JPH02250947A - Corrosion-resistant zirconium-base alloy - Google Patents

Abstract

PURPOSE:To improve the nodular corrosion resistance of the zirconium alloy by irradiating the surface of a Zr-base alloy as a final manufactured article manufacture from a pure Zr sponge as raw material with corpuscular rays. CONSTITUTION:Alloy elements are added to a pure Zr sponge 1 as raw material, which is subjected to arc melting, thereafter undergoes beta-forging, solution treatment and alpha-forging, then is subjected to hot working and annealing, repeatedly to cold rolling and annealing and is thereafter irradiated with corpuscular rays to obtain the final manufactured article. As the irradiating corpuscular rays, electron rays and ions can be considered. As the temp. at which the corpuscular rays are emitted, the temp. range where the reformation of solid soln. of precipitated substance by the irradiation preferentially occurs and precipitation is hard to occur is needed to select. In this way, the depleted zone in the solid soln. components of the alloy elements produced in the Zr-base alloy at the manufacturing stage such as annealing and rolling can be eliminated. Thus, nodular corrosion caused by the above depleted zone can be suppressed.

Description

Detailed Description of the Invention [Field of Industrial Application] The present invention relates to core constituent materials for nuclear reactors, and in particular to zirconium, which has excellent corrosion resistance and can be used as fuel cladding tubes or core structural materials suitable for nodular corrosion resistance. Regarding base alloys.

[Prior Art] In boiling water type light water reactors, zirconium-based alloys are used as core constituent materials such as fuel cladding tubes, channel boxes, and spacers. This was developed considering neutron economy and corrosion resistance in high temperature water or steam. However, localized corrosion called nodular corrosion occurs in these fuel cladding tubes and core structural materials during operation of the nuclear reactor. Nodular corrosion grows as irradiation progresses, and as the corrosion layer becomes thicker, it may peel off. The occurrence of such nodular corrosion not only leads to thinning of the fuel cladding and core structural materials, but also increases the radioactivity concentration in the reactor water due to flaking, increasing the radiation exposure of workers during regular reactor inspections. There is a fear.

In order to improve the economic efficiency of nuclear reactor fuel in the future, plans are underway to extend the useful life of fuel and core structural materials, but the safety of fuel cladding and core structural materials for longer-term use than currently is in progress. The nodular corrosion resistance of zirconium-based alloys is attracting attention from the viewpoint of safety, reliability, and reduction of radiation exposure during periodic inspection work.

As an example of such symmetry, as shown in Japanese Patent Application Laid-Open No. 57-116739, the intermetallic compound is segregated in a chain along the grain boundaries or sub-boundaries throughout the zirconium-based alloy substrate to prevent nodular corrosion. Methods have been proposed to improve sex. However, the stability of this intermetallic compound segregated in a chain under the operating conditions in a nuclear reactor has not been confirmed, and there are doubts as to whether it can demonstrate its performance in a reactor, especially over a long period of time. It cannot be said that it is a sufficient means for use.

[Problem to be solved by the invention] The present invention has been made in view of the above-mentioned circumstances, and its purpose is to solve the problem in core structural materials such as fuel cladding tubes, channel boxes, and spacers. The object of the present invention is to provide a suitable zirconium-based alloy with excellent nodular corrosion resistance.

[Means for Solving the Problems] In order to solve the above problems, the structure of the corrosion-resistant zirconium-based alloy according to the present invention is such that, in the corrosion-resistant zirconium base metal used as a core constituent material of a nuclear reactor, pure zirconium sponge as a raw material The surface of the final product of the zirconium-based alloy, which has been manufactured through a multi-step manufacturing process, is subjected to particle beam irradiation to suppress nodular corrosion of the zirconium-based alloy.

[Works] Zircaloy is a zirconium-based alloy currently used as fuel cladding and core structural material for light water-cooled nuclear reactors. Zircaloy is made by adding small amounts of tin, iron, chromium, nickel, etc. to zirconium to improve strength and corrosion resistance. Among these alloying elements, iron, chromium,
Since nickel has a low solid solution concentration in zirconium and the amount added exceeds the solid solubility limit, under normal conditions it forms an intermetallic compound with zirconium and precipitates in the Zircaloy base material.

