Patent Number: 054935920
Section: summary

CROSS-REFERENCE TO RELATED APPLICATION This application is a International continuation of application Ser. No. PCT/DE93/00151, filed Feb. 22, 1993. BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a pressurized-water-cooled fuel assembly and to a fuel rod of such a fuel assembly having a cladding tube that encloses a fuel filling and which includes a first thicker inner layer that faces toward the fuel filling and is composed of a first zirconium alloy, and a second thinner outer layer which is metallurgically bound to the inner layer and is composed of a second zirconium alloy, the two zirconium alloys each containing at least the metals tin, iron and chromium as alloying constituents. Zirconium is a comparatively soft metal which is particularly suitable for structural parts of nuclear reactors because of its low neutron absorption. It is technically generally manufactured as "sponge zirconium" having maximum impurities which are standardized for use in nuclear reactors. Since the fuel rods of reactor fuel assemblies are only finger-thin but are several meters long, a high strength which is constant even after prolonged irradiation and which is achieved by adding tin to the alloy is necessary for the fuel rods that are filled with fuel and for the guide tubes, spacers and other structural parts of the fuel assembly. In water, pure zirconium forms a thin oxide layer which protects the metal from further oxidation, but at the same time impurities incorporated in the structure of the zirconium or the oxide layer, particularly nitrogen, may considerably accelerate corrosion. Although the tin addition neutralizes the corrosive action of nitrogen, particularly in conjunction with small additions of iron, it promotes the corrosion itself at higher tin contents (as explained in the publication by B. Lustman and Frank Kerze entitled: "The Metallurgy of Zirconium", New York, 1955, page 538, FIG. 10.34 and page 628, FIG. 11.35). The iron addition hardens the alloy so that, even at comparatively low iron contents, processing the alloy to produce thin, long cladding tubes is virtually no longer possible. In addition, such an iron content results in an increased diffusion and absorption in the metal of hydrogen produced during the corrosion in water, with hydrogenated regions being formed (as explained in the publication by W. Berry, D. Vaughan and E. White entitled: "Hydrogen Pickup During Aqueous Corrosion of Zirconium Alloys" in: Corrosion, Vol. 17, No. 3, March 1961, page 109 t, FIG. 1) which are very brittle and drastically reduce the mechanical robustness of the material. Therefore, on the basis of corrosion experiments in the laboratory in which the temperature was raised to accelerate the corrosion experiments, an optimum range was specified for the contents of tin and iron, which range was optimized as far as possible with regard to corrosion by additionally taking chromium and nickel alloying constituents into account, while attention was also being paid at the same time to an adequate mechanical and thermal robustness. The robustness investigations were partly carried out under reactor conditions in order to obtain adequate strengths even after prolonged irradiation. In the meantime, the alloys "Zircaloy 2" and "Zircaloy 4" developed in that process have proved essentially satisfactory in structural parts of fuel assemblies of water-cooled reactors. Table 1 shows the grades of sponge zirconium, Zircaloy 2 and Zircaloy 4 which are permitted as materials in reactor technology. In that publication, the amounts of the alloying constituents are specified as percentages by weight, based on the alloy. Upon prolonged use in the reactor core, the fuel releases iodine and other fission products, with the result that an aggressive atmosphere with gradually increasing pressure builds up therein and the volume of the fuel itself also increases. Particular mechanical-thermal-chemical stresses which may result in the destruction of the cladding tube and make it necessary to interrupt the reactor operation to replace the fuel rod therefore occur on the inside of the cladding tube. In relation to boiling-water reactors in particular, use is therefore frequently made of composite tubes which have, on their inside, a thin layer of pure zirconium or of an alloy which is optimized in relation to higher ductility and a resistance which is adapted to the chemical, mechanical and thermal conditions. The remaining, thick outer layer of the cladding tube on one hand provides for the necessary mechanical robustness of the entire cladding tube and on the other hand is proof against corrosion under the conditions of the coolant. A liquid/vapor mixture mainly includes water at moderate pressure and moderate temperature, and includes the Zircaloy 2 already mentioned. Modern pressurized-water reactors have a fuel assembly structure and reactivity distribution which make it possible to increase the cost effectiveness as a result of long service lives, high burn-ups and increased operating temperatures. However, a requirement