JP4761803B2 - Corrosion acceleration test method for components made of zirconium alloys for boiling water reactors - Google Patents

Description

  The present invention relates to an accelerated test method effective for evaluating the corrosiveness of a constituent material made of a zirconium (Zr) alloy used in a light water reactor, particularly a boiling water light water reactor.

  Light water reactors have come to be widely used as nuclear power reactors. These light water reactors include pressurized water light water reactors (PWR) and boiling water light water reactors (BWR), of which the majority occupies in Japan. The latter is a boiling water reactor (BWR). As is well known, a zirconium alloy is adopted as a constituent element of the fuel cladding tube of this light water reactor.

  In recent years, in order to reduce power generation costs, reduce the amount of radioactive waste generated, and effectively use uranium resources, these light water reactors have been increased in burnup.

  Conventionally, in a boiling water type light water reactor (BWR), the corrosion of the cladding tube does not proceed rapidly as in the case of a pressurized water type light water reactor (PWR), and even if the burnup increases (the residence time in the furnace), the oxide film The increase in thickness is slight, and until now it has been considered that there is no particular problem with the corrosion of Zircaloy-2 cladding. However, when the hydrogen concentration of the fuel cladding tube irradiated for 4 cycles and 5 cycles was analyzed, it was found that the hydrogen concentration was greatly increased compared to the cladding tube irradiated for 1 to 3 cycles. For this reason, in BWR, the hydrogen absorption amount of the fuel cladding tube rapidly increases on the high burnup side, and there is a concern about hydrogen embrittlement that degrades the mechanical properties of the cladding tube. Development of materials with a small amount is desired.

  In general, in the development of materials for fuel cladding tubes, materials are screened by a high temperature water corrosion test at 300 ° C. to 360 ° C. and a high temperature steam corrosion test at 400 ° C. to 530 ° C. using an autoclave. However, at a test temperature of 300 to 400 ° C., a long-term corrosion test of several thousand to several tens of thousands of hours is required until an evaluation result is obtained.

  BWR has reported two types of in-furnace corrosion behaviors called local corrosion and uniform corrosion called nodular corrosion. For local corrosion behavior called nodular corrosion, corrosion resistance is evaluated by a relatively high temperature steam corrosion test at 475 ° C to 530 ° C. On the other hand, the corrosion resistance is evaluated by a high temperature steam test at 400 ° C. for uniform corrosion behavior. Further, a two-step corrosion test in which the temperature is once increased to 500 to 525 ° C. after being held at 400 to 425 ° C. for several hours is also performed.

  For local corrosion behavior, it is said that the evaluation based on the out-of-furnace corrosion test and in-furnace corrosion behavior are in good agreement, but for uniform corrosion behavior, Corrosion behavior in the out-of-core accelerated corrosion test does not always match, and the in-core accelerated corrosion test method cannot be fully evaluated by the currently implemented out-of-core accelerated corrosion test method, which is a major obstacle in developing materials. It has become.

  In general, the uniform corrosion characteristics of zirconium alloys are evaluated under high-temperature and high-pressure steam conditions of 400 ° C. and 10.3 MPa as specified in JIS (JIS H 4751). In JIS, test conditions for evaluating the sample appearance and corrosion weight increase are maintained for 72 hours (3 days) or 336 hours (14 days) in water vapor at 400 ° C. Generally, several thousand hours are used. The actual situation is that the long-time corrosion test described above has been conducted and evaluated, and it takes a very long time to evaluate the corrosion resistance, which is an obstacle to the progress of material development.

  For example, the report by Eto et al. (See Non-Patent Document 1) shows the data on the increase in corrosion after a water vapor corrosion test at 400 ° C and two cycles of irradiation with BWR, as shown in FIG. 4, FIG. 3 and TABLE 1. However, the test result at 400 ° C. shows that the increase in corrosion of the sample to which 0.2 wt% of Nb is added is equal to or less than that of the additive-free material. In other words, the 400 ° C. steam corrosion test shows that the corrosion resistance is equivalent or better at an Nb addition amount of about 0.2 wt%.

