WO2012121390A1 - Matériau pour équipement nucléaire, tuyau chauffant pour générateur de vapeur, générateur de vapeur et centrale nucléaire - Google Patents

Matériau pour équipement nucléaire, tuyau chauffant pour générateur de vapeur, générateur de vapeur et centrale nucléaire Download PDF

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WO2012121390A1
WO2012121390A1 PCT/JP2012/056178 JP2012056178W WO2012121390A1 WO 2012121390 A1 WO2012121390 A1 WO 2012121390A1 JP 2012056178 W JP2012056178 W JP 2012056178W WO 2012121390 A1 WO2012121390 A1 WO 2012121390A1
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Prior art keywords
steam generator
test
heat transfer
content
transfer tube
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PCT/JP2012/056178
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English (en)
Japanese (ja)
Inventor
貴治 前口
孝文 廣
重満 大塚
横山　裕
英仁 三牧
俊介 清水
庄司 木ノ村
岡田　浩一
神崎　学
Original Assignee
三菱重工業株式会社
関西電力株式会社
四国電力株式会社
九州電力株式会社
北海道電力株式会社
日本原子力発電株式会社
住友金属工業株式会社
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Application filed by 三菱重工業株式会社, 関西電力株式会社, 四国電力株式会社, 九州電力株式会社, 北海道電力株式会社, 日本原子力発電株式会社, 住友金属工業株式会社 filed Critical 三菱重工業株式会社
Priority to JP2013503632A priority Critical patent/JP5675958B2/ja
Publication of WO2012121390A1 publication Critical patent/WO2012121390A1/fr

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    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C30/00—Alloys containing less than 50% by weight of each constituent
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C19/00—Alloys based on nickel or cobalt
    • C22C19/03—Alloys based on nickel or cobalt based on nickel
    • C22C19/05—Alloys based on nickel or cobalt based on nickel with chromium
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C19/00—Alloys based on nickel or cobalt
    • C22C19/03—Alloys based on nickel or cobalt based on nickel
    • C22C19/05—Alloys based on nickel or cobalt based on nickel with chromium
    • C22C19/051—Alloys based on nickel or cobalt based on nickel with chromium and Mo or W
    • C22C19/053—Alloys based on nickel or cobalt based on nickel with chromium and Mo or W with the maximum Cr content being at least 30% but less than 40%
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C19/00—Alloys based on nickel or cobalt
    • C22C19/03—Alloys based on nickel or cobalt based on nickel
    • C22C19/05—Alloys based on nickel or cobalt based on nickel with chromium
    • C22C19/051—Alloys based on nickel or cobalt based on nickel with chromium and Mo or W
    • C22C19/055—Alloys based on nickel or cobalt based on nickel with chromium and Mo or W with the maximum Cr content being at least 20% but less than 30%
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C19/00—Alloys based on nickel or cobalt
    • C22C19/03—Alloys based on nickel or cobalt based on nickel
    • C22C19/05—Alloys based on nickel or cobalt based on nickel with chromium
    • C22C19/058—Alloys based on nickel or cobalt based on nickel with chromium without Mo and W
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C38/00—Ferrous alloys, e.g. steel alloys
    • C22C38/001—Ferrous alloys, e.g. steel alloys containing N
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C38/00—Ferrous alloys, e.g. steel alloys
    • C22C38/02—Ferrous alloys, e.g. steel alloys containing silicon
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C38/00—Ferrous alloys, e.g. steel alloys
    • C22C38/04—Ferrous alloys, e.g. steel alloys containing manganese
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C38/00—Ferrous alloys, e.g. steel alloys
    • C22C38/06—Ferrous alloys, e.g. steel alloys containing aluminium
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C38/00—Ferrous alloys, e.g. steel alloys
    • C22C38/18—Ferrous alloys, e.g. steel alloys containing chromium
    • C22C38/40—Ferrous alloys, e.g. steel alloys containing chromium with nickel
    • C22C38/42—Ferrous alloys, e.g. steel alloys containing chromium with nickel with copper
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C38/00—Ferrous alloys, e.g. steel alloys
    • C22C38/18—Ferrous alloys, e.g. steel alloys containing chromium
    • C22C38/40—Ferrous alloys, e.g. steel alloys containing chromium with nickel
    • C22C38/44—Ferrous alloys, e.g. steel alloys containing chromium with nickel with molybdenum or tungsten
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C38/00—Ferrous alloys, e.g. steel alloys
    • C22C38/18—Ferrous alloys, e.g. steel alloys containing chromium
    • C22C38/40—Ferrous alloys, e.g. steel alloys containing chromium with nickel
    • C22C38/48—Ferrous alloys, e.g. steel alloys containing chromium with nickel with niobium or tantalum
