Austenitic Cr-Ni-Mn-steel excellent in resistance to neutron irradiation embrittlement

Abstract: An austenitic steel excellent in resistance to neutron irradiation embrittlement which contains, by weight, not more than 0.03% carbon, not more than 1% silicon, 5 to 25% manganese, 15 to 26% chromium, and 10 to 20% nickel, the ratio of atomic volume of chromium to the average atomic volume of matrix of the steel being from 0.900 to 1.030. It is possible to add to the austenitic steel, besides the above-mentioned alloying elements, at least one element selected from the group consisting of niobium, titanium, zirconium, tantalum and vanadium which are effective in corrosion resistance and irradiation embrittlement under neutron irradiation in total amounts of not more than 1.0%. At least one of components composing the inside of a nuclear reactor or nuclear fusion reactor is made of the austenitic steel.

Description: This invention relates to new austenitic steel for making members which are irradiated with radio-active rays and are in contact with high-temperature, high-pressure water and, more particularly, to an austenitic steel which is advantageous especially in intergranular corrosion resistance and new nuclear reactors and nuclear fusion reactors made of this steel. More in particular the invention relates to a new austenitic steel used in neutron irradiation conditions according to the preamble of claim 1. At present, structural members irradiated with neutrons in the core of a light-water nuclear reactor are made of the stainless steels SUS 304 and 316 specified in JIS. The chromium contained in stainless steels contributes to the improvement in intergranular corrosion resistance. However, neutron irradiation at high temperatures forms layers deficient in chromium near the grain boundaries; this results in the formation of what is called a sensitized structure, which increases sensitivity to intergranular corrosion or boundary-type stress corrosion cracking. Methods for reducing or preventing intergranular corrosion and boundary-type stress corrosion cracking (hereinafter collectively referred to as intergranular corrosion) in steel by improving the properties of austenitic stainless steels, are disclosed in Japanese Patent Examined Publication No. 1-18143, Japanese Patent Unexamined Patent Publication Nos. 62-238355 and 62-238353. These methods involve reducing the carbon contained in stainless steels and adding carbide stabilizing elements, such as niobium and titanium, to prevent the precipitation of chromium carbides. The above-mentioned conventional arts were developed to suppress the formation of layers deficient in chromium near grain boundaries (what is called the sensitized structure), which is caused by the precipitation of chromium carbides, and to improve intergranular corrosion resistance thereby. In particular, these techniques are intended for stainless steels used in structural members of a core of a nuclear reactor and the first wall of a nuclear fusion reactor. However, it has become evident that when stainless steels which have undergone solution treatment are irradiated with neutrons, layers deficient in chromium are formed near the gain boundaries along with the knocking-out of atoms from the base metal. Layers deficient in chromium are formed by neutron irradiation according to the following mechanism: (1) Atoms of the base metal are knocked out into steel by neutron irradiation, and pairs of an interstitial atom and a vacancy are formed. (2) Both the interstitial atom and the vacancy move through the steel. During this movement, they join a solute atom, causing selective diffusion of the solute atom. (3) In the stainless steels of JIS SUS 304 and 316, neutron irradiation causes chromium to move away from grain boundaries, generating layers deficient in chromium at grain boundaries. Further, the improvement of corrosion resistance of the JIS SUS 304 stainless steel for structural members of the core of a nuclear reactor is disclosed in Japanese Patent Unexamined Publication No. 62-107047 as follows: (1) Neutron irradiation causes silicon and phosphorus to segregate at grain boundaries, reducing corrosion resistance. (2) Neutron irradiation forms chromium carbides near grain boundaries, sensitizing the structure. To prevent these phenomena, the following are disclosed in the above patent publication: 1) The silicon and phosphorus contents are lowered. 