However, as shown in the same FIG.4, the corrosion data in the furnace shows a tendency for the corrosion amount to increase sharply as the amount of Nb added increases, and in the actual operating environment at BWR, Nb has an effect of improving corrosion resistance. It is not allowed. Regarding the influence of Nb, these contradictory facts revealed by Figure 9 and Table 2 in Fukuzawa et al.'S report (Non-patent Document 2) investigating the corrosion situation in the actual furnace were also found in Fe. In other words, in the accelerated corrosion test outside the furnace, the tendency of the corrosion rate of the zirconium alloy to increase (corrosion resistance deteriorates) was reported as the amount of Fe added increased, but the report by Eto et al. According to Reference 1: TABLE 2, FIG. 6), it has been confirmed that the corrosion resistance of zirconium alloys improves with increasing Fe concentration in an actual BWR environment.

  As described above, the in-reactor corrosion behavior in the BWR environment cannot be evaluated by the accelerated external corrosion test under the 400 ° C / steam environment, which is specified by JIS and generally performed. For this reason, development of an out-of-core accelerated corrosion test method capable of simulating BWR in-core corrosion characteristics is strongly desired.

  As a method corresponding to such a request, an accelerated test method has been proposed in which a test material is placed in supercritical water and corroded while being irradiated with radiation (see Patent Document 1).

However, since this method needs to be irradiated with radiation, there is a problem in safety management as a test method outside the furnace, and it becomes a large-scale test apparatus and is disadvantageous in terms of cost. I couldn't.
[Problems to be solved by the invention]
ASTM STP 1294, pp.825-849 (1996) ANS International Topical Meeting, pp.240-249 (1997) Japanese Patent Laid-Open No. 11-352277

  Therefore, the present inventors solved the above-mentioned conventional problems, and easily reproduced the corrosion status of the material when used for a long time in a BWR environment, outside the furnace and in a relatively short period of time. As a result of diligent research and research to develop an accelerated test method that can be evaluated, by using a supercritical water environment and controlling the dissolved oxygen concentration (Do) of the water used for the test to an appropriate range The inventors have found out the fact that the corrosion in the furnace can be accurately simulated, and have completed the present invention.

Accordingly, an object (issue) of the present invention is to compare the uniform corrosion characteristics in a reactor of a constituent member or element made of a zirconium alloy such as a fuel cladding tube used in a nuclear reactor, particularly in a BWR, without using radiation. It is to provide a method for evaluating in a short time outside the furnace by a simple and safe means.

The present invention was made to solve the above-mentioned problems in a test method for performing uniform corrosion test of a zirconium alloy constituent material used in a boiling water light water reactor without using radiation. Zirconium for boiling water type light water reactors, characterized by immersing a test material of an alloy constituent material in supercritical water whose dissolved oxygen concentration is adjusted to 0.1 ppb or more and less than 10 ppb for a predetermined period to evaluate its uniform corrosiveness This is a method for accelerating corrosive testing of alloy constituent materials.

According to the present invention, the evaluation of the uniform corrosiveness of a zirconium alloy constituent material used in a boiling water light water reactor is performed with a relatively simple and safe method, with the same accuracy as in an actual use environment in a boiling water light water reactor. In addition, it provides an excellent accelerated corrosion resistance test method that can be carried out in a short period of time, and contributes greatly to the development and commercialization of major components such as cladding tubes in boiling water reactors. It can be said that the invention has great technical and industrial value.

  Hereinafter, the content of the corrosion evaluation method of the present invention, that is, the accelerated corrosion test method for constituent materials made of zirconium alloy will be described in detail.

  First, in this method, the same autoclave (a kind of high-pressure kettle) as that used in the uniform corrosion evaluation of zirconium alloy as defined in the above JIS (JIS H 4751) is used as a test apparatus. There are two types of autoclaves: the batch type and the loop type. In the case of the loop type, the adjustment and control of the water quality of the test water is simple and it is possible to bring it closer to the reactor water environment compared to the batch type. Although advantageous, any of them may be employed in carrying out the method.

  In this method, using an autoclave, the temperature of the test water is about 374 ° C. or higher and the test pressure is about 218 atm or higher, and the oxygen concentration (Do) in the test water is 0.1 ppb or more and 10 ppb. The corrosion resistance is evaluated in a state adjusted to less than, preferably 4.5 ppb or more and less than 8.5 ppb. If this concentration exceeds 10 ppb, the actual environment of BWR cannot be simulated or reproduced, the accuracy of the corrosive test and evaluation becomes insufficient, and it is not practical to make it lower than 0.1 ppb. The adjustment and control of the dissolved oxygen concentration (Do) is extremely important in the present invention.