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C38/00—Ferrous alloys, e.g. steel alloys
    • C22C38/18—Ferrous alloys, e.g. steel alloys containing chromium
    • C22C38/40—Ferrous alloys, e.g. steel alloys containing chromium with nickel
    • C22C38/50—Ferrous alloys, e.g. steel alloys containing chromium with nickel with titanium or zirconium
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C38/00—Ferrous alloys, e.g. steel alloys
    • C22C38/18—Ferrous alloys, e.g. steel alloys containing chromium
    • C22C38/40—Ferrous alloys, e.g. steel alloys containing chromium with nickel
    • C22C38/52—Ferrous alloys, e.g. steel alloys containing chromium with nickel with cobalt
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22F—CHANGING THE PHYSICAL STRUCTURE OF NON-FERROUS METALS AND NON-FERROUS ALLOYS
    • C22F1/00—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22F—CHANGING THE PHYSICAL STRUCTURE OF NON-FERROUS METALS AND NON-FERROUS ALLOYS
    • C22F1/00—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working
    • C22F1/10—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working of nickel or cobalt or alloys based thereon
    • F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
    • F22—STEAM GENERATION
    • F22B—METHODS OF STEAM GENERATION; STEAM BOILERS
    • F22B1/00—Methods of steam generation characterised by form of heating method
    • F22B1/02—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers
    • F22B1/023—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers with heating tubes, for nuclear reactors as far as they are not classified, according to a specified heating fluid, in another group
    • F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
    • F22—STEAM GENERATION
    • F22B—METHODS OF STEAM GENERATION; STEAM BOILERS
    • F22B37/00—Component parts or details of steam boilers
    • F22B37/02—Component parts or details of steam boilers applicable to more than one kind or type of steam boiler
    • F22B37/10—Water tubes; Accessories therefor
    • F22B37/107—Protection of water tubes
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D1/00—Details of nuclear power plant
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin

Definitions

  • the present invention relates to a material for nuclear equipment, a heat transfer tube for a steam generator, a steam generator, and a nuclear power plant, and particularly suitable for a heat exchanger material used in a pressurized water reactor (PWR).
  • PWR pressurized water reactor
  • a heat transfer tube for a steam generator (SG) used in a pressurized water reactor at a nuclear power plant is used in a high-temperature water environment where the maximum temperature is 300 ° C. or higher. Since heat transfer tubes used in such a high temperature water environment may cause stress corrosion cracking (SCC) in the presence of residual stress, etc., the heat transfer tube materials for steam generators have excellent corrosion resistance. It is essential from the viewpoint of safety and the like to adopt a new material.
  • TT600 alloy NCF600 (TT600 alloy) and its substitute, and has excellent durability against stress corrosion cracking (PWSCC) in primary water.
  • Special heat treatment GNCF690 (TT690 alloy) is used.
  • This TT690 alloy has not been confirmed to be cracked so far, either in an actual machine (pressurized water reactor) or in a PWSCC test for tens of thousands of hours in an accelerated environment. For this reason, it is considered that the problem that the heat transfer tube is damaged due to PWSCC has been eradicated by applying TT690 alloy to the material of the heat transfer tube for the steam generator. Further, even in the secondary system environment of the heat transfer tube for the steam generator of the pressurized water reactor, at present, problems such as corrosion damage in an actual plant using the TT690 alloy have not become apparent.
  • the possibility of becoming an alkaline environment of 9 or more cannot be completely denied.
  • Experimental investigations by the present inventors have revealed that TT690 alloy is also susceptible to intergranular corrosion cracking (IGA) in an alkaline environment where the degree of concentration is significantly increased. This environment does not affect the soundness of the TT690 alloy in the current temperature environment of pressurized water reactors, but it is required to extend the design life to about 80 years under higher cooling water temperature conditions. In the next generation of light water reactors, it will be necessary to use materials that are less sensitive to IGA.