2) The formation of chromium carbides is suppressed by adding elements that form stable carbides, such as molybdenum, niobium and titanium. It is, however, impossible to prevent the abovementioned decrease in intergranular corrosion resistance caused by sensitization due to layers deficient in chromium formed by neutron irradiation, only by controlling concentrations of impurities. In the above conventional techniques, consideration is not given to the phenomenon that layers deficient in chromium are formed at grain boundaries due to the knocking-out and diffusion of chromium by neutron irradiation, and corrosion resistance in water at high temperature poses a problem. EP-A- 342 574 discloses an entirely austenitic steel consisting of up to 0.04 % C, up to 0.89 % Si, 5.4 to 8.9 % Mn, 15.1 to 30 % Cr, 10.1 to 24.9 % Ni, 2.01 to 7 % Mo, 0.31 to 0.8 % N, balance Fe and impurities, with high corrosion resistance as a material used under conditions in which the material is exposed in a high mechanical load and various acids. In such environments, the material is required to have excellent corrosion resistance in its entire surface and high 0.2 %-yield-strength. In the austenitic steel, Mn is added to improve 0.2%-yield-strength and wear resistance. In contrast, the primary object of the present invention is to improve stainless steel, which is used in high-temperature and high-pressure pure water under neutron irradiation, in properties of resistance to stress corrosion cracking (SCC) and corrosion resistance. EP-A 142 015 discloses an austenitic steel consisting of up to 0.03 % C, 0.1 to 2 % Si, 8 to 15 % Mn, 15 to 30 % Cr, 12 to 20 % Ni, 3.5 to 10 % Mo, 0.35 to 0.55 % N, balance Fe and impurities, which is used in an environment of sea water or another cooling water in nuclear reactor systems. The steel is extremely resistant to pitting and crevice corrosion. The steel is developed specifically for and intended for use within the off-shore industry, e.g. in piping and tanks, and in other environments where the steel is exposed to sea water or to other chloride-containing liquids. A discussion is directed to improving corrosion resistance property of steels used in the above-mentioned environments, which are required to have excellent properties of corrosion resistance in entire surface and of pitting resistance. According to the discussion, Mn is advantageously added to the austenitic steel as a skilled person knows. DE-B-1 194 587 discloses the use of austenitic steel alloys consisting of up to 0.07 % C, up to 0.7 % Si, 4 to 12 % Mn, 22 to 15 % Cr, 16 to 6 % Ni, 0.23 to 0.4 % N, up to 3 % Mo, balance Fe and impurities, for providing non-magnetic steels, e.g. structural materials applied to submarine boats, having excellent property of stress-corrosion-cracking (SCC) resistance in environments where the material is exposed to sea water or to other chloride-containing liquids and having excellent mechanical strength, and mentions that Mn contributes to the steels in increasing mechanical property even if the steels are not subjected to cold rolling, but is silent about the positive relationship between Mn in austenitic steels and improvement of SCC resistance. SUMMARY OF THE INVENTION It is an object of the present invention to provide an austenitic steel used under neutron irradiation in which manganese, chromium and nickel contents are adjusted so that chromium content at grain boundaries is not virtually below chromium content of matrix due to neutron irradiation. It is another object of the present invention to provide structures and devices in the core of a light-water nuclear reactor, heavy-water nuclear reactor and nuclear fusion reactor made of an austenitic iron-base alloy which is excellent in stress corrosion cracking resistance in an environment irradiated with not less than 10¹⁹ n/cm² of fast neutrons or equivalent corpuscular beams and in contact with high-temperature water. The first object is achieved, according to the present invention, by the features of claim 1. The second object is achieved, according to the present invention, by the features of claims 3 and 4, respectively. BRIEF DESCRIPTION OF THE DRAWINGS Figure 1 is a graph showing the chromium contents near grain boundaries of steels after electron irradiation; Figure 2 is a graph showing the relationship between the volume size factor (VSF) and change in chromium content at grain boundaries; Figure 3 is a partially cutaway perspective view of a core of a boiling water reactor; and Figure 4 is a sectional view of a nuclear fusion reactor. DETAILED DESCRIPTION OF THE DRAWING The reasons for the addition of each element of the austenitic steel of the present invention will be described in the following. Because the atomic volume of manganese is larger than those of iron, nickel and chromium, manganese increases the average atomic volume of matrix. Therefore, addition of manganese increases the average atomic volume of matrix and prevents movement of chromium away from grain boundaries due to corpuscular beam irradiation. Although manganese serves to stabilize austenite structure like nickel, it should not be added excessively, because excessive addition results in worsening of workability. Therefore, the manganese content selected is from 5 to 25% as