JPS6211058B2 - - Google Patents

Description

[Detailed description of the invention]

The present invention is a Ni-based alloy with excellent stress corrosion cracking resistance that is used in gap structure members for nuclear reactors in which gaps are formed in high-temperature, high-pressure pure water and under high stress in contact with other members. The present invention relates to manufactured parts, and particularly to those suitable for use in springs and bolts for nuclear reactors. Conventional springs for nuclear reactors have been produced primarily by subjecting them to solution treatment after melting, followed by cold rolling to give them 30% to 40% processing, and then subjecting them to aging treatment. These are manufacturing methods that use cold working and aging to ensure the high-temperature strength and spring properties required for spring materials for nuclear reactors, but stress corrosion resistance in the crevices is a particular problem when applied to actual equipment. There was a drawback in that cracking resistance (hereinafter referred to as SCC resistance) had not always been sufficiently studied. In particular, reactor springs (e.g. C-springs in BWR control rod drive mechanisms) are often used in locations where gaps are formed under high stress action, and SCC
Countermeasures are important. Currently, Inconel (trade name) is used as a spring for nuclear reactors because it has high strength and high corrosion resistance.
×750 is used, and most of them are produced by applying 30 to 40% cold working after solution treatment, followed by aging treatment (direct aging).
The purpose of cold working after solution treatment is to refine the crystal grains, but on the other hand, the aging treatment that follows cold working contributes to improving the spring properties and high-temperature strength of spring materials for nuclear reactors. be. However, even if 30 to 40% cold working is performed between solution treatment and aging treatment, it is unclear whether 30 to 40% cold working is effective for improving clearance SCC resistance, which is a practical problem. It has not been sufficiently considered. Further, although JP-A-54-69517 mentions intergranular corrosion of solid solution strengthened alloys, it does not mention SCC resistance of precipitation strengthened alloys. The interstitial SCC susceptibility of precipitation strengthened alloys is different from that of solid solution strengthened alloys, and no correlation between intergranular corrosion and SCC is observed. The present invention solves these conventional problems and improves durability.
The object of the present invention is to provide a Ni-based alloy member used for a gap structure member for a nuclear reactor that has excellent SCC properties. The present inventors changed the degree of cold working carried out between the solution treatment and the aging treatment, and investigated the effect of the degree of cold working on the gap SCC resistance using constant strain tests with gaps in high-temperature, high-pressure pure water. This was investigated from both aspects of microscopic tissue observation. At the same time, a similar study was conducted regarding aging treatment conditions (direct aging and two-stage aging). As a result, we discovered the following new facts. If 10 to 30% cold working is applied between solution treatment and direct aging treatment, the gap SCC susceptibility is greatly increased. However, if 60% cold working is applied, the gap SCC resistance is improved, but in this case, it is difficult to determine the magnitude of the gap SCC susceptibility from the metal structure. 10 ~ between solution treatment and two-stage aging treatment
When 20% cold working is applied, the gap SCC susceptibility is noticeable, but when the cold working is 25% or more, the gap SCC susceptibility decreases markedly. In this case, the relationship between the degree of cold work and clearance SCC susceptibility can be easily determined from the metallographic structure. In other words, primary recrystallized grains are observed in cold working of 25% or more, where the gap SCC susceptibility is extremely small, whereas in cold working of 10 to 20%, where the gap SCC susceptibility is large. , that is not accepted. From this, the high-temperature underwater gap SCC that occurs in the two-stage aged material is different from that in the directly aged material.
It is presumed that the mechanism of occurrence of Cr is different, and it was found that it is largely related to the primary recrystallization behavior, apart from the Cr-deficient type caused by the precipitation of Cr carbides. The present invention has been made based on such knowledge, and the first invention is based on Cr14 to Cr14 in weight%.
25%, Fe30% or less, Al0.2~2%, Ti0.5~3%,
An alloy consisting of 0.7 to 4.5% Nb, the balance Ni and unavoidable impurities, which is subjected to solid solution treatment, then subjected to cold plastic working with an area reduction rate of 25% or more, and then a primary recrystallized structure is formed. It is precipitation-strengthened by heating at a high temperature and then subjected to a two-stage aging treatment at a lower temperature, and has a primary recrystallized structure, and has excellent stress corrosion cracking resistance for use in nuclear reactor gap structure members. The gist is a member made of a Ni-based alloy. Preferably, it is substantially free of Cr carbides. In addition, the second invention has Cr14 to 25% in weight%,
