Patent Number: 050248090
Section: description

DETAILED DESCRIPTION OF THE INVENTION Referring now more particularly to FIG. 1, there is shown a partially cutaway sectional view of a nuclear fuel assembly 10. This fuel assembly consists of a tubular flow channel 11 of generally square cross section provided at its upper end with lifting bale 12 and at its lower end with a nose piece (not shown due to the lower portion of assembly 10 being omitted). The upper end of channel 11 is open at 13 and the lower end of the nose piece is provided with coolant flow openings. An array of fuel elements or rods 14 is enclosed in a channel 11 and supported therein by means of upper end plate 15 and a lower end plate (not shown due to the lower portion being omitted). The liquid coolant ordinarily enters through the openings in the lower end of the nose piece, passes upwardly around fuel elements 14, and discharges at upper outlet 13 in a partially vaporized condition for boiling reactors or in an unvaporized condition for pressurized reactors at an elevated temperature. The nuclear fuel elements or rods 14 are sealed at their ends by means of end plugs 18 welded to the cladding 17, which may include studs 19 to facilitate the mounting of the fuel rod in the assembly. A void space or plenum 20 is provided at one end of the element to permit longitudinal expansion of the fuel material and accumulation of gases released from the fuel material. A nuclear fuel material retainer means 24 in the form of a helical member is positioned within space 20 to provide restraint against the axial movement of the pellet column, especially during handling and transportation of the fuel element. The fuel element is designed to provide an excellent thermal contact between the cladding and the fuel material, a minimum of parasitic neutron absorption, and resistance to bowing and vibration which is occasionally caused by flow of the coolant at high velocity. A nuclear fuel element or rod 14 is shown in a partial section in FIG. 1 constructed according to the teachings of this invention. The fuel element includes a core or central cylindrical portion of nuclear fuel material 16, here shown as a plurality of fuel pellets of fissionable or fertile material positioned within a structural cladding or container 17. In some cases, the fuel pellets may be of various shapes, such as cylindrical pellets or spheres, and in other cases, different fuel forms such as particulate fuel may be used. The physical form of the fuel is immaterial to this invention. Various nuclear fuel materials may be used, including uranium compounds, plutonium compounds, thorium compounds and mixtures thereof. A preferred fuel is uranium dioxide or a mixture comprising uranium dioxide and plutonium dioxide. Referring now to FIG. 2, the nuclear fuel material 6 forming the central core of the fuel element 14 is surrounded by a cladding 17. The cladding container encloses the core so as to leave a gap between the core and the cladding container during use in a nuclear reactor. The cladding is comprised of a corrosion-resistant zirconium alloy tube 21. In one embodiment of this invention the alloy tube 21 is made from a first zirconium alloy, consisting essentially of by weight percent 0.5 to 2.0 percent tin, 0.5 to 1.0 percent of a solute composed of molybdenum, tellurium, and the balance zirconium. In a second embodiment of this invention the alloy tube 21 is made from a second zirconium alloy, consisting essentially of by weight percent 0.5 to 2.0 percent tin, 0.5 to 1.0 percent of a solute, the solute consisting of niobium, the amount of niobium being less than 0.5 percent of the alloy, and a member selected from the group consisting of molybdenum, tellurium and mixtures thereof, and the balance zirconium. In a third embodiment of this invention, the alloy tube 21 is made from a third zirconium alloy, consisting essentially of by weight percent 0.5 to 2.0 percent tin, 0.5 to 1.0 percent of a solute composed of niobium and tellurium, and the balance zirconium. In a fourth embodiment of this invention the alloy tube 21 is made from a fourth zirconium alloy, consisting essentially of by weight percent 0.5 to 2.0 percent tin, 0.6 to 1.0 percent of a solute, the solute being composed of niobium, molybdenum and tellurium, the amount of niobium being at least 0.5 percent, and the balance zirconium. In a fifth embodiment of this invention, the alloy tube 21 is made from a fifth zirconium alloy, consists essentially of by weight