JP5978095B2 - High corrosion resistance austenitic stainless steel - Google Patents

Description

The present invention relates to an austenitic stainless steel capable of suppressing stress corrosion cracking, irradiation-induced stress corrosion cracking, crevice corrosion, and the like in an environment exposed to a neutron irradiation environment, and a reactor internal structure using the austenitic stainless steel.

  In light water reactor cores, low carbon austenitic stainless steels such as SUS316L and SUS304L with reduced carbon content are used to suppress sensitization by welding heat. However, in recent years, stress corrosion cracking, which is considered to be due to irradiation induction, has been confirmed in SUS316L sheaths and tie rods of control rods used in environments with high neutron irradiation. This is considered to be caused by hardening due to neutron irradiation of the material or segregation due to irradiation-induced grain boundary, and it is necessary to suppress neutron irradiation damage.

  Patent Document 1 discloses a low-carbon austenitic stainless steel to which Nb, Ta, Ti, Zr, and Hf are added for suppressing neutron irradiation damage. However, they cannot achieve the strength characteristics required as control rods due to the low carbon concentration.

On the other hand, Patent Document 2 discloses an austenitic stainless steel containing 0.04% or less of carbon and containing Nb and Ta. This is because Nb and Ta stabilize the carbon even after solution heat treatment by adding Nb and Ta, thereby suppressing stress corrosion cracking and radiation-induced stress corrosion cracking, and reducing the Nb addition amount and increasing the Ta addition amount. By doing so, it has the effect of facilitating handling as Ta182 having a short half-life instead of Nb94 having a long half-life generated by neutron absorption. However, there is a possibility that the corrosion resistance and strength characteristics required as a control rod cannot be satisfied with the solution.
Similar to Patent Document 2, there are Type 321, Type 347, and Type 348 as standard materials that stabilize carbon and improve corrosion resistance. Type 321 is added with Ti, and Type 347 and 348 are added with Nb + Ta.

JP 63-259055 A Japanese Patent Publication No. 06-089437

  In a nuclear reactor of a nuclear power plant, austenitic stainless steel is mainly used as a constituent material. In general, austenitic stainless steel is a material with excellent corrosion resistance, but it has been pointed out that the susceptibility to irradiation-induced stress corrosion cracking increases when exposed to high-temperature and high-pressure water and neutron irradiation environments. In addition, irradiation-induced stress corrosion cracking cases have been reported as material deterioration events in nuclear reactors.

  An object of the present invention is to provide an austenitic stainless steel with reduced susceptibility to irradiation-induced stress corrosion cracking.

Austenitic stainless steel is 0.03 to 0.04% by weight for C, 1.0 to 2.0% by weight for Mn, 9.0 to 13.0% by weight for Ni, 17.0 to 20.0% by weight for Cr, 0.50 to 0.65% for Ta more than 13 times C In the austenitic stainless steel with the balance being Fe and inevitable impurities, 30% or more of Ta is precipitated as TaC.

By applying the austenitic stainless steel of the present invention to a reactor internal component or structure used in an environment where neutrons having an energy of 0.1 MeV or more are irradiated by 0.5 × 10 ²¹ n / cm ² or more, It is possible to reduce the susceptibility of irradiation-induced stress corrosion cracking, improve the reliability of the nuclear power plant, and extend the life.

Comparison of reactivation rates of materials of the present invention. The correlation between the reactivation rate of the material of the present invention and the Ta / C ratio. Comparison of tensile strength of the material of the present invention at 100 ° C. Correlation between tensile strength at 100 ° C. and C content of the material of the present invention. Correlation between tensile strength at 100 ° C. and grain size of the material of the present invention. Crevice corrosion test piece. Comparison of maximum intergranular corrosion depth due to crevice corrosion of materials of the present invention and comparative materials. Ta concentration contained in precipitates in the material of the present invention subjected to various heat treatments. The ratio of the amount of Ta contained in the precipitate to the total amount of Ta in the material of the present invention subjected to various heat treatments. Reactivation rate of the material of the present invention after various heat treatments. Number density of TaC precipitates in the material of the present invention subjected to various heat treatments. A control rod to which the material of the present invention is applied.

