JPH06230161A - Zirconium based alloy - Google Patents

Abstract

PURPOSE:To provide a zirconium based alloy excellent in corrosion resistance and suitable as a material for fuel cladding pipe or a core structure such as a channel box or a spacer. CONSTITUTION:In the production process of a fuel cladding pipe made of zircaloy, an alloy of Zr added with 0.5-2.0wt.% of at least one kind of Y and Pb with at least 1wt.% thereof being present in the form of solid solution, is employed as the compositional material for an original pipe 1.

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a zirconium-based alloy used for a fuel cladding tube or core structural material of a nuclear reactor.

[0002]

2. Description of the Related Art In light water reactors, zirconium-based alloys are used for fuel cladding tubes and core structural materials such as channel boxes or spacers. Zirconium based alloys
It was developed in consideration of neutron economy and corrosion resistance / strength in high temperature / high pressure water or steam.

The conventionally used zirconium-based alloy is mainly zircaloy. In the method for manufacturing a fuel cladding tube made of zircaloy, the steps after the raw tube are shown in the flow chart of the fuel cladding tube manufacturing step of FIG. explain. In FIG. 3, 1 is a blank tube, 2 is an outer surface quenching step, 3 is an annealing step,
4 is a cold rolling process, 5 is an intermediate annealing process, 6 is a final rolling process, and 7 is a final annealing process.

In this manufacturing method, first, in the outer surface quenching step 2 of the shell 1, the distribution of alloying elements near the outer surface of the shell 1 is made uniform. Next, in the annealing step 3, annealing is performed at 620 ° C. for 1.25 hours to remove strain by quenching, and then in the cold rolling step 4, the outer diameter is reduced and the wall thickness is reduced.

Then, in order to adjust to a predetermined size, 620
Cold rolling is repeated twice with the intermediate annealing step 5 of 1.25 hours at ℃ performed, then the final rolling is performed in the final rolling step 6, and finally 10 to ⁴ in the final annealing step 7.
Recrystallization annealing is performed at 577 ° C. for 2.5 hours under a high vacuum of -10 to ⁶ Torr.

However, in the fuel cladding tube and the core structural material produced in this way, an oxide film is formed on the contact surface with the cooling water during the operation of the nuclear reactor. Then, the oxide film grows as the irradiation progresses, and may peel off when it becomes thick.

The generation of such an oxide film not only reduces the thickness of the fuel cladding tube and core structural material, but also increases the radioactivity concentration in the reactor water due to peeling, resulting in a radiation dose to workers during periodic inspections of the nuclear reactor. May increase.

In order to improve the economical efficiency of reactor fuel,
Plans are underway to extend the period of use of fuel cladding tubes and core structural materials, but the safety and reliability of fuel cladding tubes and core structural materials for longer-term use than before, or exposure during regular inspection work From the viewpoint of reducing the amount, the corrosion resistance of zirconium-based alloys has attracted attention.

As an example of this measure, Japanese Patent Laid-Open No. 57-11
Japanese Patent No. 6739 discloses a method in which an intermetallic compound is segregated in a chain form along a grain boundary or a subgrain boundary over the entire zirconium-based alloy substrate to improve corrosion resistance.

[0010]

However, the stability of such intermetallic compounds segregated in a chain form under the conditions of use in the reactor has not been confirmed, and the stability of the intermetallic compound in the reactor has not been confirmed. It is doubtful whether or not the performance can be exhibited, and it is hard to say that it is a sufficient means especially under the operating conditions in the reactor for a long time.

The present invention has been made in view of the above situation, and an object thereof is zirconium having excellent corrosion resistance which constitutes a fuel cladding tube or a core structural material such as a channel box and a spacer. To provide a base alloy.

[0012]

The above object can be achieved as follows.

(1) In a zirconium-based alloy which is composed of Zr and an additive element to Zr and is used for a fuel cladding tube of a nuclear reactor or a core structure material, the additive element has a valence of less than 4, It is an element that exists in an oxidized state in an oxide film produced under the conditions of use of a furnace and is less likely to be oxidized than Zr.

(2) In (1), the additive element is α-
It must be an element that forms a solid solution with at least 1 wt% in Zr.

(3) In (1), the additive element is α-
At least 1 wt% or more of Y and Pb having a thermal neutron absorption cross-section of about 1b or less in solid solution in Zr and having a total addition amount of 0.5 to 2.0 wt%.