a range which poses no problem with workability. The manganese content is preferably from 7 to 20% and is especially preferably from 10 to 16%. Chromium serves to improve corrosion resistance at grain boundaries. In JIS SUS 304 and 316 steels, irradiation flaws are formed in matrix due to neutron irradiation, and along with diffusion of the flaws chromium diffuses from near the grain boundaries into grains, with the result that chromium depleted layers are formed near the grain boundaries. As a consequence, chromium content near the grain boundaries decreases to below 12%, which is the lower limit of corrosion resistance, and this results in a decrease in intergranular corrosion resistance. Therefore, corrosion resistance at grain boundaries under neutron irradiation can be maintained if chromium content is higher than in the JIS SUS 304 and 316 steels and the average atomic volume of matrix is adjusted so that the ratio of the average atomic volume of the matrix to the atomic volume of chromium is not more than 1.030 to ensure that chromium content at grain boundaries is virtually not lower than the chromium content of matrix. To obtain this effect, it is necessary that chromium content of the alloy be from 15 to 26%. However, because chromium makes the structure of austenite unstable, it is desirable to consider the correlation to the manganese which contributes to the stabilization of austenite. Further, it is desirable that the iron content be not less than 50% so as to form an iron-base alloy. It is desirable that total content of chromium and manganese be not more than 45%. Especially, total content of these elements is preferably from 30 to 41%, with chromium content preferably ranging from 20 to 26%. Nickel serves to stabilize austenite structure in an alloy. It is desirable that the alloy used under neutron irradiation have a stable austenite phase from a standpoint of irradiation embrittlement. Therefore, desirable nickel content is from 10 to 20% so that both the above-mentioned effect of the adjustment of the average atomic volume of matrix and effect of stabilization of austenite structure are achieved. The nickel content is especially preferably from 10 to 15%. Iron is a base element of the alloy of the present invention and its desirable content is from 40 o 70% from the results of use of core materials. The iron content is preferably from 45 to 65% and is especially preferably from 48 to 60%. Titanium, niobium, tantalum, hafnium, vanadium and zirconium are effective in corrosion resistance under neutron irradiation and irradiation embrittlement resistance. All of these elements have larger atomic radii than iron, nickel, chromium and manganese. In the alloy containing these elements, therefore, the average atomic volume of matrix increases and VSF-value decreases. As a result, chromium content at grain boudnaries does not decrease even after irradiation and the alloy shows excellent corrosion resistance and phase stability. These elements form especially fine carbides, preventing precipitation of chromium carbides and a decrease in chromium content near grain boundaries. However, in consideration of the limit to dissolution of these elements to the Fe-Ni-Or-Mn alloy and to ensure good effects of addition of these elements, it is preferable to add one or two or more elements selected from the group consisting of niobium, titanium, tantalum, hafnium, vanadium and zirconium in total amounts of not more than 1.0% and it is especially preferable to add one or two or more elements selected from the above group in total amounts of from 0.1 to 0.6%. Carbon reacts with chromium upon neutron irradiation and heating to precipitate chromium carbides near grain boudnaries, reducing the chromium at the grain boundaries. Because the formation of chromium carbides lowers the corrosion resistance of grain boundaries and initiates stress corrosion cracking, carbon content should be not more than 0.03%. However, presence of no carbon lowers strength and this is, therefore, undesirable. Therefore, the carbon content is preferably not less than 0.003% and is especially preferably from 0.008 to 0.020%. Silicon serves as deoxidizer. It is contained in the austenitic steel in amount of not more than 1 %, preferably from 0.05 to 1.0 %, more preferably from 0.05 to 0.5 %. Silicon within the above range does not greatly affect on the determination of the ratio of the atomic volume (VSF) of chromium to the average atomic volume of the matrix of the austenitic steel. Further, when the embrittlement by neutron irradiation and the industrial processes are taken into consideration, silicon content is preferably from 0.1 to 0.3 %. The austenitic iron-base alloy of the above composition can be produced by melting (vacuum melting is especially preferable), casting, forging and solution treatment. Coarse precipitated