JP2007031803A - Highly corrosion-resistant reactor core material - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide a highly corrosion-resistant reactor core material having excellent corrosion resistance, swelling resistance and stress corrosion cracking resistance under radiation exposure and under a high-temperature and high-pressure water environment. <P>SOLUTION: As a member subjected to neutron irradiation in high-temperature water, austenitic steel containing at least either of Ti and Zr is used. To be concrete, the steel has a composition consisting of 0.005 to 0.04% C, ≤1% Si, ≤2% Mn, 23 to 28% Cr, 16 to 25% Ni, 0.5 to 3% Mo, 0.1 to 2% of at least either of Ti and Zr and the balance ≥50% Fe and further containing, other than the above, 1.0%, in total, of at least one or more elements among Pd, Pt and Hf. The steel is used as a constituent material for a reactor core of a nuclear reactor. <P>COPYRIGHT: (C)2007,JPO&INPIT

Description

  The present invention relates to an austenitic steel, and more particularly to an austenitic steel excellent in general corrosion resistance, swelling resistance and intergranular stress corrosion cracking resistance.

Currently, JIS standard 304 or 316L stainless steel is used as the structural member for core structures and equipment of light water reactors, and these are used to improve corrosion resistance and intergranular stress corrosion cracking resistance in the environment of use. Measures are being taken. For example, as a method for reducing or preventing intergranular corrosion and intergranular stress corrosion cracking in steel used in high-temperature water, carbon contained in stainless steel as disclosed in Japanese Patent Publication No. 1-18143. Some of them reduce the amount and add a carbide stabilizing element such as Nb or Ti to suppress or prevent the Cr depletion phenomenon near the grain boundary. Japanese Patent Laid-Open No. 62-107047 discloses a method for preventing intergranular corrosion caused by neutron irradiation by adding elements that reduce Si and P and generate stable carbides such as Mo, Nb, and Ti. ,
It is disclosed to suppress the formation of Cr carbide. Further, JP-A-3-72054 discloses that the component arrangement is adjusted so that the ratio of the atomic volume of Cr to the average atomic volume of the parent phase is 0.900 to 1.030 in stainless steel, and irradiation-induced grain boundary Cr deficiency It is disclosed to suppress the generation mechanism of the layer. These technologies were developed for the purpose of suppressing the formation of Cr-depleted layers near grain boundaries and improving the intergranular corrosion resistance in high-temperature and high-pressure water, and are especially considered to be applied to stainless steel for reactor core structural parts. Yes.

  The technologies described above are for structural members applied to the core structure and equipment of the current light water reactor, and these are used in high-temperature and high-pressure water (subcritical water) at a maximum of 360 ° C at the cooling water temperature and a maximum of 16 MPa at the pressure. Is only assumed.

Japanese Patent Publication No. 1-18143 JP 62-107047 A JP-A-3-72054

On the other hand, in recent years, for the purpose of increasing the efficiency of the current light water reactor, cooling water is being further increased in temperature and pressure, and in some cases, conditions exceeding the critical point of water (374 ° C, 22.1 MPa) (super Development of nuclear reactors that use critical water as cooling water is underway. Cooling water in the reactor core is designed to be about 290 ° C at the inlet, about 550 ° C at the outlet, and a pressure of 25MPa or higher. Both the temperature and pressure are higher than the core conditions of the current light water reactor, and the cooling of the core region Supercritical water is used as water. The members constituting the fuel cladding tube and the core structure used in such a core environment are required to have excellent overall corrosion resistance, swelling resistance and intergranular stress corrosion cracking resistance under irradiation of radiation. The above prior art focuses only on the improvement of various characteristics for core constituent materials that are in contact with high-temperature and high-pressure water under radiation irradiation up to about 350 ° C. and 16 MPa, which is the operating temperature and pressure of the current light water reactor. No consideration has been given to increasing the temperature and pressure of cooling water for the purpose of efficiency. In addition, zirconium alloys used as fuel cladding tubes in current light water reactors are difficult to apply because the mechanical strength is extremely low in the region exceeding 400 ° C. Furthermore, the material developed assuming that it is used for the fuel cladding tube in such an environment is shown in the Annual Meeting of the Atomic Energy Society of Japan 2003 (Summary No. 746). The focus is only on improving the corrosion resistance in water, and there is no consideration for improving the swelling resistance. In view of the above, in the prior art, for the purpose of improving the efficiency of the light water reactor, the material used for the fuel cladding tube and the internal structural component material of the high efficiency light water reactor in which the cooling water is in the supercritical pressure water condition However, there is a problem in providing a material excellent in all of corrosion resistance, stress corrosion cracking resistance, and swelling resistance.

