JPS60255945A - Zirconium base alloy - Google Patents

Abstract

PURPOSE:To obtain a Zr base alloy having high corrosion resistance by which nodular corrosion is prevented in water at high temp. and pressure for a long period by specifying a composition contg. Sn, Fe, Ni, oxygen and other trace impurities. CONSTITUTION:This Zr base alloy contains 1.0-2.0wt% Sn, Fe, Ni, oxygen and other trace impurities. The amounts of Fe and Ni in the Zr base alloy are in the range defined by four points (0.15, 0), (0, 0.1), (0, 0.5) and (0.5, 0), preferably four points (0.25, 0), (0, 0.15), (0, 0.4) and (0.4, 0) on the plane of orthogonal coordinates with the Fe content (%) as the (x) coordinate axis and the Ni content (%) as the (y) coordinate axis. The Zr base alloy may further contain 0.1% Cr. The Zr base alloy has high corrosion resistance and is suitable for use in water at high temp. and pressure especially in a nuclear reactor.

Description

DETAILED DESCRIPTION OF THE INVENTION [Field of Application of the Invention] The present invention relates to a zirconium-based alloy, and more particularly to a composition of a zirconium-based alloy that has high corrosion resistance and is suitable for use in high-temperature, high-pressure water in a nuclear reactor.

[Background of the invention]

Zirconium-based alloys have excellent corrosion resistance and a small neutron absorption cross section, so they are used in the reactor internal structural members such as the fuel cladding tube 1, channel box 2, spacer 3, etc., as shown in Figure 1. It is used. Zirconium alloys containing tin used in these applications include Zircaloy-2 (Sn: 1.20-1.70 wt%, F
e: 0.07-0.20 wt%.

Gr: 0.05-0.15wt%, Ni
: 0.03~0.08wt%, O: 900~1
400ppm, remaining Zr, however, Fe+Cr+N l:
0.18-0.38wt%), Zircaloy-4 (S
n: 1.20-1.70 wt%, Fe:
0.18-0.24wt%, Cr: 0.07-0.1
3wt%, O: 1000-1600ppm, remaining Z
r, however, F a +Cr: 0.28-0.37w
t%) etc.

Among the alloying elements, Sn is added to improve mechanical properties and to prevent the adverse effect of nitrogen contained in the zirconium sponge, which is a melted raw material, on corrosion resistance. Addition of oxygen improves tensile strength.

Fa, Cr and Ni are added to improve corrosion resistance.

Among Fe, Cr, and N1, which have a remarkable effect on improving corrosion resistance, it is said that when the amount of Ni added increases, the amount of hydrogen absorbed in high-temperature, high-pressure water or high-temperature, high-pressure steam increases.For example, "The Metallurgy of Zi"
rconium” (written by D. L. DOUGLAS) p.
360. The absorbed hydrogen precipitates in the form of a plate as a hydride, causing a decrease in the strength of the material. Therefore, Ni is approximately 0.05Wt% in Zircaloy-2 material.
The amount added is small, and it is not added in Zircaloy-4 material.

It is said that corrosion resistance is improved by adding 0.1 wt% to 0.5 wt% of Fe and Cr.
allurgy of The Rare Metal
s-2, Zirconium (by Miller) p3
25. Fe.

The neutron absorption cross section of Cr and Ni is larger than that of Zr, and it is preferable that the amount added be as small as possible.

For the reasons stated above, the composition of the current zirconium alloy has been selected.

However, even these commercially available zirconium alloys, which have excellent corrosion resistance, develop papular localized corrosion (hereinafter referred to as nodular corrosion) when exposed to high-temperature, high-pressure water in a furnace for a long time. The occurrence of nodular corrosion reduces the wall thickness of healthy parts, causing a decrease in strength.

If nodular corrosion penetrates the entire wall thickness, radioactive materials inside the cladding will leak into the reactor water. In order to achieve higher burn-up and longer operating cycles of nuclear fuel, it is necessary to further improve the corrosion resistance of currently used zirconium alloys.

A heat treatment technique called β quench described in JP-A-51-110411 and JP-A-51-110412 is known as a technique for increasing the corrosion resistance of currently used Zircaloy-2 and Zircaloy-4 materials. β-quenching refers to rapid cooling of a zirconium-based alloy from the [α+β] phase temperature range or β-phase temperature range (cooling rate: 30°C/sec to 300°C/second).
sec) is a heat treatment.

Zr precipitated in the alloy by β-quenching
Intermetallic compound phases such as (Cr, Fe), Zr, (Ni, Fe), etc. are dissolved in the matrix, and the intermetallic compound phases precipitated during the cooling process are finer than those before β-quenching. β-quenching improves corrosion resistance, but
Since the matrix is a supersaturated solid solution of Fe, Cr, and Ni, the ductility is significantly reduced, and cracking occurs when strong working is performed after β-quenching.

