JPS6050859B2 - Corrosion resistant zirconium alloy for nuclear reactors - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION [Technical Field of the Invention] The present invention relates to a corrosion-resistant zirconium alloy for nuclear reactors used as a structural material for nuclear reactors.

[Technical background of the invention and its problems]

For example, in boiling water type light water reactors, zirconium alloys for nuclear reactors called Zircaloy 2, Zircaloy 4, etc. are used as structural materials for fuel cladding tubes, channel boxes, spacers, and the like.

In other words, so-called reactor core structures such as fuel cladding tubes and channel boxes are generally compliant with ASTM standards G2-74 or B3 from the viewpoint of neutron economy and corrosion resistance at high temperatures.
Zirconium alloys that have passed the corrosion resistance test according to No. 53-77a are in practical use.

By the way, when we look at the core structure made of the above-mentioned zirconium alloy, during mounting operation, speckled white products due to a corrosion reaction called nodular corrosion are generated on the surface of the structure. The white corrosion products mentioned above gradually grow as the nodular corrosion progresses.
Occasionally, it may flake off, and this flaking may lead to a decrease in the mechanical strength of core structures such as channel boxes and fuel cladding tubes. In order to improve the safety and reliability of core structures, corrosion resistance against the nodular corrosion described above is attracting attention, and attempts have also been made to coat the surface of structures made of zirconium alloys with a thin layer of electronically conductive material. (Unexamined Japanese Patent Publication No. 1973-
No. 5629). However, this method of preventing or reducing the occurrence of nodular corrosion (providing corrosion resistance to nodular corrosion) by coating with an electron conductive material cannot be said to be a sufficient means from the viewpoint of coexistence with dissimilar metals, contact corrosion, etc. [Object of the Invention] The present invention was made in view of the above circumstances, and it is an object of the present invention to provide a zirconium alloy for nuclear reactors that has excellent corrosion resistance against nodular corrosion and is suitable for reactor core materials such as channel boxes, fuel cladding tubes, and spacers. It is something.

[Summary of the Invention] The present invention is based on the fact that the corrosion resistance of a welded part of a zirconium alloy for a nuclear reactor against nodular corrosion is superior to that of the base metal.

When the present inventors closely observed the structure of the welded part of such a zirconium alloy with excellent corrosion resistance, it was found that the entire structure contained intermetallic compounds with a grain size of about 0.04 to 1.5 μm and grain sizes of about 0.01 to 1. Metallic tin of .5 μm was segregated in a chain along grain boundaries or sub-grain boundaries. Therefore, if the above-mentioned precipitates are precipitated in the zirconium alloy in a chain along the grain boundaries or sub-grain boundaries by a suitable heat treatment or processing method, the corrosion resistance of the zirconium alloy against nodular corrosion will be improved. Zirconium alloys for nuclear reactors are generally used as structural materials for nuclear reactors, and contain Zr as a main component, Fe, Nl, Cr, Sn, and Nb, for example, Snl to 1.8 wt%, FeO. l
~0.2V/t%NiO~0.1Wt%, CrO-0.
Examples include Zircaloy 1, Zircaloy 2, Zircaloy 3, Zircaloy 4, Zircaloy 4, Zircaloy 4, Zircaloy 4, and Zircaloy 4, which are generally known for use in nuclear reactors, although the composition is not particularly limited. Examples include Auzenite 1.0 and Zr-2.5%Nb.

Furthermore, the intermetallic compound according to the present invention is composed of Zr, Sn (zircaloy 1
), Zr, Sn, Fe, Ni, Cr (Zircaloy 2,
3), Zr, Sn, Fe, Ni, Nb (auzenite 0
．． 5,1.0), Zr, Nb (Zr-2.5%Nb),
Al is the main impurity in the above zirconium alloy for nuclear reactors.
, C, Cn, Hf, O, Mn, Si, Ti, W, for example, Zrc, zrcr2
,zrFe2,Zr2NlO. 4FeO. 69zrcr
l. lFeO. 99ZrXFe5Cr2, Zr-Ni-
Fe, Zr22FelONi5Cr, zr29Fe7,
Examples include Nil2, zr63Fe, cr4, and the like. The composition is not particularly limited to this type. 3 The alloy of the present invention improves the corrosion resistance of a zirconium-based alloy for nuclear reactors by segregating the above-mentioned intermetallic compounds and metallic tin along grain boundaries or sub-grain boundaries. Nodular corrosion is formed when a zirconium alloy and water come into contact with each other.4 If the pressure of H2 gas generated at the interface between the ZrO2 film and the zirconium alloy exceeds the withstand pressure of the ZrO2 film, the ZrO2 film will be destroyed. caused by something. However, due to the structure in which metallic tin and intermetallic compounds precipitate in a chain at grain boundaries or sub-grain boundaries, H2
Gas is generated on the outer surface of the ZrO2 film (the surface not in contact with the zirconium alloy), no nodular corrosion occurs, and corrosion resistance is improved. In the present invention, any method may be used as long as it provides such a structure.