percent 0.5 to 2.0 percent tin, 0.3 to 1.4 percent of a solute composed of tellurium, and the remainder zirconium. In a sixth embodiment of this invention, the alloy tube 21 is made from a sixth zirconium alloy, consisting essentially of by weight percent 0.5 to 1.0 percent bismuth, approximately 0.5 to 1.0 percent of a solute composed of a member selected from the group consisting of molybdenum, niobium, tellurium and mixtures thereof, and the balance being zirconium. In a seventh embodiment of this invention, the alloy tube 21 is made from a seventh zirconium alloy, consisting essentially of by weight percent 0.5 to 2.5 percent of a mixture of tin and bismuth, approximately 0.5 to 1.0 percent of a solute composed of a member selected from the group consisting of molybdenum, niobium, tellurium and mixtures thereof, and the balance being zirconium. In an eighth embodiment of this invention, the alloy tube 21 is made from an eighth zirconium alloy, consisting essentially of by weight percent 0.5 to 2.5 percent bismuth, 0.3 to 1.0 percent of a solute composed of tellurium, and the balance zirconium. In a ninth embodiment of this invention, the alloy tube 21 is made from a ninth zirconium alloy, consisting essentially of by weight percent 0.5 to 2.5 percent of a mixture of tin and bismuth, 0.3 to 1.0 weight percent of a solute composed of tellurium, and the balance zirconium. Container claddings made from one of the first through the ninth zirconium alloys described above provide an increased resistance to nodular corrosion over prior zirconium alloy tubes. It should be noted that the first through the ninth zirconium alloys described herein will also optionally contain from about 0.09 to 0.16 weight percent of oxygen. Most commercial grade sponge zirconium which would be used in making alloys such as the ones in the present invention will contain small amounts of oxygen, roughly on the order of about 800-1300 parts per million. In some instances, it will be desirable to increase the concentration of oxygen in the alloy. Adding oxygen is one way to increase room temperature yield strength. Thus, the alloys of the present invention may be produced with or without the additional oxygen, as this will have little or no effect on the corrosion resistance of the alloys. Another embodiment of this invention is shown by referring to FIG. 3. The nuclear fuel material 16 forming the central core of the fuel element 14 is surrounded by a composite cladding 17. The composite cladding container encloses the core so as to leave a gap 23 between the core and the cladding container during use in a nuclear reactor. The composite cladding is comprised of a zirconium alloy tube 21 made from one of the first through the ninth zirconium alloys described above. The alloy tube has bonded on the inside surface thereof a metal barrier 22 so that the metal barrier forms a shield between the alloy tube 21 and the nuclear fuel material held in the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding, and is comprised of a low neutron absorption material, namely, moderate purity zirconium. One moderate purity zirconium is sponge zirconium. The metal barrier 22 protects the alloy tube portion of the cladding from contact and reaction with gasses and fission products from the nuclear fuel, and prevents the occurrence of localized stress and strain. The content of the metal barrier of moderate purity zirconium is important and serves to impart special properties to the metal barrier. Generally, there is at least about 1,000 parts per million (ppm) by weight and less than about 5,000 ppm impurities in the material of the metal barrier and preferably less than about 4,200 ppm. Of these, oxygen is kept within the range of about 200 to about 1,200 ppm. All other impurities are within the normal range for commercial, reactor-grade sponge zirconium. In another embodiment of this invention, a corrosion-resistant nuclear fuel element is shown by referring to FIG. 4. The nuclear fuel material 16 forming the central core of fuel element 14 is surrounded by a composite cladding 17. The composite cladding container encloses the core so as to leave a gap 23 between the core and the cladding container during use in a nuclear reactor. The composite cladding is comprised of a zirconium alloy tube 30 made from a Zircaloy alloy. The alloy tube has bonded on the outside surface thereof a metal layer 32 so that the metal layer forms a corrosion protective shield over the alloy tube. The outer