KR100296952B1 - New zirconium alloys for fuel rod cladding and process for manufacturing thereof - Google Patents

Abstract

The present invention relates to a zirconium alloy composition for a nuclear fuel cladding tube and a method for manufacturing the same, specifically, 0.8-1.2 wt% of niobium (Nb), 0.1-0.3 wt% of copper (Cu), 600-1400 ppm of oxygen (O) and the balance of zirconium. (Zr) alloy of the present invention and niobium (Nb) 0.8-1.2 wt%, tin (Sn) 0.8-1.2 wt%, iron (Fe) 0.2-0.4 wt%, copper (Cu) 0.1-0.3 wt% The alloy of the present invention, consisting of 600-1400 ppm of oxygen (O) and the balance of zirconium (Zr), can be usefully used as fuel cladding, support grids and structures within the reactor core of light and heavy water reactors. According to the production method of the present invention in which the β-heat treatment (beta heat treatment) and cold working are performed in order before the final heat treatment at 450 to 600 ° C. for 10 to 100 hours to manufacture the alloy, a zirconium alloy with improved corrosion resistance can be obtained. have.

Description

New zirconium alloys for fuel rod cladding and process for manufacturing description

The present invention relates to zirconium alloy compositions for nuclear fuel cladding and to a method of manufacturing the same, specifically, 0.8-1.2 wt% niobium (Nb), 0.1-0.3 wt% copper (Cu), 600-1400 ppm oxygen (O) and the balance Alloy composition composed of zirconium (Zr) and 0.8-1.2 wt% of niobium (Nb), 0.8-1.2 wt% of tin (Sn), 0.2-0.4 wt% of iron (Fe), 0.1-0.3 wt% of copper (Cu), and oxygen (O) Final heat treatment for 10 to 100 hours at an alloy composition consisting of 600-1400 ppm and residual zirconium (Zr) and 450-600 ° C. temperature in which β-heat treatment (beta heat treatment) and cold working were performed in sequence before final heat treatment. The present invention relates to a method for producing an alloy.

Zirconium alloys are widely used as fuel cladding and structural materials in nuclear reactor cores due to their low neutron absorption area and excellent corrosion resistance and mechanical properties. Most widely used. The development of zircaloy-based alloys is described in detail in ASTM STP-368 (p. 3-27, 1963). Zircaloy-1 (Sn: 2.5%, Zr: balance), Zircaloy-2 (Sn: 1.20-1.70%, Fe: 0.07-0.20%, Cr: 0.05-1.15%, Ni: 0.03-0.08% , O: 900-1500 ppm, Zr: residue, where (Fe + Cr + Ni): 0.16-1.70%), Zircaloy-3A (Sn: 2.5%, Fe: 0.25%, Zr: residue), Zircaloy-3B (Sn: 0.5%, Fe: 0.4%, Zr: remainder), Zircaloy-3C (Sn: 0.5%, Fe: 0.2%, Ni: 0.2%, Zr: remainder), Zircaloy-4 (Sn: 1.20- 1.70%, Fe: 0.18-0.24%, Cr: 0.07-0.13%, O: 900-1500 ppm, Ni: <0.007%, where (Fe + Cr): 0.28-0.24%) Introducing. However, alloys except Zircaloy-2 and Zircaloy-4 could not be commercialized due to poor mechanical strength and corrosion characteristics in the furnace. According to ASTM STP-368 described above, the role and effect of alloying elements is different, and Sn improves mechanical properties or removes the adverse effects of corrosion resistance that may be caused by nitrogen contained in sponge zirconium. In order to improve the corrosion resistance, the main purpose is to add Fe, Cr and Ni.

Current operating conditions of nuclear power plants are developing into a situation that is difficult to overcome with conventional Zircaloy-4 fuel cladding materials. Zircaloy-based alloys are no longer used as fuel cladding materials as operating conditions change with high pH operation to increase combustion, increase operating temperature, and reduce the radiation levels of the primary system of the plant. It became difficult to use. Therefore, many nuclear fuel-related organizations have studied a lot of new Zr alloys for many years to improve the corrosion resistance and strength of Zr alloys.

In US Pat. No. 5,254,308, Nb and Fe were added to maintain mechanical properties due to Sn reduction. This alloy contains 0.45-0.75% Sn (typically 0.6%), 0.4-0.53% Fe (typically 0.45%), 0.2-0.3% Cr (typically 0.25%), 0.3-0.5% Nb (typically 0.45%) , 0.012-0.03% Ni (typically 0.02%), 50-200 ppm Si (typically 100 ppm), and 1000-2000 ppm oxygen (typically 1600 ppm). At this time, the Fe / Cr ratio was 1.5, and the amount of Nb added was determined according to the amount of Fe affecting the hydrogen absorption, and the amount of Ni, Si, C, O was determined to have excellent corrosion resistance and strength. And US Pat. No. 5,278,882 discloses 0.4-1.0% Sn (typically 0.5%), 0.3-0.6% Fe (typically 0.46%), 0.2-0.4% Cr (typically 0.23%), 0.012- without adding Nb. Consists of 0.03% Ni (typically 0.02%), 50-200 ppm Si (typically 100 ppm) and 1200-2500 ppm oxygen (typically 1800 ppm). It is considered for. U.S. Patent No. 5334345 discloses alloy compositions of 1.0-2.0% Sn, 0.07-0.70% Fe, 0.05-0.15% Cr, 0.16-0.40% Ni, 0.015-0.30% Nb (typically 0.015) to improve corrosion and hydrogen absorption resistance. -0.20%), 0.002-0.05% Si (typically 0.015-0.05%), and 900-1600 ppm oxygen. U.S. Patent No. 5366690 controls the amount of Sn, N, and Nb added to produce alloy compositions of 0-1.50% Sn (typically 0.6%), 0-0.24% Fe (typically 0.12%), and 0-0.15% Cr (typically 0.10%), 0-2300 ppm N, 0-100 ppm Si (typically 100 ppm), 0-1600 ppm oxygen (typically 1200 ppm), 0-0.5% Nb (typically 0.45%).

US 4863685, 4986975, 5024809 and 5026516 are for Zr alloys containing 0.5-2.0% Sn and approximately 0.5-1.0% of other solute atoms. These alloys contain 0.09-0.16% oxygen. The alloy of US Pat. No. 48,636,85 is composed of Mo, Te and mixtures thereof or Nb-Te, Nb-Mo in addition to Sn. The alloy composition of US Pat. No. 4986975 limited the content of Cu, Ni, Fe solute atoms to 0.24-0.40%, and Cu was added by adding at least 0.05%. In US Patent Nos. 5024809 and 5026516, the range of addition of solute atoms was 0.5-1.0%, the same as US Pat. No. 4863685, and Bi or (Bi + Sn) was added in 0.5-2.5%. Mo, Nb, Te.

In US Patent No. 4938920, Sn was reduced to 0-0.8% and 0-0.3% V and 0-1% Nb were added to develop Zircaloy-4 to develop an alloy with better corrosion resistance. At this time, the addition range of Fe was 0.2-0.8% and Cr was 0-0.4%, and the range of (Fe + Cr + V) was limited to 0.25-1.0%. In addition, the range of oxygen added was 1000-1600 ppm. 0.8Sn-0.22Fe-0.11Cr-0.14-O, 0.4Nb-0.67Fe-0.33Cr-0.15O, 0.75Fe-0.25V-0.1O and 0.25Sn-0.2 after 200 days of corrosion test in 400 ℃ steam atmosphere The corrosion of the alloy having a composition of Fe-0.15V-0.1O was very good, about 60% of Zircaloy-4, and the tensile strength was similar to that of Zircaloy-4. U.S. Patent No. 4735768 relates to a sheath of duplex type. That is, a second alloy having excellent corrosion resistance is bonded to the outside of the first alloy having excellent creep resistance to about 5-20% of the thickness of the entire coating tube to extend the life of the coating tube. The outer layer to improve the corrosion resistance was limited to Fe 0-1.0% and added one of V (0.1-1%), Pt (0.1-1%), Cu (1-3%).

In US Patent No. 4963323, an attempt was made to modify an existing alloy component of Zircaloy-4 in order to develop a fuel cladding material having improved corrosion resistance. In other words, the alloy composition was designed by reducing the content of Sn, compensating for the reduction of Sn by adding Nb, and controlling nitrogen below 60 ppm. Sn: 0.2-1.15%, Fe: 0.19-0.6% (preferably 0.19-0.24%), Cr: 0.07-0.4% (preferably 0.07-0.13%), Nb: 0.05-0.5%, N: ≤ It was a Zr alloy composed of 60 ppm. In US Patent No. 5017336, Nb, Ta, V, Mo was added to modify the existing alloy components of Zircaloy-4, alloy composition range of Sn: 0.2-0.9%, Fe: 0.18-0.6%, Cr: Zr base alloy consisting of 0.07-0.4%, Nb: 0.05-0.5%, Ta: 0.01-0.2%, V: 0.05-1%, Mo: 0.05-1%. U.S. Patent No. 5196163 also shows Zr alloys (Sn: 0.2-1.15%, Fe: 0.19-0.6) in which Ta is added and Nb is selectively added as well as Sn, Fe, and Cr of the existing Zircaloy-4 alloying components. % (Preferably 0.19-0.24%), Cr: 0.07-0.4% (preferably 0.07-0.13%), Ta: 0.01-0.2%, Nb: 0.05-0.5%, N: ≤ 60 ppm) have.

U.S. Patent 5560790 discloses an alloy consisting of 0.5-1.5% Nb, 0.9-1.5% Sn, 0.3-0.6% Fe, 0.005-0.2% Cr, 0.005-0.04% C, 0.05-0.15% O, 0.005-0.15% Si I ask for the furtherance. In this patent, the distance between Sn or Fe-containing precipitated phases (Zr (Nb, Fe) ₂ , Zr (Fe, Cr, Nb), (Zr, Nb) ₃ Fe) is set to 0.20-0.40 μm, and Fe is The precipitated phase contained was limited to 60 vol.% Of the total precipitated phase.

US Pat. No. 5125985, on the other hand, creep materials using various manufacturing processes for alloys with a composition range of 0.5-2.0Nb, 0.7-1.5Sn and 0.07-0.28 (at least one of Fe, Ni and Cr). The characteristics could be adjusted. One of the features of the manufacturing process was the introduction of β-quench heat treatment in the middle.

In general, the reactor core of a nuclear power plant using thermal neutrons (eg, pressurized water reactor, boiling water reactor, pressurized water reactor) is circulated with high temperature hard water or heavy water as a coolant or neutron moderator. A sheath of Zr alloy encloses a heat generating UO ₂ nuclear fuel. The reason why the Zr alloy can be used as a fuel cladding tube is because the neutron absorption cross section is small and the corrosion resistance at the reactor operating temperature is relatively excellent. Zr alloys currently used as fuel cladding materials are Zircaloy-based alloys containing Sn, Fe, Cr, and Ni.

However, considering the trend of longer nuclear fuel loading and increased target combustion for improving the economics of power plants, the existing zircal alloys are reaching the limit of their use. It is no longer possible to ensure the health of nuclear fuel. Therefore, the development of Zr alloy with improved corrosion resistance and strength to be used in such a power plant environment is urgently needed. In order to obtain a Zr alloy with improved corrosion resistance and strength than the existing zircal alloy alloy, processing conditions and heat treatment conditions are required. In some cases, as in the aforementioned patents, the type and amount of added elements were changed.

Accordingly, the present inventors have been conducting research to develop a new Zr alloy with improved corrosion resistance and strength, and Zr-1Nb-0.2Cu, which is expected to have excellent corrosion resistance in the reactor core of a light water reactor and a heavy water reactor type nuclear power plant. Zirconium alloy and Zr-1Nb-1Sn-0.3Fe-0.1Cu zirconium alloy and manufacturing process for obtaining excellent corrosion resistance were developed to complete the present invention.

It is an object of the present invention to provide zirconium alloy compositions for nuclear fuel cladding with excellent corrosion resistance and methods for their preparation.

1 is a process diagram showing a manufacturing process (manufacturing process 2) of the present invention and a manufacturing process (manufacturing process 1) of a comparative example,

2A is a graph showing the effect of heat treatment on the corrosion of Zr-1% Nb-0.2% Cu alloy;

-■-: Comparative Example 1-●-: Comparative Example 2

-▲-: Example 1-▼-: Example 2

------: Zircaloy-4-tube

2b is a graph showing the effect of heat treatment on the corrosion of Zr-1% Nb-1% Sn-0.3% Fe-0.1% Cu alloy;

-▲-: Example 3-▼-: Example 4

3 is a graph showing the change in hardness value according to the heat treatment of Zr-1% Nb-0.2% Cu alloy,

-■-: Comparative Example 1-●-: Comparative Example 2

-▲-: Example 1-▼-: Example 2

Figure 4 shows a SEM micrograph of the precipitate formed in the specimen heat-treated for 3 hours,

Fig. 4A: Comparative Example 1, Fig. 4B: Comparative Example 2, Fig. 4C: Example 1, Fig. 4D: Example 2

Figure 5 shows a TEM tissue picture of the specimen heat-treated for 50 hours,

5A: Example 1, FIG. 5B: Example 2

FIG. 6 shows typical precipitates and EDS analysis results of specimens heat-treated at 580 ° C. for 30 minutes by the preparation method of Example 2. FIG.

Figure 6a and 6b: precipitates, Figure 6c: EDS analysis results

7 shows typical precipitates and EDS analysis results of specimens heat-treated at 580 ° C. for 50 hours by the preparation method of Example 2;

Figure 7a: precipitates, Figure 7b and Figure 7c: EDS analysis results

8 shows typical precipitates and EDS analysis results of specimens heat-treated at 480 ° C. for 30 minutes by the method of Example 1,

Figure 8a: precipitates, Figure 8b and Figure 8c: EDS analysis results

9 shows typical precipitates and EDS analysis results of specimens heat-treated at 580 ° C. for 30 minutes and 50 hours by the method of Example 4,

Figure 9a: precipitate (30 minutes heat treatment), Figure 9b: precipitate (50 hours heat treatment),

9c: EDS analysis results

FIG. 10 shows typical precipitates and EDS analysis results of specimens heat-treated at 480 ° C. for 50 hours by the preparation method of Example 3. FIG.

Figure 10a: precipitates, Figure 10b: EDS analysis results

In order to achieve the above object, in the present invention, an alloy composition composed of Nb 0.8-1.2 wt%, Cu 0.1-0.3 wt%, O 600-1400 ppm and the balance Zr and Nb 0.8-1.2 wt%, Sn 0.8-1.2 wt% , Alloy composition consisting of 0.2-0.4 wt% Fe, 0.1-0.3 wt% Cu, 600-1400 ppm O and balance Zr, and a method for producing a zirconium alloy in which β-heat treatment and cold working are performed sequentially before final heat treatment. do.

Hereinafter, the present invention will be described in detail.

The biggest stumbling block in high-combustion cladding is corrosion acceleration, and the next problem is creep caused by neutron irradiation. In the present invention, the alloy was developed with the focus on improving the corrosion resistance of the high-combustion cladding tube among these phenomena. First of all, in order to select alloying elements, neutron effect, manufacturing cost, processability, alloying properties with parent Zr are considered first, and based on the literature studies published so far, how each additive element affects corrosion resistance, mechanical properties, and creep characteristics. The alloy of the present invention was designed based on this study. In addition, the corrosion resistance of the alloy is closely related to the microstructure that is changed by processing or heat treatment during the manufacturing process, and thus the alloy manufacturing process is optimized through corrosion test and structure observation.

The present invention provides a zirconium alloy composition for nuclear fuel cladding consisting of Nb 0.8-1.2 wt%, Cu 0.1-0.3 wt%, O 600-1400 ppm and the balance Zr, 1.0 wt% Nb, 0.2 wt% Cu, 600 oxygen Most preferred is the configuration of -1400 ppm and the balance Zr.

In addition, in the present invention, a zirconium alloy composition for a nuclear fuel cladding tube comprising Nb 0.8-1.2 wt%, Sn 0.8-1.2 wt%, Fe 0.2-0.4 wt%, Cu 0.1-0.3 wt%, O 600-1400 ppm and the balance Zr Most preferably, the composition of 1.0 wt% Nb, 1.0 wt% Sn, 0.3 wt% Fe, 0.1 wt% Cu, 600-1400 ppm oxygen and the balance Zr.

In addition, the present invention provides a method for producing the alloy composition of claim 1, the method of the present invention is subjected to β-heat treatment and cold working in order before the final heat treatment, and then the final heat treatment for 10 to 100 hours at 450 ~ 600 ℃ temperature (The sixth to eighth steps described later) has the greatest feature.

The method for producing a zirconium alloy composition of the present invention specifically comprises the steps of dissolving each metal having the composition of the present invention (first step); β-heat treatment (second step); Hot rolling (third step); Heat treatment step (fourth step); First cold rolling (fifth step); β-heat treatment step (sixth step); The second cold rolling step (seventh step) and the final heat treatment step (eighth step).

At this time, the process is the step of dissolving each metal having a composition of the present invention (first step); Β-heat treatment at a temperature of 1000 to 1100 ° C. for 10 to 30 minutes (second step); Preheating at 670-690 ° C. for 10-30 minutes and then hot rolling at a reduction ratio of 50-80% (third step); Heat treatment at 670-690 ° C. for 1-3 hours (fourth step); Primary cold rolling at a reduction ratio of 40 to 60% (fifth step); Β-heat treatment at a temperature of 1000 to 1100 ° C. for 10 to 30 minutes (sixth step); It is preferred to proceed to the second cold rolling step (seventh step) at a reduction ratio of 40 to 60% and the final heat treatment (step 8 step) at a temperature of 450 to 600 ° C. for 10 to 100 hours.

The zirconium alloy compositions prepared by the composition and method of the present invention are excellent in corrosion resistance and can be usefully used as support grids and other structures in nuclear reactor cladding as well as nuclear fuel cladding in the reactor core of a light water reactor and a heavy water reactor type nuclear power plant.

Hereinafter, the present invention will be described in detail by examples. However, the following examples are only for exemplifying the present invention, and the present invention is not limited by the examples.

Example 1 Preparation of Zirconium Alloy Compositions (1)

The alloy of the present invention was prepared in the following order to consist of 1.0% by weight of Nb, 0.2% by weight of Cu, 600-1400 ppm of oxygen and the balance Zr.

(1) Ingot manufacturing

Small ingots of 200 g were prepared by using a vacuum arc remelting (VAR) method. Five repeated dissolutions were performed to prevent impurities from segregating or unevenly distributed alloy composition. In order to prevent oxidation during dissolution, vacuum is sufficiently drawn up to 1x10 ^-4 torr and then Ar gas is injected into the chamber to apply 500 A of applied current at a weak negative pressure of 2x10 ^-1 torr, and the cooling water pressure is 1 kg _f / cm ² , Dissolution was carried out in a water-cooled copper crucible 60 mm in diameter for about 3.5 minutes. After dissolution, in order to suppress oxidation of the surface of the specimen while the specimen was cooled, the vacuum was maintained at 1 × 10 ⁻⁴ torr and then cooled by injecting Ar gas.

(2) β-heat treatment

β-heat treatment was performed to homogenize the alloy composition in the ingot by solution treatment in the β region. In order to prevent oxidation of the specimen, the specimen was coated with a 1 mm thick stainless steel plate and maintained at 1050 ° C. for 20 minutes, dropped into a bath containing 100 L water, and stirred with a rod to cool water. And sufficiently dried for 24 hours at 150 ℃ to remove the moisture remaining in the coating.

(3) hot rolling and heat treatment

Hot rolling was performed using the rolling mill of 100 ton scale. After preheating at 680 ° C. for 20 minutes, it was rolled in a pass rate of about 70% in one pass. After the hot rolling, the cladding was removed and the oxide film generated during β-heat treatment or hot rolling was removed using a pickling solution having a volume ratio of hydrofluoric acid: nitric acid: water = 5: 45: 50. In addition, the oxide film remaining locally after pickling was completely removed mechanically using an electric wire brush.

(4) cold rolling and heat treatment

In order to remove the residual stress after hot rolling and to prevent the specimen from being damaged during the first cold working, it was annealed at 680 ° C for 2 hours and then reduced by 50% by a 0.5 mm thickness reduction in a pass using a 70 ton rolling mill. First cold rolling was carried out. Then, β-heat treatment was performed at 1050 ° C. for 20 minutes as in Preparation Step 2 of FIG. 1, and the second cold rolling was performed at a reduction ratio of 30%. The last heat treatment was then performed at 480 ° C. for up to 500 hours.

Example 2. Preparation of Zirconium Alloy Compositions (2)

A zirconium alloy composition was manufactured using the same method as in Example 1, except that the final heat treatment was performed at a temperature of 580 ° C. for up to 500 hours.

Example 3 Preparation of Zirconium Alloy Compositions (3)

Zirconium using the same method as in Example 1, except that the composition of the alloy is composed of 1.0 wt% Nb, 1.0 wt% Sn, 0.3 wt% Fe, 0.1 wt% Cu, 600-1400 ppm oxygen and the balance Zr. An alloy composition was prepared.

Example 4 Preparation of Zirconium Alloy Compositions (4)

A zirconium alloy composition was manufactured using the same method as in Example 2, except that the final heat treatment was performed at a temperature of 580 ° C. for up to 500 hours.

Comparative Examples 1 and 2

As shown in the manufacturing process 1 of FIG. 1, after the first cold rolling, the intermediate recrystallization heat treatment was performed at 580 ° C. for 2 hours, and the second cold rolling was performed at a rolling reduction of 50%. The β-heat treatment was performed at 1050 ° C. for 20 minutes. It was. Then, a zirconium alloy composition was prepared by the same method as Example 1 and Example 2, except that the final heat treatment was performed at two temperatures of 480 ° C. (Comparative Example 1) and 580 ° C. (Comparative Example 2) for up to 500 hours. .

Experimental Example 1. Corrosion Test

The specimens of Examples 1 to 4 and Comparative Examples 1 and 2 after heat treatment were processed into 15 x 20 x 1 mm plates and polished up to 2000 times with SiC abrasive paper to evaluate corrosion characteristics, and were prepared at 400 ° C. water vapor (1500 psi). The corrosion test was carried out outside the furnace for 120 days. Corrosion resistance was evaluated by measuring the change in weight increase with exposure time.

The results of the corrosion test for 120 days in a steam atmosphere of 400 ℃ is shown in Figure 2, it can be seen that the effect of the manufacturing process and heat treatment has a very significant effect on the corrosion. According to Figure 2a shows that the manufacturing process of Examples 1 and 2 is more advantageous to the corrosion resistance than the manufacturing process of Comparative Examples 1 and 2, and also in the embodiment, the aging heat treatment at 480 ℃ (Example 1) is heat treated at 580 ℃ It showed better corrosion resistance than that (Example 2). It was observed that the Zr-1Nb-0.2Cu alloy has very different corrosion characteristics according to the heat treatment, and it can be seen that the proper manufacturing process and heat treatment can provide better corrosion resistance than commercial Zircaloy-4 tube specimens.

On the other hand, in the case of the Zr-1Nb-1Sn-0.3Fe-0.1Cu alloy of Examples 3 to 4, the specimen produced by the manufacturing process of the comparative example was very severely corroded and excluded during the experiment, Examples 3 to 4 in FIG. Only the results of specimens prepared by the manufacturing process are shown. The Zr-1Nb-1Sn-0.3Fe-0.1Cu alloys of Examples 3 to 4 were shown to be more advantageous in corrosion resistance than the manufacturing process of Comparative Example, as in the case of Zr-1Nb-0.2Cu alloy. Aging treatment at 480 ℃ also showed better corrosion resistance than heat treatment at 580 ℃. In addition, it was found that corrosion resistance can be improved by giving an appropriate aging heat treatment time.

Experimental Example 2. Hardness Test

Hardness measurements were performed using a Vickers microhardness meter to indirectly investigate the microstructure change due to heat treatment. Each specimen was measured 10 times with a load of 1 kg and the average of the remaining values except the maximum and minimum values was taken as the hardness value.

A microhardness test was performed to indirectly measure the formation of precipitates by heat treatment of the Zr-1Nb-0.2Cu alloy, and the results are shown in FIG. 3. At 480 ° C., the hardness of the sample was slightly increased after aging for 3 hours in the comparative example, whereas the hardness was increased in about 30 minutes in the case of the example, in which case the cold rolling was formed before the aging treatment. This is because it was promoted. The increase in hardness due to precipitates was not observed at 580 ° C because the increase in hardness due to precipitation was offset by the decrease in hardness due to the recovery process.

Experimental Example 3. Observation of Microstructure

The microstructure and precipitates were observed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM), and the relationship between the observed tissues and corrosion test results was analyzed. SEM specimens were prepared by etching gold with a solution of HF (10%) + HNO ₃ (45%) + H ₂ O (45%) and coating them with gold coating. Mechanically polished to 3 mm discs, then twin-jet at -40 ° C and 13 V using a mixed solution of ethanol (90%) and perchrolic acid (10%). -jet) to prepare a thin film specimen. The component of the precipitate was analyzed by EDS in TEM.

In the case of the Zr-1Nb-0.2Cu alloys of Examples 1 to 2, it can be seen in FIG. 4 that the formation of precipitates is promoted in Examples as described above, and the formation is more accelerated at 580 ° C. than at 480 ° C. . When the specimen prepared by the comparative method is heat-treated for 50 hours, the change of the structure according to the heat treatment temperature can be seen in Figure 5, the first β-heat treatment tissue is still present at 480 ℃, but twins disappeared at 580 ℃ plate It was observed that the plates changed into the shape of polygons.

In the Zr-1Nb-0.2Cu alloys of Examples 1 to 2, the shape of the precipitate formed during the heat treatment and the EDS analysis results are shown in FIGS. 6 to 8. When the specimen prepared in Example was heat-treated at 580 ° C. for 30 minutes, a precipitate containing only Cu and Zr without Nb was observed (FIG. 6), but when the heat treatment was further progressed, the precipitate contained Nb in 50 hours. Was observed (FIG. 7). However, the relative amount of Nb was higher in the precipitates at grain boundaries than in grains. On the other hand, when the specimen prepared in Example was heat-treated at 480 ° C. for 50 hours, two precipitates were observed. Only Cu and Zr were found without Nb in a relatively large precipitate, and Nb was found in a finer precipitate (FIG. 8). .

The shape of the precipitate formed during the heat treatment in the Zr-1Nb-1Sn-0.3Fe-0.1Cu alloys prepared in Examples 3 to 4 and the EDS analysis results are shown in FIGS. 9 and 10. As a result of EDS analysis, one precipitate of Zr-Nb-Fe type was observed in the specimen heat-treated at 580 ° C., and the size and Nb content of the precipitate increased as the heat treatment time progressed (FIG. 9). Meanwhile, even when the specimen prepared in Example was heat-treated at 480 ° C. for 50 hours, one precipitate in the form of Zr-Nb-Fe was observed, and its composition was similar to that at 50 ° C. for 50 hours (FIG. 10).

As a result of the experiment, it was found that the zirconium alloys of the present invention can obtain very excellent corrosion resistance by controlling the appropriate manufacturing method and heat treatment, and the alloy obtained by the manufacturing process of the example was superior in terms of corrosion resistance than the process conditions of the comparative example, The optimum aging heat treatment was performed for about 100 hours at 480 ℃. On the other hand, comparing the observations of the precipitates formed during the heat treatment and the corrosion test results, it was found that the corrosion resistance was more excellent in the case of finely distributed precipitates containing a large amount of Nb in both alloys.

As described above, it was found that the zirconium alloys of the new composition of the present invention can obtain very excellent corrosion resistance by controlling the appropriate manufacturing method and heat treatment. Therefore, the zirconium alloy compositions prepared by the manufacturing method of the present invention can be usefully used as nuclear fuel cladding, support grid and structure in the reactor core of the light and heavy water reactor type nuclear power plant.

Claims (6)

A zirconium alloy composition for a nuclear fuel cladding tube composed of 0.8-1.2 wt% niobium (Nb), 0.1-0.3 wt% copper (Cu), 600-1400 ppm oxygen (O), and balance zirconium (Zr). The zirconium alloy composition according to claim 1, comprising 1.0 wt% Nb, 0.2 wt% Cu, 600-1400 ppm oxygen and the balance Zr. Niobium (Nb) 0.8-1.2 wt%, Tin (Sn) 0.8-1.2 wt%, Iron (Fe) 0.2-0.4 wt%, Copper (Cu) 0.1-0.3 wt%, Oxygen (O) 600-1400 ppm and balance A zirconium alloy composition for a nuclear fuel cladding tube composed of zirconium (Zr). 4. The zirconium alloy composition for a nuclear fuel cladding according to claim 3, comprising 1.0 wt% of Nb, 1.0 wt% of Sn, 0.3 wt% of Fe, 0.1 wt% of Cu, 600-1400 ppm of oxygen, and the balance Zr. Dissolving the metal having the composition of claim 1 (first step); Β-heat treatment at a temperature of 1000 to 1100 ° C. for 10 to 30 minutes (second step); Preheating at 670-690 ° C. for 10-30 minutes and then hot rolling at a reduction ratio of 50-80% (third step); Heat treatment at 670-690 ° C. for 1-3 hours (fourth step); Primary cold rolling at a reduction ratio of 40 to 60% (fifth step); Β-heat treatment at a temperature of 1000 to 1100 ° C. for 10 to 30 minutes (sixth step); Zirconium for nuclear fuel cladding, characterized in that it comprises a second cold rolling step (seventh step) at a reduction ratio of 40 to 60% and a final heat treatment (step 8 step) at a temperature of 450 to 600 ° C. for 10 to 100 hours. Method for producing an alloy composition. Dissolving the metal having the composition of claim 3 (first step); Β-heat treatment at a temperature of 1000 to 1100 ° C. for 10 to 30 minutes (second step); Preheating at 670-690 ° C. for 10-30 minutes and then hot rolling at a reduction ratio of 50-80% (third step); Heat treatment at 670-690 ° C. for 1-3 hours (fourth step); Primary cold rolling at a reduction ratio of 40 to 60% (fifth step); Β-heat treatment at a temperature of 1000 to 1100 ° C. for 10 to 30 minutes (sixth step); Zirconium for nuclear fuel cladding, characterized in that it comprises a second cold rolling step (seventh step) at a reduction ratio of 40 to 60% and a final heat treatment (step 8 step) at a temperature of 450 to 600 ° C. for 10 to 100 hours. Method for producing an alloy composition.