  • IGA intergranular corrosion cracking
  • a method of optimizing the component composition design in the material is known. For example, by adopting a material with an increased Cr content in a Ni-based alloy (Ni—Cr—Fe alloy), the stress corrosion cracking susceptibility and intergranular corrosion susceptibility of heat transfer tubes in the primary and secondary systems of a light water reactor are reduced. Can improve the corrosion resistance. However, if the amount of Cr in the material is too high, the thermal conductivity and plastic workability are lowered, so that the desired heat transfer characteristics cannot be obtained, the mechanical properties are lowered, and the steam generator transmission of a small diameter and thin wall is not achieved. There is a problem that the manufacture of the heat tube itself becomes difficult.
  • Patent Document 1 In order to improve the workability when producing a heat transfer tube for a steam generator used in a pressurized water reactor, Ca and Mg are added to a Ni—Cr—Fe alloy having a high Cr content as a heat transfer tube material, and B, It has been proposed to limit the contents of O, N, and S to a small amount (see Patent Document 1). According to Patent Document 1, it is said that plastic workability is improved by the limitation of such components. However, when the material having the component composition described in Patent Document 1 is used for the heat transfer tube, the workability is improved, but the amount of Cr is not in an appropriate range, so that the heat transfer characteristics are deteriorated when the amount of Cr is excessive. And there is a problem that the mechanical properties are lowered when the Cr amount is too large.
  • the present invention has been made in view of the above problems, and is particularly suitable for a heat generator tube for a steam generator used in a pressurized water reactor, excellent in corrosion resistance and workability, a material for nuclear equipment, a heat generator tube for a steam generator, steam It aims to provide a generator and a nuclear power plant.
  • the present inventors have intensively studied to solve the above problems.
  • the corrosion resistance in the water of the primary system of pressurized water light water reactor is remarkably improved. investigated.
  • the performance exceeds conventional TT600 alloy and TT690 alloy (that is, lower than TT690 alloy). It was found to show (IGA sensitivity).
  • the amount of Cr is excessive, a harmful intermetallic compound phase appears in the metal structure, and there is a problem that mechanical properties deteriorate.
  • the present inventors limit the amount of Cr and Ni in the alloy to appropriate ranges and optimize the heat treatment conditions, thereby reducing the IGA sensitivity without deteriorating mechanical properties.
  • the inventors have found that it can be suppressed, and completed the present invention.
  • the material for nuclear equipment according to the present invention is in mass%, Cr: 34 to 38%, Ni: 44 to 56%, C: 0.015 to 0.025%, Si: more than 0%, 0.5 %: Mn: 0.05 to 0.5%, S: 0.003% or less, P: 0.015% or less, N: 0.05% or less, Ti: 0.5% or less, Al: 0.00%. It is characterized in that it contains 0.5 to 0.5%, and the balance is composed of Fe and inevitable impurities.
  • the nuclear equipment material according to the present invention is subjected to a heat treatment for 30 minutes or less at a temperature of 1050 to 1150 ° C. with respect to the nuclear equipment material having the above composition, followed by water cooling or air cooling, and further 680 to 750 ° C. It can manufacture by the method of air-cooling, after performing the heat processing for 20 hours or less at this temperature.
  • the Cr and Ni contents in the alloy are limited to the above ranges, and more preferably, the heat treatment conditions are optimized to impair mechanical properties and workability.
  • action which can suppress IGA sensitivity is acquired without this.
  • the heat exchanger tube for steam generators according to the present invention has the above-mentioned material for nuclear equipment. According to the heat transfer tube for a steam generator having such a configuration, since the material for nuclear equipment according to the present invention is used, it has high thermal conductivity, suppresses IGA sensitivity, and has excellent corrosion resistance. .
  • a steam generator according to the present invention includes the above-described heat transfer tube for a steam generator. According to the steam generator having such a configuration, since the heat transfer tube for a steam generator according to the present invention is used, it has high thermal conductivity, IGA sensitivity is suppressed, and the corrosion resistance is excellent.
  • the nuclear power plant which concerns on this invention comprises said steam generator.
  • the steam generator according to the present invention since the steam generator according to the present invention is provided, the thermal conductivity and the corrosion resistance are excellent.
  • the secondary of the pressurized water reactor is IGA sensitivity in a high-temperature alkaline environment that can occur on the system side is suppressed, excellent SCC resistance can be ensured, and corrosion resistance is excellent.