Taking the manufacturing process of fuel cladding tubes as an example, the molten ingot is hot forged (1000℃) and solution treated (approximately 100℃).
00℃ for several hours), (Hot forging (700℃~750℃)
After that, it is formed into a cylindrical billet by hot extrusion.
Usually, after annealing, this cylindrical billet is formed into a fuel cladding tube by repeating cold rolling and annealing three times alternately.

In order to obtain a highly corrosion-resistant fuel cladding tube, β-quenching is performed in the final step, resulting in a decrease in ductility and the cladding tube specifications no longer being met. In order to impart ductility, β-quenching is performed before any cold rolling process, and after β-quenching, cold rolling and annealing are alternately repeated, so that the metal structure becomes a recrystallized structure. A process is also proposed. However, since β-quenched material cannot be subjected to strong working, the number of repetitions of cold rolling and annealing is increased by one to two times compared to the normal manufacturing process.

After β-quenching, when annealing is performed for a long time, Fe becomes a supersaturated solid solution in the matrix.

Since Cr and Ni precipitate as intermetallic compound phases and become coarse, corrosion resistance gradually decreases.

Therefore, it is desirable that zirconium alloys used as internal structural members of nuclear reactors, such as channel boxes, fuel cladding tubes, and spacers, have high corrosion resistance and do not change in corrosion resistance due to heat treatment.

[Purpose of the invention]

An object of the present invention is to provide a zirconium-based alloy that does not cause nodular corrosion and has high corrosion resistance even when used in high-temperature, high-pressure water or high-temperature, high-pressure steam for a long period of time.

[Summary of the invention]

The present invention is based on the fact that among the alloying elements Sn, Fe, Cr, and Ni, Fa and Ni have a remarkable effect on improving corrosion resistance, and that Zr containing 0.3 wt% of Ni is added.
This is based on the discovery that the amount of hydrogen absorbed in the -Sn-Ni alloy is approximately the same as or lower than that of conventional zirconium-based alloys.

Figure 2 shows the growth mechanism of an oxide film formed on the surface of a zirconium-based alloy.The dioxide film is a metal-rich (oxygen-deficient) n-type semiconductor, and its composition deviates from the stoichiometric composition. It is X. Excess metal ions are compensated by equivalent electrons to maintain electrical neutrality,
Oxygen-deficient areas are present in the oxide film as anion defects. Oxygen ions diffuse into the interior by exchanging their positions with these anion defects, bond with zirconium at the interface between the oxide film and the metal, and oxidation progresses inward. At this time, electrons with a charge equivalent to that of oxygen ions move from the inside of the oxide film to the surface, and hydrogen ions are reduced by these electrons to generate hydrogen gas. Therefore, the amount of oxidation and the amount of hydrogen gas generated are in a proportional relationship, and a portion of the hydrogen gas is absorbed inside the zirconium alloy, causing the formation of hydrides. From this.

It is thought that zirconium-based alloys with high corrosion resistance have a low absorption amount of hydrogen gas.

The growth rate of the oxide film is determined by the diffusion rate of oxygen in the oxide film, and the diffusion rate is proportional to the number of anion defects and their ease of movement. In order to suppress oxidation and improve corrosion resistance, it is effective to reduce the number of anion defects. F
Elements that improve corrosion resistance such as e, Cr and Ni are ZrO,
It is thought that it penetrates between the ion lattices, becomes ionized, and becomes a donor for missing electrons, thereby having the effect of reducing anion defects.

The amount of alloying elements added is set at 0.5 to avoid increasing the neutron absorption cross section. It is preferable that it is less than wt%.

Table 1 Chemical composition of alloy [Example of the invention] <Example 1> Figure 3 shows the melting, heat treatment and processing method of a zirconium-based alloy. Zirconium sponge for a nuclear reactor was used as the melting raw material. Z with the composition shown in Table 1 by vacuum arc melting
r-8n alloy, Zr-8n-Fe alloy, Zr-8n-C
An r alloy and a Zr-3n-Ni alloy were melted. Each ingot is hot rolled (700℃), annealed at 700℃,
4 hours) and then divided into 4 parts. Three of the divided plates have (α+β) phase temperature ranges (840°C and 90°C).
0° C.) and the β phase temperature (1000° C.) for 5 minutes, the mixture was cooled with water and subjected to β quenching. The remaining one sheet was not subjected to β-quenching. These four plates were cold-rolled three times and intermediately annealed at 600°C for 2 hours, resulting in a plate thickness of 1+.
I made it with cod. Each board was further divided into 3 parts and heated at 530℃ and 620℃.
℃ and annealed for 2 hours at a temperature of 730℃. T and P were cut out from plate materials with different alloy compositions and heat treatments and subjected to corrosion tests. The corrosion test was conducted in high-temperature, high-pressure steam at a pressure of 10.3 MPa. Corrosion test temperature and time are 4
The temperature was 10°C for 8 hours and 510°C for 16 hours, and the temperature was changed continuously without cooling in between.