(4) In (1), the additive element is α-
At least 1 wt% or more of Y and Pb having a solid solution in Zr of 1 wt% or more and having a thermal neutron absorption cross section of about 1 b or less, and the total addition amount is 0.5 to 2.0 wt%,
And at least one of Zn, Co and Ti,
The total amount of addition should be 0.1-0.5 wt% element.

[0017]

First, the corrosion mechanism of the zirconium-based alloy, which is the basis of the present invention, will be described.

When Zr is oxidized, a ZrO ₂ type oxide is formed. ZrO ₂ has a monoclinic system (low temperature type) and a tetragonal system (high temperature type), and the transition from the former to the latter reversibly occurs at around 1000 ° C.

Monoclinic ZrO ₂ is porous and has a brittle property, and even if it occurs on the surface of a zirconium-based alloy, it does not form a barrier against the invasion of an oxidant, that is, a protective film, and therefore it is a monoclinic system. Zirconium-based alloys containing crystalline ZrO ₂ have very poor corrosion resistance.

On the other hand, Zr containing a few% of MgO, CaO, or a rare earth oxide is called stabilized zirconia or partially stabilized zirconia, which has a cubic system or a tetragonal system and has a phase of Does not cause transition. Particularly, partially stabilized zirconia has excellent mechanical properties, and when an oxide film of stabilized zirconia is formed on a zirconium-based alloy, this oxide film becomes a dense protective film excellent in corrosion resistance.

The operating temperature of the zirconium-based alloy in the nuclear reactor is usually about 200 to 400 ° C., and the oxide film formed is monoclinic ZrO ₂ . However, zircaloy, which is a zirconium-based alloy that has been mainly used in the past, has relatively good corrosion resistance despite the formation of monoclinic ZrO ₂ .

This is because the oxide produced at the interface between the oxide film and the metal is tetragonal zirconia, which plays the role of a protective film. However,
Since this tetragonal zirconia undergoes a phase transition to monoclinic zirconia when the oxide film becomes thick,
The protective film does not become so thick, and thick monoclinic zirconia, which has no effect on corrosion resistance, is formed on the outside.

The reason why the tetragonal oxide film is formed on the surface of zircaloy is that MgO, CaO, or Y ₂ O _{3 is} contained when Fe, Cr, or Ni is incorporated into the oxide film.
It is thought that this is because it has the property of stabilizing tetragonal zirconia as described above.

That is, it is considered that stabilization of tetragonal zirconia is achieved because oxygen vacancies are introduced into zirconia by incorporating an element having a valence smaller than Zr into zirconia. To be

However, in the case of zircaloy, the oxide film stabilizes the tetragonal system to a sufficient thickness due to the low concentration of these alloying elements and the fact that it is used in a very severe oxidizing atmosphere. It is thought that a phase transition occurs. Further, a compressive stress is also required to stabilize the tetragonal system due to a low concentration of alloying elements, but when the oxide film becomes thicker, the compressive stress becomes smaller, so it is considered that a phase transition occurs.

There is a correlation between the critical thickness at which this phase transition occurs and the alloy element concentration or oxygen vacancy concentration in zirconia, and it is expected that the higher the oxygen vacancy concentration, the greater the critical thickness at which phase transition occurs. To be done.

In Zircaloy, most of Fe and Cr are contained.
It is known that Ni and Ni are precipitated by forming an intermetallic compound with Zr. When Zircaloy is oxidized,
From these precipitates, some alloying elements form a solid solution in the oxide matrix.

However, originally, when these alloy elements are in solid solution in the matrix of the zirconium-based alloy, the concentration of the alloy elements in the matrix of zirconia becomes higher, and the oxygen vacancy concentration can be increased. It is possible to stabilize tetragonal zirconia up to a thicker oxide film and improve the corrosion resistance of the zirconium-based alloy.

Therefore, in order to produce a zirconium-based alloy, it is necessary to add an element having an oxidation number of less than 4 and form a solid solution in the matrix at a concentration as high as possible.

In the present invention, studies have been carried out based on the above-described zirconium-based alloy corrosion mechanism, and it has become possible to manufacture a zirconium-based alloy having improved corrosion resistance by using the conventional manufacturing process. That is, when the zirconium-based alloy according to the present invention is used by being loaded in the core of a nuclear reactor, tetragonal zirconia, which is a protective film having excellent corrosion resistance, can be stabilized for a long period of time, and therefore the corrosion resistance of the zirconium-based alloy is improved. Therefore, it is possible to obtain a zirconium-based alloy core structural material suitable for increasing the burnup of fuel.

