JPH0449919B2 - - Google Patents

Description

[Detailed description of the invention]

[Field of Application of the Invention] The present invention particularly relates to a cladding tube for a nuclear fuel element having an improved structure for preventing stress corrosion cracking. [Background of the Invention] Nuclear reactors currently being designed, manufactured, and operated typically use nuclear fuel elements in which nuclear fuel material is enclosed in a cladding tube that is corrosion-resistant, non-reactive, and has excellent thermal conductivity. . These nuclear fuel elements are assembled in a lattice shape at regular intervals within a coolant flow channel to form a fuel assembly, and an appropriate number of these fuel assemblies are combined to form a nuclear fission chain reaction assembly or reactor core capable of performing a fission reaction. This reactor core is placed in a reactor vessel through which coolant flows. Cladding tubes are used for several purposes, the first of which is to prevent contact and chemical reaction between the nuclear fuel and the coolant or moderator. The second purpose is to prevent radioactive fission products, which are partially gaseous, from escaping from the fuel into the coolant or moderator, or into both coolant and moderator if both are present. be. Usually, stainless steel, zirconium, zirconium alloy, etc. are used as the cladding material. A rupture of these claddings, i.e., leakage or loss of seal, could contaminate the coolant or moderator and their associated systems with radioactive long-lived products to the extent that blunt operation is impaired. There is. Zirconium and zirconium alloys are excellent cladding materials under normal conditions. The reason is that zirconium and zirconium alloys have a small neutron absorption cross section and, furthermore, at temperatures below about 400°C, they are strong and ductile in the presence of water vapor, commonly used as reactor coolant and moderator. This is because it is extremely stable and non-reactive. Years of operating experience using such nuclear reactor fuel rods has shown that the fuel rods currently in use have a high degree of reliability. However, as shown in Japanese Unexamined Patent Publication No. 55-33037, when a nuclear reactor needs to operate with a fluctuating load or when the output needs to be increased rapidly, fuel failure can easily occur. It has become clear that this occurs frequently. The fluctuating load in this case results in fluctuations in the temperature of the reactor fuel rods and therefore
The fuel rods are also subject to periodic thermal expansion. In this case, it became clear that stress corrosion cracking (SCC) could occur on the inner wall of the cladding tube due to interaction between the fuel pellets, the cladding tube, and the fission products. Iodine was identified as the corrosive medium in that case. One way to solve the above problem is to make the cladding a double structure, so that the inner side of the Zircaloy-2 cladding is
A method of lining with zirconium, which has good SCC properties, was proposed. According to this method, it was confirmed that the SCC resistance performance was superior to that of the Zircaloy 2-coated tube as it was. However, depending on the operating method of the nuclear reactor, it is not always possible to completely prevent SCC damage. One of the reasons for this is that even though zirconium is used, SCC damage occurs in corrosive media environments. [Object of the Invention] The present invention was made in view of the above-mentioned situation, and an object of the present invention is to provide a cladding tube for a nuclear fuel element that improves stress corrosion cracking resistance and has strength and corrosion resistance equivalent to that of conventional cladding tubes. This is the purpose. [Summary of the Invention] The cladding tube of the nuclear fuel element of the present invention is composed of an inner zirconium layer and an outer zirconium alloy layer formed to surround the outside of the inner zirconium layer, and has a nuclear fuel material sealed inside. The grain shape of the inner zirconium layer is
It is formed to a thickness of 5 μm or less. In general, pure zirconium such as crystal bar zirconium or sponge zirconium has better SCC resistance than zirconium alloys such as Zircaloy-2. Conversely, Zircaloy-2 has better corrosion resistance than pure zirconium. On the other hand, the mechanism of SCC caused by iodine in zirconium is that first, the grain boundaries of zirconium are preferentially attacked due to a chemical reaction between iodine and zirconium, and the bonding strength of the grain boundaries is reduced. When stress is applied to zirconium, this decrease in bonding strength causes grain boundary fracture. Inside the zirconium layer, which has not reacted with iodine, grain boundaries do not initially deteriorate, but as grain boundary destruction progresses, it is attacked by iodine and grain boundary destruction occurs. In this way, SCC destruction due to iodine occurs sequentially from the inner surface of the cladding tube along the grain boundaries. Therefore, the SCC resistance of the cladding tube
In order to improve the performance, it is sufficient to increase the area of the grain boundaries of the zirconium layer. The simplest way to increase the grain boundary area of the zirconium layer is to increase the thickness of the zirconium layer. However, zirconium has lower strength than zircaloy, and if the thickness of the zirconium layer is increased, the design stress of the cladding tube cannot be satisfied. Another way to increase the grain boundary area of zirconium layer is to
The goal is to reduce the crystal grain size. Assuming that crystal grains are approximately spheres, the grain boundary area per unit volume is inversely proportional to the grain size. Therefore, the reduction rate of grain size, α {α = (D ₀ - D)/D ₀ , D ₀ and D are the grain sizes before and after reducing the grain size, respectively)
and the increase rate of grain boundary area per unit volume β{β
=(A-A ₀ )/A ₀ , A ₀ , and A are the grain boundary areas before and after reducing the grain size, respectively} and the relationship is as shown in the following equation (1). β=1/(1-α)-1 ...(1) Table 1 shows the relationship between α and β.

