JPS6050133A - Corrosion resistant zirconium alloy - Google Patents

Abstract

PURPOSE:To obtain a Zr alloy having superior nodular corrosion resistance and mechanical characteristics and causing hardly bending and warping by leaving compressive stress on the surface of a Zr alloy having an alpha-phase crystal structure. CONSTITUTION:A Zr alloy is melted, forged, not extruded, and subjected to repeated cold working and annealing, and the final annealing is carried out at a prescribed temp. for a prescribed time to obtain a Zr alloy contg. no residual strain and having an alpha-phase crystal structure. After the final annealing, the Zr alloy is passed through a heat treatment stage to leave residual stress on the surface. In the stage, the surface part of the alloy is rapidly heated to a temp. in the alpha-phase region by high frequency heating, and it is rapidly cooled after holding at the temp. for several sec. Thus, the desired Zr alloy as a structural material is obtd.

Description

DETAILED DESCRIPTION OF THE INVENTION [Technical Field of the Invention] The present invention relates to a corrosion-resistant zirconium alloy used as a structural material that requires corrosion resistance, such as a core member of a nuclear reactor or a structural material of a chemical device.

[Technical background of the invention and its problems]

In general, zirconium alloys such as Zircaloy-2 and Zircaloy-4fx have a small thermal neutron absorption cross section, excellent corrosion resistance in the reactor environment, and have sufficient mechanical properties as a structural material. For this reason, it is often used as a core member of nuclear reactors.

For example, a fuel assembly for a light water reactor has a structure as shown in FIG. This fuel assembly consists of nuclear fuel pellets' between an upper tie plate 1 and a lower tie plate 2.
f! : A plurality of nuclear fuel rods 4 loaded inside the cladding tube 3 are attached.

These nuclear fuel rods 4 are inserted and supported by spacers 5 arranged horizontally at appropriate intervals, and are aligned so that movement in the lateral direction is suppressed and a coolant flow path is formed between the nuclear fuel rods 4. It is supposed to be done. All of these are housed in a channel box 6 and are integrated. Several hundred of these fuel assemblies are loaded into the nuclear reactor.

As shown in FIG. 2, the spacer 5 has a grid-like structure in which a plurality of pars 8 and a plurality of dividers 9 are arranged vertically and horizontally in a rectangular plane surrounded by an outer frame 7 and intersect with each other in a horizontal plane. is formed.

Further, the ends of the pars 8 and the dividers 9 are welded to the outer frame 7, and the intersections between the pars 8 and the dividers 9 are also fixed by welding. par 8
A lantern spring 10 is supported in the shape of a rectangular frame at the intersection of C and par 8, respectively. In this way, a plurality of small partitions are formed, into which the fuel rods 4 shown in FIG. 1 are inserted.

Among the core members forming these fuel assemblies, for example,
Core members such as the channel box 6, fuel cladding tube 3, and components of the lantern springs 5 are generally made of zirconium alloys such as Zircaloy-2 and Zorcaloy-4. .

However, as the core members made of these zirconium alloys are used over time, white corrosion products due to a corrosion reaction called so-called nodular collisions tend to form on their surfaces in spots. be. This occurs when the zirconium alloy reacts with high-temperature water, forming an oxide film on its surface, and the generated hydrogen accumulates between the alloy base material and the oxide film on the surface, forming corrosion products. It is. This corrosion product accumulates on the surface over time and eventually peels off from the surface, which may lead to a decrease in the strength of the alloy member. In addition, when the generated hydrogen penetrates into the alloy, zirconium hydrides are formed, and when these are formed perpendicular to the surface, there is a problem of so-called hydrogen embrittlement due to continuous hydrides. In order to solve the problem, we have developed a method (Japanese Patent Application Laid-Open No. 100947/1983) in which zirconium alloy is subjected to a series of processes such as melting, forging, hot extrusion, cold working, and annealing, and then β-quenched in the final annealing process (Japanese Patent Application Laid-open No. 55-100947) and a zirconium alloy member. Method of rapid cooling after rapidly melting the surface of
53) etc. have been proposed. All of these methods improve nodular corrosion resistance by changing the crystal structure of at least the surface of the zirconium alloy member into a β phase (body-centered cubic lattice) of needle-shaped crystal grains through quenching. .

