from the existence of water-gaps in-between the fuel assemblies in a BWR. The uranium-zirconium hydride fuel is illustrative only of the principles of this invention and is not intended to limit the invention to the particular fuel described. In view of this disclosure those skilled in the art will be able to use a variety of fissile isotopes containing fissionable materials and hydride materials to form a hydride fuel suitable for use in the fuel rod. Other hydride fuel materials include, but are not limited to, uranium and plutonium containing hydrides of thorium, titanium, cerium and yttrium. In another promising embodiment of this invention, the hydride fuel used in undermoderated regions of the core is uranium-thorium hydride (U-ThH.sub.x). The number of hydrogen atoms per thorium atom in this fuel material can range from about one to about three and is preferably about 2. Details about the properties and fabrication of the uranium-thorium hydride fuel U-ThH.sub.x can be found in U.S. Pat. No. 4,493,809 by Simnad (1/85). The uranium-thorium hydride fuel U-ThH.sub.x is incorporated in the BWR fuel rods as pellets, just as the uranium-zirconium hydride fuel U-ZrH.sub.x described above was incorporated. Relative to uranium-zirconium hydride, the uranium-thorium hydride fuel offers a better neutron economy, as the neutrons captured in thorium convert it to the fissile isotope uranium-233. The longer the uranium-thorium hydride fuel is irradiated in the BWR, the higher will be the U-233 concentration and the larger will be the contribution of this U-233 to the chain reaction, or reactivity of the core. So far there is only little experience in the fabrication of uranium-thorium hydride. Little is known also on the behavior of the uranium-thorium hydride fuel in BWR operating conditions. While the advantages of this invention have been demonstrated with respect to BWRs where the problems of undermoderation are most profound, the principles of this invention are applicable to undermoderation in other type of nuclear reactors. For example, in pressurized water reactors (PWRs) hydride fuel rods would replace control rod thimbles in those fuel assemblies which do not house control rods. The power attainable from the fuel assemblies of the present invention can be higher by nearly 10% than the power attainable from conventional PWR fuel assemblies. Alternatively, for the same power output, the PWR fuel assemblies of the present invention can reside longer in the PWR core and generate more energy than conventional PWR fuel assemblies. In above described embodiments of this invention, the hydride fuel pellet is homogenous in the sense that the fissionable material and the hydride are highly mixed. In another embodiment, the hydride fuel pellet is heterogeneous. One illustration of such an heterogenous fuel is a fuel made of zirconium hydride ZrH.sub.x into which small grains of uranium oxide UO.sub.2 are imbedded. A more extreme heterogeneity is illustrated in FIGS. 8 and 9 which show two-zone fuel pellets 850 having a cylindrical hydride inner zone 854 and a uranium oxide outer zone 852. Outer zone 852 is an annulus surrounding inner zone 854. Except for the specifics of the design of pellet 850, fuel rod 828A (FIG. 8) is the same as fuel rod 328 of FIG. 7A. Hydride materials suitable for inner zone 854 of two-zone pellet 850 includes but are not limited to zirconium hydride ZrH.sub.x. The volume fraction of inner hydride zone 854 is adjusted in accordance with the design requirements for fuel rod 828A. Outer region 852 of the two-zone pellet 850 may include any fissionable material. The replacement of the oxide fuel with hydride fuel in the upper part of fuel rods adds a significant amount of hydrogen to the BWR core regions in which the water substantially boils which in turn eliminates or, at least, highly reduces the undermoderation. Moreover, the addition of hydrogen to the otherwise undermoderated core regions is done along with the inclusion of fissile fuel with the solid moderator thus maximizing the weight of fuel and the total length of fuel rods which can be loaded into the BWR core while using BWR fuel assemblies that have a mechanical design similar to that of existing BWRs. Other prior art designs for alleviating the undermoderation compromised fuel volume for increased moderation. The improvement in the performance of a BWR made possible by the present invention is illustrated by considering an infinite array of BWR hydride fuel rods surrounded by water, to be referred to as the hydride fuel lattices, in comparison to an infinite array of BWR oxide fuel rods. The dimensions and composition of the oxide fuel lattices in the array of BWR oxide fuel rods are of a typical BWR design. The fuel pellet outside diameter is 1.0566 cm. The zircaloy cladding inside diameter is 1.0795 cm. The fuel rod outside diameter is 1.2522 cm. The distance between fuel rod centers is 1.6256 cm. The dimensions and composition of the hydride fuel lattices in the array of BWR hydride fuel rods are identical in all respects to the oxide fuel lattices with the exception to the fuel composition, which is U-ZrH.sub.1.6 having 45 wt. % uranium. The barrier to hydrogen permeation in the hydride fuel is taken to be a 0.1 mm layer of zirconium oxide. The oxide fuel has a characteristic beginning-of-life enrichment of 2.07%, whereas the uranium enrichment of the hydride fuel is taken as 2.07% in one embodiment and 3.0% in another embodiment. The characteristics of the three fuel rod lattices as a function of the void fraction in the water moderator surrounding the fuel rods in the lattice were determined using the WIMS lattice code using 69 energy groups of the WIMS cross-section library. The WIMS lattice code and the WIMS cross-section library is available from the Radiation Shielding Information Center of the Oak