Hereinafter, the present invention will be described in detail.
In the present invention, C is 0.03 to 0.04% by weight, Mn is 1.0 to 2.0% by weight, Ni is 9.0 to 13.0% by weight, Cr is 17.0 to 20.0% by weight, Ta is 13 times or more of C, and 0.50 to 0.65% by weight. In the austenitic stainless steel, the balance is Fe and inevitable impurities, and 30% or more of the added Ta is precipitated as TaC.
Since the atomic radius of Ta is larger than the average atomic radius of the material constituent elements, when Ta is dissolved in the material, it captures atomic vacancies generated by irradiation and recombines with interstitial atoms. Probability can be increased and irradiation damage can be suppressed. In addition, when atomic vacancies flow into the grain boundary by diffusion, Cr atoms near the grain boundary diffuse away from the grain boundary, and the Cr concentration near the grain boundary is lower than the Cr concentration of the matrix. Deficiency occurs, but when Ta captures atomic vacancies, diffusion of atomic vacancies into grain boundaries is suppressed, and grain boundary Cr depletion is suppressed. On the other hand, even when Ta is present as TaC, the interface between TaC and the matrix becomes an atomic vacancy disappearance site, so that grain boundary Cr deficiency can be suppressed as in the case where Ta is dissolved.
As described above, the presence of Ta can suppress irradiation-induced grain boundary Cr deficiency, whether it is dissolved alone or as TaC, and is effective in suppressing irradiation-induced stress corrosion cracking. it is conceivable that. However, when C is present in the matrix more than necessary, in locations where it has been subjected to welding heat, Cr carbide is formed on the grain boundary, resulting in Cr deficiency, so-called thermal sensitization, and stress corrosion cracking susceptibility is increased. . Therefore, it is necessary to form TaC in order to reduce the concentration of dissolved C.
Therefore, by precipitating more than a predetermined amount of TaC with respect to the C concentration present in the material and stabilizing C, it is possible to suppress both grain-boundary Cr segregation due to irradiation induction and grain boundary Cr due to heat. Corrosion cracking sensitivity can be reduced. As a material constituting the furnace, it is preferable that the tensile strength at 100 ° C. is 470 MPa or more. To achieve this, 0.03% by weight of C is necessary, but if the amount of C is too large, heat sensitivity Therefore, the upper limit was made 0.04% by weight.
Further, Ta is required to be 13 times the amount of C and 0.5% by weight or more, and the upper limit is set to 0.65% by weight in view of deterioration of characteristics such as weldability. Further, when Ta stabilizes C, if 30% or more of the total amount of Ta is precipitated as TaC, both the stress corrosion cracking resistance and strength are good. The amount of Ta precipitated as TaC can be measured by quantitatively analyzing the electrolytic extraction residue collected by the constant current electrolysis method.

In order to secure the strength more stably with the material, the crystal grain size number is preferably 4.0 or more. In addition, since the effect of improving the strength can be obtained when the number density of precipitated TaC is a certain amount or more, the average number density of TaC precipitated particles having an average particle diameter of 10 nm or more in the stainless steel cross section is 1.0 × 10 ^12. It is preferable that the number of particles / m ² or more. The number density of TaC particles can be evaluated by observing a specimen collected by a replica method using C vapor deposition with STEM (Scanning Transmission Electron Microscopy).
Such a material structure needs to be adjusted depending on the thickness of the material, but was subjected to a solution heat treatment at 1,050 to 1,150 ° C. for 1 to 60 minutes and an aging treatment at 850 to 950 ° C. for 0.5 to 6 hours. Can be obtained. By applying the above-mentioned predetermined heat treatment to the above-mentioned chemical component material and obtaining the above-mentioned material structure, it is possible to obtain a tensile strength of 470 MPa or more at 100 ° C. necessary for the material, and electrochemical reactivation The thermal sensitization can be suppressed by setting the conversion rate to 30% or less.
The above materials can suppress irradiation damage and irradiation-induced stress corrosion cracking when used in an environment where neutrons with energy of 0.1 MeV or higher are irradiated by 0.5 × 10 ²¹ n / cm ² or higher. . In particular, by applying it to tie rods, sheaths, and handles that constitute control rods used in boiling water reactor cores, the possibility of radiation-induced stress corrosion cracking is reduced and control rod replacement cycles are extended. As well as being able to improve soundness and reliability.
In addition to the control rod, the above-mentioned materials can be applied not only to control rods but also to equipment or structures that constitute the reactor and are irradiated with neutrons with an energy of 0.1 MeV or more of 0.5 × 10 ²¹ n / cm ² or more. The reliability of the nuclear reactor, and hence the nuclear power plant, can be improved and the life can be extended.