However, although these hardening methods improve the nodular corrosion resistance, they deteriorate the mechanical properties of the single core member, and since the wall thickness is a thin plate with a wall thickness of about 1 mm or less and is long, during hardening, There was a problem of bending and twisting, and the result was not satisfactory.

[Purpose of the invention]

The present invention has been made in view of these conventional problems.
The present invention provides a corrosion-resistant zirconium alloy that has excellent nozzle flow resistance, excellent mechanical properties, and little bending or twisting, and is suitable as a structural material for nuclear reactor core members and chemical equipment. .

[Summary of the invention]

The present invention is characterized in that the crystal structure of the zirconium alloy is formed in the alpha phase, and compressive stress remains on the surface, thereby improving nodular corrosion resistance and mechanical properties.

The present invention will be explained in detail below.

The zirconium alloy used in the present invention includes, for example, tin 1.2 to 1.7% and iron 0.07 to 0.2% by weight.
0%, 05-0.15%, nickel 0.03
Zircaloy consisting of ~0.08%, balance zirconium
2, 1.2-1.7% tin, 0 iron
．． Zircaloy-4, which consists of 18-0.24% chromium, 0.07-0.13% chromium, and the balance zirconium, or zirconium-2.5% niobium-based, zirconium-ltiniobic-based, or ozenite It can be applied to zirconium alloys such as.

Next, a method for manufacturing the zirconium alloy of the present invention will be explained.

After the zirconium alloy is melted, forged, and heated, cold working and annealing are repeated several times.The final annealing temperature is usually around 580°C and heated for about 2 hours, or as described above. The obtained zirconium alloy has no residual strain,
Moreover, the crystal structure is α phase (hexagonal crystal structure).

Up to this point, the method is the same as the conventional method, but in the present invention, the following heat treatment step is added after the final annealing, so that all the compressive stress remains on the surface.

All 11 final annealed zirconium alloys are heated rapidly to 780 to 860°C in the entire alpha region of the surface using, for example, high-frequency heating, held for a few seconds, and then rapidly cooled to form fine spherical particles while keeping the crystal structure in the alpha phase. It is possible to leave the crystal grains as they are and to have all the compressive stress remain on the surface.

When rapid heating and cooling are performed in this manner, a large temperature difference occurs inside the product, and the surface portion cannot contract freely because it envelops and cools the high-temperature portion inside. As a result, the surface part cools while taking up all the tension due to the internal high temperature part. In the early stage of cooling, the temperature is relatively high, so the yield point is low, and the surface portion undergoes some permanent deformation. Moreover, when the center reaches room temperature,
The surface area is overextended only outside the deformation area compared to the center area, and this is subjected to contraction force from within, and ultimately,
The surface portion is compressed by the center portion. That is, compressive stress remains on the surface, and tensile stress acts on the center N, resulting in a balanced state.

Although it is not clear why the nodular corrosion resistance improves even when the crystal structure remains in the α phase when compressive stress remains on the surface, the crystal lattice This is thought to be because oxygen is difficult to diffuse from the surface to the inside because the particles are packed tightly together. When oxygen diffusion is inhibited in this way, an oxide film is less likely to be formed, hydrogen is prevented from accumulating between the oxide film and the alloy surface, and the nodular corrosion resistance is improved.

The heating temperature in the α region is 780 to 860 degrees Celsius, and after being held there for a short time, when it is rapidly cooled, a temperature of 8 to 46 k is applied to the surface.
A compressive stress of g/tea2 remains, and in particular, when the residual compressive stress is 20 kgZ2 or more, effective nodular corrosion resistance can be obtained.