[0031]

Embodiment An embodiment of the present invention will be described with reference to the drawings. FIG. 1 is a process flow diagram of a zircaloy fuel cladding tube according to an embodiment of the present invention after a raw tube. Differences between this embodiment and FIG. In FIG. 1, the outer surface quenching step 2 is not included, and the constituent material of the shell 1 is different.

In the case of FIG. 3 of the conventional example, the distribution of the alloying elements made uniform by the outer surface quenching step 2 becomes non-uniform again by the intermediate annealing step 5 and the final annealing step 7 after the outer surface quenching step 2. This is because Fe, Cr or Ni is α-
This is because it hardly forms a solid solution in Zr.

Therefore, at least 1 w in α-Zr
If an element that forms a solid solution of about t% and has the same corrosion resistance improving effect as Fe, Cr or Ni is added, the outer surface quenching step of the blank tube becomes unnecessary. Since these elements are added in large amounts, it is necessary to have a small thermal neutron absorption cross section from the viewpoint of neutron economy, and a suitable thermal neutron absorption cross section is about 1b or less. Elements satisfying such conditions include Y and Pb.

In this embodiment, an alloy in which at least one of Y and Pb is added to Zr in a total amount of 0.5 to 2 wt% is used, and as shown in FIG. Thus, a fuel cladding tube having excellent corrosion resistance was manufactured.

Next, another embodiment of the present invention will be described. FIG. 2 is a process flow diagram of a zircaloy-made cladding tube according to another embodiment of the present invention after the raw pipe, and FIG. 3 of the above-mentioned conventional example shows only the constituent material of the raw pipe 1. Different.

Further, in the manufacturing process of the conventional zircaloy cladding tube (see FIG. 3), Fe and Ni are sufficiently precipitated by the annealing temperature and the annealing time after the outer surface quenching step 2, but Cr is not contained. There was a possibility that the solution was slightly supersaturated.

Therefore, if an element having a diffusion coefficient smaller than that of Cr and an oxidation number of less than 4 is added as an alloy element, even if the manufacturing process of the conventional example is used, the alloy element on the outer surface is more than that of the conventional example. Further, it can be made into a solid solution state in supersaturation, and the corrosion resistance can be further improved.

Since the diffusion coefficient of Cr is about 10,000 times that of Zr, the self diffusion coefficient of Zr is used as a reference.
Elements having a diffusion coefficient less than 10,000 times Zr are effective.
Examples of such elements include Ti and Zn in addition to Cr.

Further, from the observation result of the microstructure of the zircaloy after irradiation, the intermetallic compound precipitated by the irradiation of fast neutrons forms a solid solution, and the corrosion resistance of the zirconium-based alloy is improved during use in the nuclear reactor. I understood it. It was also found that the rate of this irradiation-induced solid solution varies depending on the type of precipitate, and further depends on the type of element within the same precipitate.

For example, Zr-F in Zircaloy-2
Two types of intermetallic compounds, an e-Ni intermetallic compound and a Zr-Fe-Cr intermetallic compound, have been observed. From the observation result of Zircaloy-2 after irradiation, the Zr-Fe-Ni-based precipitate is more preferable. , Zr-Fe-Cr-based precipitates form a solid solution faster, and both Zr-Fe-Ni-based and Zr-Fe-Cr-based precipitates have the fastest solid solution of Fe. .

For irradiation-induced solid solution, the rapid ejection of alloying elements from the precipitate to the matrix by the fast neutrons, the recovery of the concentration gradient caused by the diffusion of the alloying elements inside the precipitate, and the removal of the alloying elements in the matrix It is thought that there are three diffusion processes.

In order to accelerate the irradiation-induced solid solution, it is necessary that the diffusion rate of the alloying element in the precipitate and the matrix is high. In the case of Zircaloy-2, comparing the irradiation-induced solid solution rates of alloying elements, the order is Fe> Ni >> Cr, which can be explained from the magnitude of the diffusion rate in the matrix.