  An object of the present invention is to provide an austenitic steel that is particularly excellent in overall corrosion resistance, swelling resistance, and intergranular stress corrosion cracking resistance with respect to constituent materials of fuel cladding tubes and internal structures of high-efficiency reactors. is there.

  The highly corrosion-resistant core material of the present invention is a member that comes into contact with water and is used under radiation irradiation, and is composed of austenitic steel containing at least one of Ti and Zr. Corrosion resistance, swelling resistance, and stress corrosion cracking resistance It is characterized by excellent properties.

Further, the high corrosion resistance core material of the present invention is, by weight, C: 0.005% to 0.04%, Si: 1% or less, P: 0.02% or less, N: 0.05% or less, Mn: 0.5-16%, Cr: 20-28%, Ni: 16-25%, Mo: 0.5-3%, Zr: 0.1-2% or Ti:
It has at least one kind of 0.1 to 2%, 50% or more of Fe and inevitable impurities, and is excellent in corrosion resistance, swelling resistance and stress corrosion cracking resistance.

  Further, the high corrosion resistance core material of the present invention is characterized by containing at least one kind of Pd, Pt and Hf in the above-mentioned steel in a total of 1.0% or less in total, and having corrosion resistance, swelling resistance and stress corrosion cracking resistance. It is characterized by superiority.

In addition, the high corrosion resistance core material of the present invention has a Cr content of 20 to 28 so that the metal thinning rate does not exceed 0.05 mm / year when immersed in high-temperature high-pressure water at 400 ° C. or higher and 25 MPa in the aforementioned steel. %, Ni content is added in an amount of 16 to 25%, and it is particularly preferable that the thinning rate is 0.03 mm / year, and the Cr content is 23 to 28% so as to have an austenite structure.
It is characterized by adding 18 to 23% of the amount of Ni.

  Further, the high corrosion resistance core material of the present invention has a Zr or Ti content of 0.1 to 2% so that the amount of swelling is within 1% after irradiation to a temperature irradiation damage of 10 dpa at 300 ° C. or higher in the above steel. It is characterized by adding.

  Further, in the high corrosion resistance core material of the present invention, the amount of swelling in the steel described above is preferably not more than 0.5%, and in consideration of the influence on the mechanical properties, the addition amount of Zr and Ti is It is preferable to set it as 0.5 to 1%, respectively. When used as a solution heat treatment, the C content is preferably 0.02% or less so as not to form carbides of Zr and Ti. In addition, after adding Zr or Ti amount of 1 to 2%, heat treatment may be performed at a temperature of 650 ° C. to 750 ° C. to form these carbides in the parent phase. When this heat treatment is performed, the C content is preferably 0.02 to 0.03% for the purpose of promoting the precipitation of carbides.

  In addition, the high corrosion resistance core material of the present invention is characterized in that the above-mentioned steel is processed into the shape of a double U-bend test piece, and no intergranular cracking occurs when immersed in high-temperature high-pressure water at 300 ° C. or higher and 25 MPa. It is what.

Further, in the high corrosion resistance core material of the present invention, after the above steel is irradiated at a temperature of 300 ° C. or more to a temperature irradiation damage of 10 dpa, the Cr deficiency at the grain boundary is within 5% as compared with the Cr concentration of the parent phase. Thus, 0.1 to 2% of Zr or Ti amount is added. In particular, it is preferable that the Cr deficiency at the grain boundary does not exceed 2%, and considering the influence on the mechanical properties, the addition amounts of Zr and Ti are each preferably 0.5 to 1%. When used as a solution heat treatment, the C content is preferably 0.02% or less so as not to form carbides of Zr and Ti. In addition, after adding 1-2% of Zr or Ti amount,
These carbides may be formed in the matrix by heat treatment at a temperature of 750 ° C. When this heat treatment is performed, the C content is preferably 0.02 to 0.04% for the purpose of promoting the precipitation of carbides.

