JPS6126738A - Zirconium alloy - Google Patents

Abstract

PURPOSE:To improve the corrosion resistance is steam at high temp. and pressure by adding prescribed percentages of Sn, Ni and Fe to Zr without adding Cr. CONSTITUTION:This Zr alloy consists of, by weight, 1-2% Sn, <=0.3% Ni, <=0.55% Fe+oxygen+impurities (Fe+Ni>=0.15%), 0% Cr and the balance Zr. The alloy has superior corrosion resistance and undergoes no nodular corrosion even when it is used in steam at high temp. and pressure for a long period.

Description

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

[Background of the invention]

'Zirconium base metal has excellent corrosion resistance and small neutron absorption cross section, so it is used in the internal structural components of nuclear reactors such as fuel cladding tubes and channel boxes.

Used for spacers, etc. The tin-containing zirconium alloy used for these purposes is Zircaloy-2 (
Sn: 1.20 to 1.70 wt%.

Fe: 0.07-1.2 wt%, Cr; 0.05
~0.15wt%, Ni: 0.03~0.08
wt%yO: 900-1400ppm, residual Zr,
However, Fa+Cr+Ni: 0.18~0.38w
t%).

Zircaloy-4 (Sn: 1.20-1.70-wt
%, Fa: 0.18-0.24 wt%, Cr:
0.07-0.13-wt%, O: 1000-16
00pp+a.

Remaining Zr, however, Fe+Or: 0.28~0.37w
t%) etc.

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

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

Among Fe, Cr, and Ni, 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.1For example, "The Metallurgy of Z"
irconium" (by D. L. Douglas) P360. Therefore, the amount of Ni added in Zircaloy-2 material is as small as about 0.05 wt%, and it is not added in Zircaloy-4 material.

Soviet Zircaloy improved alloy "Zirkaloy" is 0.
8wt%Sn, 0.2wt%Fe, 0.3wt
%Ni and does not contain Car, as described in "Heat-Resistant Materials Handbook" (authored by Imai and Kawashima), Shokusho Shoten (S, published on November 30, 40). This alloy contains a large amount of Ni, which has the largest neutron absorption cross section among alloying elements, so the neutron absorption cross section of the alloy also differs from that of ordinary Zircaloy-2 or Zircaloy-2.
greater than 4.

Since the neutron absorption cross section of Fee Cr and Ni is larger than that of Zr, it is preferable to keep the amount as small as possible.
For example, “Metalurgy of The Rar”
eMetalg-2, Zirconium” (Mil
ler), these alloying elements are added to current zirconium-based alloys because they are effective in improving corrosion resistance.

These commercially available zirconium alloys, which have excellent corrosion resistance, also develop papular localized corrosion (hereinafter referred to as nodular corrosion) when exposed to high-temperature, high-pressure water in a furnace for a long period of time. Occurrence of nodular corrosion causes a decrease in strength because the wall thickness of healthy parts is reduced, and if nodular corrosion penetrates the entire wall thickness, radioactive materials in the cladding tube will leak into the reactor water. Higher burnup of nuclear fuel,
In order to extend the operating cycle, it is necessary to further improve the corrosion resistance of currently used zirconium alloys.

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

The intermetallic compound phase such as Fe) is dissolved in the matrix,
The intermetallic compound phase precipitated during the cooling process becomes finer than that before β-quenching. Corrosion resistance is improved by β quenching, but the matrix is Fe.

Since it is a supersaturated solid solution of Cr and Ni, its 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 melted ingot undergoes hot forging (approximately 1,000°C) and solution treatment (approximately 1,000°C).
000℃ 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 cold rolling and annealing are alternately repeated after β-quenching, so that the metal structure becomes a recrystallized structure. An m* process has also been proposed. However, since β-quenched material cannot be subjected to strong working, the number of repetitions of cold rolling and annealing is increased by 1 to 2 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]

The object of the present invention is to provide a zirconium-based alloy that does not cause nodular corrosion even when used for long periods in high-temperature, high-pressure water or high-temperature, high-pressure steam, has high corrosion resistance, and has a neutron absorption cross section that does not exceed that of currently used zirconium-based alloys. Our goal is to provide the following.

[Summary of the invention]

FIG. 1 shows the alloy composition range of Fe and Ni according to the present invention. Alloys in this composition range have the excellent properties of having a neutron absorption cross section equal to or lower than that of current Zircaloys, being able to withstand long-term use, and absorbing less hydrogen than current Zircaloys. This will be explained in detail below.

Figure 2 shows the growth mechanism of an oxide film formed on the surface of a zirconium-based alloy. It is. 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. Reducing the number of anion defects is effective in suppressing oxidation and increasing corrosion resistance++
It is thought that elements that improve corrosion resistance, such as Fsg Cr and Ni, penetrate into the ion lattice of ZrO, become ionized, and become donors for missing electrons, thereby having the effect of reducing anion defects.

FIG. 3 shows the relationship between the neutron absorption cross section of the Zr-1,5wt%5n-Fe-Ni alloy and the amounts of Fe and Ni added. From the figure, the additive amount that does not exceed the neutron absorption cross section of the current Zircaloy-2 material is Fe50.55 wt%, Ni≦0
．． 3 wt%.

(Embodiments of the invention) Examples will be described in detail below.

Example 1 No. 4W4 shows a method of melting, heat treatment, and processing a zirconium-based alloy. Zirconium sponge for nuclear reactors was used as the melting raw material. Zr-8n alloy 5Zr-8n-Fe alloy, Zr-8n-
A Cr alloy and a Zr-8n-Ni alloy were melted. Each ingot is hot rolled (700℃) and annealed to 700'
C, 4 hours) and then divided into 4 pieces.Three of each plate was divided into 4 parts, and three of them were subjected to (α+β) phase temperature range (840"C and 900℃) and β phase temperature (1000"C). After being held for a minute, it was cooled with water and β-quenched. The remaining one was subjected to β quenching. These four boards are 3w!
J cold rolling and intermediate annealing at 800'C for 2 hours to give a plate thickness of 1. Divide each plate into 3 parts, 58
Test pieces were cut out from each plate material with different alloy compositions and heat treatments annealed at temperatures of 0°C, 620'e, and 730°C for 2 hours and subjected to corrosion tests. Corrosion test at pressure 10.3
The test was conducted in high-temperature, high-pressure steam of MPa. The corrosion test temperature and time were 410° C. for 8 hours and 510° C. for 16 hours, and were continuously changed without cooling in between.

FIG. 5 shows the influence of alloying element content, β quench temperature and final annealing temperature on corrosion resistance. The mark in the figure indicates that nodular corrosion has occurred.

The O mark indicates that nodular corrosion did not occur.

In all Zr-8n alloys, nodular corrosion occurs regardless of the amount of Sn added and the heat treatment. Z, r-8n-F
The corrosion resistance of e-alloys is almost unaffected by the final annealing temperature, and when Fe is alloyed with 0.25 wt% or more,
It can be seen that nodular corrosion does not occur even without β-quenching. When β-quenched at 1000℃, Fe becomes 0.
．． It can be seen that the occurrence of nodular corrosion can be prevented when alloyed with a content of 15 wt% or more.

The corrosion resistance of Zr-8n-Cr alloy decreases as the final annealing temperature increases, and when annealing is performed at 600°C or higher, the occurrence of nodular corrosion cannot be prevented even if the β-quench temperature and alloying amount are changed. I understand that.

The corrosion resistance of the Zr-8n-Ni alloy is not affected by the final annealing temperature, and the corrosion resistance of the Zr-8n-Ni alloy is not affected by the final annealing temperature. When alloyed with I5 wt% or more, nodular corrosion can be prevented without β-quenching. When the β-quench temperature is 900°C, the amount of Ni alloyed to prevent nodular corrosion is 0.1
It can be seen that wt% is sufficient. From the above results, it can be seen that the alloying elements effective in improving corrosion resistance are Fe and Ni, and the effect of Cr is small.

Example 2 Zr was produced by the same melting method, processing, and heat treatment as in Example 1.
-8n-Fe-Ni alloy plates were manufactured.

The final annealing temperature was 730°C. Figure 6 shows that the corrosion resistance of each alloy was examined by the same corrosion test as in Example 1 to determine whether or not nodular corrosion occurred. In the region where the alloyed amount of Ni satisfies formula (1), the heat treatment is 0.25 ・X1lt: +o, 15 Xya≧0
．． 0375 (1) X*t: Alloying amount of Ni.

XFII! It can be seen that nodular corrosion can be prevented regardless of the amount of F8 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 x□+0.1 x, ≧0.01j
(2) Example 3 Figure 7 shows the relationship between the corrosion weight increase and the hydrogen absorption amount for alloys with the respective compositions shown in Table 1. There is a proportional relationship between the hydrogen absorption amount and the corrosion weight gain as shown in the figure. There is.

The figure shows that materials with high corrosion resistance absorb less hydrogen, and the addition of Ni does not increase the amount of hydrogen absorbed.

Example 4 Sn was 3. When Owt% 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 1.0 to 2. Owt
A range of % 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 one component is improved and the lifetime in the reactor can be significantly extended, making it possible to increase the burnup of nuclear fuel.

Even in the manufacturing process h'- of zirconium alloy members,
Since the heat treatment temperature can be selected relatively freely, it has the effect of facilitating its 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 drawings]

Figure 1 shows the optimum alloy composition range, Figure 2 shows the mechanism of oxygen diffusion in the oxide film, Figure 3 shows the neutron absorption cross section of the alloy, and Figure 4 shows the zirconium alloy plate material. Diagram showing the manufacturing process, No. 51! 1 and 6 are diagrams showing the effects of alloying elements and heat treatment temperatures on the occurrence of nodular corrosion, and FIG. 7 is a diagram showing the relationship between corrosion weight increase and hydrogen absorption amount for alloys having the respective compositions shown in Table 1. 1...Fuel cladding tube, 2...Channel box, 3
...Spacer, 4...Water rod, 5...
fuel handle.

Claims (1)

[Claims] 1. In zirconium-based alloy, 1.0 to 2.0wt
% tin, 0.3 wt% or less nickel, 0.55 wt%
A zirconium-based alloy containing the following iron, oxygen, and other impurities, having a total content of iron and Ni of 0.15 wt% or more, and containing no Cr. 2. In claim 1, the iron content is expressed on the x-axis (
Unit: wt%), (0.55, 0), (0.2, 0.
2), (0.09, 0.09), (0.25, 0) 4
A zirconium-based alloy characterized by having a composition surrounded by dots.