  • the steam generator heat transfer tube since sufficient mechanical characteristics and workability can be ensured, it is also possible to manufacture the steam generator heat transfer tube as a thin-walled thin tube, which improves the heat transfer characteristics and productivity.
  • (A) is a graph showing a time-temperature-phase transformation curve from an austenite single phase state when the Cr content is 35%
  • (b) is a Cr content of 40%.
  • 4 is a graph showing a time-temperature-phase transformation curve from a single austenite state in the case of FIG. It is a graph which shows the relationship between content of Cr, and the precipitation time of the ⁇ phase which reduces M 23 C 6 carbide contributing to improvement of corrosion resistance and mechanical properties.
  • FIG. 1 shows an environment simulating the Cr content and the primary system and secondary system of a pressurized water reactor. It is a graph which shows the relationship with the corrosion weight loss below.
  • FIG. 2 is a graph showing the relationship between the content of Cr and Ni and the grain boundary fracture surface ratio in a low strain rate tensile test.
  • FIG. 3 (a) is a graph showing a time-temperature-phase transformation curve from an austenite single phase state when the Cr content is 35%.
  • FIG. 3 (b) is a graph showing a time-temperature-phase transformation curve from the austenite single phase state when the Cr content is 40%.
  • FIG. 4 is a graph showing the relationship between the Cr content and the precipitation time of the M 23 C 6 carbide that contributes to the improvement of corrosion resistance and the ⁇ phase that reduces the mechanical properties.
  • the material for nuclear equipment according to the present invention is applied, for example, as a material for a heat transfer tube for a steam generator (SG) in contact with both primary and secondary environments in a pressurized water reactor (PWR) of a nuclear power plant.
  • the steam generator heat transfer tube is at a high temperature of 300 ° C. or higher, alkali concentration occurs in the gap between the steam generator heat transfer tube and the tube support plate, and the steam generator heat transfer tube has a pH of 9 or more.
  • the material for nuclear equipment which is excellent in corrosion resistance by suppressing intergranular corrosion cracking (IGA) in an alkaline environment and has excellent heat transfer characteristics while ensuring high plastic workability is required. It is like that.
  • the nuclear equipment material of the present invention is, in mass%, Cr: 34 to 38%, Ni: 44 to 56%, C: 0.015 to 0.025%, Si: More than 0%, 0.5% or less, Mn: 0.05 to 0.5%, S: 0.003% or less, P: 0.015% or less, N: 0.05% or less, Ti: 0.5 %, Al: 0.05 to 0.5%, respectively, with the balance being Fe and inevitable impurities.
  • the material for nuclear equipment having the above composition was subjected to a heat treatment at a temperature of 1050 to 1150 ° C. for 30 minutes or less, then water-cooled or air-cooled, and further a heat treatment at a temperature of 680 to 750 ° C.
  • Cr: Chrome 34-38% Cr is an indispensable element for maintaining the SCC resistance, and the effect of remarkably improving the SCC resistance and pitting corrosion resistance is obtained by the protective action of the dense oxide film.
  • IGA resistance and SCC resistance corrosion resistance
  • the Cr content is too high, the plastic workability (hot workability), thermal conductivity, and weldability all decrease. .
  • the present inventors have found that when the Cr content is 34% or more, the corrosion resistance necessary as a nuclear equipment material is secured.
  • the present invention is completed by performing the above-described heat treatment, but the present inventors have found that when the Cr content exceeds 38%, an embrittled phase that deteriorates mechanical properties appears in the course of this heat treatment. I found out. Therefore, the Cr content is specified in the range of 34 to 38% as a region satisfying all of the heat transfer characteristics, corrosion resistance, workability and mechanical properties. The Cr content is preferably 35 to 36%.
  • Ni Nickel 44-56%
  • Ni is an element that is generally effective for improving the corrosion resistance. In particular, the effect of improving the acid resistance and the SCC resistance in high-temperature water containing chloride ions is obtained. If the Ni content is too small, such a corrosion resistance improving effect cannot be obtained.
  • the present inventors have found that when Ni is 44% or more, the corrosion resistance necessary as a nuclear equipment material is ensured, and the content is specified in the range of 44 to 56%.
  • the upper limit of Ni: 56% is not defined from the viewpoint of corrosion resistance and thermal conductivity, but is determined in consideration of the content of other elements.