  In addition, the high corrosion resistance core material of the present invention is subjected to forging, hot rolling, and solution heat treatment in the above steel, and then heat-treated at 650 ° C. to 750 ° C. so that either Zr or Ti is added to the parent phase. It is characterized by forming a carbide containing.

  According to the present invention, an austenitic steel used in an environment exposed to high-temperature and high-pressure water under irradiation is expected to exhibit excellent performance particularly in terms of overall corrosion resistance, swelling resistance, and intergranular stress corrosion cracking resistance. For example, it is possible to provide a highly corrosion-resistant core material that greatly contributes to the practical application of a high-efficiency light water reactor.

  Hereinafter, the steel of the present invention will be described focusing on the components.

  Zr and Ti promote recombination of point defects (interstitial atoms and atomic vacancies) caused by radiation irradiation when added to the austenitic steel according to the present invention. As a result, it is an element that suppresses the formation of dislocation loops and voids caused by accumulation of point defects, and also suppresses irradiation-induced grain boundary segregation caused by diffusion of point defects to grain boundaries. This action can suppress the generation of grain boundary Cr deficiency and voids that occur in austenitic steel under irradiation. However, addition of a large amount promotes the formation of brittle intermetallic compounds, which may lead to deterioration of workability and mechanical strength. Therefore, excessive addition of Zr and Ti should be avoided, and the range in which there is no such problem is set to 0.1 to 2%. 0.5 to 1% is particularly preferable.

C reacts with Cr by irradiation and heating to precipitate Cr carbide near the grain boundary, and lowers the Cr concentration near the grain boundary. The formation of a grain boundary Cr-depleted layer with a reduced Cr concentration lowers the corrosion resistance of the grain boundary and causes grain boundary type stress corrosion cracking, so the C content is preferably as low as possible. On the other hand, C is beneficial for increasing the mechanical strength at room temperature and high temperature, considering that the C content of 0.002% or less is not preferable because the strength is lowered and the irradiation brittleness is deteriorated.
It should be 0.005% or more and 0.04% or less.

  In relation to the above Zr, Ti and C amounts, the following two points need to be considered. First, when the steel of the present invention is used after solid solution heat treatment, Zr and Ti are used as a solid solution in the matrix phase, in order to suppress the precipitation of Zr and Ti carbides as much as possible, the C content is 0.005% to 0.02. %, And 0.008 to 0.02% is particularly preferable. Next, the Zr or Ti carbide uniformly dispersed in the matrix phase promotes recombination between point defects more than the solid solution Zr and Ti. After adding ˜2%, heat treatment may be performed at a temperature of 650 ° C. to 750 ° C. to disperse these carbides in the parent phase. When this heat treatment is performed, the C content is preferably 0.02 to 0.04%, particularly preferably 0.025 to 0.035% for the purpose of promoting the precipitation of carbides.

Cr works to improve corrosion resistance by forming a strong oxide film in high-temperature water,
In order to maintain the corrosion resistance in high-temperature and high-pressure water at 300 ° C. or higher and 25 MPa, less than 20% is insufficient. On the other hand, if it exceeds 28%, the σ phase is easily formed and the mechanical properties deteriorate, so it should be 20% to 28%. Further, in the heat treatment for forming 650 ° C. to 750 ° C., a part of Cr in the vicinity of the grain boundary contributes to carbide formation, and there is a possibility that the corrosion resistance in the vicinity of the grain boundary may be impaired. -28% is preferable, and 23% -26% is particularly preferable.

In the steel of the present invention, it is desirable that the alloy used under irradiation is stable in the austenite phase from the viewpoint of irradiation embrittlement, and Ni and Mn are each added to obtain an austenitic structure in the alloy. To do. In particular, Ni should be at least 16%, and Mn should be added at 2% or less. In order to form a stable austenite layer based on the Cr equivalent and Ni equivalent of the steel of the present invention and referring to the Schaeffler diagram, etc., it is particularly necessary to add 16% or more of Ni. On the other hand, excessive addition of Ni and Mn is not preferable because it induces strength reduction and precipitation of an embrittlement phase. Therefore, Ni: 16 to 25% and Mn 0.5 to 2% are preferable as addition amounts satisfying the effect of stabilizing the austenite structure of the base material described above, and Ni is particularly preferably 20 to 25%.

