Patent Application
20170167005
The present invention relates to dispersion-strengthened austenitic stainless steel, and more particularly to highly irradiation-resistant austenitic stainless steel applicable to a reactor internal structure used under a neutron irradiation environment.
Austenitic stainless steel has high corrosion resistance under a corrosive environment due to a Cr passive film formed on the surface thereof. Austenitic stainless steel is frequently used as a component of a structural material, and is often used also in a nuclear power plant. However, the following possibility has been pointed out. That is, under a high-temperature, high-pressure water environment such as in a reactor, stress corrosion cracking occurs, and furthermore, the stainless steel is irradiated with neutrons within the reactor and an irradiation defect is introduced. As a result, Cr deficiency occurs at a crystal grain boundary, leading to irradiation-induced stress corrosion cracking. Therefore, in order to lower the susceptibility to irradiation-induced stress corrosion cracking, development of a material having excellent resistance to irradiation and stress corrosion cracking is required.
PTL 1 describes that corrosion resistance, strength, and resistance to irradiation are improved by using austenitic stainless steel having an average crystal grain diameter of 1 μm or less and containing at least one kind selected from Ti at 1.0% or less, Zr at 2.0% or less, and Nb at 1.0% or less.
PTL 2 describes ferritic stainless steel in which toughness and workability as well as strength are improved by bulking powder having an ultrafine grain structure obtained by mechanical alloying or the like while maintaining the structure.
PTL 1: JP 8-337853 A
PTL 2: JP 2002-285289 A
A purpose of the austenitic stainless steel described in PTL 1 is to improve corrosion resistance, strength, and resistance to irradiation by making crystal grains finer, and to improve strength and resistance to stress corrosion cracking by forming oxides and carbides by Ti, Zr, or Nb. Meanwhile, an element additive amount relative to the amount of oxygen and carbon is not clear, and there maybe a risk of a decrease in mechanical strength due to the formation of an intermetallic compound and residual oxygen and carbon. In addition, the ferritic stainless steel described in PTL 2 largely improves toughness by stabilizing oxygen, which adversely affects the mechanical properties of steel, by forming an oxide of the added element. Meanwhile, since the parent phase is a ferrite phase, there is a concern about corrosion resistance under the environment inside a reactor.
There is a problem with the fact that, while austenitic stainless steel is excellent in corrosion resistance, irradiation-induced stress corrosion cracking may occur under an irradiation environment. Ferrite martensitic oxide dispersion-strengthened steel has been developed as a material having excellent resistance to irradiation, but still has a problem with corrosion resistance under the environment inside a reactor.
An object of the present invention is to provide austenitic stainless steel having a high strength and high corrosion resistance while improving resistance to irradiation and reducing irradiation-induced stress corrosion cracking.
Austenitic stainless steel of the present invention contains Cr: 16 to 26%, Ni: 8 to 22%, O: 0.02 to 0.4%, C: 0.08% or less, and N: 0.1% or less by weight, and further contains at least one kind of Zr: 0.2 to 2.8%, Ta: 0.4 to 5.0%, and Ti: 0.2 to 2.6%, the balance being Fe and unavoidable impurities, wherein Zr, Ta, and Ti are precipitated as precipitates of one or more kinds of oxide particles, carbide particles, nitride particles, and composite particles thereof.
A method for producing austenitic stainless steel of the present invention includes: alloying, through mechanical alloying treatment, alloy powder containing Cr: 16 to 26%, Ni: 8 to 22%, O: 0.02 to 0.4%, C: 0.08% or less, and N: 0.1% or less by weight, and further containing at least one kind of Zr: 0.2 to 2.8%, Ta: 0.4 to 5.0%, and Ti: 0.2 to 2.6%, the balance being Fe and unavoidable impurities, or mixed powder satisfying the composition; encapsulating the powder in a container under vacuum; and solidifying and molding the powder at 800° C. to 1200° C.
According to the present invention, austenitic stainless steel having a high strength and high corrosion resistance while improving resistance to irradiation and reducing irradiation-induced stress corrosion cracking, and a method for producing the austenitic stainless steel can be provided.