Examples in which the material of the present invention is confirmed to be a material having good characteristics are shown below.
First, a material having the chemical composition shown in Table 1 was produced by vacuum melting.

Note that No. 10 is SUS316L as a comparative material. All the materials were subjected to a solution heat treatment with water cooling at 1050 ° C. for 30 minutes after hot rolling. Moreover, the solution material of material No. 1-7 was divided into two ingots, and one of them was subjected to air-cooled aging treatment at 900 ° C. for 1 hour, that is, stabilization heat treatment for precipitating TaC and stabilizing C. . As a result, the materials No. 1 to No. 7 produced a solution heat treatment material and a solution heat treatment + stabilization heat treatment material. On the other hand, No. 10 as a comparative material was produced only by a solution heat treatment material.

In order to evaluate the thermal sensitization characteristics that correlate with the stress corrosion cracking characteristics, the solution heat treatment materials of material Nos. 1 to 7 and solution heat treatment + stabilization heat treatment materials were further sensitized by water cooling at 700 ° C for 30 minutes. After heat treatment, the reactivation rate was measured according to JIS G 0580 “Method for measuring electrochemical reactivation rate of stainless steel”. The measurement was repeated 3 times for each material. The reactivation rate is a value corrected by the crystal grain size according to the equation (1). Here, R is the reactivation rate corrected for particle size, Rm is a measured value of the reactivation rate, and N is the crystal grain size number of the test surface.
(1)
The average value of the reactivation rate measured for each material is shown in FIG. In JIS G 0580, a reactivation rate of less than 31% is defined as non-sensitization, 31% or less and less than 156% is defined as slight sensitization, and 156% or more is defined as sensitization. Material No. 6 is in a state of sensitization by solution heat treatment + sensitization heat treatment, slight sensitization by solution heat treatment + stabilization heat treatment + sensitization heat treatment, and is considered to be a heat sensitization material. The possibility of stress corrosion cracking is high. Material Nos. 1 to 3 were in a region of slight sensitization by solution heat treatment + sensitization heat treatment, but could be brought into a non-sensitization state when subjected to stabilization heat treatment. Material Nos. 4, 5, and 7 were in a non-sensitized state for both solution heat treatment + sensitization heat treatment and solution heat treatment + stabilization heat treatment + sensitization heat treatment. From these results, the materials No. 1 to 5 and 7 have good characteristics against thermal sensitization, and Nos. 3, 4, and 7 have excellent characteristics.
Next, in order to confirm the correlation between the reactivation rate and the Ta / C ratio, the relationship between the average value of the reactivation rate shown in FIG. 1 and the Ta / C ratio is shown in FIG. From this figure, the higher the Ta / C ratio tends to reduce the reactivation rate. If the Ta / C ratio is 13 or more, it has good characteristics for thermal sensitization, and especially the stabilization heat treatment. When it was applied, even better characteristics were exhibited.
The present invention is directed to structures used in nuclear reactors. Since some materials require high strength, a tensile strength of at least 100 ° C. is required to be 470 MPa. Accordingly, FIG. 3 shows the tensile strength data when the solution heat treatment material and the solution heat treatment + stabilization heat treatment material of each material were subjected to a tensile test at 100 ° C. in accordance with JIS G 0567.
