JPH0640136B2 - Nuclear fuel cladding - Google Patents

Description

Description: FIELD OF THE INVENTION The present invention relates to a nuclear fuel element used in the core of a nuclear fission reactor, and more particularly to a nuclear fuel cladding tube for a light water cooled nuclear reactor.

[Background of the Invention]

FIG. 6 is a sectional view of a conventional nuclear fuel element. In the figure, 1 is a cladding tube, 2 is a fuel pellet, 3a and 3b are end plugs, 4 is a plenum, 5 is a spring, and 6 is a liner layer. A large number of fuel pellets 2 made of uranium oxide are stacked and housed in the cladding tube 1, and both end openings have end plugs 3a,
It is sealed at 3b. A gas storage plenum 4 is provided above the nuclear fuel element, and a spring 5 for stably supporting the fuel pellet 2 is provided.

In the nuclear fuel element configured as described above, the cladding tube 1
Is required to prevent the radioactive fission products released from the fuel pellets 2 from entering the coolant. However, according to the operation experience to date, the cladding tube 1 is based on the superposition action of the stress generated in the cladding tube 1 due to the chemical reaction with the corrosive fission products and the thermal expansion of the fuel pellet 2 at the time of high burnup. It has been discovered that stress corrosion cracking occurs.

In order to prevent the stress corrosion cracking of the cladding tube 1 as described above, as shown in FIG.
So-called zirconium liner cladding has been developed which is provided with a pure zirconium liner layer 6 of ˜100 μm. The pure zirconium liner layer 6 is expected to prevent contact between the cladding tube 1 and corrosive fission products and to alleviate local stress generated in the cladding tube 1 to prevent stress corrosion cracking.

However, the pure zirconium liner layer 6 contains impurities substantially more or less, and the effect of reducing stress corrosion cracking is impaired depending on the types of impurities and their contents. Therefore, conventionally, a zirconium liner material has been used in which a high-purity crystal bar zirconium or a high-purity portion of sponge zirconium is used and the content of impurities is reduced as much as possible.

[Object of the Invention]

The present invention has been made in view of the above circumstances, and an object of the present invention is to provide a nuclear fuel cladding tube that can reduce the risk of stress corrosion cracking of the cladding tube and improve reliability.

[Outline of Invention]

The nuclear fuel cladding tube of the present invention comprises an inner zirconium liner layer and an outer zirconium alloy cladding tube, and the ratio of oxygen concentration a to iron concentration b in the zirconium liner layer, a / b is 1. Impurities larger than 0 and contained in the zirconium liner layer are solid-dissolved in the zirconium matrix.

According to the present invention, the total amount of impurities in the zirconium liner layer is 5000 p.
It is preferably pm or less, and regarding oxygen and iron contained therein, oxygen concentration a (ppm) and iron concentration b (p
The value of the ratio a / b of pm) is larger than 1.0 as described above, and the impurities contained in the zirconium liner layer are solid-solved in the zirconium matrix, so that the stress corrosion cracking A zirconium liner with low sensitivity is used to prevent stress corrosion cracking.

Among the impurities, oxygen is conventionally considered to be a particularly important factor because oxygen enhances the mechanical strength of zirconium. On the other hand, the technical idea of keeping the oxygen concentration below a certain value is Japanese Patent Publication No. 33037 and JP-A-54-59600
It is disclosed in the publication. However, according to the results of experiments recently conducted by the present inventors, it has been found that the iron concentration is a more important factor than the oxygen concentration. Fig. 5 shows the effect of oxygen concentration and iron concentration on the stress corrosion cracking susceptibility of zirconium according to the results of a recent experiment in which the horizontal axis represents the impurity content and the vertical axis represents the stress corrosion cracking susceptibility. Is. It shows that the oxygen concentration has no significant effect, while the stress corrosion cracking is likely to occur as the iron concentration increases. It was found that this was due to the presence of particles of an intermetallic compound of zirconium and iron.

By the way, in the case of the zirconium liner layer 6 of the cladding tube 1 of the currently manufactured nuclear fuel element, the oxygen concentration a (ppm)
And the iron concentration b (ppm), the value of a / b is about 1.
It is close to 0, and the oxygen concentration and the iron concentration are almost in the same ratio.
In other words, the current zirconium liner materials are extremely high purity crystal bar zirconium having an oxygen concentration and an iron concentration of 200 ppm or less, respectively, or an oxygen concentration and a zirconium concentration of 500 ppm to 1000 ppm, respectively.
Sponge zirconium is considered. The present invention has been made based on the experimental results shown in FIG. 5, and by removing iron in sponge zirconium to reduce the iron concentration, that is, the ratio a of the oxygen concentration a and the iron concentration b. By making the value of / b larger than 1.0, it has the same low stress corrosion cracking susceptibility as crystal bar zirconium. And the present invention is
In order to further improve the susceptibility to stress corrosion cracking due to iron content, a solution treatment is performed. The solution treatment is effective not only for iron but also for impurity elements forming an intermetallic compound with zirconium, that is, Cr, Ni, Si, C, W, Al and the like.

Further, the present inventors have found that zirconium matrix contains an intermetallic compound with zirconium (hereinafter referred to as second phase particles).
It has been discovered that the more the amount of the present is present, the more the stress corrosion cracking reducing effect of the zirconium liner is impaired. That is, under a stress corrosion environment in iodine (fission product), about 0.1 μm
When the above-mentioned second phase particles are randomly dispersed regardless of the crystal grain boundaries of zirconium and the crystal grains, the precipitation region of the second phase particles becomes a stress concentration field, and the initiation and propagation of cracks are significantly increased. I knew it would help. Therefore, paying attention to the impurities that generate the second phase particles, if they are forcibly solid-dissolved in the zirconium matrix, the amount of the second phase particles will be reduced, and the original purpose of the pure zirconium layer will be achieved. A certain stress corrosion relaxation effect is exhibited.

Example of Invention

The nuclear fuel cladding tube of the present invention will be described below with reference to the drawings using the same reference numerals for the same parts as those of the conventional one. FIG. 1 shows a manufacturing process of a cladding tube having a pure zirconium liner layer 6.
The zirconium liner layer 6 is a liner material having a ratio of oxygen concentration a and iron concentration b, a / b> 1, and during the manufacturing process of FIG. 1, the following (A), (B), (C) It adds a process.

(A) Solid solution treatment of zirconium hollow billet (liner layer 6) (B) After hot extrusion, solution treatment of raw tube (C) After final annealing, solution treatment of zirconium liner cladding tube In this example, It is most effective to apply all of the steps (A), (B) and (C). However, any one of steps (A), (B), and (C) may be added. In particular, it is effective to add the step (C).

The smaller the precipitation amount of the second phase particles, the more preferable, but
As described above, stress corrosion cracking is a phenomenon that occurs stochastically with the portion where the second phase particles are present as the nucleus, so the number of those that can be substantially observed is approximately 0.1 μm or more. By controlling the amount to 50% or less of the conventional one to which the present invention is not applied, a sufficient improvement effect can be obtained.

In the solution treatment, since the above impurity elements are re-dissolved in zirconium, the solution treatment is performed on zirconium used as a liner from 800 ° C. or 860 ° C., that is, near the transformation temperature (862 ° C.) of zirconium, and then uniformly Re-dissolve the second phase particles. Further, in order to remove the quenching strain due to the solution treatment, strain relief annealing is performed in a low temperature range where re-precipitation of the second phase particles is not performed.

FIG. 2 shows zirconium having different ratios of iron and oxygen concentrations described in the prior application of the present inventors (Japanese Patent Application No. 59-46300) in which the horizontal axis represents iron concentration and the vertical axis represents oxygen concentration. It is a stress corrosion test result of a liner pipe. The samples 7, 8 and 9 in FIG. 2 added thereto are the samples in which the present invention was carried out, and the samples 7, 8 and 9 were evaluated as ○ (excellent in performance). The evaluation result of the present invention was ⊚ (excellent in performance), and the effect of the present invention is clear. Then, the samples 7, 8 of FIG.
⊚ other than 9 indicates that zirconium, which is a good material, was used. The test conditions are as follows. The impurity analysis values of Samples 7, 8 and 9 are as shown in Table 1.

Test conditions Iodine concentration: 0.20 torr Strain rate: 10 ^-6 to 10 ^-3 sec ^-1 Test temperature: 350 ° C Samples 7, 8 and 9 were manufactured by the process of FIG. 1 as described above, and after the final annealing, they were heated in vacuum at 850 ° C. for 2 hours and rapidly cooled to room temperature. In addition, 530 samples
Annealed in vacuum for 2 hours at ° C. FIG. 3 shows a scanning electron micrograph of the inner surface of the zirconium liner layer of Sample 7.
(A) is a photograph before the solution treatment, and (B) is a photograph after the solution treatment. It can be seen that the number of second phase particles precipitated is extremely reduced by the solution treatment.

FIG. 4 shows the total amount of impurities contained in pure zirconium, that is, (Fe + Cr + Ni + Si + W + Al + C).
FIG. 3 is an explanatory diagram showing the relationship between the two, with the amount on the horizontal axis and the average circumferential breaking strain as a stress corrosion cracking susceptibility index on the vertical axis. The results shown in FIG. 4 show the results of the stress corrosion cracking test by the strain concentration method, and the test conditions are as follows.

Test conditions Iodine concentration: 1 mg / cm ² Strain rate: 1 × 10 ⁻³ min ⁻¹ Test temperature: 350 ° C. Also, in Table 2, the three types of zirconium liner tubes A, B, and C used for the liner layer 6 were used. Shows the impurity concentration of. In both cases, the ratio of oxygen concentration to iron concentration is greater than 1.0. Further, characteristic 10 is the one which is subjected to the solution treatment of the present invention, and characteristic 11 is the one which is not subjected to the solution treatment. As a result, the breaking strain decreases as the amount of impurities increases, but the breaking strains of the present invention are high.
That is, it can be seen that the stress corrosion cracking susceptibility is lowered.
The solution treatment is the same as in the case of FIG.

As described above, in the nuclear fuel cladding tube of the present embodiment, the ratio of the oxygen concentration to the iron concentration in the zirconium liner layer is greater than 1.0, and the impurity zirconium matrix contained in the zirconium liner layer is contained. Since it is a solid solution, the function of preventing the stress corrosion cracking by the zirconium liner layer is sufficiently exerted, the risk of the stress corrosion cracking of the cladding can be significantly reduced, and the reliability can be improved.

〔The invention's effect〕

As described above, the nuclear fuel cladding tube of the present invention has an effect of significantly reducing the risk of stress corrosion cracking of the cladding tube and improving reliability.

[Brief description of drawings]

FIG. 1 is an explanatory view of a manufacturing process of an embodiment of a cladding tube for nuclear fuel of the present invention, and FIG. 2 is an explanatory view of a relationship between iron and oxygen concentrations and stress corrosion cracking susceptibility of a cladding tube manufactured in the manufacturing process of FIG. , Third
Figures (a) and (b) are scanning electron micrographs of the surface of the liner layer before and after the solution treatment in the step of Figure 1, respectively.
FIG. 5 is an explanatory diagram of the relationship between the total amount of impurities in the liner layer manufactured in the process of FIG. 1 and the average circumferential breaking strain, and FIG. 5 is an explanatory diagram of the relationship between the impurity contents of iron and oxygen and the stress corrosion cracking susceptibility, FIG. 6 is a vertical sectional view of a conventional nuclear fuel element. 1 ... cladding tube, 6 ... liner layer.

Claims (3)

[Claims] 1. A method comprising an inner zirconium liner tube and an outer zirconium alloy cladding tube, wherein the ratio of oxygen concentration a to iron concentration b in the zirconium liner layer, a / b is 1.0 or more. A cladding tube for nuclear fuel, which is large and in which impurities contained in the zirconium liner layer are solid-solved in zirconium matrix. 2. The cladding tube for nuclear fuel according to claim 1, wherein the impurities are elements that form an intermetallic compound with zirconium such as iron, chromium, nickel, silicon, carbon, tungsten and aluminum. 3. The zirconium liner layer is subjected to a solution treatment in which it is heated near the transformation temperature of zirconium and then rapidly cooled to room temperature, and a strain relief annealing treatment after the solution treatment. A nuclear fuel cladding tube according to claim 1.