The above-mentioned intermetallic compounds are mainly generated during annealing in the Zircaloy product manufacturing process. FIG. 2 is a flowchart of a conventional manufacturing process for manufacturing a raw fuel cladding tube from pure Zr sponge as a raw material. The unevenly distributed alloying elements become uniform through solution treatment, but during the subsequent hot extrusion and annealing process, the alloying elements that were supersaturated in solid solution in zirconium precipitate.0 The precipitation of alloying elements becomes uniform. does not occur, and as a result, the distribution of solid solution components also becomes non-uniform. For this reason, regions where solid solution components of alloying elements are locally depleted occur in the Zircaloy base material.

The formation mechanism of nodular corrosion is explained below.
When Zircaloy is corroded in which there is a region deficient in solid solution components of alloying elements, the depleted region exhibits corrosion behavior similar to that of zirconium since no alloying elements are present. Therefore, a porous ZRO2 type oxide film similar to that when zirconium is oxidized is formed on the surface of the depleted region. In other regions, alloy elements are incorporated into the oxide film. The alloying elements iron, chromium, and nickel are divalent or trivalent, and zirconium is tetravalent, so in order to balance the charge, zirconium oxide containing alloying elements contains Oxygen vacancies are generated, resulting in a 2rO,x type oxide. The ZrO,-x type oxide has a much more dense structure than the ZrO□ type oxide, and plays the role of a protective film against corrosion. For this reason, in the deficient region, water, which is an oxidizing agent, comes into direct contact with the metal surface, and the corrosion reaction becomes the rate-limiting reaction, whereas in the region where a protective film has formed, the diffusion of oxygen in the protective film becomes the rate-limiting reaction. 1. The corrosion rate differs greatly between the two. As a result, oxidation progresses locally in the depleted region, forming nodular corrosion.

However, when we investigated the growth rate of nodular corrosion inside the reactor, we found that the growth rate decreased as the irradiation progressed. Figure 3 shows the burnup dependence of nodular corrosion thickness. In Fig. 3, 14 indicates the range of variation in the thickness of nodular corrosion, and solid line 13 indicates a fitting curve based on the least squares method. It can be seen that the thickness of nodular corrosion tends to be saturated with respect to burnup. . That is, the growth rate of nodular corrosion occurring on the surface of a zirconium-based alloy decreases during use in a furnace. The microstructure of the Zircaloy-2 cladding tube after use in the furnace was examined using a transmission electron microscope and an energy dispersive X-ray microscope.
As a result of observation and analysis using a line analyzer, it was found that alloying elements were re-dissolved in the Zircaloy-2 base material from precipitates of intermetallic compounds. The base material structure after irradiation! As expected, Zr (Cr, Fe), type and Zr2 (Ni
, Fe) type precipitates were observed, and this was analyzed using an X-ray analyzer.

Figure 4 shows the elemental concentrations of Fe, Cr, and Ni in the precipitate,
FIG. 3 is a relationship diagram summarizing the relationship with the distance (μm) from the center position of the precipitate.

That is, it became clear that Fa, Cr, and Ni were re-dissolved in the Zircaloy-2 base material after being used in the furnace (after irradiation). Due to this re-solid solution phenomenon of alloying elements,
The depletion region disappeared and the growth of nodular corrosion was suppressed.

This discovery revealed that irradiation with fast neutrons causes the precipitated alloying elements to form a solid solution to a supersaturated state, which is thermodynamically impossible. As a result, the corrosion resistance of Zircaloy-2 was improved and the growth rate of nodular corrosion was reduced by irradiation.

The present invention provides the above-mentioned irradiation to cause the alloying element to become a supersaturated solid solution in the Zircaloy base material, to eliminate the alloying element-deficient region,
This was born out of the discovery that the nodular corrosion resistance was improved.

When the zirconium-based alloy of the present invention is loaded and used in the core of a nuclear reactor, at least the region deficient in solid solution components of alloying elements on the surface disappears, local corrosion does not occur, and the occurrence of nodular corrosion is prevented. can do.