The structure, composition and production conditions of austenitic stainless steel of the present invention will be described below. The austenitic stainless steel of the present invention has the following basic composition. The austenitic stainless steel contains Cr: 16 to 26%, Ni: 8 to 22%, O: 0.02 to 0.4%, C: 0.08% or less, and N: 0.1% or less by weight, and further contains at least one kind of Zr: 0.2 to 2.8%, Ta: 0.4 to 5.0%, and Ti: 0.2 to 2.6%, the balance being Fe and unavoidable impurities. In the austenitic stainless steel of the present invention, Zr, Ta, and Ti are precipitated as precipitates of one or more kinds of oxide particles, carbide particles, nitride particles, and composite particles thereof.
Zr, Ta, and Ti are added in order to stabilize oxygen present in the steel and to be precipitated as oxide particles. Oxygen contained in the steel is known as a factor that may deteriorate the mechanical properties, particularly toughness, and also lower the intergranular corrosion resistance of the stainless steel. Therefore, Zr, Ta, and Ti having a high affinity for oxygen are added to stabilize oxygen and improve the mechanical properties. Furthermore, Zr, Ta, and Ti have a high affinity for carbon and nitrogen as well as oxygen, and stabilize these elements in the steel and form precipitates. By receiving heat generated by, for example, welding, carbon in the stainless steel becomes a factor that precipitates Cr carbides on a grain boundary, and lowers the Cr concentration near the grain boundary to cause stress corrosion cracking. Nitrogen lowers the resistance to irradiation of the stainless steel, and accelerates irradiation-induced intergranular segregation. Stabilizing these elements by the addition of Zr, Ta, and Ti contributes to an improvement in the mechanical properties, resistance to irradiation, and corrosion resistance of the stainless steel.
Preferably, oxygen, carbon, and nitrogen are stabilized by Zr, Ta, and Ti and do not remain in the parent phase. Thus, Zr, Ta, and Ti are added in enough amounts to stabilize all the oxygen, carbon, and nitrogen, the amounts being neither stoichiometrically excessive nor deficient. The addition in stoichiometrically excessive amounts is not desirable, because such addition may form intermetallic compounds, leading to deterioration of mechanical properties.
The additive amounts of Zr, Ta, and Ti are determined within the following ranges in accordance with correlation formulas, where the amounts of oxygen, carbon, nitrogen contained in steel are expressed by X %, Y %, and Z % respectively.
(Correlation formulas)
Zr:0.2≦2.85×+7.60Y+6.52Z≦2.8
Ta:0.4≦4.52×+15.0Y+12.9Z≦5.0
Ti:0.2≦1.50×+3.99Y+3.42Z≦2.6
The additive amounts may be increased, as necessary, from the above amounts in order to stabilize oxygen, carbon, and nitrogen inevitably mixed during the production process. In the present invention, the amounts of oxygen, carbon, and nitrogen mixed during the production process are 0.1 wt % or less, 0.07 wt % or less, and 0.02 wt % or less respectively. The minimum value of the additive amount is 0 wt %, and the maximum value thereof is an enough amount to stabilize all the oxygen, carbon, and nitrogen, and is neither stoichiometrically excessive nor deficient. Preferably, by a preliminary experiment, an oxygen amount α%, a carbon amount β%, and a nitrogen amount Υ% to be mixed during the production process are estimated and determined according to the following correlation formulas. Excessively added Zr, Ta, and Ti may each form intermetallic compounds and cause deterioration of mechanical properties.
(Correlation formulas)
Zr:2.85α+7.60β+6.52Υ
Ta:4.52α+15.0β+12.9Υ
Ti:1.50α+3.99β+3.42Υ
Furthermore, if necessary, predetermined amounts of Zr, Ta, and Ti are made into a solid solution state in addition to the above additive amounts. When these oversized elements are made into solid solutions in the parent phase, the elements serve as a trap site of irradiation defects, and have an effect of promoting recombination of the irradiation defects and suppressing the irradiation-induced segregation. Therefore, it is expected to suppress intergranular Cr deficiency due to irradiation-induced intergranular segregation, and to lower susceptibility to the irradiation-induced stress corrosion cracking. If these elements are excessively added, on the other hand, the elements form intermetallic compounds, which is not preferable.