  Fe is an element which is the base of this alloy, and is preferably 50 to 70%, more preferably 50 to 65%, and more preferably 52 to 60% based on the past use record of the core material.

  Mo is preferably added in an amount of 3% or less from the viewpoint of improving the corrosion resistance of the matrix in high-temperature water. However, if it exceeds 3%, the formation of the σ phase is promoted to significantly impair the mechanical properties or accumulate at the grain boundary, so 1.0 to 2.0% is preferable.

  Hf is effective for corrosion resistance and irradiation brittleness under neutron irradiation. In particular, fine carbides can be formed to prevent precipitation of Cr carbides and to prevent a decrease in Cr concentration near the grain boundary. However, in consideration of the solid solubility limit of the element of Hf in the Fe—Ni—Cr alloy, it is preferable to add 1.0% or less as an addition amount with which a sufficient addition effect can be obtained, and particularly 0.1 to 0.5. % Is preferred.

  Pd and Pt are elements that are completely dissolved in Fe, and are elements that occlude hydrogen in the parent phase. Therefore, by adding Pd and Pt, they combine with the hydrogen generated and invaded in the steel and within the grain boundaries. Contributes to the suppression of intergranular stress corrosion cracking. These elements have a function of stabilizing the austenite phase, but a large amount of addition may lead to deterioration of workability. Therefore, it is preferable to add 1.0% or less of one or two of Pd and Pt as an addition amount with which a sufficient addition effect can be obtained, and 0.1 to 0.6% is particularly preferable.

N, like C, is a component that dissolves in the matrix and raises the strength of the steel, so a very low content of 0.003% or less is not preferable because the strength is lowered and the irradiation brittleness deteriorates. Therefore, 0.05
% Or less, particularly 0.03% or less, more preferably 0.01 to 0.02%.

These elements of P and S are contained as inevitable impurities. These impurities are accumulated at the grain boundaries by the irradiation-induced grain boundary segregation mechanism, and the corrosion resistance of the grain boundaries is significantly impaired. Therefore, P is 0.05
% Or less and S is preferably 0.01% or less. Si acts to suppress crevice corrosion susceptibility in neutron irradiation and high-temperature and high-pressure water environments, and further acts as an oxygen scavenger in the steelmaking process, so it should be 1% or less, especially 0.1-0.5%. preferable. As described above, considering irradiation embrittlement by neutron irradiation and industrial processes, Si is more preferably 0.3 to 0.6%, P is 0.02% or less, and S is 0.005% or less.

  The austenitic steel having the above components is manufactured through melting, forging, hot rolling and solution heat treatment, and the melting atmosphere is preferably a vacuum. Further, since coarse precipitate phases such as carbides and σ phases are formed in the manufacturing process, in order to suppress this, after solution heat treatment at a temperature of 1100 ° C. or higher, cold rolling of 50% or lower and 1000 to 1150 ° C. By repeating the annealing at a temperature of 1 or more times, formation of a coarse precipitate phase can be suppressed, and workability can be improved. The steel according to the present invention is preferably a solid solution having a uniform austenite phase.

  The austenitic steel thus manufactured can be heat-treated at 650 to 750 ° C. particularly when Zr and Ti carbides are dispersed in the matrix. In this case, the heat treatment temperature is preferably 700 to 750 ° C. in order to suppress the precipitation of Cr carbide at the grain boundaries as much as possible.

The steel of the present invention preferably has a yield strength at room temperature of 18 kg / mm ² or more, a tensile strength of 49 kg / mm ² or more, an elongation of 40% or more, a drawing ratio of 60% or more, and a Vickers hardness of 200 or less. 50 kg / mm ^2, a tensile strength 55~80kg / mm ^2, elongation of 40 to 75% are preferred.

  By applying the austenitic steel as described above to the fuel cladding tube, the reactor internal structure, and the equipment of the nuclear reactor that achieves high efficiency by increasing the temperature and pressure of the cooling water, compared to the conventional JIS 304 and 316 stainless steel, It is expected to show excellent corrosion resistance.

