JPS60242342A - Method for discriminating corrosion resistance of zirconium alloy - Google Patents

Abstract

PURPOSE:To discriminate the relative nodular corrosion resistance in the environment of a water cooling type nuclear reactor by exposing respectively a test piece for specific time to steam of a specific temp. and for another specific time to steam of another specific temp. in succession thereto and ascertaining the presence or absence of generation of corrosion after exposing. CONSTITUTION:The test piece is removed of stains, etc. by pickling and washing, etc. and after the weight thereof is measured, the test piece is heated to about 300-420 deg.C for about 5-15hr and is heated to about 510-550 deg.C; thereafter, the test piece is exposed to the steam for about 12-30hr in an autoclave in which about 70-105kg/cm<2> pressure is maintained. The test piece is taken out of the autoclave and is cooled down to the ambient conditions. The test piece subjected to the steam treatment is weighed and the weight increase is checked; at the same time, the presence or absence of the trace for the nodular corrosion on the surface is visually inspected.

Description

DETAILED DESCRIPTION OF THE INVENTION [Field of Application of the Invention] The present invention relates to a method for determining the corrosion resistance of zirconium alloys, and more particularly to an inspection method for determining the superiority or inferiority of relatively hard zirconium alloys in corrosion resistance.

[Background of the invention]

Zirconium alloys are widely used as core materials and structural materials for water-cooled nuclear reactors because they have properties that make them suitable for such applications, particularly a small neutron cross section.

In this regard, please refer to the Reactor Materials Book. Several zirconium alloy compositions have been developed and are commercially available, primarily for nuclear reactor applications. Typical examples of such luconium base metal compositions are commercially available materials called Zircaloy-2 and Zircaloy-4, which Zircaloys contain at least about 95 (by weight) zirconium and up to about 2.0 (by weight) zirconium. of tin, about 0.5 (
(weight) iron, up to approximately 0.5% (by weight) chromium,
and O to about 0.15% (by weight) nickel. In addition, as another zirconium alloy composition for nuclear reactors, a niobium-containing zirconium alloy containing at least about 95 (by weight) Zr and about 2 to 3 Nb has been put into practical use. Regarding the alloy components of the commercially available zirconium alloy compositions shown above, please refer to the descriptions of a, sTM-8350, nuclear reactor materials, etc.

However, the susceptibility of material i to corrosion is an important factor for its use or performance in water-cooled nuclear reactors. In the environment of a nuclear reactor, zirconium alloys typically develop a relatively harmless dark surface oxide that uniformly covers the surface of the zirconium alloy, providing protection for the underlying metal and protecting the atoms. It gradually thickens as the residence time in the furnace increases. However, zirconium alloys also suffer from fragrant nodular corrosion.
n) (sometimes nail-like corrosion (pu51ular corr)
Also called osion). ) may occur. Such nodular corrosion rapidly increases in area and depth on the alloy surface and can therefore, under certain conditions, impair the properties of the alloy. Nodular corrosion consists of white oxides that grow several times faster than harmless black oxides to produce thick white oxide layers, thereby reducing heat transfer rates and mechanical strength due to thinning. This may cause other problems.

The susceptibility of zirconium alloys to nodular corrosion when exposed to the environment of a water-cooled nuclear reactor is determined not only by environmental factors (temperature, pressure, etc.) in the operating reactor, but also by the alloy phase composition of the zirconium alloy and its microstructure. It has been found that it is highly dependent on the organization.

For the above reasons, many improvements have been made to the processing conditions, heat treatment conditions, etc. in the manufacturing process of zirconium alloys, but it has been found that the corrosion susceptibility of zirconium alloys often fluctuates in the current process. That is, incomplete or insufficient annealing, or metalworking or fabrication operations such as drawing, drawing, shaping, cutting, or welding, can lead to differences and non-uniformity in the microstructure within the alloy. may occur.

Therefore, large variations or uncertainties exist regarding the corrosion susceptibility of nuclear reactor components such as fuel cladding, channels, and spacers made from zirconium alloy compositions. Therefore, it is necessary to distinguish between those with poor corrosion resistance and those with good corrosion resistance, and in order to eliminate those with poor corrosion resistance from the standpoint of quality control, a high-temperature corrosion test at 400C, 72 hours, as described in ASTM-02, has been carried out. . Also, patent application publication 1982-9524
A high-temperature corrosion test at 490-520C as described in No. 7 has also been carried out.

On the other hand, it is necessary to further improve the corrosion resistance of conventional zirconium alloys in order to improve the economic efficiency of nuclear reactors, increase the burnup of reactor fuel, and extend the operating period.