in that connection is that the fuel assembly failure probability due to cladding tube defects is kept extremely low. In that connection, the emphasis is on efforts to eliminate primary damage to the outer surface of the cladding tube since, as a result of such primary defects at any point in the long cladding tube, the high pressure in the coolant flow can force water into the cladding-tube interior where it reacts with the hot fuel. The consequence may then be devastating secondary damage on the inside of the cladding tube. Therefore, for the first time, Published European Application No. 0 212 351 A1, corresponding to U.S. Pat. No. 4,735,768, proposes, as a cladding tube for a water-cooled reactor fuel assembly, a double-layer composite tube having an inner layer, adjacent the fuel, which occupies 80 to 95% of the total wall thickness of the cladding tube and is formed of Zircaloy 4, while the thin outer layer is formed of zirconium containing 0.5% iron and 0.25% vanadium. The so-called "duplex tube", which is used as the carrier that determines the mechanical properties of the entire cladding tube, generally has a thick layer of Zircaloy 2 or Zircaloy 4 to which a thin, outer protective layer of a second zirconium alloy is metallurgically bonded (for example by combined extrusion of two concentric tubes). The alloy contains 0.1 to 1% vanadium and/or 0.1 to 1% platinum and/or 1 to 3% copper, and optionally up to 1% iron. All of the percentages are based on the weight of the alloys. Such cladding tubes exhibit an excellent behavior and, in particular, even at burn-ups of between 40 and 60 MWd/kg of uranium, only oxide layers having thicknesses below 20 .mu.m occur at the outer surface which is exposed to the pressurized water, while the mechanical behavior, such as the increase in length and the shrinkage of the fuel-rod diameter, is within the range of the most favorable values obtained with one-piece tubes of Zircaloy 4. However, those cladding tubes are comparatively expensive, and in particular, the outer alloy is difficult to process mechanically because of its hardness and requires lengthy, careful processing steps with an increased reject occurrence. In addition, the alloying constituents of the second alloy are not permitted per se as reactor materials because of the high neutron absorption and can only be tolerated because they are used only in the thin outer layer in low concentrations. However, the reject material produced during the cladding-tube manufacture cannot readily be fed back into the cladding-tube production since the alloying constituents contained in the second alloy result in impurities in the zirconium or zircaloy which cannot be tolerated. The excellent results of that duplex cladding tube is attributed, in particular, to the absence of tin in the outer layer. Published European Application No. 0 301 295 A1, corresponding to U.S. Pat. No. 4,963,316, describes a duplex cladding tube which is easier to process and in which the outer layer is formed of a tin-free zirconium alloy containing 2.5% niobium or at least a low-tin alloy containing 0.25% tin, 0.5% iron and 0.05% chromium. Good results are expected for a niobium content between 0.2 and 3% and/or a total content of iron, chromium, nickel and tin of between 0.4 and 1.0%. U.S. Pat. No. 5,023,048 describes a similar fuel rod in which the outer layer is formed of Zr, (0.35 . . . 0.65) % Sn, (0.2 . . . 0.65) % Fe, (0.24 . . . 0.35) % Nb and (0.09 . . . 0.16) % O and contains no chromium. Whereas the fuel assemblies of boiling-water reactors are exposed to a coolant temperature of about 280.degree. C. at a pressure of 70 bar, the surface temperature of the cladding tubes of pressurized-water reactors is about 340.degree. C., with the coolant having an outlet temperature of about 320.degree. C. at 170 bar. Despite the fact that the differences in the operating conditions at first appear relatively small, the corrosion processes occurring under those circumstances differ markedly. Laboratory experiments which were carried out at elevated temperatures (for example 360.degree. to 550.degree. C.), that is to say in the vicinity of or above the critical temperature of water, in order to shorten the test times and to intensify the corrosion conditions in the development of Zircaloy 2 and Zircaloy 4, are therefore only of limited meaningfulness for the corrosion behavior. In addition, measures which may have an effect on the corrosion processes occurring at the fuel rods are taken by individual power-station operators for other reasons, for example to prevent corrosion of heat exchangers or other components in the coolant circuit. Measures which may result in an alteration in the corrosion occurring at the fuel-rod claddings are also taken during start-up or during certain temporary operating states. Thus, for example, fuel rods of pressurized-water reactors which are operated at a coolant outlet temperature of 316.degree. C. on average may exhibit an excellent behavior but may exhibit a higher failure probability even in reactors having mean coolant outlet temperatures of 