Example 1
Table 1 shows the chemical composition (% by weight) of the alloys related to the present invention and the conventional steel, and experiments (1) to (5) were conducted on these. These steels are subjected to forging and hot rolling after vacuum melting, solution heat treatment at 1100 ° C, repeated cold rolling and annealing to the desired specimen thickness, and finally solution heat treatment at 1100 ° C for 30 minutes. It is what went. Nos. 6 and 12 are those which were kept at 700 ° C. for 1 hour as a heat treatment for dispersing carbides of Zr and Ti in the matrix after the solid solution heat treatment and then cooled in the atmosphere. Further, Nos. 17, 18, and 19 are conventional materials of JIS 304, 310, and 316L stainless steel, respectively.
(1) As for austenite phase stability, Cr equivalents and Ni equivalents were determined from the chemical components of each specimen, and the phase stability was evaluated with reference to the Schaeffler diagram, and the specimens heat-treated under the respective conditions. Were compared using a ferrite scope.
(2) Mechanical strength (0.2% proof stress, tensile strength, total elongation) of specimens heat-treated under the respective conditions by a tensile test at a strain rate of 5 × 10 ⁻³ / sec at 550 ° C. in the atmosphere. Were evaluated and compared.
(3) Similar to (2), a surface corrosion test piece was prepared from the specimens heat-treated under the respective conditions, and high temperature and high pressure water (supercritical water) at a temperature of 400 and 550 ° C. and a pressure of 25 MPa and a dissolved oxygen content of 8 ppm. It was immersed for 500 hours, and the weight change after immersion was measured. Based on the result, it was converted into a thinning rate (mm / year), and corrosion resistance was evaluated.
(4) Similarly to (2), the specimens heat-treated under the respective conditions were irradiated with 1 MeV electron beams at 450 ° C. and 550 ° C. to 10 dpa, and the amount of void swelling was observed from the subsequent microstructure observation. Were evaluated and compared.
(5) Similar to (2), double U-bend specimens were prepared from the specimens heat-treated under the respective conditions, and high-temperature and high-pressure water (supercritical) with a temperature of 300 and 550 ° C., a pressure of 25 MPa, and a dissolved oxygen content of 8 ppm. The sample was immersed in water for 500 hours, and the presence or absence of stress corrosion cracking after immersion was evaluated and compared.
(6) Similar to (2), the specimens heat-treated under the respective conditions were irradiated with 1 MeV electron beams at 300 ° C. and 550 ° C. to 10 dpa, and then the Cr content at the grain boundaries irradiated. Were analyzed using an energy dispersive X-ray analyzer, and the difference in Cr concentration between the parent phase and the grain boundary was used as a deficiency and compared.

Table 2 summarizes the results of the experiments (1) to (5). The results of the experiment are described below.
(1) It was found that the austenite phase stability of each specimen was comparable to or more stable than SUS304 (conventional steel 17). The same level as SUS304 is located in the 5% ferrite region of the Schaeffler diagram and is quasi-austenite stable. Further, those having better phase stability than SUS304 were located in a region that becomes all austenite in the Schaeffler diagram.
(2) It turned out that the high temperature strength of each test material is all better than SUS304 (conventional steel 17). Since the steel of the present invention contains a large amount of Ni, it is considered that good high-temperature strength was obtained.
(3) The thinning rate of each test material was smaller than that of SUS304 (conventional steel 17) or SUS310 (conventional steel 18), and the thinning rates were both less than 0.05 mm / year. In particular, the steel thinning rate of the invention steel having a high Cr content is less than 0.03 mm / year, and it was revealed that the steel has good corrosion resistance.
(4) The amount of swelling of each test material is smaller than SUS304 (conventional steel 17) and SUS310 (conventional steel 18), and shows a value of less than 1% even at 450 ° C., where the swelling is particularly noticeable in austenitic steel. It was. Invented steel with a large amount of Ni added and invented steel in which carbides are dispersed by heat treatment, the amount of swelling was 0.5% or less, and it was revealed that the steel had good swelling resistance.
(5) The stress corrosion cracking susceptibility of each test material was observed in SUS304 (conventional steel 17) at 300 ° C., but was not observed in the steel of the present invention.
(6) The Cr deficiency under irradiation at the grain boundaries of each specimen is SUS304 (conventional steel 17),
While it was 5% or more in SUS310 (conventional steel 18) and SUS316L (conventional steel 19), it was 3% or less in the steel of the present invention. The amount of Cr deficiency under irradiation can be used as an index for evaluating SCC sensitivity under irradiation, and it was shown that the steel of the present invention exhibits good characteristics in terms of SCC sensitivity under irradiation.