For example, ZrFe2, It has a structure in which sn etc. are segregated. After that, the density of the sintered body can be increased by rolling or the like. Furthermore, when a tin-containing zirconium plate and a stainless steel plate such as SUS3O4 are heat-treated in close contact with each other, grain boundary diffusion is faster than intragranular diffusion, so the diffused substances such as Fe, Ni, Cr, etc. from the stainless steel plate are absorbed into the zirconium plate. Concentrated at grain boundaries. At this time, Zr. ! :． Fe, Ni, Cr, etc. form an intermetallic compound. In addition to the above method, a method may also be used in which the added elements in the zirconium alloy for nuclear reactors are segregated. For example, ozenite 0.5. Zircaloy 4
Additive elements are dissolved in solid solution by applying heat treatment to zirconium alloys such as zirconium alloys. At this time, some of the added elements are not completely dissolved and precipitate at grain boundaries and sub-grain boundaries. At this time, it is also possible to increase the grain boundary density and dislocation density by performing cold milling. Then, by annealing, the elements that were in solid solution precipitate in a chain shape using the precipitates already existing at the grain boundaries and sub-grain boundaries as nuclei, and at the same time combine with Zr to form an intermetallic compound, and the metal tin and the metal It is possible to obtain an alloy of the present invention in which intermediate compounds are segregated.
[Embodiments of the Invention] Hereinafter, embodiments of the present invention will be described in detail with reference to the drawings, taking as an example a zirconium alloy (Zircaloy 4) constituting a channel box.

FIG. 1 shows the structure of a conventional zirconium alloy. As is clear from the figure, the intermetallic compound 1 and the tin-containing inorganic compound 2 (both have particle sizes of about 0.2 to 1.5 μm) are precipitated almost uniformly over the entire structure. FIG. 2 shows the structure of the corrosion-resistant zirconium alloy for nuclear reactors according to the present invention. The zirconium alloy shown in Figure 2 is Zircaloy 4 with 1120
℃ for 5 minutes, then 1120℃ → 1000℃ for 50 minutes.
0 to 700 C/SllOOO0C → 800℃ to 100
It was formed by cooling at a rate of ~500°C/S and then annealing in vacuum at 750°C for 30 minutes.

By the cooling, most of the added elements become a solid solution, but some elements are dispersed and precipitated at the grain boundaries and sub-grain boundaries. During the subsequent annealing, the solid-dissolved elements precipitate at the grain boundaries or sub-grain boundaries using the precipitates that have already appeared as nuclei, forming intermetallic compounds. As is clear from the figure, the particle size is approximately 0.04 to 1
．． 5 μm intermetallic compound 3 and particle size approximately 0.01-1.5 μm
Metallic tin 4 having a square body center of m is segregated in a chain along grain boundaries or sub-grain boundaries. Note that this intermetallic compound is Zr, C
It contained r and Fe, and its crystal structure was a face-centered cubic Hcr2 type. For example, Zircaloy 2, which has undergone similar treatment, contains Zr, Fe, and Cr in addition to metal tin, and is an intermetallic compound with a face-centered cubic Zrcr2 type crystal structure, while it contains Zr and Sn and has an orthorhombic HSn type crystal structure. An intermetallic compound containing Zr and Sn and having a hexagonal Zr5sn3 crystal structure was precipitated. Next, the nodular corrosion behavior of the corrosion-resistant zirconium alloy according to the present invention will be specifically shown in comparison with that of a conventional zirconium alloy.

First, corrosion-resistant zirconium alloy pieces according to the present invention (27 orders x 2
0Twt×3? ) and conventional zirconium alloy piece (27T
I0rL×2−×3? ) and have a particle size of about 2
After surface polishing with 5μm diamond powder, 500℃, 10
7k9/c! It was maintained in a water vapor environment of L.

The test environment is 290°C176k9/d'S,
This is an accelerated test of a nodular corrosion that simulates a real reactor environment in a rising water atmosphere and taking into account the effects of neutron irradiation.
In the above test, in the case of the corrosion-resistant zirconium alloy according to the present invention, no speckled white products were observed on the surface after a holding time of 4 [phases], indicating excellent corrosion resistance.

On the other hand, speckled white products appeared on the surface of the conventional zirconium alloy after several hours, and gradually grew larger over time. In addition, the weight change (corrosion amount) of the corrosion-resistant zirconium alloy according to the present invention and the conventional zirconium alloy
The trends were as shown in curves a and b in FIG. 3, respectively. Furthermore, zirconium alloys for nuclear reactors other than Zircaloy 4, such as Auzenite O2, also exhibited superior corrosion resistance against nodular corrosion compared to corresponding conventional zirconium alloys.

[Effects of the Invention] As is clear from the above, the present invention provides a zirconium alloy in which at least one of the intermetallic compound and metallic tin is segregated in a chain along grain boundaries or sub-grain boundaries. has excellent corrosion resistance against nodular corrosion.

Thus, it can be said that the corrosion-resistant zirconium alloy according to the present invention can perform the required functions as a structural material for a long time even when used as a material for core structures such as channel boxes, fuel cladding tubes, and spacers.

[Brief explanation of the drawing]

FIG. 1 is a conventional zirconium alloy structure, FIG. 2 is a zirconium alloy structure according to the present invention, and FIG. 3 is a curve diagram showing the state of weight change due to corrosion. 1, 3...Intermetallic compound, 2,4...Tin-containing inorganic compound.

Claims (1)

[Claims] 1 Intermetallic compounds with a particle size of 0.04 to 1.5 μm and a particle size of 0.0 throughout the zirconium alloy substrate for nuclear reactors
A corrosion-resistant zirconium alloy for nuclear reactors, characterized in that metallic tin of 1 to 1.5 μm is segregated in a chain along grain boundaries or sub-grain boundaries.