  • Fe: Iron (remainder) Fe has an effect of improving hot workability in the alloy constituting the material for nuclear equipment of the present invention, but its content is not particularly limited.
  • a nickel-base alloy when Fe exceeds 6%, pitting corrosion resistance under an environment containing chloride ions and crevice corrosion resistance under a weak alkaline environment may be lowered.
  • seawater is not mixed into the secondary water, and the corrosion resistance in an environment containing chloride ions is not so important. For this reason, in this invention, about Fe, after evaluating heat conductivity, weldability (solid-liquid coexistence temperature range), IGA sensitivity, hot workability, etc., as the remainder in the alloy which added other elements Yes.
  • M 23 C 6 carbide M is mainly Cr
  • C necessary for the formation of M 23 C 6 carbide is: This element is necessary from the viewpoint of PWSCC resistance.
  • the content of C is excessive, it appears as a carbide other than M 23 C 6 , which makes the precipitation state of M 23 C 6 carbide an undesirable state from the viewpoint of corrosion resistance.
  • the PWSCC resistance may decrease. Therefore, it is necessary to appropriately define the upper limit of the C content.
  • the present inventors derived the content by which C is completely dissolved and no unnecessary carbide is formed in the following heat treatment at a temperature of 1050 to 1150 ° C. by the method of the calculation state diagram.
  • the C content is specified to be in the range of 0.015 to 0.025%.
  • Si: silicon more than 0% and 0.5% or less Si is an element effective as a deoxidizer during steelmaking, and it is necessary to contain it in a predetermined amount or more. Si also has the effect of reducing intergranular cracking during hot working and improving hot workability. However, if the Si content exceeds 0.5%, the weldability and cleanliness of the alloy are reduced, so this is preferably the upper limit. On the other hand, if the Si content is too small, the deoxidation effect becomes insufficient, so the lower limit is preferably made 0.05%.
  • Si having high oxidizability forms an oxide film preferentially over Cr.
  • the protection by the Cr oxide film is reduced.
  • Si is not added more actively than mixing at the time of deoxidation, and its content is limited to a range of more than 0% and 0.5% or less.
  • Mn Manganese 0.05-0.5% Mn, like Si, is an element that acts as a deoxidizer and must be contained in a predetermined amount or more. Moreover, Mn has the effect of reducing the grain boundary cracking during hot working and improving hot workability. However, if the Mn content exceeds 0.5%, the weldability and cleanliness of the alloy are lowered, so this is preferably made the upper limit. On the other hand, if the content of Mn is too small, the deoxidation effect becomes insufficient, so the lower limit is preferably made 0.05%. For this reason, in the present invention, Mn is not added more actively than mixing at the time of deoxidation, and the content is limited to a range of 0.05 to 0.5% unless there is a special intention. To do.
  • S Sulfur 0.01% or less S is an unavoidable impurity inevitably mixed from pig iron and the like in a normal manufacturing process. Since S is a harmful impurity element that lowers the plastic workability (hot workability), its content is preferably suppressed to 0.01% or less.
  • P Phosphorus 0.015% or less
  • P is an unavoidable impurity inevitably mixed from pig iron and the like in a normal manufacturing process.
  • P has a characteristic that it easily segregates at the grain boundaries, and the segregation causes SCC without thermal sensitization. If P is completely dissolved in the alloy, no significant effect on corrosion resistance is observed, but as the amount of P increases, pitting corrosion sensitivity increases. Thus, since the beneficial effect by containing P cannot be expected, the P content is preferably suppressed to 0.015% or less.
  • N Nitrogen 0.05% or less N has a function of delaying the precipitation of grain boundary carbides in the nickel-base alloy by delaying the diffusion rate of C, and thus may reduce the PWSCC resistance. Further, N is known to be harmful to general SCC resistance, and it is necessary to suppress the content from the viewpoint of ensuring hot workability. N is important as an element that increases the strength and stabilizes the austenite phase by suppressing the formation of the ferrite phase, but it is suppressed to 0.05% or less from the viewpoint of hot workability and SCC resistance. It is necessary. As described above, in the present invention, the upper limit of the N content is limited to 0.05% from the viewpoint of securing strength, hot workability, and SCC resistance. The N content is preferably 0.001 to 0.03%, more preferably 0.001 to 0.01%.