In addition to the zirconium alloy materials that are normally manufactured using current processes, there are also known techniques for manufacturing highly corrosion-resistant zirconium alloy materials, including a special heat treatment process called β-quenching, and techniques for manufacturing highly corrosion-resistant zirconium alloy materials using low-temperature heat treatment. . Details of the method for manufacturing highly corrosion-resistant zirconium alloys are shown in Japanese Patent Application Laid-open No. 58-22364 and No. 58-22365. Also, β quench is described in Japanese Patent Application Laid-Open No. 51-11
0412, JP-A-52-70917, the temperature range in which the α phase and the β phase coexist or the temperature at which the β phase becomes a single phase [7. A heat treatment in which the intermetallic compound phase that was uniformly dispersed in the matrix is made into a solid solution in the matrix, and is selectively reprecipitated at the grain boundaries and subgrain boundaries by the cooling process after step 7 and annealing in the alpha phase affinity range. However, in the conventional high-temperature corrosion test described above, the highly corrosion-resistant zirconium alloy material manufactured by applying the above-mentioned special heat treatment, etc., and the ordinary zirconium alloy material manufactured by the current process that does not include special heat treatment, etc. Even among materials, clear test results cannot be obtained compared to alloy materials, which have relatively excellent corrosion resistance. Furthermore, it is not possible to evaluate the variation in corrosion resistance of alloy materials that have been subjected to special heat treatment, and therefore quality control of highly corrosion resistant alloy members becomes a chore.

[Purpose of the invention]

The purpose of the present invention is to determine the relative nodular corrosion resistance of zirconium alloys in the environment of a water-cooled nuclear reactor, and to determine the high corrosion resistance of zirconium alloys with respect to conventional products and the manufacturing history of high corrosion resistance. An object of the present invention is to provide a method for evaluating variations in corrosion resistance.

C. Summary of the Invention] The present invention provides that the nodular corrosion that occurs in zirconium alloys is accelerated by increasing the ambient temperature, and that under certain conditions, the susceptibility to nodular corrosion decreases after the test due to the increase in corrosion mass and the external appearance of the zirconium alloy. It is based on the fact that it is prominently seen in the upper cisterna. The details will be described below.

Figure 1 shows the relationship between test temperature and corrosion weight increase. This figure shows various test pieces estimated to have different corrosion resistance.
5 for 8 hours under 4/cd conditions (first step)
1 under the same pressure and temperature condition of 480C to 550C.
・This shows the correlation between the test temperature and corrosion weight increase of various test pieces obtained after 6 hours of exposure (second step). However, the data at the test temperature of 400C in the figure is for the AsT
This is based on the M corrosion test method.

As can be seen in this figure, when the test temperature in the second step becomes higher than approximately 520℃, it is not only possible to determine the presence or absence of nodular corrosion in the test material, which is a zirconium alloy, but also to distinguish between highly corrosion-resistant and ordinary materials. It is possible to determine irregularities in the corrosion resistance of highly resistant materials due to manufacturing history.

Furthermore, the test time is preferably within a range that can simulate the long-term operation of an actual reactor.

Regarding this matter, Fig. 2 shows the relationship between the time in the furnace and the test time when the corrosion weight increase is the same. As can be seen here, assuming that the residence time in the furnace is 2000 days (approximately 5 years), the test temperature and test time are preferably approximately 530C and 10 hours or more.

Incidentally, from the viewpoint of accelerated testing, it is desirable that the test temperature be high, but the corrosion characteristics of zirconium alloy materials exposed to high temperature atmosphere change. Figure 3 is 400
? These are the results of a corrosion test of a zirconium alloy material (Zircaloy-2 material) annealed at r~700C in steam at 500C for 124 hours. As can be seen from this figure, it is shown that the corrosion characteristics of the zirconium alloy material change when the annealing temperature is set to 550C or higher. Therefore, the temperature range for the corrosion test is up to 550C.

[Embodiments of the invention]

Hereinafter, preferred embodiments of the present invention will be described.

When applying the present invention, test pieces of zirconium alloy (
After removing all dirt and foreign matter by washing the sample (or a suitable specimen), its weight is accurately determined. The above cleaning may be carried out by the usual procedure consisting of immersion in an acid bath or "pickling" followed by washing with water.

The specimen is then exposed to water vapor in an autoclave under pressure in the range of about 70-105 Kg/CJ. Such water vapor is initially about 300 to 4'20
The temperature is increased to a temperature level of tZ' and maintained at that temperature level for at least about 5 hours, and then increased to a level of about 520 to about 550 C and maintained at that temperature level for at least about 12 hours. More specifically, the time to achieve effective water vapor exposure is about 5 to about 15 hours after heating for a first temperature level of about 300 to about 4200, and about 510 to about 550 tZ'. For the second temperature level of approximately 1 after heating
2 to about 30 hours.

According to one preferred embodiment of the invention, the test specimen is first
Exposure to water vapor at about 10r for about 8 to 10 hours, then exposure to water vapor at about 530C for about 16 hours.
Exposure for ~24 hours.

After removal from the autoclave and cooling to ambient conditions, the steam treated specimens are weighed and checked for weight gain. Such specimens are also visually inspected for evidence of nodular corrosion on their surfaces.

Next, a preferred embodiment of the present invention will be described in detail.

The following three types of zirconium alloys were used in the corrosion test.

