Patent Number: 053735412
Section: summary

BACKGROUND OF THE INVENTION The present invention relates to fuel rods for use in fuel assemblies for a water cooled and moderated nuclear reactor, in particular in assemblies for a pressurized water reactor. These rods are made of fuel pellets enclosed in cladding of an alloy having low neutron absorption. The cladding must satisfy numerous conditions, some of which are difficult to reconcile. It must remain watertight, it must conserve its mechanical properties under irradiation at high temperature, and its amount of creep must be low. It must resist corrosion by the aqueous medium in which it is immersed. It must have little interaction with the fuel contained inside the cladding. Until now, cladding has been made above all from a zirconium based alloy known as "Zircaloy 4" which contains: 1.20% to 1.70% tin; PA1 0.18% to 0.24% iron; PA1 0.07% to 0.13% chromium; PA1 0.35% to 0.65% tin; PA1 0.18% to 0.25% iron; PA1 0.07 to 0.13% chromium, and PA1 0.19% to 0.23% oxygen with the sum of the iron, chromium, tin, and oxygen content being less than 1.26%; PA1 up to 200 ppm silicon PA1 and/or 0.80% to 1.20% niobium, the oxygen content then being in range of 0.10% to 0.16% by weight, PA1 the thickness of the outer layer being in the range of 10% to 25% of the total thickness of the cladding. the total iron plus chromium content being in the range 0.28% to 0.37%. Standards concerning "Zircaloy 4", also known under the reference UNSR 60804, place a limit on the content of elements other than zirconium and those specified above, except with respect to oxygen, where it is merely stated that the oxygen content must be specified in each case. The usual oxygen content of "Zircaloy 4" does not exceed 0.12%, and is generally much less. While the mechanical strength of "Zircaloy 4" claddings has been found satisfactory, it has been observed that corrosion by the surrounding high temperature aqueous medium considerably reduces the length of time they can be kept in a reactor. Proposals have already been made to avoid this defect by using "duplex" or "triplex" claddings (see FR-A-1 547 960; EP-A-212 351; U.S. Pat. No. 4,649,023) which have at least an inner layer of "Zircaloy 4", or of a similar alloy, and an outer layer which is considerably thinner than the inner layer and which is made of a zirconium-based alloy that withstands corrosion better than "Zircaloy 4". In particular, cladding has been proposed that has an inner layer of "Zircaloy 4" and an outer layer made of a zirconium-based alloy having a reduced or zero tin content, but containing additional elements such as niobium, vanadium and nickel, which improve corrosion resistance. It has long been known (e.g., U.S. Pat. No. 4,717,534) that Zr--Nb alloys having about 2.5% niobium have good corrosion resistance in a high temperature aqueous medium. The composition of the alloy constituting the outer layer must be such that the cladding can be obtained by co-rolling or co-extrusion, with a high thickness reduction ratio at each manufacturing step. In addition, the presence of the outer layer must not significantly degrade the mechanical characteristics of the cladding as a whole. Unfortunately, to a first approximation, the mechanical properties of the cladding is the result of summing the properties of both layers, weighted by a factor representing the fraction of the total thickness applicable to each layer. It is well-known that ordinary zirconium-niobium alloys, having a very low oxygen content, have mechanical properties that are greatly inferior to those of Zircaloys. SUMMARY OF THE INVENTION An object of the present invention is to provide cladding for a nuclear fuel rod where the cladding includes at least one inner layer of "Zircaloy 4" and an outer layer that is thinner than the inner layer, which better fulfils practical requirements than previously known cladding, in particular by having considerably increased resistance to corrosion in the ambient aqueous medium while conserving mechanical characteristics that are quite comparable to those of cladding of solid "Zircaloy 4". To this end, the invention provides a nuclear fuel rod whose cladding comprises at least an inner portion of "Zircaloy 4" and an outer portion of a zirconium-based alloy that contains by weight, besides zirconium and unavoidable impurities: (a) (b) In a modification, a 0% to 0.05% content of iron, chromium or niobium is replaced by an equivalent content of vanadium. When the outer layer contains tin and does not have an appreciable niobium content, an oxygen content that is much higher than in ordinary Zircaloy 2, 3 and 4 type alloys makes it possible to obtain mechanical characteristics that come close to those of Zircaloy 4, providing the cladding is in relaxed condition (stress-relieved). When the outer layer has only one metal additive (ignoring unavoidable impurities) constituted by niobium, then the alloy in relaxed condition, even when its oxygen content is high, suffers from very poor resistance to hot creep. This drawback is avoided by having an alloy which is simultaneously doped with oxygen and by subjecting the cladding to a final recrystallization heat treatment. The invention also provides a method of manufacturing cladding suitable for use in a rod of the type defined above. In order to obtain a duplex tube for a fuel rod in accordance with the present invention, a composite billet is made having an inner portion of "Zircaloy 4", and in particular having a tin content in the low end of the range specified by the standard, and an outer portion made of a zirconium--niobium--oxygen alloy or of a zirconium-based alloy containing tin, iron, chromium and oxygen. These two billets are assembled together by welding their ends. The billets obtained in this way for the two types of alloy are hot extruded, typically at 650.degree. C. It is during this coextrusion operation that metallurgical bonding is obtained between the two zirconium alloys. The duplex tube blanks obtained in this way are transformed into finished duplex tubes by a succession of thermomechanical cycles. Typical dimensions of finished duplex-tubes are 9.50 mm outside diameter and 0.625 mm thickness, or 10.75 mm and 0.725 mm, with an outer layer of zirconium-based alloy containing niobium and oxygen (or iron, chromium, tin, oxygen) having a thickness in the range of about 80 .mu.m to about 140 .mu.m. Cold rolling steps, typically in a pilgrim step machine, of the thermo-metallurgical sequence are identical for both alloys studied, for all passes, in terms of cross-section reduction ratio and of Q factor (ratio between variation in thickness and variation in diameter), even at high deformation ratios. Transformation takes place without difficulty, and without creating crack-type defects. However, intermediate recrystallization annealing operations and the final annealing operation are adapted to each of the two alloys. For the zirconium-tin-iron-chromium-oxygen alloy, intermediate recrystallization annealing takes place in the range of 700.degree. C. to 750.degree. C. If there are five steps, the first two are advantageously at about 735.degree. C. and the last two at about 700.degree. C., while final annealing is performed at about 485.degree. C. As for the zirconium-niobium-oxygen alloy, the intermediate recrystallization annealing operations between rolling passes are performed first at about 580.degree. C..+-.15.degree. C. in order to avoid corrosion during the respective rolling phases. The last three annealing operations may be carried out at 700.degree. C..+-.15.degree. C. for Zircaloy 4 to have satisfactory resistance in a PWR. The final annealing operation is performed at about 580.degree. C. For a finished duplex-tube, the thermal mechanical transformation sequences selected for the two alloys give rise to sizes and distributions of intermetallic precipitates that are optimal mainly with respect to generalized corrosion, namely, precipitation that is fine (precipitate diameter of about 50 nm) and uniform for the zirconium alloy containing niobium and oxygen, and precipitation that is uniform and of sufficient size (intermetallic particle diameter greater than 0.18 .mu.m) for the zirconium-tin-iron-chromium- and oxygen-alloy.