The number of repetitions of the tensile test was 2, and the average value was shown in FIG. The required tensile strength at 100 ° C is 470 MPa or more. As a result, in the solution heat treated material, the materials No. 4, 6, and 7 were able to satisfy the required strength. No. 1 and 3 satisfied the required strength. On the other hand, Nos. 2 and 5 could not satisfy the required strength. Moreover, when it applied to stabilization heat processing, the tendency which an intensity | strength improved compared with the solution heat processing material was shown. From this, it is possible to further increase the strength by dispersing TaC by stable heat treatment.
FIG. 4 shows the correlation between the tensile strength at 100 ° C. and the C content shown in FIG. Both the solution heat treatment material and the solution heat treatment + stabilization heat treatment material have improved strength with increasing C content. Here, the solution heat-treated material was able to satisfy 470 MPa when containing 0.03% by weight or more of C. On the other hand, the solution heat treatment + stabilization heat treatment material was able to satisfy 470 MPa with a C content of 0.02 wt% or more.
In order to satisfy the strength, it is possible to increase the strength by processing. However, when subjected to strong processing, the stress corrosion cracking sensitivity is increased, which is not preferable. Therefore, high strength by crystal grain refinement was examined. Material No. 1 was subjected to 40% cold rolling, heat treated at 1050 ° C. for 10 minutes or 1 hour, then water cooled, and 1100 ° C. for 1 hour, then water cooled. The crystal grain size and tensile properties at 100 ° C. were measured for three types.
The result is shown in FIG. As a result, it was confirmed that the strength increased as the crystal grain size increased. Here, it can be seen that if the crystal grain size is 4.0 or more, it exceeds the required 470 MPa. For this reason, the material of the present development preferably has a crystal grain size of 4.0 or more.
Next, crevice corrosion characteristics in high-temperature and high-pressure water were evaluated. Four test pieces of 10 × 50 × 1.5 t were collected from each of the materials No. 1, 2, 4, 7, and 10, and the surface was finished up to 600 with water-resistant abrasive paper.
The test pieces 1 of the same material face each other as shown in FIG. Thereby, a gap was formed between the two test pieces. For each material, two spot weld specimens were prepared. They were subjected to a 1000 hour immersion test in an environment of 288 ° C., dissolved oxygen concentration of 8 ppm, conductivity of 1.0 to 1.2 μS / cm ² and γ-ray dose rate of 20 kGy / h. After the immersion test, the specimen was cut at the center of the specimen parallel to the longitudinal direction of the specimen, and the corrosion state of the cross section was observed.
FIG. 7 shows the maximum intergranular corrosion depth in each test piece. For all materials, the corrosion state was observed for four test pieces. As a result, in No. 10 SUS316L, corrosion occurred in all the test pieces, and the maximum intergranular corrosion depth exceeded 100 μm. In Nos. 1, 2, and 4, intergranular corrosion occurred in two of the four test pieces, and the maximum depths were 27 μm, 35 μm, and 15 μm, respectively. In No. 7, no intergranular corrosion was observed in any of the test pieces, which is considered to have good crevice corrosion resistance.
Furthermore, the stress corrosion cracking susceptibility in high temperature and high pressure water was evaluated. A Creviced Bent Beam (CBB) test was adopted as the test method. Seven test pieces of 10 × 50 × 2 t were collected from each of the materials No. 1, 2, 4, 7, and 10, and the surface was finished up to 600 with water-resistant abrasive paper. These were set together with graphite wool in a jig having an arc-shaped aspect with a radius of 100 mm, to obtain a test piece with a gap provided with a constant strain of 1%. The test piece set on the jig was subjected to a 500 hour immersion test at 288 ° C., a dissolved oxygen concentration of 8 ppm, and a conductivity of 1.0 μS / cm 2. After the immersion test, remove the specimen from the jig, cut it at the center of the specimen parallel to the longitudinal direction of the specimen, observe the cracking state on the cross section, and determine the number of cracked specimens and the maximum crack depth. evaluated. The results are shown in Table 2.