Figure 4 shows the influence of alloying element content, β quench temperature, and final annealing temperature on corrosion resistance. In Figure 0, the mark ・ indicates that nodular corrosion has occurred, and the mark 0 indicates that nodular corrosion has not occurred. to show that

In the Zr-5n alloy, nodular corrosion occurs in all 5Zr-8n-Fe alloys, regardless of the Sn addition amount and heat treatment.
It can be seen that the corrosion resistance of the alloy is hardly affected by the final annealing temperature, and when Fa exceeds the total by 0.25 wt%, nodular corrosion does not occur even without β-quenching. When β-quenched at 1000°C, Fe becomes 0.
It can be seen that the occurrence of nodular corrosion can be prevented if the total content is 15 wt% or more.

The corrosion resistance of Zr-3n-Cr alloy decreases as the final annealing temperature increases;
It can be seen that the occurrence of nodular corrosion cannot be prevented even if the quench temperature and the amount of alloying are changed.

The corrosion resistance of the Zr-3n-Ni alloy is not affected by the final annealing temperature, and if Ni is alloyed at 0.15 wt% or more, nodular corrosion can be prevented even without β-quenching. It can be seen that when the β-quench temperature is 900° C., the amount of N1 alloyed to prevent the occurrence of nodular corrosion may be O.1 wt%.

From the above results, it can be seen that the alloying elements effective in improving corrosion resistance are Fe and N1, and the effect of Cr is small.

<Example 2> Zr was melted using the same melting method, processing, and heat treatment as in Example 1.
-5n-Fa-Ni alloy plates were manufactured.

The final annealing temperature was 730°C. FIG. 5 shows the corrosion resistance of each alloy, and the presence or absence of nodular corrosion was examined by the same corrosion test as in Example 1. In the figure, the Δ mark indicates that nodular corrosion has occurred at only one point. From Figure 5, in the region where the alloying amount of Fe and Ni satisfies formula (1), the heat treatment is 0.25 x ≧0.0375
=・(1) xsl: Alloying amount of Ni* X Fs:
It can be seen that the occurrence of nodular corrosion can be prevented regardless of the amount of F a alloyed. It can be seen that when the β-quenching temperature is 1000° C., the occurrence of nodular corrosion can be prevented by alloying Fe and Ni that satisfies equation (2).

0.15 XNI +0. IXF@≧0.015 −(
2) <Example 3> FIG. 6 shows the relationship between the amount of hydrogen in each test piece and the increase in corrosion due to oxidation after the corrosion test described in Example 1. The amount of hydrogen before the corrosion test was 10 to 2 Qppm. From Figure 6,
It can be seen that the amount of hydrogen absorbed is proportional to the increase in corrosion, and the higher the corrosion resistance, the lower the amount of hydrogen absorbed. The amount of hydrogen absorbed did not increase due to the addition of Ni.

<Example 4> When 3.0 wt% of Sn was added, work hardening during cold rolling was significant, and cracking occurred when cold rolling was performed at 30% or more. From this, the alloying amount of Sn is 0.1 to 2.0 vr.
A range of t% is preferred.

【Effect of the invention〕

According to the present invention, it is possible to manufacture a zirconium alloy member with excellent corrosion resistance. As a result, the reliability of the components improves, and the lifetime in the reactor can be significantly extended, making it possible to increase the burnup of nuclear fuel.

Also in the process of manufacturing zirconium alloy members, the heat treatment temperature can be selected relatively freely, which has the effect of facilitating the manufacture. Furthermore, the neutron absorption cross section is equivalent to that of conventional Zircaloy-2 and Zircaloy-4 materials, and power generation efficiency does not decrease.

[Brief explanation of the drawing]

Figure 1 shows the fuel bundle, Figure 2 shows the mechanism of oxygen diffusion in the oxide film, Figure 3 shows the manufacturing process for zirconium alloy plates, and Figures 4 and 5 show how nodular corrosion occurs. FIG. 6 is a diagram showing the influence of alloying elements and heat treatment temperature, and FIG. 6 is a diagram showing the relationship between corrosion weight increase and hydrogen absorption amount. ■...Fuel cladding tube, 2...Channel box, 3
...Spacer, 4..Water rod, 5..Fuel bundle. Agent Patent Attorney Akio Takahashi

Claims (1)

[Claims] 1. tin of 1.0 to 2. In a zirconium-based alloy containing iron, nickel, oxygen, and other trace impurities, the iron content is plotted on the X axis (unit: wt%). (0,15,O), (0,0,1), (0°0.5), (0,5,O) on the orthogonal coordinate plane with the nickel content as the y-axis (unit: wt%). A zirconium-based alloy characterized by having a composition of a plane surrounded by four points. 2. In claim 1, preferably (0,
25,O) , (0,0,15) , (0,0,4)
. A zirconium-based alloy characterized by having a composition within a plane surrounded by four points (0,4,O). 3. A zirconium-based alloy according to claims 1 and 2, characterized in that it contains 0.1 wt% or less of chromium.