metal layer is about 5 to 20 percent of the thickness of the alloy tube and is comprised of one of the first through the ninth zirconium alloys described above. The outer metal layer protects the Zircaloy alloy tube portion of the cladding from nodular corrosion. Another improved nuclear fuel element is shown by referring to FIG. 5. The nuclear fuel material 16 forming the central core of fuel element 14 is surrounded by a composite cladding 17. The composite cladding container encloses the core so as to leave a gap 23 between the core and the cladding container during use in a nuclear reactor. The composite cladding is comprised of a Zircaloy alloy tube 30. The alloy tube has bonded on the inside surface thereof a metal barrier 22 so that the metal barrier forms a shield between the alloy tube 30 and the nuclear fuel material held in the cladding. The metal barrier is about 1 to about 30 percent of the thickness of the alloy tube and is comprised of a low neutron absorption material, namely, moderate purity zirconium as described above. The metal barrier 22 protects the alloy tube portion of the cladding from contact and reaction with gases and fission products from the nuclear fuel, and prevents the occurrence of localized stress and strain. The outer surface layer is bonded on the outside surface of the alloy tube 30. The outer metal layer is about 5 to 20 percent of the thickness of the alloy tube and is comprised of one of the first through the ninth zirconium alloys described above. The outer metal layer protects the Zircaloy alloy tube portion of the cladding from modular corrosion. Sponge zirconium metal forming the metal barrier in the composite cladding is highly resistant to radiation hardening, and this enables the metal barrier after prolonged irradiation to maintain desirable structural properties such as yield strength and hardness at levels considerably lower than those of conventional zirconium alloys. In effect, the metal barrier does not harden as much as conventional zirconium alloys when subjected to irradiation, and this together with its initially low yield strength enables the metal barrier to deform plastically and relieve pellet-induced stresses in the fuel element during power transients. Pellet induced stresses in the fuel element can be brought about, for example, by swelling of the pellets of nuclear fuel at reactor operating temperatures (300.degree. C. to 350.degree. C.) so that the pellet comes into contact with the cladding. It has further been discovered that a metal barrier of sponge zirconium of the order preferably about 5 to 15 percent of the thickness of the cladding and a particularly preferred thickness of 10 percent of the cladding bonded to the alloy tube of a zirconium alloy provides stress reduction and a barrier effect sufficient to prevent failures in the composite cladding. The corrosion resistant nuclear fuel rod cladding used in the nuclear fuel elements of this invention can be fabricated from a billet comprised of a zirconium alloy made from one of the first through the ninth zirconium alloys described above. The billet is heated to 590.degree. to 650.degree. C. and extruded. The extruded tubing is then subjected to a process involving conventional tube reduction until the desired size of tubing is achieved. In another method, a hollow collar of the sponge zirconium selected to be the metal barrier is inserted into a hollow billet of one of the first through the ninth zirconium alloys described above. The assembly is heated to 590.degree. to 650.degree. C. and extruded. The extruded tubing is then subjected to a process involving conventional tube reduction until the desired size of cladding is achieved. In another method, a tube blank is made from a Zircaloy alloy and an outer tube is placed on this tube blank. The outer tube is composed of the first through the ninth zirconium alloys described above. This assembly is then heated to a temperature in the range of 590.degree. to 650.degree. C. and is extruded. The extruded tubing is then subjected to a process involving conventional tube reduction until the desired size of tubing is achieved. In another method, a tube blank is made from a Zircaloy alloy and an outer tube is placed on this tube blank. The outer tube is composed of the first through the ninth zirconium alloys described above. A hollow collar of the sponge zirconium selected to be the metal barrier is inserted into the tube blank. The assembly is heated to a temperature in the range of 