[Example] An example of the present invention will be described below with reference to FIG. FIG. 1 is a flowchart of the manufacturing process of a fuel cladding tube made of a corrosion-resistant zirconium-based alloy according to the present invention.

In the configuration shown in Figure 1, an alloying element 2 is added to a pure Zr sponge 1 as a raw material, and after arc melting, it is passed through β-forging 4, solution treatment 5, and α-forging 6, followed by hot processing and sintering. After annealing 8, cold rolling 9 and annealing 10 are repeated, and then particle beam irradiation 11 according to the present invention is performed to obtain a final product 12.

Next, the operation of the manufacturing process will be explained according to the flow.

Predetermined alloying elements (Sn, Fa
, Cr, Ni, etc.) 2 and compression molded using a press to make cylindrical briquettes. This is welded under an inert atmosphere to form an electrode, which is arc melted (3) to form an ingot. For molding, the ingot is heated to about 1000°C and β-forged 4. This is held at 1000° C. or higher for several hours, and then rapidly cooled and subjected to solution treatment 5. This solution treatment 5 makes the distribution of the unevenly distributed alloying elements uniform. In order to remove the surface oxide film produced by the solution treatment and adjust the dimensions, α forging 6 is performed after preheating within the α region temperature range of around 700°C. This is made into a blank tube by hot processing at around 700°C. In order to remove distortion due to processing, 1
Annealing 8 at 650°C under a high vacuum of 0"'-10-' Torr. The outer diameter is reduced by cold rolling 9 to reduce the wall thickness. Annealing 10 is performed in the middle until the predetermined dimensions are reached. Repeat cold rolling 9 several times with a sandwich.Finally, 10-'-10
-'T.

Recrystallization annealing is performed at around 580° C. under a high vacuum of rr.

During the hot processing F after the solution treatment 5 and annealing steps 8 and 10, the alloying elements that were in solid solution are precipitated, and a depletion region is generated. In order to eliminate this depletion region, as shown in FIG. 1, particle beam irradiation 11 according to the present invention is performed as a final step.

Possible particle beams to be irradiated include electron beams and ions. Electron beams with high energy and high current density can be obtained relatively easily. Furthermore, since interactions with substances are less likely to occur compared to ions, the range is longer and the irradiation effect can be exerted more deeply when compared with the same energy.For example, an electron beam of 100 electron volts is The maximum range is about 600μm,
This is approximately 100 times more powerful than when protons are irradiated with the same energy.

When irradiating with ions, it is advantageous to use ions of alloying elements. In this case, the effect of increasing the concentration of the solid solution component of the alloying element by implanting the alloying element is greater than the effect of redissolving the alloying element as a solid solution by ion irradiation.

In addition, in order to rapidly re-dissolve the precipitate by particle beam irradiation, it is preferable to select an element that diffuses quickly in hexagonal zirconium as the alloying element. A fast-diffusing element is an element that can diffuse at interstitial sites in hexagonal zirconium, and this corresponds to an element whose atomic radius is 0.85 times or less than that of zirconium.6 Figure 5 shows the atomic radius and hexagonal zirconium. The relationship between the diffusion coefficient and the diffusion coefficient was shown. A broken line 15 indicates a position where the atomic radius corresponds to 0.85 times that of zirconium. From Figure 5, the elements whose atomic radius is 0.85 times or less than zirconium are:
It can be seen that the diffusion coefficient is dramatically large. Among them, cobalt, nickel, and iron have large diffusion coefficients and are suitable as additive elements for zirconium-based alloys.

As the temperature at which the particle beam is irradiated, it is necessary to select a temperature range in which re-solid solution of precipitates due to irradiation occurs preferentially and precipitation is difficult to occur. Zircaloy-2 used in the furnace
From the observation and analysis results, it can be seen that iron, chromium, and nickel re-dissolve in solid solution at an irradiation temperature of around 300°C (573K).

In addition, the alloying elements that had been re-dissolved in supersaturation during post-irradiation annealing at 550°C (823K) were precipitated. The melting points of intermetallic compounds of zirconium, iron, chromium, and nickel are each 1600°C (1873K).