Fe30% or less, Al0.2-2%, Ti0.5-3%, Nb4.5
~8% Mo, less than 8% Mo, the balance Ni and unavoidable impurities, and is solution treated, then subjected to cold plastic working with an area reduction rate of 25% or more, and then a primary recrystallized structure is formed. Heat to a temperature that
Ni-based alloy members with excellent stress corrosion cracking resistance used in nuclear reactor gap structure members that are precipitation-strengthened through two-stage aging treatment at a lower temperature and have a primary recrystallized structure. The main points are as follows. Cr: Cr needs to be at least 14% for SCC resistance, while more than 25% impairs hot workability and inhibits cold workability due to the formation of a harmful phase known as the TCP phase. The content was set at 14 to 25% because the properties, mechanical properties, and corrosion resistance deteriorate. Fe: Fe stabilizes the base structure and improves corrosion resistance. However, if the Fe content is too high
30% ≧ Fe to generate harmful phases such as Laves phase.
And so. Al, Ti, and Nb all form intermetallic compounds with Ni and contribute to precipitation strengthening. At least 0.2% Al and 0.5% to give age hardenability
The above combination of Ti is necessary, and by increasing the amounts of Al and Ti and adding Nb, a high-strength alloy suitable for the purpose can be obtained. On the other hand, if the content is too high, the properties will deteriorate, so Al should be 0.2~
2%, Ti is 0.5-3%, and in the first invention, Nb is 0.7-4.5%. Moreover, in the second invention, Mo is further contained. Mo coexists with Cr to ensure sufficient SCC resistance. However, if the amount of Mo is too large, workability deteriorates and harmful phases that impair corrosion resistance and mechanical properties are likely to be generated, so the Mo content was set to 8% or less. Also in the second invention, Nb has a greater effect on precipitation strengthening than Al or Ti, and it is necessary to add Nb to obtain the high hardening ability required for springs, but if too much Nb is added, coarse carbides and Deterioration of mechanical properties and reduction in workability may occur due to the formation of intermetallic compounds. Therefore, in the second invention
Nb was set at 4.5 to 8%. In the present invention, the temperature for the solution treatment is preferably 950 to 1150°. The solution temperature is important, and it is necessary to maintain a temperature range that provides good SCC resistance and prevents coarsening of crystal grains. In addition, as a two-stage aging treatment, after holding at 800 to 900°C for 1 to 30 hours, cooling, and then at 600 to 750°C for 1 to 30 hours.
It is preferable to carry out the treatment for 30 hours. Therefore, the present invention is particularly suitable for springs and bolts for nuclear reactors. Figure 1 shows various springs for nuclear reactors.
This shows the shape of the bolt. This spring, bolt, and method of manufacturing the same will be explained below. FIGS. 1a and 1b show an expansion spring 12 used to attach a graphite seal 11 to the inner wall of an index tube 10. FIG. This expansion spring 1
2 is a band-shaped ring having a notch 13, and has a width of 10 mm and a diameter of 60 mm. This expansion spring 12 is produced by melting, rolling, solution treatment, cold rolling, forming (processing of 25% or more), and two-stage aging. FIGS. 1c and 1d show a garter spring 22 for attaching a graphite seal 21 to a piston tube 20. FIG. This garter spring 22 has a coil shape, and the length of the coil portion is
166mm, core wire diameter 0.36mm. This is manufactured by melting, solid solution treatment, wire drawing, coiling (at this time, 25% or more processing is added), and two-stage aging treatment. 1e shows a spring 32 provided between the tie plate 30 and the channel box 31, and FIG. 1f shows a spring 41 provided in the cap screw 40. These springs 32 and 41 are manufactured in the same manner as the expansion spring 12 shown in FIGS. 1a and 1b, respectively. In FIG. 1f, 42 is a guard. After being forged or rolled, the cap screw 40 is solid-solutionized, and threaded by rolling or machining. It is then aged in the two stages mentioned above. An embodiment of the present invention will be described below. The materials are Inconel x 750 (product name) and Inconel 718 (product name). Its main chemical components are Inconel x
750, 72.92% Ni, 15.48% Cr, 6.91% Fe, 0.57
%Al, 2.60%Ti, 0.95%Nb+Ta, 0.04%C,
Inconel 718 52.47% Ni, 18.37% Cr, 0.40%
Al, 0.85% Ti, 5.06% Nb+Ta. Table 1 shows the results of a constant strain test with a gap in high-temperature, high-pressure pure water. The test conditions are as follows. Test temperature: 288
°C, pressure: 86Kg/cm ² , dissolved oxygen: 8ppm, gap forming material: graphite fiber wool, strain: approximately 1.0%, test time: 500 hours.