590.degree. to 650.degree. C. and is extruded. The extruded tubing is then subjected to a process involving conventional tube reduction until the desired size of tubing is achieved. Intermediate and final anneals are used during the tube reduction processes described above. Anneals range between 570.degree. to 590.degree. C. The invention includes a method of producing a nuclear fuel element comprising making a cladding or a composite cladding container comprised of a zirconium alloy, or a zirconium alloy and a barrier layer, or a Zircaloy alloy and a surface layer, or a Zircaloy alloy and an outer surface layer and an inner barrier layer. The container is open at one end and filled with a core of nuclear fuel material leaving a gap between the core and the container and leaving a cavity at the open end. A nuclear fuel material retaining means is inserted into the cavity and an enclosure is applied to the open end of the container, leaving the cavity in communication with the nuclear fuel. The end of the clad container is then bonded to the enclosure to form a tight seal therebetween. The present invention offers several advantages promoting a long operating life for a nuclear fuel element. A greater resistance to nodular corrosion protects the strength and integrity of the cladding. On cladding having a barrier layer, the reduction of chemical interaction on the cladding, the minimization of localized stress on the zirconium alloy tube portion of the cladding, and the minimization of stress corrosion and strain corrosion on the zirconium alloy tube portion of the cladding, all reduce the probability of a splitting failure occurring in the zirconium alloy tube. The invention further reduces expansion or swelling of the nuclear fuel into direct contact with the zirconium alloy tube, and this reduces the occurrence of localized stress on the zirconium alloy tube, initiation or acceleration of stress corrosion of the alloy tube and bonding of the nuclear fuel to the alloy tube. An important property of the composite cladding of this invention is that the foregoing improvements are achieved with no substantial additional neutron absorption penalty. Further, the composite cladding has a very small heat transfer penalty in that there is no thermal barrier to transfer of heat, such as results in a situation where a separate foil or a liner is inserted in a fuel element. Also, the composite cladding of this invention is inspectable by conventional non-destructive testing methods during various stages of fabrication and operation. The following examples are offered to further illustrate the improved nodular corrosion resistance of the alloys used in this invention. EXAMPLES Tests for both uniform corrosion resistance and nodular corrosion resistance have been conducted on alloys representative of the first through the ninth zirconium alloys used in the present invention. These tests have shown that a dramatic decrease in susceptibility to nodular corrosion can be attained in an alloy which is relatively insensitive to heat treatment while retaining essentially the same uniform corrosion resistance of a Zircaloy 2 alloy. Table 1 lists several examples of alloys of the present invention, along with three entries at the bottom of the table which are Zircaloy 2 alloys in three different heat treatment states. These alloys were tested in water containing 8 ppm oxygen at 288.degree. C. and 1500 psig, conditions similar to a reactor operating temperature and pressure but absent a radiation source, to evaluate the resistance to uniform corrosion. It can be seen from the results in this table that the tested alloys of the present invention exhibit excellent resistance to uniform corrosion. The uniform corrosion rates in most cases being comparable to those of the Zircaloy 2 specimens. None of the specimens tested under these conditions exhibit any sign of the formation of nodular corrosion products. Table 2 reports the results of tests conducted to determine the susceptibility of the Zirconium alloys of the present invention to nodular corrosion. The test conditions used were those which induce in the laboratory the formation of the nodular corrosion products on Zircaloy alloys that have been annealed at 750.degree. C. for forty-eight hours and are identical to that nodular corrosion sometimes found on Zircaloy after being used in reactor service. For purposes of comparison, the weight gains of the annealed Zircaloy under the same test conditions were on the order