Ridge National Laboratory having Post Office Box 2008 Oak Ridge, Tenn., 37831-6362. Curve 1076 (FIG. 10) is multiplication constant K for the lattice of hydride fuel rods with 3.0% uranium enrichment. Curve 1080 is multiplication constant K for the lattice of hydride fuel rods with 2.07% uranium enrichment. Curve 1078 is multiplication constant K for the lattice of oxide fuel rods with 2.07% uranium enrichment. The curves in FIG. 10 represent the multiplication constant at different axial elevations in that part of the BWR core where boiling occurs. In the high void fraction regions of the core, the multiplication constant of the hydride fuel lattices, curves 1076 and 1080, exceeds that of the oxide fuel lattice, curve 1078. The higher the void volume fraction, the larger is the reactivity improvement offered by the hydride fuel. The multiplication constant of the hydride fuel lattices vary only slightly with the change in the steam volume fraction in comparison to the variation of the multiplication constant with the void fraction in the oxide fuel lattices. In view of the large amount of zirconium present in the hydride fuel lattices relative to the amount of zirconium in the oxide fuel lattices, it was surprising to find that the multiplication constant of U-ZrH.sub.1.6 fuelled lattices can be comparable and even higher than the multiplication constant of similar UO.sub.2 fueled lattices, when both the oxide and hydride fueled lattices use LEU of the same enrichment. As described above, only MEU and HEU was used so far for uranium-zirconium hydride fuel. This unexpected finding is due to the fact that in the high void fraction lattices, the improvement in the neutron moderation provided by the hydrogen of the hydride fuel more than compensates for the increase in the probability for neutron absorption in the zirconium. Table 2 compares the multiplication constant of a lattice of identical BWR fuel rods with and without hydride fuel. Six lattice compositions are considered for this comparison; they represent the fuel composition and steam volume fraction found in six elevations, or axial locations, in the core. The fuel design is as described in connection with FIG. 10. In a first embodiment, "all-oxide", the array contained uranium oxide fuel UO.sub.2 having an enrichment of 2.07%. In a second embodiment, "mixed hydride-oxide", the uranium oxide in the upper half of the fuel rod is replaced by the hydride fuel (U-ZrH.sub.1.6) in which the uranium is enriched to 3.0%. Table 2 shows that whereas with the all-oxide fuel the multiplication constant strongly declines towards the top of the fuel rod (or core), the use of hydride fuel brings the multiplication constant at the upper part of the core to practically its value at the bottom of the core. The improved moderation and increased reactivity resulting from the substitution of hydride fuel for oxide fuel in the upper region of the core make the axial power distribution more uniform without the use of either burnable poisons or control rods. One benefit of the flatter axial power distribution is an improved fuel utilization. Another benefit of the flatter axial power distribution is a larger safety margin against fuel meltdown in case of accidents. Yet another benefit of the flatter axial power distribution is a higher effectiveness of the BWR control rods TABLE 2 ______________________________________ Axial Distribution of the Multipication Constant With and Without Hydride Fuel Axial core zone (from Bottom) 1/6 2/6 3/6 4/6 5/6 6/6 ______________________________________ All-oxide 1.21 1.18 1.11 1.07 1.04 1.02 Mixed hydride-oxide 1.21 1.18 1.11 1.21 1.20 1.19 ______________________________________ towards the end of the irradiation cycle and, hence, a smaller period of time for scramming the reactor. Moreover, the reactivity swing associated with the transition from full power operation to low or zero power operation is substantially smaller with hydride fuel rather than oxide fuel occupying the high void fraction part of the core, i.e., the most undermoderated region of the core. This behavior is demonstrated by curves 1076 and 1080 of FIG. 10 which show that the void fraction dependence of the hydride fuel lattice multiplication constant is significantly flatter than the corresponding dependence for the oxide fuel lattice, curve 1078 of FIG. 10. The smaller the reactivity swing, the larger the cold shutdown reactivity margin which is a positive safety feature. Another improvement in the BWR economics and safety is derived from the increase in the total length of fuel rods in the BWR core made possible by using hydride fuel in place of the water rods used in the prior art BWR fuel assemblies. (Note: any particular fuel rod is not being increased in length, but as fuel rods substitute for water rods, the total length of fuel pellets in the assembly and, therefore, in the core, is increased.) The increase in the total length of fuel in the core lowers the maximum power density the reactor needs to operate at if the reactor is to deliver a given total power output. One benefit of the increase in the total length and mass of fuel is an increase in the amount of electricity which the BWR can generate in between refuellings. Another benefit of the increase in the total length and mass of fuel in the assembly is an increase in the fuel residence time in the core. The increased fuel residence time in the core increases the reactor operation time between refueling outages thereby increasing the reactor availability and reducing the cost of generating electricity. Yet another benefit of the increase in the total length of fuel in the assembly is lowering of the maximum power density and linear-heat-rate, thus increasing the safety margin against fuel meltdown accidents. Finally another advantage of the present invention is that the leakage of energetic