【table】

[Table] Inconel x 750, solid solution treatment (1066℃
x 1h → water cooling) and direct aging treatment (704℃ x 4 or
When cold working with a cross-section reduction rate of 10 to 60% is performed between 20h → air cooling), the susceptibility to gap SCC is large at 10%, 20%, and 30% cold working, and when the cold working is 30% or less, It is recognized from Table 1 that machining is harmful to the gap SCC property. Even in the case of direct aging treatment, when the degree of cold work reaches 60%, the gap SCC property becomes significantly smaller, but the work becomes stronger. Note that it is difficult to judge the degree of crevice SCC susceptibility of aged materials directly from the metallographic structure. Solid solution treatment (1066℃ x 1h → water cooling) and two-stage aging treatment (843℃ x 24h → air cooling + 704℃ x 4 or 20h
→ Air cooling) When the same cold working as described above is applied, the susceptibility to gap SCC is large at 10% and 20% cold working, but the gap SCC resistance is low at cold working of 30% or more. Significantly improved. Figures 2 and 3 are Inconel x 750 gaps.
This is a micrograph showing the relationship between SCC _{susceptibility} and metallographic structure.
It is corroded by immersion in H ₂ SO ₄ ). Second
The figure shows the one subjected to direct aging treatment, and the one shown in Fig. 3 is the one subjected to two-stage aging treatment. 60%
It is. In the case of direct aging, the grain size becomes finer as the degree of cold working increases, but there is no correlation between crevice SCC susceptibility and metallographic structure. On the other hand, in the case of two-stage aging treatment, only intergranular corrosion is observed in metal structures that show a high susceptibility to interstitial SCC, whereas metal structures that show a significantly low susceptibility to interstitial SCC show intergranular corrosion. A large number of primary recrystallized grains are observed. No Cr carbide is seen. This result shows that the gap SCC in high-temperature water has a good correlation with the primary recrystallization behavior, and in order to improve the gap SCC resistance of this alloy, the final This shows that it is effective to make the metal structure a primary recrystallized structure. This primary recrystallization is determined by the degree of cold working and intermediate heat treatment conditions, progresses as the degree of cold working increases, and significantly improves the gap SCC resistance. For these reasons, when manufacturing nuclear reactor springs and bolts with excellent gap SCC resistance, it is practical to perform cold working of 25% or more between solution treatment and two-stage aging treatment. It was found to be both powerful and beneficial. In this case, we have also discovered that the superiority or inferiority of the gap SCC resistance can be easily determined from the metal structure (primary recrystallization behavior). In the case of Inconel 718, the above-mentioned Inconel x 750
As in the case of , it was found that if the metal structure after the final aging treatment was made into a recrystallized structure, the gap SCC resistance would be extremely good. As described above, according to the present invention, by performing cold plastic working with a cross-section reduction rate of 25% or more between the solution treatment and the two-stage aging treatment, it has a precipitation-strengthened primary crystal structure and is resistant to crevices. It is possible to provide a Ni-based alloy member used for a gap structure member for a nuclear reactor that has even better SCC properties. Furthermore, the gap SCC resistance of Ni-based alloy members used in nuclear reactor gap structure members manufactured by cold plastic working between solution treatment and two-stage aging treatment is determined by the metallographic structure. It can be easily determined from the recrystallized structure. As described above, according to the present invention, it is possible to provide a nuclear reactor gap structure member with excellent SCC resistance.

[Brief explanation of the drawing]

Figures 1a to 1f are schematic diagrams showing the shape of a nuclear reactor spring, and Figures 2a to d and 3a to 3d are microscopic photographs of the metal structure of the Ni-based alloy. 10... Index tube, 12... Expansion spring, 20... Piston tube, 22... Garter spring, 30... Tie plate, 32... Spring, 40... Cap screw.

Claims (1)

[Claims] 1% by weight: Cr14-25%, Fe30% or less, Al0.2
~2% Ti, 0.5~3% Ti, 0.7~4.5% Nb, the balance Ni and unavoidable impurities, and is solution treated and then subjected to cold plastic working with an area reduction rate of 25% or more. A gap structure member for a nuclear reactor, characterized in that it is precipitation-strengthened by heating at a temperature at which a primary recrystallized structure is formed, and then subjected to a two-stage aging treatment at a lower temperature, and has a primary recrystallized structure. A Ni-based alloy member with excellent stress corrosion cracking resistance used in 2 Weight%: Cr14-25%, Fe30% or less, Al0.2
~2% Ti, 0.5~3% Nb, 4.5~8% Nb, 8% Mo or less, the balance Ni and unavoidable impurities, and is solution treated and then has a cross-section reduction rate of 25%.
It is characterized by being subjected to the above cold plastic working, further heated at a temperature at which a primary recrystallized structure is formed, and then precipitated strengthened by being subjected to a two-step aging treatment at a lower temperature to have a primary recrystallized structure. A Ni-based alloy member with excellent stress corrosion cracking resistance used in gap structure members for nuclear reactors.