  As described above, according to the present invention, there are provided a fuel cladding tube and a core constituent material used in the core of a high-efficiency light water reactor in which the cooling water conditions are 290 ° C. or higher and 550 ° C. or lower and the pressure is 25 MPa. can do.

According to the present invention, the above-described austenitic steel can be applied to, for example, a nuclear reactor fuel cladding tube, a reactor internal structure, and equipment that achieve high efficiency by increasing the temperature and pressure of cooling water.
Table 1 Chemical composition and heat treatment of inventive steel and conventional steel Table 2 Experimental results according to Examples (1) to (5)

Claims (9)

  High corrosion resistance core with excellent corrosion resistance, swelling resistance and stress corrosion cracking resistance, characterized by being composed of austenitic steel in contact with water and containing at least one of zirconium and titanium used under irradiation material. By weight, C: 0.005% to 0.04%, Si: 1% or less, P: 0.02% or less, N:
0.05% or less, Mn: 2% or less, Cr: 20 to 28%, Ni: 16 to 25%, Mo: 0.5 to 3%, Zr: 0.1 to 2% and Ti: 0.1 A highly corrosion-resistant core material comprising an austenite structure comprising at least one of ˜2% and the balance of Fe and impurities of 50% or more. By weight, C: 0.005% to 0.04%, Si: 1% or less, P: 0.02% or less, N:
0.05% or less, Mn: 2% or less, Cr: 20 to 28%, Ni: 16 to 25%, Mo: 0.5 to 3%, Zr: 0.1 to 2% and Ti: 0.1 It includes at least one of ˜2% and a total of at least one of Pd, Pt, and Hf of 1.0% or less, and the balance has an austenitic structure composed of 50% or more of Fe and impurities. High corrosion resistance core material. In the highly corrosion-resistant core material according to claim 2 or 3,
The austenitic steel, after undergoing forging, hot rolling, and solution heat treatment, forms a carbide containing either Zr or Ti in the matrix phase. In the highly corrosion-resistant core material according to claim 2 or 3,
The austenitic steel is subjected to forging, hot rolling, and solution heat treatment, followed by heat treatment at 650 ° C. to 750 ° C., a highly corrosion-resistant core material. In the highly corrosion-resistant core material according to claim 2 or 3,
The austenitic steel is characterized by having a thickness reduction rate of 0.05 mm / year or less when immersed in high-temperature and high-pressure water at 400 ° C. or higher and 25 MPa in a member that is in contact with water and used under radiation irradiation. High corrosion resistance core material with excellent corrosion resistance, swelling resistance and stress corrosion cracking resistance. In the highly corrosion-resistant core material according to claim 2 or 3,
The austenitic steel is in contact with water, and is used under irradiation of radiation. After being irradiated to 300 dpa to 10 dpa at an irradiation damage amount, the amount of swelling is less than 1%. Core material with excellent corrosion resistance and stress corrosion cracking resistance. In the highly corrosion-resistant core material according to claim 2 or 3,
When the austenitic steel is in contact with water and subjected to an SCC susceptibility evaluation test using a double U-bend test piece in a high-temperature high-pressure water at 300 ° C. or higher and 25 MPa in a member used under radiation irradiation, a grain boundary crack occurs. A highly corrosion-resistant core material with excellent corrosion resistance, swelling resistance and stress corrosion cracking resistance, characterized by the absence of corrosion. In the highly corrosion-resistant core material according to claim 2 or 3,
The austenitic steel is in contact with water, and after being irradiated to a radiation damage of 10 dpa at a temperature of 300 ° C. or higher in a member used under radiation irradiation, the decrease in Cr concentration at the grain boundary is compared with the concentration of the parent phase. High corrosion resistance core material with excellent corrosion resistance, swelling resistance and stress corrosion cracking resistance, characterized by being within 5%.