In No. 10 SUS316L, cracks occurred in four of the seven test pieces, and the maximum crack depth was 62 μm. In Nos. 1, 2, 4, and 7, no cracks were observed. From this, it is considered that the material of the present invention has good stress corrosion cracking resistance.
Next, the precipitation state of TaC was investigated. Here, material No. 1 was used. This material was subjected to 40% cold rolling, followed by solution heat treatment by heat treatment at 1050 ° C., 1100 ° C., and 1150 ° C. for 30 minutes and water cooling. After solution treatment, each of stabilization heat treatment, sensitization heat treatment, stabilization heat treatment + sensitization heat treatment was performed. Moreover, the material as a solution was also prepared. Here, the stabilization heat treatment conditions were air cooling at 900 ° C. for 2 hours. The sensitizing heat treatment was air-cooled at 650 ° C. for 2 hours. First, the amount of Ta contained in the precipitates of these materials was investigated. A test piece of 10 × 50 × 10 mmt was taken from each heat-treated material. After electrolyzing it with a constant current electrolysis method, the residue was collected with a 0.2 μm filter, and the Ta concentration was analyzed with an emission spectroscopic analysis (ICP) method.
FIG. 8 shows the relationship between the amount of Ta in the precipitate and various heat treatment conditions. From this figure, it is considered that part of Ta is precipitated as TaC in the solution solution at 1050 ° C., but most of Ta is dissolved in the material in the solution solution at 1100 ° C. and 1150 ° C. When the stabilization heat treatment is performed, Ta in the material forms TaC and precipitates, so the Ta concentration in the precipitate increases. The sensitizing heat treatment hardly changes the Ta concentration in the precipitate, and the sensitizing heat treatment only forms Cr carbide, and it is considered that TaC is not formed.
FIG. 9 shows the proportion of Ta that forms precipitates out of Ta contained in the material. While in solution and in solution plus sensitization, 10 to 20% of TaC was precipitated at 1050 ° C., but hardly precipitated at 1100 ° C. and 1150 ° C. On the other hand, when the stabilization heat treatment was performed after the solution heat treatment, 30% or more was precipitated as TaC, although there was a difference depending on the solution temperature. However, a little less than 60% was deposited at the maximum. In order to evaluate the sensitization state of the material subjected to such heat treatment, the reactivation rate was measured according to JIS G 0580.
The result is shown in FIG. From this result, it can be seen that, after the solution treatment, when the sensitizing heat treatment is performed, all the test pieces are in a slightly sensitized state. On the other hand, when a stabilization heat treatment is performed after solution treatment, the reactivation rate increases as the solution temperature rises, but the solution was not sensitized at any solution temperature. This reactivation rate test result correlates with the proportion of precipitated Ta, and it can be seen that thermal sensitization can be suppressed by precipitating TaC by stabilizing heat treatment. When stabilizing heat treatment is performed to precipitate TaC, thermal sensitization can be suppressed and strength can be improved by precipitation strengthening as shown in FIGS. Therefore, the number density of TaC precipitates in the material subjected to the stabilization heat treatment was compared with that of the solution heat treatment material. The precipitate was observed as follows. First, after lightly etching the mirror-polished test piece surface, carbon deposition is performed to fix the precipitate to the carbon film. Thereafter, the base material is eluted, and the carbon film on which the precipitate is fixed is peeled off. The carbon film on which the precipitate is fixed is sufficiently washed and dried to obtain a specimen for observation of the precipitate. This test piece is observed with a field emission type (FE) STEM (Scanning Transmission Electron Microscopy). In this way, the state of the precipitate was investigated, particularly focusing on the size and number density. Here, the precipitate on the carbon film is observed as a precipitate existing on the surface of the test piece. In order to measure the number density of the precipitates, 50 visual fields were observed with FE-STEM at a magnification of 50,000 times so as not to be biased, and precipitates of 10 nm or more were evaluated.
The result is shown in FIG. As a result, with regard to the solution material, TaC precipitates were recognized at a certain number density in the case of 1050 ° C. solution formation, but were hardly observed in the 1100 ° C. and 1150 ° C. solution treatment material, and the number density was also extremely small. On the other hand, in the solution-treated / stabilized heat-treated material, the higher the solution temperature, the higher the number density of precipitates was observed. Since it is shown in FIG. 10 that the solution + stabilized heat treatment material is not heat-sensitized and the strength can be improved by the stabilization heat treatment, the number density of TaC precipitates is 1 × 10 ¹² m ⁻² or more. Is considered preferable.