Therefore, in order to promote the irradiation-induced solid solution, it is necessary to add an element having a large diffusion rate in Zr under the conditions of use of a nuclear reactor, and such an element should be F
There is Co in addition to e and Ni, and the diffusion coefficient of Co is Zr.
Is more than 100,000 times the self diffusion coefficient of. Further, these elements have an oxidation number smaller than Zr. The susceptibility of irradiation-induced solid solution due to the difference in the type of intermetallic compound can be roughly judged by its melting point, and when a substance having a low melting point and a substance having a high melting point are compared, the lower melting point shows the point defect. The mobility is high, and therefore the diffusion coefficient of the element in the compound is also high. In the case of Zircaloy-2 irradiated material, Zr-Fe-Ni-based precipitates are Zr-Fe-
Irradiation-induced solid solution is faster than Cr-based precipitates.

The melting points of Zr ₂ Ni and ZrCr ₂ having the same crystal structure as those of these intermetallic compounds are 120 respectively.
It is 0 ° C. and 1675 ° C., and it can be said that it is desirable to use an alloying element that forms an intermetallic compound having a melting point of 1200 ° C. or less with Zr in order to accelerate the irradiation-induced solid solution.

Among the alloying elements having a high diffusion rate in Zr under the operating conditions of the reactor described above, Ni and Co (Z
r ₂ Co) satisfies this condition. Fe (ZrFe ₂ ) does not satisfy this condition, but Fe in zircaloy does not directly form an intermetallic compound with Zr, and Zr ₂ Ni or ZrC
It can be said that irradiation-induced solid solution is likely to occur because it is contained in these intermetallic compounds by substituting Ni or Cr of r ₂ and can be present in the precipitate having a low melting point.

Therefore, in another embodiment of the invention, Z
In r, at least one of Y and Pb, the total amount of addition of which is 0.5 to 2.0 wt%, and Ti, Zn and C
At least one type out of o and the total addition amount is 0.1
~ 0.5 wt% element was added. As a result, the effect of improving corrosion resistance by quenching and the effect of improving corrosion resistance during irradiation could be improved.

[0047]

According to the present invention, it is possible to increase the concentration of a solid solution alloy element that contributes to the improvement of corrosion resistance, stabilize a tetragonal oxide film excellent in corrosion resistance for a long time, and oxidize a zirconium-based alloy. It is possible to provide a core structure material such as a fuel cladding tube and a spacer or a channel box, which is made of a zirconium-based alloy and has a reduced speed and is excellent in safety and reliability.

[Brief description of drawings]

FIG. 1 is a flow chart of a fuel cladding tube manufacturing process according to an embodiment of the present invention.

FIG. 2 is a flow chart of a fuel cladding tube manufacturing process of another embodiment of the present invention.

FIG. 3 is a flow chart of a conventional fuel cladding tube manufacturing process.

[Explanation of symbols]

DESCRIPTION OF SYMBOLS 1 ... Element pipe, 2 ... External quenching process, 3 ... Annealing process, 4 ... Cold rolling process, 5 ... Intermediate annealing process, 6 ... Final rolling process, 7
… Final annealing step.

─────────────────────────────────────────────────── ─── Continuation of the front page (51) Int.Cl. ⁵ Identification code Internal reference number FI Technical display location G21C 3/34 9216-2G G21C 3/34 Y

Claims (4)

[Claims] 1. A zirconium-based alloy which is composed of Zr and an additive element to the Zr and is used for a fuel cladding tube of a nuclear reactor or a core structure material, wherein the additive element has a valence of less than 4, A zirconium-based alloy, which is an element that is present in an oxidized state in an oxide film produced under the conditions of use of the nuclear reactor and is less likely to be oxidized than Zr. 2. The zirconium-based alloy according to claim 1, wherein the additional element is an element which forms a solid solution in α-Zr in an amount of at least 1 wt%. 3. The additive element is at least 1 wt% or more solid-dissolved in α-Zr and has a thermal neutron absorption cross section of about 1 b or less, and at least one kind of Y and Pb, the total addition amount of which is 0. The zirconium-based alloy according to claim 1, which contains 5 to 2.0 wt% of the element. 4. The additive element is at least 1 wt% or more solid-soluted in α-Zr and has a thermal neutron absorption cross-section of at least about 1b. 5 to 2.0 wt% element and at least one kind of Zn, Co and Ti, and the total addition amount is 0.
The zirconium-based alloy according to claim 1, which contains 1 to 0.5 wt% of the element.