phases, such as carbides and sigma-phase, are formed during such producing. The formation of coarse precipitated phases can be prevented by conducting solution treatment at about 1,050°C followed by cold rolling at reductions of not more than 50% and annealing at temperatures between 950°C and 1,050°C at least twice, whereby an improvement in workability can be accomplished. Further, in the fabrication process of the core structures of a light-water nucelar reactor, heavy -water reactor and nuclear fusion reactor, the formation of coarse precipitated phases can also be suppressed by conducting annealing at temperatures between 950°C and 1,050°C at least twice. Structures and devices in a core of a light-water nuclear reactor, heavy-water reactor and nuclear fusion reactor made of the above-mentioned austenitic iron-base alloy are expected to show better corrosion resistance than those made of the conventional stainless steels JIS SUS 304 and 316 when they are irradiated with not less than 10¹⁹ n/cm² of neutrons in the core and installed in an environment in contact with pure water. In the steel of the present invention, the ratio of the atomic volume of chromium to the average atomic volume of the matrix must be between 0.900 and 1.030. If this ratio is not more than 0.900 on one hand, the manganese content of the alloy is too high, with the result that workability, strength and toughness decrease. On the other hand, if this ratio exceeds 1.030, corpuscular beam irradiation reduces the chromium content at the grain boundaries to below the chromium content of the matrix, causing stress corrosion cracking. This ratio should be between 0.900 and 1.030 to prevent the decrease in the chromium content at the grain boundaries by corpuscular beam irradiation. This ratio is especially preferably between 0.95 and 1.000, and at 1.000 or less the chromium content at grain boundaries can be raised by corpuscular beam irradiation. The atomic volume of matrix can be calculated by multiplying the contents (%) by weight of each of the elements chromium, nickel, manganese and iron to the atomic volumes of chromium, nickel, manganese and gamma-iron as matrix respectively and dividing the total by 100. If there are other added elements, they are similarly calculated. Incidentally, carbon and impurity elements, such as phosphorus and sulfur, may be excluded from the calculation because they are contained in trace amounts. The steel of the present invention preferably has proof stress of 176.6 N/mm² (18 kg/mm²) or more, tensile strength of 480.7 N/mm² (49 kg/mm²) or more, elongation of 40 % or more, reduction of area of 60 % or more, and Vickers hardness of 200 or less at room temperature, and especially preferably has proof stress of from 245.3 to 490.5 N/mm² (25 to 50 kg/mm²), tensile strength of from 539.6 to 784.8 N/mm² (55 to 80 kg/mm²), and elongation of from 40 to 75 %. [Example 1] An examination was made into the chromium precipitation behavior by neutron irradiation of a steel with whole austenite phase which contains, by weight, 0.020% carbon, 15.0% manganese, 10.2% nickel, 23.1% chromium, 0.45% silicon and the balance essentially iron. After vacuum melting, this steel was hot forged and then hot rolled, and was subjected to solution treatment at 1,050°C. After that, cold rolling to a thickness of 0.25 mm and annealing were repeated and finally, solution treatment was conducted by heating to 1,050°C for 15 min. The ratio of the atomic volume of chromium to the average atomic volume of the matrix (volume size factor: VSF) of this alloy determined by the following calculation is 0.966. The atomic radius of gamma-iron, chromium, nickel and manganese is 1.26A, 1.29A, 1.25A and 1.49A, respectively. The volume of one atom of each of these elements was determined, and each volume was multiplied by the content of each element (% by weight) to obtain the sum total, which was divided by 100. The quotient thus obtained was regarded as the average atomic volume of the matrix. Nickel was added in the range of 10 to 20% and the nickel amount added was such that it was immediately above 0% ferrite shown in Schaeffler diagram of nickel equivalent and chromium equivalent so as to form whole austenite phase. Carbon, silicon, phosphorus and sulfur were not taken into account because of their trace amounts. The VSF-value of the conventional steel JIS SUS 304 is 1.057. Figure 1 is a graph showing chromium content near grain boundaries after the above alloy was irradiated with electrons. Irradiation conditions were such that the alloy was irradiated with 1 MeV electrons at an irradiation temperature of 500°C in an amount corresponding to 5 x 10²² n/cm² in terms of neutron irradiation, and neutron irradiation was simulated. After electron irradiation, the chromium content at the gain boundaries was analyzed using a energy dispersion