326.degree. C. which, although slight, is undesirable. One of the causes thereof may be the strong temperature dependence of hydrogen diffusion during very long service lives. At the start of the operation of some nuclear reactors, a small amount of dissolved lithium hydroxide, which may considerably affect the corrosion of the cladding tubes, is present in the coolant circuit. In particular, an increase in fuel-rod power may result in a fuel-rod outer temperature which is in fact only a few degrees higher but at which local boiling resulting in substantially more severe corrosion conditions occurs in the pores of the oxide layer. However, the Li content in the cooling water, which is not enough to cause problems in conventional fuel-assembly structures, may concentrate in the pores during local boiling and make it necessary either to change over to other alloys or to dispense with the increase in output. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a nuclear-reactor fuel rod with a double-layer cladding tube and a fuel assembly containing such a fuel rod, which overcome the hereinafore-mentioned disadvantages of the hereto-fore-known devices of this general type and which provide a fuel assembly having fuel-rod cladding tubes that can be produced more efficiently and are adapted to the operating conditions of modern pressurized-water reactors. In order to achieve this object, the invention proceeds from the requirement that no completely new alloys whose approval as a material in reactor technology is possible only by means of extensive, time-consuming and expensive proceedings, are used in this connection. In the case of the inner tube surface, it is assumed that Zircaloy 2 and, in particular, Zircaloy 4 are adequately adapted to the conditions prevailing there during the pressurized-water operation and, with an appropriate thickness of an inner layer of a cladding-tube manufactured therefrom, also ensure the load capacity which is to be required for the cladding tube as a whole. If possible, no alloying constituents that are not permitted for Zircaloy 2 and Zircaloy 4 are to be used for the alloy of the outer layer either. As a result, although it is possible to have recourse to the earlier, extensive experience with zircaloy in pressurized-water reactors, the proportions by weight of the individual alloying constituents of the outer layer must be redetermined, in particular, in relation to the long-term behavior during corrosion and hydrogen take-up, and opposing, mutually exclusive trends may occur which have to be compensated for by a fresh optimization. At the same time, a substantial deviation from the concentration ranges which are valid for zircaloy is also avoided for the alloy of the outer layer. On one hand, this facilitates the approval and acceptability of the cladding tubes for use in nuclear reactors and on the other hand, it also makes it possible to recirculate waste and reject material in the cladding-tube manufacture. In particular, the thick supporting layer can comply strictly with the zircaloy specifications if this is necessary, but on the other hand, certain, slight deviations also appear possible and justifiable in order, for example, to adapt the inner layer even better to the mechanical-thermal-chemical stresses prevailing at the inner surface of the cladding tube. In addition, it is also possible, if necessary, to increase the resistance of the outer layer to a lithium-containing coolant during long operating times. At the same time, the serviceability of the novel fuel rods is also to include pressurized-water reactors which operate, for example, at slightly higher coolant temperatures and/or with lithium additions in the coolant. With the foregoing and other objects in view there is provided, in accordance with the invention, a fuel rod of a pressurized-water-cooled fuel assembly, comprising a fuel filling, a cladding tube enclosing the fuel filling and including a first thicker inner layer facing toward the fuel filling and being formed of a first zirconium alloy, and a second thinner outer layer being metallurgically bound to the inner layer and being formed of a second zirconium alloy, the two zirconium alloys each containing at least the metals tin, iron and chromium as alloying constituents, and: a) the first alloy containing 1-2% by weight of Sn, 0.05-0.25% by weight of Fe and 0.05-0.2% by weight of Cr; b) the second alloy having 0.5-1.3% by weight of Sn, 0.15-0.5% by weight of Fe and 0.05-0.4% by weight of Cr; c) the second alloy having a total content of tin, iron and chromium of more than 1.0% by weight and a content of tin having a ratio to the content of tin in the first alloy being between 0.35 and 0.7; and d) a ratio of the content of iron and chromium in the second alloy to the tin content of the first alloy being between 0.2 and 0.5. With the objects of the invention in view, there is also provided a fuel assembly of a pressurized-water reactor, comprising fuel rods as described above. The experimental finding, which is discussed below in even greater detail, shows that a higher tin