As described above, the additive amounts of Zr, Ta, and Ti are determined so as to satisfy the following relational formulas, where the amounts of oxygen, carbon, nitrogen are X %, Y %, and Z % by weight respectively, and the amounts of Zr, Ta, Ti are X %, Y %, and Z % by weight respectively.
By determining the additive amounts of Zr, Ta, and Ti so as to satisfy the relationship:
2.85×7.60Y+6.52Z≦A (Formula 1)
4.52×15.0Y+12.9Z≦B (Formula 2)
1.50×3.99Y+3.42Z≦C (Formula 3)
the amounts of oxygen, carbon, and nitrogen remaining in the parent phase without being stabilized can be reduced, making it possible to improve the mechanical properties, resistance to irradiation, and corrosion resistance of the stainless steel.
In austenitic stainless steel of the present invention, the precipitates that are precipitated as one or more kinds of the oxide particles, carbide particles, nitride particles, and composite particles thereof serve as a pinning site that suppresses the coarsening of crystal grain boundaries and the movement of dislocations and as a trap site of irradiation defects, and contribute to an improvement in mechanical strength and resistance to irradiation. According to the models of Orowan and Ansell, as for precipitated particles, the finer the size and the higher the precipitation density, the better, in terms of mechanical strength. From the viewpoint of suppressing coarsening of crystal grains, a pinning effect of the crystal grains is observed when the grain diameter is 0.1 μm or less and, allegedly, the grain diameter of 0.05 μm or less is more preferable. From the viewpoint of resistance to irradiation, the higher the number density of precipitates, the better. An effect of suppressing irradiation defect aggregation can be observed when the number density is on the order of 1022 m−3. It is assumed in the present invention that precipitates having a grain diameter of 0.1 um or less are dispersed at a number density of 1.0 ×1022 m−3 or higher.
The parent phase structure in the austenitic stainless steel according to the present invention is an austenite single-phase structure in order to improve corrosion resistance.
Cr contained in the austenitic stainless steel of the present invention is an element that improves the corrosion resistance of alloy and stabilizes an austenite structure, and the additive amount thereof is preferably 16% or more. If excessively added, Cr may form the σ phase and deteriorate the material properties, and thus the additive amount of Cr in the present invention is 16 to 26%. To obtain a more stable austenite phase, the Cr content is preferably 18 to 20%.
Ni contained in the austenitic stainless steel of the present invention contributes to the corrosion resistance, and the stability of the austenite phase. However, excessively added Ni is not preferable because such Ni may cause deterioration of the mechanical properties or precipitation of an embrittling phase. Therefore, the additive amount of Ni in the present invention is 8 to 22%. To obtain a more stable austenite phase, the Ni content is preferably 8 to 12%.
The hall-petch equation empirically holds for the relationship between the crystal grain diameter and the mechanical strength. The finer the crystal grains, the higher the tensile strength. When the grains in iron are made finer up to about 5 μm, the tensile strength becomes about 1.5 times as high as a normal material (average crystal grain diameter of several tens of micrometers). There is also a characteristic of not losing ductility, and it can be said that the smaller the average crystal grain diameter, the better, in terms of the mechanical strength. The miniaturization of the crystal grains can increase the total length of the crystal grain boundaries that are defect sink sites. Thus, the smaller the grain diameter, the better, also in terms of resistance to irradiation. The miniaturization of the crystal grains is also known to improve corrosion resistance, and is preferable in terms of any of mechanical strength, resistance to irradiation, and corrosion resistance. In the present invention, the average crystal grain diameter is 5 μm or less, and preferably ultrafine grains of 1 μm or less are applied.