neutrons out of the upper part of the core is reduced relative to the neutron leakage from oxide fuel in the prior art BWR fuel assemblies. The reduced leakage of energetic neutrons increases the core reactivity and reduces the neutron induced damage to structural components located in the vicinity of the core. In another embodiment of the present invention, rather than operating the BWR at a reduced peak power density, the BWR with the novel fuel rods of this invention is operated with the same maximum power density as in the prior art and thereby benefit from the flatter power distribution and larger accumulated fuel rod length by running the hydride fueled core at a higher power level. This may not be achievable in certain of the existing BWRs in which the heat transport and energy conversion system are not capable of accommodating an increase in the power output. However, the design of new BWRs could take full benefit from this feature. That is, the hydride fuel makes it possible to design more compact BWRs than existing BWRs, i.e., the same total power output is obtained from a smaller size core, making the reactor more economical. The maximum linear-heat-rate BWR oxide fuel rods are designed to deliver is, typically, 450 watts per cm. When operating at such a linear-heat-rate, the temperature at the center of the oxide fuel is in the vicinity of 1800.degree. C. The thermal conductivity of the uranium-zirconium hydride fuel, 17.6 watts/m.degree. K., is significantly higher than that of an oxide fuel, which is typically about 2.8 watts/m.degree. K. As a result, if the hydride fuel is to operate at 450 watts per cm, its central temperature will be only about 700.degree. C. Part of the improvements and benefits described above can be achieved by replacing only water rods in the central part of conventional all-oxide BWR fuel assemblies by all-hydride, or predominantly hydride fuel rods. By using hydride rods instead of the water rods, the improved moderation and power flattening across the fuel assembly are achieved without a reduction in the overall length of fuel rods in the assembly. Consequently, the optimal number of all-hydride fuel rods in the BWR fuel assembly that is designed in accordance with the present invention is likely to be larger than the number of water rods in conventional designs of BWR fuel assemblies. Thus, hydride fuel rods can flatten the power density distribution across the BWR fuel assembly better than water rods or than rods containing fuel free solid moderator. The consequences are improved fuel utilization, longer fuel residence time, improved safety or, alternatively, increased power output per assembly. The inclusion of fissionable material with the hydride of the hydride fuel reduces the probability that moderated neutrons will be absorbed in the hydride and in the cladding relative to the absorption probability in fuel assembly designs which have the same hydride moderator but without fissionable material. For example, relative to the zirconium hydride moderator sections proposed in the above-identified Weitzberg patent, the probability of neutron loses through absorption in the zirconium of the solid moderator is reduced by two orders of magnitude. The reduction in the neutron absorption in the zirconium of the solid moderator increases the reactivity of the core and improves the fuel utilization relative to that achievable with solid moderators. Accordingly, the use of hydride fuel in combination with oxide fuel according to the principles of this invention improves the moderation of BWR cores while maximizing the total length of fuel rods in these cores. As a result, BWR cores can be designed to have a flatter power distribution; a smaller amount of burnable poisons; a reduced neutron absorption in control rods; and a reduced leakage of neutrons from the core. The benefits from these improvements in BWR cores include a better fuel utilization; a longer fuel residence time and a larger amount of energy generation between refuelling periods, and, therefore, an increased availability; or a higher total power output from a core of a given size. These advantages improve the economics of generating electricity from BWRs. Additional improvements of the present invention include a larger cold shutdown reactivity margin, a shorter time for scramming the reactor, and a larger safety factor against fuel meltdown accidents. These advantages improve the safety of BWRs. In addition, the reduced leakage of energetic neutrons out of the upper part of the core reduces the damage rate to structural components. Although the description is illustrative of the principles of this invention, the description should not be construed as limiting the scope of the invention, but as merely providing illustrations of some of the presently preferred embodiments of this invention. In view of this disclosure, it will be apparent to those skilled in the art that the hydride fuel can be included in the fuel rod in different combinations with oxide fuel and with segments of water or of solid moderators; the uranium in the hydride and oxide fuel can be of different enrichment; the hydride fuel can be made of thorium hydride rather than of zirconium hydride; the hydride fuel can contain different number of hydrogen atoms per zirconium or thorium atom; the cladding of the fuel rod which contains hydride pellets can be made of a material different than zirconium or its alloy, such as stainless steel, or of stainless steel or glass enamel coated zircaloy; the hydride fuel can be made of small fuel particles imbedded in the solid moderating material; and the fuel rod and fuel assembly can be of different dimensions and different designs. Accordingly, the scope of this invention should be determined not by the embodiments illustrated, but by the appended claims and their legal equivalents.