[Embodiments of the invention]

DESCRIPTION OF THE PREFERRED EMBODIMENTS The cladding tube for a nuclear fuel element according to the present invention will be explained below using an example with reference to FIG. FIG. 1 is a sectional view. In the figure, reference numeral 1 denotes a cladding tube, which is composed of an outer zirconium alloy layer 2 and an inner zirconium layer 3. The outer diameter of the cladding tube 1 is 12.5 mm. The inner zirconium layer 3 is made of sponge zirconium and has a thickness of 100 μm, and the outer zirconium alloy layer 2 is made of Zircaloy-2 and has a thickness of 100 μm.
It is 760μm. The cladding tube 1 is filled with a large number of fuel pellets (not shown), and both ends are sealed to form a fuel rod. The outer peripheral surface of the fuel pellet faces the inner zirconium layer 3. SCC tests were carried out on the cladding tube 1 in which the zirconium crystal grain size of the inner zirconium layer 3 was changed to 3 μm, 4 μm, 5 μm, 6 μm, and 8 μm. The test results were compared with the test results of 12 μm in diameter.
The test was conducted in an iodine atmosphere (iodine concentration 1 mg/cm ³ ) using an internal pressurization method using argon gas. The test temperature was 350°C, and the applied stress was 25 Kg/mm ² (stress in the circumferential direction of the cladding tube). The test results are shown in FIG. 2, with the crystal grain size plotted on the horizontal axis and the rupture time plotted on the vertical axis. In FIG. 2, the black circle 4 is the rupture time for each crystal grain size of this example, and the white circle 5 is the rupture time for a conventional double-clad tube with a crystal grain size of 12 μm. Figure 2 shows that when the grain size of zirconium is 6 μm or more, there is not much difference in the rupture time by SCC, but when the grain size is 5 μm or less, there is a significant difference in the rupture time.
Moreover, the smaller the crystal grain size, the longer the rupture time. In other words, it is durable in the range where the crystal grain size is 5 μm or less.
Improvement in SCC performance is observed. In this way, the cladding tube of the nuclear fuel element of this example is
By forming the inner zirconium layer with a crystal grain size of 5 μm or less, it has significantly improved SCC resistance compared to conventional double-clad tubes made of zirconium and zirconium alloy, while maintaining the same strength and corrosion resistance as conventional tubes. have. [Structure of the Invention] As described above, the cladding tube for a nuclear fuel element of the present invention has the effect of significantly improving stress corrosion cracking resistance and having strength and corrosion resistance equivalent to that of conventional cladding tubes.

[Brief explanation of the drawing]

FIG. 1 is a cross-sectional view of an embodiment of the cladding tube for a nuclear fuel element of the present invention, and FIG. 2 is an explanatory diagram of the relationship between the crystal grain size and rupture time of the inner zirconium layer in FIG. 1. 1... Cladding tube, 2... Outer zirconium alloy layer, 3... Inner zirconium layer.

Claims (1)

[Claims] 1 In a cladding tube consisting of an inner zirconium layer and an outer zirconium alloy layer formed to surround the outside of the inner zirconium layer, and in which nuclear fuel material is sealed, the crystal grain size of the inner zirconium layer is 5 μm or less. A cladding tube for a nuclear fuel element, characterized in that: