JPS62194490A - Nuclear fuel element - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

[Detailed description of the invention] [Field of application of the invention] TECHNICAL FIELD This invention relates to nuclear fuel elements.

[Background of the invention]

ASTM for elements of cladding materials of nuclear fuel elements
The standard ASTM B353-77a is known.

Generally, in a nuclear fuel element, a plurality of nuclear fuel pellets are stacked and stored in a cladding tube, and both openings of the cladding tube are sealed with end plugs. Nuclear fuel pellets are made by molding and sintering fissile oxide fuel powder into cylindrical pellets with a length-to-diameter ratio of about 1, for example. In addition, the cladding tube of the nuclear fuel element configured in this way has the function of preventing the coolant from coming into contact with the nuclear fuel pellets and preventing chemical reactions from occurring, and the function of preventing radioactive fission products released from the fuel. A function is required to prevent the coolant from entering the coolant. Therefore, in the event that the cladding tube does not satisfy these functions, that is, the cladding tube is damaged, the radioactivity level in the cooling system plant will increase, and reactor operation must be stopped to ensure safety. It disappears.

On the other hand, cladding tubes of nuclear fuel elements used in water-cooled nuclear reactors are generally made of zirconium and alloys thereof. Zirconium and its alloys have a small neutron absorption cross section, are strong and ductile at temperatures below about 400° C., and have characteristics that do not react with water vapor used as a coolant.

However, according to operational experience to date, even cladding made of zirconium and its alloys is susceptible to brittleness due to a decrease in material strength due to neutron irradiation and corrosion due to chemical reactions with nuclear fission products. Cracks have occurred. It is thought that such an undesirable phenomenon occurs as follows.

That is, in order to efficiently transfer the heat generated by the nuclear fuel pellets to the outer surface of the cladding tube, it is necessary to set the gap formed between the inner surface of the cladding tube and the nuclear fuel pellets to several tens of microns or less. On the other hand, as nuclear fuel pellets generate heat during operation, the pellets themselves crack due to thermal stress, resulting in misalignment of the fracture surfaces, and furthermore, fission products accumulate within the nuclear fuel pellets during combustion, resulting in volumetric expansion. , the cladding tube is expanded by the nuclear fuel pellets and subjected to stress. Although the average value of the strain in the circumferential direction that the tube 'III is subjected to is not very large, not only does the strain locally concentrate on the wall near the crack that occurs in the nuclear fuel pellet, but this strain exceeds the yield stress. reach. 3) In addition, corrosive gases such as iodine and iodine compounds, cesium and cesium compounds are generated from nuclear fuel pellets as a result of nuclear fission, and these corrosive gases collect in free spaces within the cladding, such as cracks. In particular, corrosive gas tends to collect near parts of the cladding where strain is particularly concentrated.

Generally, when stress (especially above yield stress) acts in a corrosive gas atmosphere, the ductility of the material decreases.

A brittle fracture phenomenon called stress corrosion cracking occurs.

Stress corrosion cracking depends on temperature, stress, and concentration of corrosive gas.

The mechanism by which dissolved oxygen occurs depends on the composition of the dissolved oxygen 9 alloy, heat treatment, degree of processing, etc., and there is no single mechanism for its occurrence. The concept of lining the cladding tube for the purpose of preventing such undesirable destruction is well known, and is disclosed in U.S. Pat. No. 3,502,549,
U.S. Patent No. 3,625,821, Japanese Patent Application Publication No. 51-6979
No. 2, JP-A-51-69795, JP-A-Sho 51
In JP-A-69796 and JP-A-51-71497, Mo+WtNb, Ni,
Fe, Mg, Cu, pure Zr, AQ.

Ni--Cr alloys, aluminized coatings, silicided coatings, etc. are shown.

However, some liner materials used as barrier materials have drawbacks such as a large neutron absorption cross section, which reduces the economic efficiency of the reactor. Additionally, the use of liner materials not only increases the manufacturing process of the cladding, but also presents inherent technical problems and economic disadvantages.

[Purpose of the invention]

The present invention has been made in view of the above points, and it is an object of the present invention to provide a nuclear fuel element that can increase stress corrosion cracking resistance and improve reliability and economic efficiency.

[Summary of the invention]

That is, the present invention provides a nuclear fuel element in which nuclear fuel pellets are sealed inside a cladding tube, wherein the cladding tube has a carbon concentration of 80M ppm or less and an oxygen concentration of 600 ppm by weight.
It is characterized by being formed from the following zirconium alloy, which makes stress corrosion cracking of the cladding less likely to occur.

The inventor investigated how stress corrosion cracking resistance could be improved. Among the elements constituting the zirconium alloy, the elements that increase the strength at high temperatures (approximately 350° C.) are mainly elements such as oxygen, which are dissolved in zirconium. In addition, carbon is a major element that is dispersed within the crystal grains and increases the strength of the cladding. Conventionally, carbon is contained at about 140 ppm by weight, but it is industrially possible to further reduce this concentration. Furthermore, in the past, the oxygen concentration was set at about 1100 to 1300 ppm by weight to ensure the strength of the cladding tube in the initial stage of use, but in the present invention, in order to keep the high temperature strength of the cladding tube low, the concentration of oxygen as well as carbon We focused on the point that by restricting the cladding tube, a cladding tube with excellent resistance to stress corrosion cracking caused by the interaction between the fuel and the cladding tube can be obtained6 [Embodiments of the Invention] Table 1 shows examples of the present invention. The chemical composition excluding zirconium of the zirconium alloy used in the above is shown.

The mouth is an example. First, the chemical composition standard for nuclear reactor grade Zircaloy-2 (ASTM*R353+ Gra
The following amounts of alloying elements satisfying HRA-1,) are added to zirconium. i.e. tin 1.20 to 1
−． 70 wt%, iron 0.07 to 0.20 wt%, chromium 0.05 to 0.15 wt%, nickel 0.03 to 0.08 wt%, iron + chromium-nickel 0.18 to 0.38 wt% %. Next, according to the above-mentioned conventional standards, the carbon content is 270 ppm by weight or less, but in Example 42, the carbon concentration was 50.80 ppm by weight, and the oxygen concentration was 4 ppm by weight.
00,600 ppm by weight, and other impurity elements were made into a zirconium ingot that satisfied the above specifications. Conventional examples C and 2 have carbon concentrations of 140 and 170 ppm by weight, respectively.
, the oxygen concentration is 1100 and 1300 ppm by weight, respectively.''
The ingot was made to satisfy the conventional carbon concentration of about 140 ppm by weight and oxygen concentration of about 1,100 to 1,300 ppm by weight, as in the case of two series.

Table 1 Cladding tubes were manufactured from the Zircaloy-2 materials of I99 and H2 whose compositions were adjusted as described above, according to the manufacturing process shown in FIG. That is, I22 Ro, H2's Zircaloy-2
Cladding tubes were manufactured from M materials through the steps of melting (ingot), forging/forming (hollow billet), hot extrusion (base tube), cold rolling, and sintering.

In this case, the final sintering temperature was selected from within the range of 480 to 500°C. This is to reduce the strength at high operating temperatures and to ensure that a predetermined strength can be obtained even at the initial stage of use.

Furthermore, the contamination of carbon during the manufacturing process was avoided as much as possible.

In order to investigate the properties of the Zircaloy-2 cladding tubes produced in this manner, hollow nuclear fuel pellets were inserted into these cladding tubes. Then, a cylindrical pure aluminum rod was inserted into the hollow part of this nuclear fuel pellet, and the iodine concentration was 3 mg/cc.
An aluminum rod was compressed in the longitudinal direction in an atmosphere with a cladding tube temperature of 350° C., and stress in the circumferential direction was applied to the cladding tube through a hollow nuclear fuel pellet, and the elongation at break was determined.

The measurement results of elongation at break are shown in FIG. In this figure, the vertical axis shows the elongation at break in the circumferential direction, and the horizontal axis shows the sample, i.e., Example 42.
First, the cladding tubes made with the chemical compositions of conventional examples C and 2 were
9ro 0. Ha0. Taking the number 20, the example subject NI
Pipe A 02 port 0, Conventional example cladding pipe C 0. The change characteristics of the elongation at break in the circumferential direction of 20 samples were shown.

As is clear from the figure, the elongation at break of the example cladding tube 404 is 0.0 compared to that of the conventional example cladding tube. It was recognized that it was larger than that of 20.

In this way, the carbon concentration in Zircaloy-2 was reduced to 80°50 and 80% by weight ppI11, and the oxygen concentration was reduced to 600°400.
By limiting the amount to 600 ppm by weight or less, it has been confirmed that when stress is applied to the cladding tube due to interaction with fuel in corrosive gas, its elongation at break increases and stress corrosion cracking becomes less likely to occur. It was done. The same applies not only to Zircaloy-2 but also to other Zircaloys and zirconium-niobium alloys.

As shown in these examples, the cladding tube had a carbon concentration of 80 pp by weight.
A nuclear fuel element made of a zirconium alloy with an oxygen concentration of less than 600 ppm by weight and an oxygen concentration of less than 600 ppm by weight has a circumferential elongation at break that is significantly larger than that of a conventional one, has increased stress corrosion cracking resistance, and is highly reliable and economical. improves.

〔Effect of the invention〕

-1- As mentioned above, the present invention has increased stress corrosion cracking resistance, improved reliability and economical efficiency.

It is possible to obtain a nuclear fuel element that has increased stress corrosion cracking resistance and improved reliability and economic efficiency.

BRIEF DESCRIPTION OF THE DRAWINGS FIG. 1 is a process diagram for manufacturing a cladding tube according to one embodiment of the nuclear fuel element of the present invention, and FIG. 2 is a diagram showing the elongation characteristics at break in the circumferential direction of the cladding tube manufactured according to the same embodiment.

Claims (1)

[Claims] 1. A nuclear fuel element in which nuclear fuel pellets are sealed inside a cladding tube, wherein the cladding tube has a carbon concentration of 80 pp by weight.
A nuclear fuel element, characterized in that it is formed from a zirconium alloy with an oxygen concentration of 600 ppm or less and an oxygen concentration of 600 ppm or less by weight. 2. The nuclear fuel element according to claim 1, wherein the zirconium alloy is formed from Zircaloy.