type X-ray analyzer (EDX). The results of irradiation of a commercial steel (JIS SUS 304 stainless steel) were shown for a comparison. As shown in the figure, the conventional steel shows about 3% chromium deficiency (or depletion) from 18% to 15% of matrix at grain boundaries. In the Fe-15%Mn-10%Ni-23%Cr alloy of the present invention, chromium segregation corresponding to about 0.8% higher than chromium content of matrix occurred near grain boundaries. This is because this alloy has a low VSF-value of 0.966, with the result that chromium segregated at the grain boundaries in association with the formation of irradiation defects and diffusion. Incidentally, the chemical composition of JIS SUS 304 is that of No. 11 in Table 1 which will be described later. The results of research into layers deficient in chromium at the grain boundaries due to thermal sensitization by high-temperature water reveal that the critical condition under which intergranular corrosion does not occur is, by weight ratio, (chromium content at grain boundaries)/(chromium content of matrix) > 0.6 Since this ratio exceeds 1.0 in the Fe-15%Mn-10%Ni-23%Cr of the present invention, it is expected that intergranualr corrosion does not occur due to irradiation. At VSF of not more than 1.000, therefore, it might be thought that layers deficient in chromium are not formed and that stress corrosion cracking in high-temperature water does not occur. In the conventional steel JIS SUS 304, however, this ratio becomes not more than 0.6 after long hours of service, and as a result intergranular corrosion occurs. The following experiments were conducted on the various alloys (% by weight) given in Table 1. The VSF-values in the table were obtained in the same procedure as described above. Further, the commercial stainless steels JIS SUS 304 (No. 11 and No. 12) and JIS SUS 316 (No. 13) were used for a comparison. The properties of these specimens were investigated according to the following test methods. Each specimen was prepared in the same manner as mentioned above. (1) Behavior of chromium segregation near boundaries by the electron irradiation as described above (2) After solution treatment at 1,050°C for 30 min, the slow strain rate test (SSRT) was conducted at a low strain rate of 3 x 10⁻⁷/sec in the water of a boiling water type light-water reactor at 288°C and 75 atm. (3) Neutron irradiation was simulated by irradiation with 1 x 10¹⁷ ions/cm² of 20 MeV alpha-rays at room temperature and the SSRT test was conducted under the same conditions as (2) above. (4) Measurement of the stability of the austenite phase In the test (1) above, it is possible to investigate the formation of layers deficient in chromium after irradiation with a large neutron does through a structural change, which is local, however. The test (2) is suited to the evaluation of corrosion resistance in an environment of the core of a light-water reactor, and the test (3) is a test in which the effect of neutron irradiation is added to the function of the test (2). To evaluate corrosion resistance, the fracture after the SSRT was observed under a scanning type electron microscope and the intergranular fractured surface rate was calculated. Those with a lower fractured surface rate than the stainelss steel JIS SUS 304 were judged to be excellent in corrosion resistance. In the test (4) above, the stability of the austenite phase was investigated using a ferrite indicator, and a comparison was made with JIS SUS 304. Table 2 shows the results of the tests (1) to (4) conducted on the steels of the compositions given in Table 1. As shown in Table 1, all of the steels of the present invention No. 1 to No. 9 have low VSF-values of not more than 1.030 and all of the conventional steels have high values of not less than 1.056. Incidentally, carbon, silicon and other impurities are excluded from the calculation of the VSF-value. Because in the steels of the present invention No. 1 to No. 6 the chromium content at grain boundaries increases about 0.5% by weight, the VSF-value is not more than 1.000 although the chromium content at the grain boundaries increases after neutron irradiation. In the steels No. 7 to No. 9, the changes in the chromium are small and resistance to stress corrosion cracking under neutron irradiation can be expected. Therefore, to ensure a positive increase in the chromium content at the grain boundaries, it is desirable that the VSF-value be not more than 1.000. In all of the conventional steels with VSF of not less than 1.05, the chromium content at grain boundaries decreases about 3% and it might be thought that stress corrosion cracking occurs. In the corrosion test before irradiation, all of the steels of the present invention showed good corrosion resistance as good as that of the conventional stainless steel JIS SUS 304. In the corrosion test after