  • Ti Titanium 0.5% or less Ti is effective in improving hot workability by reducing solid solution N by combining with N to become TiN or Ti (C, N). It is an element. Usually, in order to obtain such an effect, the Ti content needs to be 5 times or more the N content. On the other hand, if the amount of Ti exceeds 0.5%, the effect is saturated, so the upper limit was made 0.5%.
  • Al Aluminum 0.05-0.5% Al, like Si and Mn, is an element that is effective as a deoxidizer, but if its content exceeds 0.5%, the cleanliness of the alloy is lowered, so this amount is preferably made the upper limit. However, if the Al content is too small, the deoxidation effect becomes insufficient and the hot workability is lowered, so the lower limit is preferably 0.05%.
  • the content of Al is specified in the range of 0.05 to 0.5%.
  • Heat treatment conditions The nuclear power equipment material of the present invention is subjected to a heat treatment (heating) for 30 minutes or less at a temperature of 1050 to 1150 ° C. with respect to the alloy having the above component composition, followed by water cooling or air cooling, and further 680 to 750 ° C. It is more preferable to manufacture by the method of air-cooling after performing the heat processing for 20 hours or less at this temperature.
  • a heat treatment heating
  • a heat treatment for 30 minutes or less at a temperature of 1050 to 1150 ° C. with respect to the alloy having the above component composition
  • water cooling or air cooling and further 680 to 750 ° C. It is more preferable to manufacture by the method of air-cooling after performing the heat processing for 20 hours or less at this temperature.
  • regulated by this invention is explained in full detail.
  • the alloy having the above component composition is heated at a temperature of 1050 to 1150 ° C. for 30 minutes or less, and then cooled by water cooling or air cooling.
  • a solution treatment is performed.
  • the heating temperature in the solution treatment is preferably in the temperature range of 1050 to 1150 ° C.
  • the heating temperature in the solution treatment is less than 1050 ° C., C is not sufficiently dissolved, so that the above effect is hardly obtained.
  • the heating time in the solution treatment is preferably 30 min or less. Even if this heating time is exceeded, the above effect is saturated.
  • the cooling treatment using water cooling or air cooling means in the solution treatment can be performed using a conventionally known apparatus or the like, but the cooling rate at this time is higher than the normal air cooling conditions, that is, The accelerated cooling condition is more preferable from the viewpoint of maintaining strength and corrosion resistance.
  • an aging treatment is performed by heating at a temperature of 680 to 750 ° C. for a time of 20 hours or less and then air cooling.
  • the heating temperature in this aging treatment is preferably in the temperature range of 680 to 750 ° C.
  • the heating temperature in the aging treatment is less than 680 ° C., it takes a long time to precipitate M 23 C 6 carbide necessary for improving the corrosion resistance, and it becomes difficult to obtain the effect of the aging heat treatment.
  • the heating temperature in an aging treatment exceeds 750 degreeC, the effect is saturated.
  • the heating time in an aging treatment needs to be 20 hr or less.
  • the heating time exceeds 20 hours, the above effect is saturated, and in the alloy having the above composition having a high Cr content, an embrittlement phase such as a ⁇ phase is precipitated and the mechanical properties are deteriorated. Moreover, it is desirable that the lower limit of the heating time in the aging treatment is at least 7 hours.
  • FIG. 4 is a graph showing the relationship between the precipitation time of M 23 C 6 carbide and ⁇ phase during the heat treatment at 700 ° C. with respect to the Cr content in the alloy according to the present invention heat-treated at 1100 ° C. .
  • the Cr content is increased, the time for precipitation of the ⁇ phase shifts to the short time side, and the time for precipitation of M 23 C 6 carbide shifts to the long time side, so that M 23 C 6 carbide necessary for improving corrosion resistance. It is necessary to control the Cr content to 38% or less in order to cause precipitation of slag and not to cause embrittlement phase such as ⁇ phase.
  • the content of Cr and Ni is limited to an appropriate range, and more preferably, the heat treatment conditions are optimized as described above to ensure sufficient mechanical properties and workability.
  • action that IGA sensitivity can be suppressed is acquired.
  • the Cr content is specified in the range of 34 to 38%
  • the Ni content is specified in the range of 44 to 56%.