In this case, heating below 780°C will not leave sufficient compressive stress, and heating above 860°C will result in a β region, and rapid cooling will cause quenching and the surface layer will change to β phase (body-centered cubic lattice). Furthermore, the bending and twisting may be large and the mechanical properties may deteriorate. Also, since compressive stress remains on the surface of the zirconium alloy of the present invention, for example, the tensile strength is about 45 kli' /
A compressive stress of 20 kg/tea was applied to a cladding tube made of 111112 zirconium alloy by the above method.
If left to remain, it will be able to withstand up to an external stress of 55 klil/mm2 without breaking.

Furthermore, due to residual compressive stress, the alloy of the present invention has a high yield strength and low elongation, which is particularly important for nuclear reactor core components exposed to boiling, high-temperature water for long periods of time, and structural materials for chemical equipment. , has a large effect on improving mechanical properties such as creep properties.

[Embodiments of the invention]

Using Zircaloy-4, normal melting and forging were performed to form a billet hTh, followed by full hot extrusion, and then cold rolling and vacuum annealing were repeated four times to form a plate material.Next, this plate material was Using a thyristor-type high-frequency heating furnace,
The surface layer was rapidly heated to 800°C.

In this case, the residence time in the furnace (holding time) was about 5 seconds. Thereafter, the mixture was immediately cooled with water for rapid cooling. Next, the oxide film on the surface was completely removed by polishing.

The thus obtained plate material was left in high-temperature, high-pressure steam at 500° C. and 105 atm, and subjected to an accelerated corrosion test to examine its corrosion resistance. This result is shown in the graph of Figure 3 by curve a.
As shown in , the weight increase due to corrosion was slight even after 48 hours had elapsed.

In addition, in order to investigate the mechanical characteristic sound, we investigated the yield strength and elongation, and found that the yield strength was 42.8 kg/mm2, and the elongation was 32.9.
It was confirmed that the material had excellent creep properties.

Furthermore, in order to see the entire surface stress state of the plate material of the present invention, the residual compressive stress was measured using X-rays, and it was found to be 28 kg/mm2.
Met. Note that the X-ray measurement in this case was performed using CrKα (chromium Kα) rays and diffraction line 2 from the (2022) plane.

Next, in order to compare with the present invention, all accelerated corrosion tests were similarly conducted on a conventional zirconium alloy plate material that was not subjected to any heat treatment after final annealing.

As a result, the weight increase due to corrosion drew a sharp increasing curve as shown in the graph of FIG. Also, when I checked the mechanical characteristic sound, the yield strength was 42.7 kl? / Depth 2, elongation was 32.9 inches, which was equivalent to the example of the present invention. Further, no residual was found in the measurement of compressive stress using X-rays.

〔Effect of the invention〕

As explained above, according to the corrosion-resistant zirconium alloy according to the present invention, by leaving compressive stress on the surface, it is possible to improve nodular corrosion resistance and mechanical properties, and it is possible to improve the nodular corrosion resistance and mechanical properties. Compared to the formation of the β phase, the rapid cooling from a low temperature produces remarkable effects such as less bending and twisting and improved dimensional accuracy.

[Brief explanation of drawings]

FIG. 1 is a vertical cross-sectional view showing a fuel assembly partially cut away, and FIG. 2 is a graph showing changes over time in corrosion weight increase between a flat material showing the main part of a spacer and a conventional plate material. DESCRIPTION OF SYMBOLS 1... Upper tie plate, 2... Lower tie plate, 3... Cladding tube, 4... Nuclear fuel rod, 5... Spacer, 6... Channel box, 7... Outer frame, g
... Par, 9... Divider, 10... Lantern Spring. Applicant's agent Patent attorney Takehiko Suzue @Figure 1 Figure 2 Figure 3

Claims (2)

[Claims] (1) A corrosion-resistant zirconium alloy characterized in that the crystal structure of the zirconium alloy is formed in the α phase and compressive stress remains on the surface. (2) The corrosion-resistant zirconium alloy according to claim 1, which has a residual compressive stress of 2 Q kg/mm2 or more.