  The method of adjusting the dissolved oxygen concentration (Do) is not particularly limited. For example, when a batch-type autoclave is adopted, a test material suspended in advance from a kettle lid is placed in a kettle containing test water. After setting and closing the lid (closed loop), the kettle is heated with a heater, the temperature is raised, the temperature is raised to 110-150 ° C. or higher, the valve provided in the kettle is opened, and the test water is boiled. At the same time as the air inside is degassed, dissolved oxygen in the test water is removed, and the dissolved oxygen concentration (Do) is reduced to less than 10 ppb. On the other hand, in the case of a loop type autoclave, the dissolved oxygen is removed by performing Ar bubbling or the like in advance in another test water tank, and water whose dissolved oxygen concentration (Do) is adjusted to less than 10 ppb is connected to the autoclave loop. And circulate in the hook.

  This corrosion test with supercritical test water adjusted to less than 10 ppb can sufficiently simulate the operating environment of BWR in a relatively short period of 2000 to 3000 hours.

  Next, examples are given to clarify the excellent effects of the present invention.

Example 1
Test water with a dissolved oxygen concentration adjusted to 8 ppb for high Fe-Zry-2 as a test material (test material) and high Fe-Zry-2 containing 0.18 wt% Nb by batch autoclave The corrosion test was conducted under supercritical water conditions at 400 ° C. and 250 atg. The test results are shown in Table 1 and FIG. FIG. 1 shows the case where the test time is 4200 hours.

  For comparison with the in-furnace data, the high-Fe-Zry-2 1-cycle and 2-cycle irradiation data shown in Figure 9 of Non-Patent Document 2 and Fig. 4 of Non-Patent Document 1 are used. The corrosion data of the Nb-added Zry-2 shown is also shown in FIG.

 As is clear from Table 1 and FIG. 1, according to this test method, the corrosion increase tends to increase in proportion to the increase in the Nb concentration regardless of whether the test time is 3000 hours or 4200 hours. It can be seen that this is the same trend as in-furnace data, that is, this test method produces results consistent with in-furnace corrosion behavior.

(Example 2)
Corrosion test under supercritical water conditions at 400 ° C and 250 atg. Using test water adjusted for dissolved oxygen concentration of 5 ppb for Zry-2 with three different Fe concentration levels as a test material using a loop autoclave Carried out. The test results are shown in Table 2 and FIG. FIG. 2 shows a case where the test time is 4200 hours. In addition, in order to compare with the in-furnace data, the irradiation data shown in FIG. 6 and TABLE 2 of Non-Patent Document 1 described above (data of irradiation material in which Fe concentration is changed with Zry-2 composition: 2 The cycle irradiation and the 4-cycle irradiation) are also shown in FIG.

(Conventional example)
Using a batch type autoclave, a steam corrosion test at 400 ° C. and 105 atg. Was conducted on Zry-2 in which three types of Fe concentration levels similar to those in Example 2 were changed as test materials. The results are shown in Table 3 and are shown together in FIG.

 As is apparent from Table 2 and FIG. 2, according to this test method (Example 2), the increase in corrosion at both the test time 3000 hours and 4200 hours tends to decrease as the Fe concentration increases. This is consistent with the trend of irradiation data in the furnace, and it turns out that the same result as the behavior in the furnace is obtained.

  On the other hand, in the results of the steam corrosion test of the conventional example, from Table 3 and FIG. 2, the increase in corrosion tends to increase with the increase in Fe concentration at any test time, which is the reverse trend to the data in the furnace. It can be seen that the behavior in the furnace is not correctly reflected.

The graph which shows the relationship between the Nb density | concentration of a zirconium alloy by the test method concerning Example 1 of this invention, and corrosion increase. The figure also shows the in-furnace data for comparison. Here, the ■ broken line indicates the present invention, and the ○, □, and ● broken lines indicate the in-furnace data, respectively. The graph which shows the relationship between the Fe density | concentration of a zirconium alloy by the test method concerning Example 2 of this invention, and corrosion increase. For comparison, the figure also shows the data by the conventional test method and the in-furnace data. Here, the □ broken line indicates the present invention, the ○ broken line indicates the conventional example, and the ● and ■ broken lines indicate the in-furnace data.

Claims (1)

In a test method for performing a uniform corrosion test of a zirconium alloy constituent material used in a boiling water light water reactor without using radiation , the dissolved oxygen concentration of the test material of the zirconium alloy constituent material is 0.1 ppb or more. A corrosive acceleration test method for a constituent material made of a zirconium alloy for boiling water reactors, wherein the uniform corrosiveness is evaluated by immersing in supercritical water adjusted to less than 10 ppb for a predetermined period.

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