1675℃ (1948K), 1200℃ (1473K)
It is. Re-solid solution and precipitation are phenomena controlled by the behavior and diffusion of irradiation defects, and when normalized by the absolute temperature of the melting point,
It is known to be quite generalizable. The temperature at which solid solution occurs again is 0.31 (573/
1873).

0.29 (573/1948), 0.39 (573
/1473), the normalized value of the temperature at which precipitation occurred by annealing after irradiation was 0.44 (823/1873), 0.42
(823/1948), 0.56 (823/147
3). From this, it can be seen that it is preferable to irradiate at a temperature that is 0.4 times or less the temperature normalized by the melting point of the intermetallic compound.

[Effects of the Invention] According to the present invention, it is possible to eliminate the region deficient in solid solution components of alloying elements that occurs in the zirconium-based alloy during manufacturing processes such as annealing and rolling. It is possible to suppress the nodular corrosion caused by this, and it is possible to provide core structural materials such as fuel cladding tubes made of zirconium-based alloys, spacers and channel boxes, which have excellent safety and reliability.

In summary, it is an object of the present invention to provide a zirconium-based alloy with excellent nodular corrosion resistance and suitable for core constituent materials such as fuel cladding tubes, channel boxes, and spacers.

[Brief explanation of drawings]

Fig. 1 is a flow diagram showing the manufacturing process of a zirconium-based alloy fuel cladding tube according to the present invention, Fig. 2 is a flowchart showing the manufacturing process of a conventional zirconium-based alloy fuel cladding tube, and Fig. 3 is a flowchart showing the manufacturing process of a conventional zirconium-based alloy fuel cladding tube. Figure 4 shows the relationship between the thickness of nodular corrosion formed on the outer surface of the fuel cladding tube used in the furnace and the burnup. FIG. 5 is a characteristic diagram showing the relationship between concentration and distance from a precipitate, and FIG. 5 is a characteristic diagram showing the correlation between the diffusion coefficient of a metal element in hexagonal zirconium and its atomic radius. <Explanation of symbols> 1...Pure zirconium sponge, 2...Alloying element addition, 3...Arc melting, 4...β-forging, 5...=
Solution treatment, 6...α-forging, 7... Hot working, 8
... Annealing, 9... Cold rolling, 10... Annealing, 11... Particle beam irradiation, 12... Final product, 13
...fitting curve by least squares method, 14...
・Range of variation in thickness of nodular corrosion, 15...
A position where the atomic radius is equivalent to 0.85 times that of zirconium.

Claims (1)

[Claims] 1. In a corrosion-resistant zirconium-based alloy used as a core constituent material of a nuclear reactor, the surface of the final product of the zirconium-based alloy is manufactured from pure zirconium sponge as a raw material through a multi-step manufacturing process. , by applying particle beam irradiation,
A corrosion-resistant zirconium-based alloy characterized in that nodular corrosion of the zirconium-based alloy is suppressed. 2. The corrosion-resistant zirconium-based alloy according to claim 1, wherein the particle beam to be irradiated is an electron beam. 3. The corrosion-resistant zirconium-based alloy according to claim 1, wherein the irradiated particle beam is an ion. 4. The corrosion-resistant zirconium-based alloy according to claim 3, wherein the ions to be irradiated are ions of constituent elements of the zirconium-based alloy. 5. At least one metal element with an atomic radius of 0.85 times or less than zirconium is added to the zirconium-based alloy, making it possible to quickly uniformize the precipitate distribution by particle beam irradiation. The corrosion-resistant zirconium-based alloy according to claim 1. 6. Cobalt, as a metal element with an atomic radius of 0.85 times or less than zirconium, to be added to the zirconium-based alloy
The corrosion-resistant zirconium-based alloy according to claim 5, characterized in that one or more selected from iron and nickel. 7. Set the particle beam irradiation temperature to 0, which is the melting point (absolute temperature) of the intermetallic compound formed by zirconium and the additional alloying element.
．． The corrosion-resistant zirconium-based alloy according to claim 5, characterized in that the corrosion resistance is 4 times or less.