As an example of an embodiment of the present invention, an application example to a control rod having the fastest neutron irradiation damage speed among reactor internal structures and equipment will be shown.
FIG. 12 shows a control rod using boron carbide as a neutron absorber. This control rod mainly has a tie rod 10, a handle 9, a connector 7, a sheath 6, and a neutron absorber rod 5, all of which are made of austenitic stainless steel.
Since the control rod has a gap, there is a possibility of crevice corrosion, and there is also a possibility of generating irradiation-induced stress corrosion cracking. Therefore, by using the austenitic stainless steel of the present invention, which is considered to be excellent in crevice corrosion resistance and stress corrosion cracking resistance, and also effective in suppressing irradiation damage, control rod crevice corrosion and radiation induced stress corrosion Cracks can be suppressed, and a long-life and highly reliable control rod can be obtained. In addition, since the austenitic stainless steel of the present invention can be manufactured as members of various shapes such as rods, thin plates, and tubes, not only control rods using boron carbide but also control using hafnium as a neutron absorber. It can also be applied to bars.
Conventionally used austenitic stainless steels in nuclear reactors may cause irradiation-induced stress corrosion cracking when neutrons with energy of 0.1 MeV or more are irradiated by 0.5 × 10 ²¹ n / cm ² or more. is there. Therefore, not only control rods, but also reactor internal structures and equipment such as boiling water reactor core shrouds, upper lattice plates, core support plates, etc., baffle plates, former plates, baffle formers for pressurized water reactors, etc. -By using the austenitic stainless steel of the present invention for a bolt or the like, a nuclear reactor and a nuclear power plant having excellent long-term reliability can be obtained.

1: Crevice corrosion test piece 2: Spot welding 3: Roller 4: Cooling hole 5: Neutron absorber rod 6: Sheath 7: Connector 8: Connector / socket 9: Handle 10: Tie rod

Claims (6)

C is 0.03 to 0.04% by weight, Mn is 1.0 to 2.0% by weight, Ni is 9.0 to 13.0% by weight, Cr is 17.0 to 20.0% by weight, Ta is 13 times or more of C, 0.50 to 0.65% by weight, and the balance In austenitic stainless steel consisting of Fe and inevitable impurities,
Austenitic stainless steel characterized in that more than 30% of Ta is precipitated as TaC. ^{2. The} austenitic stainless steel according to claim 1, wherein among TaC particles precipitated on the stainless steel cross section, the average number density of particles having an average particle diameter of 10 nm or more is 1.0 × 10 ¹² particles / m ² or more.   The austenitic stainless steel according to claim 1, having a grain size number of 4.0 or more and a tensile strength at 100 ° C of 470 MPa or more.   2. The austenitic stainless steel according to claim 1, wherein the electrochemical reactivation rate is 30% or less. A reactor internal structure comprising the austenitic stainless steel according to any one of claims 1 to 4 and irradiated with neutrons having an energy of 0.1 MeV or more of 0.5 × 10 ²¹ n / cm ² or more. . 6. The reactor internal structure according to claim 5, wherein the reactor internal structure is a control rod.