content, which is necessary to harden the zirconium, adversely affects the corrosion with oxide layer thicknesses of about 100 .mu.m, such as those which occur on cladding tubes in a reactor but are observed in laboratory tests at about 350.degree. C. only after long test times. This long-term corrosion decreases with decreasing tin content. Although an iron content frequently results in brittle precipitates which result in problems during mechanical processing, it does inhibit the long-term corrosion and limits the hydrogen take-up, particularly in conjunction with chromium. The invention therefore envisages a double-layer structure of the cladding tube in which the limits for the content of tin, iron and chromium in the two alloys are modified with respect to the zircaloy specification, with the tin content in the outer layer ("(Sn)(outer)") being reduced, although without falling below a minimum content of tin, iron and chromium: EQU (Sn)(outer)+(Fe+Cr) (outer)&gt;1% In the case of the tin content, it has to be borne in mind in this connection that a too-severe reduction in the outer layer entails the risk of heretofore unnoticed damage (for example, increased corrosion in the presence of lithium added to the coolant for other reasons) occurring. Although values which prove to be unfavorable in relation to corrosion and hydrogenation are provided for the tin content in the specification of the zircaloy because of the required hardness of the cladding tube, the tin in the inner layer can be enriched provided an adequate content of iron and chromium in the outer layer provides protection against corrosion and hydrogenation. Advantageously, this (Fe+Cr) content of the outer layer is higher than in the inner layer. However, in order to avoid difficulties due to the hardening of the inner layer (Sn content!) and the embrittlement of the outer layer (content of Fe+Cr!) during the processing and in order to ensure a metallurgical bond between the two layers (for example, by combined extrusion of two coaxial tubes), the tin content of the outer layer, in relation to the tin content of the inner layer ("(Sn)(inner)"), is reduced in accordance with: ______________________________________ (Sn)(outer) &gt;0.35 (preferably .gtoreq. 0.40), (Sn)(inner) (Sn)(outer) &lt;0.7 (preferably .ltoreq. 0.65), (Sn)(inner) (Sn)(inner) .gtoreq.2 .times. (Fe + Cr) (outer), (Sn)(inner) .ltoreq.5 .times. (Fe + Cr) (outer). ______________________________________ Since both layers are therefore essentially composed of the same alloys which are present only in modified quantitative ratios, a composite body of these mutually similar alloys can be manufactured and processed further largely without problems. Table 2 specifies upper limits and lower limits for the concentrations of the individual alloying constituents in the first and second alloy, with advantageous or preferred values which narrow the concentration ranges further being specified in the brackets. The different composition of the two alloys composed of the same alloying constituents may result in reject material or waste produced during the cladding tube manufacture having a mean composition which is slightly outside the limits permitted for Zircaloy 2 or Zircaloy 4 but is usable for the manufacture of the first layer. In relation to the long-term corrosion and the hydrogen take-up, the standardization of Zircaloy 2 and Zircaloy 4 in any case appears to be not completely optimum according to the test results presented below and a slight shift in the concentration limits would probably also be tolerable for approval authorities and power-station operators. The invention therefore results in the minimum and maximum proportions by weight of tin, iron and chromium specified in Table 2 for the inner layer. In addition, the lower and upper limits of the proportions by weight of these metals in the outer layer provided in accordance with the invention are specified. Narrower limits are partly the result of the requirement that the amounts of all of the constituents of the first layer are to lie within the limits permitted for Zircaloy 2 or Zircaloy 4 and are specified in brackets. In this connection, the brackets also contain in some cases advantageous or particularly preferred limit values which result, in particular for the second alloy ("outer layer"), from the test results due to optimization. Common to both layers is the fact that, in addition to tin, iron and chromium, they advantageously contain only further alloying constituents which are permitted for Zircaloy 2 and Zircaloy 4 and the amounts of the further alloy constituents, in particular nickel, silicon and oxygen, are virtually inside the limits permitted for Zircaloy 2 or Zircaloy 4. It is expedient to adjust the content of silicon and oxygen to achieve a defined alloy having a stabilized grain structure and an advantageous grain refinement. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a nuclear-reactor fuel rod with a double-layer cladding tube and a fuel assembly containing such a fuel rod, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings.