JPS63213629A - Zirconium based alloy - Google Patents

Abstract

PURPOSE:To improve corrosion resistance and mechanical strength, by specifying the contents of Sn, Fe and Cr in a Zr based alloy as well as the total amount thereof and incorporating proper quantity of Nb into said alloy. CONSTITUTION:The contents of the Zr based alloy contg. Sn, Fe, Cr and Nb are regulated, by weight, to 0.4-1.2% Sn, 0.2-0.4% Fe and 0.1-0.6% Cr, and the total amount of Sn, Fe and Cr is regulated to 0.9-1.5%; the content of Nb is further regulated to <=0.5%. Since the compounded ratios of the elements to be added in said Zr alloy are regulated, the lowering of the amount of corrosive reaction by high-temp. water or high-temp. steam is improved at the time of using said alloy as the constitutional member in a nuclear reactor of an atomic power generation plant.

Description

DETAILED DESCRIPTION OF THE INVENTION [Industrial Application Field] The present invention relates to a zirconium-based alloy used for internal reactor components of nuclear power plants.

[Prior Art] A fuel assembly used in a nuclear reactor of a nuclear power plant is as explained below.

FIG. 1 is a cross-sectional view illustrating the outline of fuel elements used in nuclear power reactors according to conventional nuclear power generation plan 1 and the present invention, and FIG. 2 shows a plurality of fuel elements shown in FIG. 1 arranged in a grid pattern. FIG. In Figures 1 and 2, columnar sintered bodies (hereinafter referred to as pellets) of uranium oxide are shown.
A rod-shaped fuel element 7 is covered with a cladding tube 2 and both ends of the cladding tube are sealed with end plugs 4.5 via a coil spring 3, and a support lattice 6 in which these fuel elements 7 are arranged in a lattice pattern. It is configured. Further, 8 is an L section nozzle, 9 is a lower nozzle, IO is a leaf spring, and 11 is a control rod cluster.

Materials for the cladding tube 2 and support grid 6 of the conventional fuel element 7 include -a, ASTM (American Society for Testing and Materials) B
UNS (Unified N) defined in 2S3
Umbering System For Me
tals and Alloys) number R6080
2 or R60804, i.e. the former is tin, iron, chromium.

A zirconium-based alloy with a small amount of N added, such as nickel, is used. Furthermore, since this zirconium-based alloy does not particularly affect nuclear fission, it contains negligible trace impurities.
Contains B, Cd, C, etc.

During the operation of a nuclear power plant, the outer surfaces of the reactor internal components of these plants are in contact with high-temperature, high-pressure cooling water, and the zirconium-based alloy that is the material of the coating v2 and the support grid 6 is exposed to high-temperature and high-pressure cooling water. Alternatively, a corrosive reaction with high-temperature steam forms a zirconium oxide-like or localized film, and a portion of the hydrogen generated by the corrosion reaction is absorbed into the zirconium-based alloy through the film.

As the corrosion reaction progresses and the zirconium oxide coating on the outer surface becomes thicker, the thickness of the zirconium-based alloy on the inside thereof decreases, and the strength of the cladding tube 2 and support grid 6 made of the zirconium-based alloy decreases. Furthermore, as the amount of hydrogen generated by corrosion reactions absorbed into the zirconium-based alloy increases, the strength and ductility of the zirconium-based alloy decrease.

For the reasons mentioned above, the strength and ductility of the cladding tube 2 and the supporting grid 6 are reduced due to corrosion, which may impair the integrity of these members. The amount of corrosion on the outer surfaces of the support grid 2 and the support grid 6 is small and does not reach the level of impairing the integrity of these members.

[Problems to be solved by the invention] As described in the above, conventional cladding tubes and support grids using zirconium-based alloys aim to improve the efficiency of nuclear fuel operation by increasing the burn-up of the fuel and increasing the atomic energy efficiency. When staying in the furnace for a long time, the corrosion reaction progresses on the outer surface, the zirconium oxide coating becomes thicker, the thickness of the zirconium-based alloy as a strength member decreases, and the hydrogen generated by the corrosion reaction increases. There was a problem that there was a risk that the zirconium-based alloy member would be absorbed by the ψ and the integrity of the zirconium-based alloy member would be impaired.

The present invention has been made to solve these problems, and is made of zirconium, which is a material with high corrosion resistance and high mechanical strength that can be used for internal reactor components such as fuel element cladding tubes and support grids. The purpose is to provide a base alloy.

[Means for Solving the Problems] In order to achieve the above object, the zirconium-based alloy of the present invention contains three metals consisting of tin, iron and chromium or three metals consisting of tin, iron and chromium and niobium. A zirconium-based alloy containing 0.4 to 1.2 trace% of tin,
0.2 to 0.4% by weight of iron and 0.1 to 0.0% of chromium.
6% by weight, and the total content of tin, iron, and chromium is 0.9 to 1.5% by weight.

[Function] In the present invention, by appropriately adjusting the blending ratio of the main additive elements of the zirconium-based alloy, even when used as a J111 member in the reactor of a nuclear power plant,
It aims to reduce the amount of corrosion reaction caused by high-temperature water or high-temperature steam, and even when the fuel stays in the reactor for a long time.

The integrity of the cladding and support grid can be prevented from being compromised.

[Example] Currently, in a pressurized wooden reactor under the operating conditions of a nuclear couplant, uniform corrosion is occurring on the outer surface of the cladding tube and support grid, and local boiling may occur as the core power increases further in the future. It is estimated that this will result in increased localized corrosion, which is often seen in current boiling water reactors. Therefore, in order to simulate uniform corrosion and local corrosion inside the reactor, temperatures outside the reactor were heated to 400°C and 500°C, respectively.
Autoclave tests in high-temperature steam are widely used.

Next, an example in which the zirconium-based alloy of the present invention is used in a reactor internal component of a nuclear power plant will be described in detail with reference to the accompanying drawings.

Table 1 shows the 4I inside the reactor of the nuclear power plant of the present invention.
This figure shows the blending ratios of tin, iron, chromium, and niobium contained in zirconium-based alloys that were trial-produced as component parts.The samples made of zirconium-based alloys with different content ratios totaled 12.9L. It is revealed.

(Note) Impurities are A-5TM B553 R60804
Equivalent to Table 1 For example, in sample 1 of Table 1, the zirconium-based alloy contains 1.58 tonne weight percent tin and 0.21 weight percent iron.

Kurumuro, tz4. This sample zirconium-based alloy, which shows H% gold content, was hot-rolled,
After the β heat treatment, cold rolling and annealing were repeated, and the final annealing temperature was 470°C for the cold-worked strain relief annealed material and the final annealing temperature for the material was 600°C.
This was prototyped as a completely annealed material at ℃.

Then, each sample of these zirconium-based alloys was
The samples were held in high-temperature steam at 00°C and 500°C for 1150 hours and 24 hours, respectively, and the corrosion weight increase was measured. However, this corrosion weight increase is based on conventionally known materials (for example, ASTM B5
53 UNS R60804 (corresponding to Sample 1) is set as 1, and represents the degree of corrosion of the zirconium-based alloy of this example.

Figures 3(a) and (b) show the tin content contained in the zirconium-based alloy of the present invention and the ii
FIG. 3(a) is a diagram showing the relationship between corrosion weight increase when held for s days, and FIG. Figure (b) shows data for a zirconium-based alloy that was similarly completely annealed.

Next, from the test data of the tin content and corrosion weight increase of the zirconium-based alloy shown in FIGS. 3(a) and (b), at a test temperature of 400°C,
If the amount of chromium added is in the range of iron: 0.2 to 0.4% by weight and chromium = 0.1 to 0.6% by weight, the correlation between iron or chromium and the amount of corrosion #lI is weak, and corrosion There is no effect on the increase in weight.Also, when the amount of niobium added in the zirconium-based alloy is in the range of 0.16 to 0.5% by weight, the corrosion increase 9 increases with the amount of niobium added, but 0.01i At Jyuken%, the amount of corrosion decreases by 1 yen compared to the non-additive which does not contain niobium.

And the tin content contained in the zirconium-based alloy is 0.
．． 3 to 1.6 layers? % range, the corrosion weighting decreases with the tin content.

As mentioned above, the range of tin content in the civilized zirconium-based alloy shown in No. 31M (a) and (b) is 0.4 to 1.2 i[%, so the conventional zirconium-based alloy The tin content is lower than the stipulated tin content of 1.2 to 1.7 i1% for the alloy ASTM B:153 R60804''t', and therefore the corrosion weight increase is small and the corrosion resistance is excellent.

'54 Figures (a) and (b) show the contents of tin, iron, and chromium contained in the zirconium-based alloy of the present invention, and the
Fig. 4(a) is a diagram showing the relationship between corrosion weight increase when held in high-temperature steam at 0°C for 24 hours;
b) is data for a fully annealed zirconium-based alloy.

Next, from the test data of tin, iron, and chromium contents and corrosion weight increase of the zirconium-based alloy shown in Fig. 4(a) and (b), we can determine the amount of tin contained in the zirconium-based alloy at a test temperature of 500''C. The content of iron, or chromium or tin is 20,4-1.6% by weight.

Iron: 0.2-0.44171%, Chromium: 0.1
~0.5 g (%), the correlation between tin, iron, or chromium and corrosion weight gain is weak, and there is no effect on corrosion weight gain. If it is within the range of 0.9 to 1.5% by weight, the corrosion weight gain will be significantly reduced compared to the corrosion weight gain of the conventional zirconium-based alloy ASTM B353R60804, and the zirconium-based alloy will have excellent corrosion resistance. When the amount of niobium added is in the range of 0.05 to 0.5% by weight, the amount of corrosion I?I decreases with the amount of niobium added.

Above, Figures 3(a), (b), and Figure 4(a).

From the test data on increased corrosion shown in (b), the tin content in the zirconium-based alloy is between 0.4 and 1.2 heavy grade, regardless of the cold working strain relief trowel or complete annealing. %
, the iron content is 0.2 to 0.4 seri%, the chromium content is 0.1 to 0.6 seri%, and the sum of the tin, iron, and chromium contents is 0.9 to 0.9%. The zirconium-based alloy of the present invention, which is 185% heavier than conventional zirconium-based alloy ASTM
B: 15 Excellent corrosion inhibition compared to R60804.

In addition, the amount of niobium added is 1 to 0.0 to 0.0 to uniformly suppress corrosion depending on the purpose of use of the zirconium-based alloy of the present invention.
Corrosion can be further suppressed by adding 6m1%, 0.06 to 0.5% by weight to suppress local corrosion, and about 06061% by weight to suppress uniform and local corrosion. .

[Effects of the Invention] As explained above, the zirconium-based alloy of the present invention prevents uniform corrosion and localized corrosion compared to the conventional zirconium-based alloy ASTM B3E3R60804 used as a nuclear reactor material. As a result, the inside of the reactor ar#,? can be significantly suppressed. When used as material A, the reliability of members made of zirconium-based alloys is improved, the time spent in the reactor can be extended, and the burnup of nuclear fuel can be increased.

[BRIEF DESCRIPTION OF THE DRAWINGS] FIG. 1 is a sectional view showing fuel elements used in reactors of conventional and inventive nuclear power plants, and FIG.
FIG. 3(a) is a cross-sectional view of a fuel assembly in which a plurality of fuel elements shown in FIG. 1 are arranged in a lattice pattern. (b) is a diagram showing the relationship with the tin content when the vertical axis is the corrosion weight increase of the zirconium-based alloy of the present invention, and FIG.
a) and (b) are line sections showing the relationship between the corrosion weight increase of the zirconium-based alloy of the present invention and the contents of tin, iron, and chromium, when the vertical axis is taken as the vertical axis. In the figure. l: Bellet 2: Cladding tube 3: Coil spring 4, 5: End column 6: Support grid 7: Fuel element 8: Upper nozzle 9: Lower nozzle Iro: Leaf spring 11: Control rod cluster agent
Patent Attorney 1) Haru Kitatake Figure 2

Claims (2)

[Claims] (1) A zirconium-based alloy containing three metals consisting of tin, iron, and chromium or three metals consisting of tin, iron, and chromium, and niobium, used for internal reactor components of nuclear power plants, etc. 0.4 to 1.2% by weight, 0.2 to 0.4% by weight of iron, 0.1 to 0.6% by weight of chromium, and the total of tin, iron, and chromium is 0.9 to 1.
A zirconium-based alloy characterized by a content of 5% by weight. (2) The zirconium-based alloy according to claim (1), which contains 0.5% by weight or less of niobium.