  • the Cr and Ni contents are appropriately controlled within the above range, and further, heat treatment conditions for this alloy are optimized under the above conditions, while ensuring sufficient mechanical properties and workability. Therefore, the IGA sensitivity can be suppressed, and a heat transfer tube material suitable for the secondary system of PWR can be obtained.
  • Table 1 shows a list of component compositions of test materials for evaluation tests, which will be described in detail in Examples described later.
  • FIGS. 1 (a) and 1 (b) show the relationship between the Cr content and the corrosion weight loss in an environment simulating the primary and secondary systems of a pressurized water reactor (relating to the overall corrosion described below) See Sensitivity Confirmation Test). As shown in FIGS. 1 (a) and 1 (b), it can be confirmed that as the Cr content in the alloy is increased, the corrosion resistance in the PWR primary and secondary environments is significantly improved (in Table 1). Specimens a and b).
  • the graph of FIG. 2 shows the relationship between the Cr and Ni content and the grain boundary fracture surface rate in the low strain rate tensile test (SSRT test). As shown in FIG. 2, it is clear that if Ni is contained at 44% or more and Cr is contained at 34% or more, IGA corrosion resistance in an alkaline environment is suppressed, and corrosion resistance higher than that of a conventional TT690 alloy is obtained ( (See specimens a and b in Table 1).
  • the graph of FIG. 3 shows the M 23 C 6 carbide and mechanical properties that improve the corrosion resistance when held at 700 ° C. after heat treatment at 1100 ° C. while changing the Cr content for the component of the specimen a.
  • the results of evaluating the time for generating the sigma phase that degrades the slag are shown.
  • Table 2 below shows the pitting corrosion potential, which is an index of Cr content and pitting corrosion resistance. As shown in Table 2, when the Cr content is within the specified range of the present invention, the pitting corrosion potential is higher than that of the conventional TT690 alloy, and the pitting corrosion resistance is excellent (Table 2). (Refer to specimens a and b).
  • the Cr content is specified in the range of 34 to 38%
  • the Ni content in the range of 44 to 56%
  • the heat treatment conditions for this alloy are as described above.
  • the nuclear equipment material excellent in corrosion resistance and workability according to the present invention, by controlling the component composition within an appropriate range and optimizing the heat treatment conditions, a high-temperature alkaline environment is achieved. Underlying IGA sensitivity is suppressed, excellent SCC resistance can be ensured, and corrosion resistance is excellent. Moreover, since sufficient mechanical characteristics and workability can be ensured, it is possible to manufacture the heat transfer tube for the steam generator as a small-diameter thin-walled tube, and the heat transfer property is improved and the productivity is improved.
  • the heat transfer tube for a steam generator according to the present invention uses the above-described material for nuclear equipment according to the present invention. According to the heat transfer tube for a steam generator of the present invention, since the material for nuclear equipment is used, it has high thermal conductivity, suppresses IGA sensitivity, and has excellent corrosion resistance.
  • the steam generator according to the present invention uses the steam generator heat transfer tube according to the present invention described above. According to the steam generator of the present invention, since the heat transfer tube for the steam generator is used, it has high thermal conductivity, IGA sensitivity is suppressed, and the corrosion resistance is excellent.
  • the nuclear power plant according to the present invention is provided with the steam generator according to the present invention described above. According to the nuclear power plant of the present invention, since it comprises the steam generator, it has excellent thermal conductivity and corrosion resistance.
  • test materials In this example, first, an alloy having the chemical composition shown in Table 1 is melted by a vacuum melting method, and then hot forging, hot rolling, and cold rolling are performed to obtain plate materials having a thickness of 14 mm and 5 mm. Finished in (sample material). Next, the plate material was subjected to a heat treatment at a temperature of 1100 ° C. and then subjected to a solution treatment by water cooling. Next, after further heat treatment for 15 hours at a temperature of 700 ° C., an aging treatment was performed by allowing to cool. And from each of these specimens, test pieces for evaluating various properties as described below were collected.
  • test materials prepared by the above procedure were subjected to various evaluation tests for items as described below.
  • the collected test piece was immersed in water inside the circulation type corrosion test apparatus, and immersed for 3000 hr or more under primary and secondary environmental conditions. Then, the weight loss was calculated by measuring the mass of each test piece before immersion and after immersion (after descaling), and data on the overall corrosion amount was obtained. At this time, when performing the descaling process, the same process is applied to the blank test piece made of the same material as the test piece, and the amount of weight reduction is measured, thereby correcting the weight loss of the test piece base material due to the descaling process. did.