The austenitic stainless steel of the present invention is produced as follows. Alloy powder or mixed powder satisfying the aforementioned composition is subjected to mechanical alloying in a ball mill or an attritor. The resultant powder is encapsulated in a vacuum container and solidified and molded by any of hot isostatic pressing, hot extrusion, and hot rolling. The solidifying and molding temperature is preferably 800° C. or higher in order to achieve a satisfactory bond between the powder, and the solidification and molding are carried out at equal to or less than 1200° C., which is a solution heat treatment temperature. The higher the treatment temperature, the easier the coarsening of the crystal grains. Therefore, solidification and molding at too high a temperature are not preferable.
Machining processes such as forging and rolling are optionally applied to the solidified and molded material. Alloy produced by solidifying and molding powder has a weak bond at a former powder boundary, leading to the deterioration of the mechanical strength. Therefore, by means of the above processes, the density is increased and the bond between the powder is strengthened, thereby improving the mechanical properties. The temperature at the time of implementing the processes is preferably from 800° C. to 1200° C. in a similar manner to the solidifying and molding process.
Furthermore, heat treatment is performed in order to homogenize the structure of, and remove distortion from, the solidified and molded alloy. The treatment temperature is 800° C. to 1200° C. in the same manner as the solidifying and molding temperature.
In an example shown below, it was confirmed that the materials of the present invention exhibited good characteristics. First, Table 1 shows the chemical compositions of the materials produced in this example.
Nos. 1 to 7 are materials that satisfy the basic composition of the austenitic stainless steel of the present invention. Zr is added to Nos. 1, 2, and 3 as a stabilizing element. Zr is added to No. 1 in an enough amount to satisfy the relationship of (Formula 1) and make a predetermined amount of solid solution. Zr is added to No. 2 in an enough amount to satisfy the relationship of (Formula 1). Zr is added to No. 3 in an amount necessary to stabilize 50% of each of oxygen, carbon, and nitrogen in the steel. Ta is added to Nos. 4, 5, and 6 as a stabilizing element. Ta is added to No. 4 in an enough amount to satisfy the relationship of (Formula 2) and make a predetermined amount of solid solution. Ta is added to No. 5 in an enough amount to satisfy the relationship of (Formula 2). Ta is added to No. 6 in an amount necessary to stabilize 50% of each of oxygen, carbon, and nitrogen in the steel. Ti is added to No. 7 as a stabilizing element, in an enough amount to satisfy the relationship of (Formula 3) and make a predetermined amount of solid solution.
These samples of Nos. 1 to 7 were obtained by subjecting powder (SUS304L powder) serving as base steel and the additive element powder (Zr, Ta, Ti) to mechanical alloying by using a planetary ball mill. The powder subjected to mechanical alloying was encapsulated in a vacuum container, and vacuum deaeration was performed for three hours at 400° C. and 10−4 torr. The resultant powder was then solidified and molded through hot extrusion at 1100° C. Thereafter, hot forging was performed at 1100° C. in order to further tighten the bond between the powder. The resultant material was subjected to a solution treatment for 30 minutes at 1100° C. as a final thermal treatment.
Meanwhile Nos. 8 and 9 are SUS304L serving as a comparative material. No. 8 was obtained as follows. SUS304L powder subjected to mechanical alloying was encapsulated in a vacuum container, and vacuum deaeration was performed for three hours at 400° C. and 10−4 torr. The resultant powder was then solidified and molded through hot extrusion at 1100° C. Thereafter, hot forging was performed at 1100° C. in order to further tighten the bond between the powder. The resultant material was subjected to a solution treatment for 30 minutes at 1100° C. as a final thermal treatment. No. 9 was obtained as follows. SUS304L atomized powder not subjected to mechanical alloying was encapsulated in a vacuum container, and vacuum deaeration was performed for three hours at 400° C. and 10−4 torr. The resultant powder was then solidified and molded through hot extrusion at 1100° C. Thereafter, hot forging was performed at 1100° C. in order to further tighten the bond between the powder. The resultant material was subjected to a solution treatment for 30 minutes at 1100° C. as a final thermal treatment.
Table 2 shows the average crystal grain diameters, the grain sizes, and the number densities of the grains of the materials of the present invention and the comparative materials.