irradiation, however, the steels of the present invention, especially No. 1 to No. 6 showed improved corrosion resistance compared with those tested before irradiation. In other words, in the test (3) above, the fractured surfaces of samples after the test were observed. In the steels of the present invention No. 1 to No. 6, the area fraction of intergranular fracture was almost 0%. The area fraction of intergranular fracture of the steels No. 7 to No. 9 was from 10 to 15%, and that of the conventional steels No. 10 to No. 12 was about 30%. It might be thought that the corrosion resistance of grain boundaries improved compared with the test before irradiation because the chromium near grain boundaries was concentrated due to alpha-ray irradiation. The same effect can be expected from neutron irradiation. In the phase stability test, the steels of the present invention show the same stability of austenite phase as the conventional steels such as JIS SUS 304 and 316 stainless steels, and irradiation embrittlement and mechanical properties of the steels of the present invention are expected to be equal or superior to those of the conventional steels. Figure 2 is a diagram showing the relationship between the VSF-values of various alloys and the change in the chromium content at the grain boundaries after irradiation with electron beams or alpha-rays. As shown in the figure, the chromium content at the grain boundaries decreases when the VSF-value exceeds 1.030; it is apparent that the chromium content decreases when the alloy is irradiated with corpuscular beams of larger exposure. Therefore, alloys used in high-temperature water may develop stress corrosion cracking. However, the decrease in the chromium content is small at VSF-value of not more than 1.030 and the chromium content at the grain boundaries is higher than that of the matrix at VSF-value of not more than 1.000; therefore, it is apparent that stress corrosion cracking does not occur. The steels of the present invention showed proof stress of 176.6 N/mm² (18 kg/mm²) or more, tensile strength of 480.7 N/mm² (49 kg/mm²) or more, elongation of 40 % or more, reduction of area of 60 % or more, and Vikkers hardness of 200 or less at room temperature. [Example 2] Figure 3 is a schematic sectional perspective view of the core of a boiling water reactor (BWR). In the figure, the numeral 1 represents a neutron source pipe, the numeral 2 a core support plate, the numeral 3 a neutron counter, the numeral 4 a control rod, the numeral 5 a shroud, and the numeral 6 an upper lattice plate. These structures and devices compose the core of the light-water reactor; they are subjected to a large neutron irradiation and are used in high-temperature, high-pressure water at 288°C and 7 MPa. The chromium content at the grain boundaries increases and corrosion reistance is improved if these structures and devices are made of the austenitic iron-base alloy of the present invention. The same effect can be expected if parts which are used in these structures and devices and are parts other than those shown in Fig. 2 are made of the alloy of the present invention. Further, the same effect can be expected if the alloy of the present invention is used in the structures and devices of the core of water-cooled reactors other than the boiling water type. If the above-mentioned neutron source pipe 1, neutron counter 3, control-rod insertion pipe, and channel box and fuel clad pipe of fuel assembly 7 are made of the alloys No. 1 to No. 6 with VSF of not more than 1.000 shown in Example 1, they provide excellent resistance to stress corrosion resistance by neutron irradiation. These members are obtained from ingots by hot working and the same repetition of solution treatment and succeeding cold working as mentioned above. Further, the core support plate 2, shroud 5, and upper lattice plate 6 are obtained from alloy ingots of No. 1 to No. 6 with VSF of not more than 1.000 by hot working and solution treatment in the same manner as mentioned above. Furthermore, the core of a reactor are composed of the following devices or members, and the alloys of the present invention can also be used in these structures of the inside of the reactor: 8: top-mirror apray nozzle, 9: bent nozzle, 10: pressure vessel cover, 11: pressure vessel flange, 12: measuring nozzle, 13: steam separator, 14: shroud head, 15: feedwater inlet nozzle, 16: jet pump, 17: recirculating water outlet nozzle, 18: steam drier, 19: steam outlet nozle, 20: feedwater sparger, 21: core spray nozzle, 22: lower core lattice, 23: recirculating water inlet nozzle, 24: baffle plate, 25: control-rod guide pipe. Further, the alloys of the present invention can also be applied to Advanced Boiling Water Reactors (ABWRs) provided with internal pumps in the reactors and Pressurized Water Reactors (PWRs). In the inside of ABWR, internal