of several thousand milligrams per square decimeter. The Zirconium alloys of the present invention were tested in various heat treatment states. The results in Table 2 also provide an indication that the corrosion-resistance properties of these alloys are relatively insensitive to the heat treatment state of the specimen. Some compositions were tested using specimens in cold rolled plate form, both with and without a subsequent anneal. Others were tested in the as-cast form, both with and without subsequent annealing. Several compositions were tested only after having been annealed. The 750.degree. C. anneal for forty-eight hours, which all of the tested alloys of the present invention were subjected to, is the heat treatment which strips the Zircaloy 2 alloy of its resistance to nodular corrosion under the laboratory steam tests. All of the weight gains reported in Table 2 are far superior to the results obtained when sensitized or annealed Zircaloy alloy is tested. Most of the alloys of the present invention produced weight gains of less than 100 mg/dm.sup.2, while the remainder produced weight gains on the order of one or two hundred milligrams per square decimeter. As previously mentioned, weight gains reported in tests of sensitized Zircaloy specimens under the same test times and conditions are on the order of several thousand milligrams per square decimeter. In addition to the reduced weight gains evidenced in the alloys of the present invention, none of these alloys showed any sign of formation of nodular corrosion products. Under the test conditions, these alloys clearly provide improvement in resistance to nodular corrosion. TABLE 1 ______________________________________ OXIDE GROWTH/WEIGHT GAIN IN WATER AT 288.degree. C., 1500 psig, 8 ppm OXYGEN Zr Alloy Composition Weight Percent Weight Gain (mg/dm.sup.2) Sn Nb Mo Te Bi 1000 hrs. 1700 hrs. ______________________________________ 1.51 0.38 0.19 13.8 14.0 1.56 0.21 0.38 9.5 10.3 1.53 0.39 0.22 11.3 12.0 1.57 0.26 0.15 0.16 13.3 10.8 1.55 0.2 0.18 0.2 11.7 12.5 1.60 0.58 9.7 11.5 1.16 0.72 9.8 11.0 1.13 0.60 10.5 12.5 1.65 0.68 10.0 12.0 1.54 0.69 12.5 20.8 1.50 0.29 0.28 15.3 19.0 1.50 0.5 13.5 16.8 1.49 0.34 0.32 11.0 13.6 1.5 0.2 0.4 21.8 50.0 1.0 0.2 0.4 21.2 40.0 1.5 0.4 0.3 16.8 33.0 1.4 18.0 33.8 1.3 0.6 4.3 b 0.5 1.4 12.0 14.5 0.5 2.0 12.0 b 0.3 0.3 0.8 20.5 32.0 0.5 0.5 0.7 39.8 65.0 0.28 0.66 11.8 15.0 0.33 0.84 9.3 11.0 0.41 1.12 15.8 22.0 0.3 0.3 1.6 85.0 150.0 1.0 0.5 1.0 9.3 11.2 Z2 (Zircaloy 2, cross-rolled 11.0 13.2 commercial plate) Z2 w/750.degree. C./16 hr. anneal 11.0 15.0 Z2 w/Beta quench 15.0 17.0 ______________________________________ b -- Test not completed. TABLE 2 __________________________________________________________________________ WEIGHT GAIN AFTER EXPOSURE TO STEAM AT 510.degree. C., 1500 psig FOR 24 HOURS Zr Alloy Composition Weight/Percent Weight Gain (mg/dm.sup.2) Cold Rolled 0.1" Bi Sn Nb Mo Te As Cast 750.degree. C./48 hr. As Rolled 750.degree. C./48 hr. 920.degree. C./3 __________________________________________________________________________ hr. 1.4 0.5 44 43 * 51 66 0.8 0.3 0.3 53 57 * 49 60 2.5 0.5 68 70 * 76 116 0.7 0.5 0.5 53 81 * 56 116 1.6 0.3 0.3 76 82 * 73 132 1.5 0.5 96 103 * 82 182 1.7 0.5 0.5 119 141 * 111 264 2.5 0.5 247 196 * 135 434 0.66 0.28 157 82 * 95 * 0.84 0.33 114 82 * 93 * 1.12 0.41 109 130 * 100 * 0.5 0.5 396 177 * * * 1.5 0.5 120 142 * * * 1.0 1.0 0.5 * * 43 51 * 1.5 0.7 0.5 * * 44 54 * 1.0 1.0 0.3 0.3 * * 67 61 * 1.5 0.7 0.3 0.3 * * 80 67 * 1.0 1.0 0.3 0.3 * * 80 55 * 1.5 0.7 0.3 0.3 * * 87 65 * 1.0 1.0 0.2 0.2 0.2 * * 92 63 * 1.5 0.7 0.2 0.2 0.2 * * 102 70 * 1.0 1.0 0.6 * * 163 104 * 1.5 0.7 0.6 * * 172 119 * 1.0 1.0 0.5 * * 109 80 * 1.5 0.7 0.5 * * 130 84 * 1.51 0.38 0.19 * * 44 61 * 1.56 0.21 0.38 * * 69 58 * 1.53 0.39 0.22 * * 47 49 * 1.57 0.26 0.15 0.16 * * 48 58 * 1.55 0.2 0.18 0.2 * * 60 64 * 1.60 0.58 * * 111 107 * 1.16 0.72 * * 94 88 * 1.13 0.60 * * 88 90 * 1.65 0.68 * * 99 95 * 1.54 0.69 * * * 83 * 1.50 0.29 0.28 * * * 70 * 1.50 0.5 * * * 68 * 1.49 0.34 0.32 * * 42/43 61/57 * 1.5 0.5 0.5 * * 52 70 * 1.0 0.5 0.5 * * 85 70 * 1.5 0.5 0.5 * * 70 68 * 1.5 0.15 0.15 0.3 * * 65 67 * 1.5 0.15 0.3 0.15 * * 65 63 * 1.0 0.30 214 117 * * * 1.5 0.30 260 107 * * * 0.5 0.50 144 88 * * * 1.3 0.60 * * 45 42 * 1.5 1.00 100 156 * * * 1.4 * 74 * * * 2.0 0.30 0.30 * 102 * * * 0.5 0.50 0.50 * 62 * * * 1.0 0.50 0.50 * 79 * * * 1.5 0.50 0.50 * 108 * * * __________________________________________________________________________ *Not tested.