(1) Highly corrosion-resistant zirconium alloy material: A highly corrosion-resistant fuel cladding tube manufactured by adding an α+β quenching process to the reactor fuel cladding manufacturing process. (Several types of cladding tubes with different quenching process insertion times) (2) Normal zirconium alloy material: Normal nuclear reactor fuel cladding tube.

(3) Zirconium alloy material with poor corrosion resistance: Normal reactor fuel cladding tube is annealed at high temperature to deteriorate its corrosion resistance.

The details of the manufacturing methods of (1) and (2) above are based on the above-mentioned Japanese Patent Application Laid-Open No. 58-22364.

Test pieces were cut from the above three types of materials, deburred, and then washed. If surface oxides are present, it is necessary to remove them using sandpaper. For cleaning, for example, 5.0 (by volume) of concentrated hydrofluoric acid (E (F), 45 (by volume) of concentrated nitric acid (HNO3)
Etching is performed in an acid solution consisting of water and residual water.

After etching, the specimen was washed with water, dried, and
Weighing was performed with an accuracy of mg.

The test piece thus obtained was suspended in an autoclave,
Water vapor is introduced and the test system is heated to about 410 Cs.
Equilibrated to conditions of 105 Ky/cti. In the first stage, such temperature-pressure equilibrium was maintained for about 8 hours. Thereafter, the temperature was further increased to equilibrate the second stage, that is, the test system, to conditions of 53° r and 105 Kt/i. In the second stage, such equilibrium was maintained for about 16 hours.

After completion of the steam treatment at the two temperature level steps, the autoclave was returned to atmospheric conditions, the specimens were removed, dried, and weighed and visually inspected. Thereby, the presence or absence of an increase in the weight of the test piece and the absence of nodular corrosion were confirmed.

As a result, compared to the corrosion increase of 1.0 in the A8TM corrosion test, the highly corrosion resistant zirconium alloy material showed a corrosion increase of 2.0 to 7.0, the normal zirconium alloy material showed a corrosion increase of 10 to 50, and the poor corrosion resistance zirconium alloy The material showed further increase in corrosion. In addition, the state of occurrence of nodular corrosion was clearly different among the materials of the above 3s, and the three materials with different corrosion resistance could be clearly distinguished. Furthermore, it was possible to identify differences in the corrosion resistance of highly corrosion-resistant zirconium alloy materials due to differences in manufacturing methods.

If the corrosion test temperature in the second stage is 500iC to 520C, the relative value of the corrosion increase for the normal zirconium alloy material and the relative value for the highly corrosion-resistant zirconium alloy material will partially overlap9, and the two can be clearly distinguished. I can't.

As shown in the above example, the second stage corrosion test temperature was set to 5
It was confirmed that the corrosion resistance of normal zirconium alloy materials and highly corrosion-resistant zirconium alloy materials can be clearly distinguished by setting the temperature to 30C. It can be said that it is possible.

〔Effect of the invention〕

According to the present invention, it is possible not only to determine the relative corrosion resistance of zirconium alloys, but also to determine whether the materials are highly corrosion resistant or ordinary materials. The long-term stay inside the reactor can be simulated through short-term tests.

[Brief explanation of the drawing]

Figure 1 is a diagram showing the relationship between test temperature and corrosion weight increase, Figure 2 is a diagram showing the relationship between residence time in the furnace and test time at which the corrosion weight gain is the same, and Figure 3
The figure shows the relationship between annealing temperature and corrosion increase. Agent Patent Attorney Akio Takahashi 'f11 Diagram'g2 [m 9 Hours of Provisions (tl) Continuation of page 1 @ Inventor Jun Nakataka 2nd Department Hitachi Wheels 3-chome @ Inventor Iwao Takase 3-chome Saiwaimachi, Hitachi City Inside the office

Claims (1)

Claims: 1. (a) exposing a zirconium alloy specimen to an atmosphere of water vapor having a temperature of about 300 to about 420C for at least about 5 hours;
The test piece is exposed to a water vapor atmosphere having a temperature of at least about 550C or less at a temperature of at least about
A method for determining the corrosion resistance of a zirconium alloy, comprising the steps of: (C) confirming the presence or absence of corrosion on the test piece after exposure to water vapor. 2. A method for determining corrosion resistance of a zirconium alloy according to claim 1, wherein the test piece is exposed to the steam atmosphere of step (a) for about 5 to about 15 hours. 3. The test piece is exposed to the steam atmosphere of step (b) for about 10 to about 30 hours. lf A method for determining the corrosion resistance of a zirconium alloy according to claim 1. 4, about 1000 to about 15001) s ig (70,3
~105Kti/crl)
A method for determining the corrosion resistance of a zirconium alloy according to claim 1, 2, or 3, characterized in that the method is performed in a furnace. 5. The method for determining the corrosion resistance of a zirconium alloy according to claim 1, 2, or 3, wherein the steam atmosphere in the autocupboard furnace is circulating or stagnation.