pumps are installed in place of the above-mentioned jet pumps 16 of BWR and the construction of the inside of ABWR is similar to that of BWR in other respects. Therefore, like the above products applied to BWR, the alloys of the present invention are applied to the devices and structures of the inside of this ABWR in the same manner as mentioned above. High safety is obtained by using the alloys of the present invention. [Example 3] Figure 4 is a schematic sectional view of a TOKAMAK nuclear fusion reactor. In the figure, the numeral 31 represents a divertor, the numeral 32 the first wall and cooling panel, and the numeral 33 a vacuum vessel. These structures and devices compose the core of the TOKAMAK fusion reactor. They are so designed as to be irradiated with large amounts of neutrons and various corpuscular beams leaking from plasma and be in contact with the cooling water. If these structures and devices are made of the austenitic iron-base alloys of the present invention, the decrease in the chromium content at grain boundaries under irradiation does not occur and, conversely, the chromium content at grain boundaries increases, whereby corrosion resistance can be improved. The divertor 31, the first wall 32 and vacuum vessel 33 are composed of water-cooled components of the alloys of the present invention. The divertor 31 and the first wall 32 have a construction in which blocks of low atomic numbers (for example, SiC, Si₃N₄, AlN, Al₂O₃ and ceramics) are mechanically or metallurgically joined to the cooled metal surfaces. The alloys of the present invention are also applied to them, and they are made of plate or pipe of the above-mentioned alloys, especially, No. 1 to No. 6 with VSF of not more than 1.000. The nuclear fusion equipment is provided with a toroidal coil and an evacuation device is addition to a toroidal coil 34 (not illustrated). There are also an open magnetic field type and an inertial confinement laser heating type as other types of nuclear fusion equipment, and the alloys of the present invention can also be applied to these types and high reliability can be obtained. As is apparent from the above, according to the present invention, it is possible to improve the corrosion resistance of structures and devices used in an environment which is irradiated with not less than 10¹⁹ n/cm² of fast neutrons or equivalent corpuscular beams and is in contact with high-temperature water or high-temperature, high-pressure water. If the austenitic steel of the present invention is used in the structures and devices of the inside of an nuclear reactor as well as in the vacuum vessel of a nuclear fusion reactor and the structures and devices within the vacuum vessel, high reliability can be obtained and the service conditions can be improved.

Claims: An austenitic steel used in high-temperature and high-pressure pure water under neutron irradiation, which is excellent in resistance to neutron irradiation embrittlement when said steel is irradiated at 500 °C with 1 MeV electron beams in an amount corresponding to 5x 10² ² n/cm² in terms of quantity of neutron irradiation, c haracterized by: a ratio (VSF) of the atomic volume of chromium to the average atomic volume of the matrix of the austenitic steel from 0.900 to 1.030 and which consists of, by weight, not more than 0.03 % carbon, not more than 1 % silicon, 5 to 25 % manganese, 15 to 26 % chromium, 10 to 20 % nickel and optionally at least one element selected from the group consisting of niobium, titanium, zirconium, tantalum, vanadium and hafnium in total amounts of not more than 1.0 %, balance iron and unavoidable impurities. The austenitic steel according to claim 1, wherein the manganese content by weight is 7 to 25 %. The use of the austenitic steel claimed in claim 1 or 2 in a nuclear reactor provided in a pressure vessel with a) a neutron source pipe (1), b) a core support plate (2), c) a neutron counter (3), d) a control-rod (4) insertion pipe, e) a shroud (5), f) an upper lattice plate (6), and g) a clad pipe and h) a channel box for a fuel assembly (7), for at least one of the above-mentioned components a) to h). The use of the austenitic steel claimed in claim 1 or 2 in a nuclear fusion reactor comprising a) a divertor (31) provided with ceramic tiles on plasma side thereof and having a water-cooled construction and b) the first wall (32) provided with ceramic tiles on plasma side thereof and having a water-cooled construction in c) a vacuum vessel (33) having a water-cooled construction, for at least one of the above-mentioned components a) to c).

Inventors: HATTORI SHIGEO, KASAHARA SHIGEKI, KUNIYA JIRO, NAKATA KIYOTOMO, HATTORI, SHIGEO, KASAHARA, SHIGEKI, KUNIYA, JIRO, NAKATA, KIYOTOMO
Applicants: HITACHI LTD, HITACHI, LTD.
Application year: 1990
Main classification: C22C38
IPC codes: G21B1, C22C38, G21C7, G21C3, G21C13