Table 3 shows the results of a Vickers hardness test carried out on the materials of the present invention and the comparative materials. The hardness of the comparative material No. 8 is 195 Hv, which is substantially equal to that of general stainless steel. It can be said that the coarse precipitates of 0.1 μm or more, which are coarsely precipitated, hardly contribute to the improvement of the mechanical strength. Meanwhile, it is found that the hardness of the materials of the present invention is significantly higher than that of the comparative materials, i.e., 313 Hv for No. 1 and 280 Hv for No. 4. In the results, the smaller the average crystal grain diameter, the higher the hardness. Also according to this result, it can be said that the additive amounts of the stabilizing elements are preferably large enough to stabilize all the oxygen, carbon, and nitrogen contained in the steel and all the oxygen, carbon, and nitrogen to be mixed during the production.
Table 4 shows the results of an entire surface corrosion test carried out on the materials of the present invention (Nos. 1, 3, 4, and 6) and the comparative material (No. 9). The test results indicate relative values in the case where the corrosion weight increase of the comparative material No. 9 is set to 1. The test was carried out for 2000 hours as an immersion test under a light-water reactor environment (at 288° C., 8 MPa). The weight increases of the materials of the present invention were all substantially equal to or slightly better than that of the comparative material, and the same applies to the sample surface shape after the immersion test. Among the materials of the present invention, the finer the crystal grains, the smaller the weight increase. The results showed that the materials of the present invention had corrosion resistance similar to that of SUS316L. It can be said that the miniaturization of the crystal grain boundaries is preferable for improving the corrosion resistance.
The susceptibility to stress corrosion cracking was evaluated under a high-temperature, high-pressure water environment. The creviced bent beam (CBB) test was employed.
In the test, an immersion test was carried out for 2000 hours at 288° C. and a dissolved oxygen concentration of 8 ppm, and this test was carried out seven times on one sample. All of the samples used were subjected to 40% cold working. Table 5 shows the results of evaluating occurrence of a crack in the test pieces. A crack having a depth of 50 μm or more was defined as the crack. The test was carried on the materials of the present invention (Nos. 1, 3, 4, and 6) and the comparative material (No. 9). In the comparative material No. 9, cracks were formed on multiple test pieces, while no crack was observed on Nos. 1 and 4 to which elements were added in enough amounts to stabilize all oxygen, carbon, and nitrogen. Meanwhile, a crack was observed on one test piece in the material of the present invention No. 6. It is known that stress corrosion cracking is reduced with a decrease in the amount of a solid solution of C. Reducing the amount of a solid solution of C by adding the stabilizing elements was shown to be effective. The crack was formed on No. 6 in which the additive amounts were suppressed to stabilize only 50% of the oxygen, carbon, and nitrogen. Therefore, it was shown that the additive amounts of the stabilizing elements are preferably large enough to stabilize all the oxygen, carbon, and nitrogen in consideration of resistance to SCC characteristics.
An application example of the austenitic stainless steel of the present invention is described below, in which example the austenitic stainless steel is applied to a reactor internal structure and a control rod. The control rod has the highest rate of neutron irradiation damage among the devices.
Irradiation-induced stress corrosion cracking is likely to occur in austenitic stainless steel that has conventionally been used in a reactor, upon irradiation with neutrons having energy of 0.1 MeV or higher at 0.5×1021 n/cm2 or more. Therefore, the austenitic stainless steel is applicable not only to the control rod but also to other reactor inner structures and devices in which irradiation-induced stress corrosion cracking is likely to occur upon neutron irradiation. For example, the austenitic stainless steel of the present invention can be used in a core shroud, a top guide, a core support plate, a baffle plate, a former plate, or a baffle former bolt, whereby the reactor and the nuclear power plant can be made highly reliable.
| Number | Date | Country | Kind |
|---|---|---|---|
| 2014-139282 | Jul 2014 | JP | national |
| Filing Document | Filing Date | Country | Kind |
|---|---|---|---|
| PCT/JP2015/057971 | 3/18/2015 | WO | 00 |