  • the specimens a and b according to the present invention in which the Cr content is in the range of 34 to 38% have a corrosion weight loss compared to other specimens. It is clear that the corrosion resistance in water of the PWR primary system is remarkably improved with 0.005 to 0.011 (mg / cm 2 ⁇ 3050 hr). In contrast, the comparative examples c to j and the conventional examples k and m in which the Cr content is not appropriate have a corrosion weight loss of 0.008 to 0.136 (mg / cm 2 ⁇ 3050 hr) and 0.012 respectively. It can be seen that the corrosion resistance is inferior to 0.325 (mg / cm 2 ⁇ 3050 hr).
  • PWSCC test Immersion test
  • the SCC resistance in an environment simulating the primary system of PWR of each test material was evaluated.
  • the temperature was raised from the actual usage environment and immersed in water of a PWR primary system environment of 360 ° C. capable of an acceleration test. .
  • the test piece was visually confirmed every predetermined time and the presence or absence of generation
  • production of PWSCC was investigated (each test material is n 4).
  • Table 3 below shows the results of a PWSCC test in which a test piece was immersed for up to 3050 hr in water simulating a PWR primary system.
  • the specimens a and b produced within the range defined by the present invention are free from cracks in the immersion test for 3050 hours and have excellent PWSCC resistance. Became clear.
  • the specimens a and b according to the present invention have at least the same PWSCC resistance as the TT690 alloy.
  • the comparative sample materials c to j having a Cr content of 20% or more and the conventional sample material k were found to have cracks in all the test pieces up to about 2500 hr.
  • Pitting corrosion sensitivity test In this example, assuming that impurities were mixed in the secondary water of PWR, pitting potential measurement was performed in a high-temperature water environment containing chloride, and the pitting corrosion sensitivity of the test materials was compared and evaluated. In the pitting corrosion sensitivity test, a disk-shaped pitting potential measurement test piece having a diameter of 10 mm and a height of 7 mm was first collected from each test material.
  • the susceptibility to pitting corrosion was evaluated by immersing the collected test piece in a tank of a circulating high-temperature electrochemical measurement test apparatus.
  • a potential measurement cell electrode tank
  • a pressure balanced external reference electrode Au / AgCl
  • a platinum electrode is used as a counter electrode.
  • the pitting potential of the test piece was measured.
  • the temperature in the electrode chamber was measured with a thermocouple.
  • the potential sweep and the current value were measured using a potentiostat.
  • the test temperature is 300 ° C.
  • the test solution is a 0.1 M sodium chloride aqueous solution
  • the pH is 5.6.
  • the measurement was performed at a potential sweep rate of 25 mV / min.
  • specimen materials a and b having a Cr content within the specified range of the present invention are higher than specimens k and m (TT690 alloy and TT600 alloy), which are conventionally known materials. It shows pitting corrosion potential and is clearly excellent in pitting corrosion resistance.
  • IGA susceptibility confirmation test in secondary environment
  • SSRT test Low strain rate tensile test
  • SSRT test Slow Strain Rate
  • IGA susceptibility confirmation test a flat plate-type uniaxial tensile specimen for SSRT having a thickness of 2 mm and a width of the test portion of 4 mm was first collected from each test material.
  • test piece was installed in an SSRT test apparatus, and an SSRT test was performed.
  • the test conditions are a 10% aqueous sodium hydroxide solution (pH 11.5), a temperature of 300 ° C., a potential of +100 mV vs Ec, and a strain rate of 8.3 ⁇ 10 ⁇ 7 s ⁇ 1 .
  • a potentiostat was used to apply a constant potential to the test piece.
  • grain boundary fracture surface ratio (%) is determined by the following formula ⁇ (area of fracture surface affected by test environment / area of total fracture surface) ⁇ 100 (%) ⁇ I asked for it.
  • the specimens a and b according to the present invention in which the Cr content is 34 to 38% and the Ni content is 44 to 56% have a grain boundary fracture surface ratio of 15 and 23, respectively. It can be seen that IGA corrosion resistance in an alkaline environment is suppressed. Thereby, it is clear that the corrosion resistance exceeding the conventionally known test material k (TT690 alloy) can be obtained.
  • test material k (TT690 alloy) (intergranular fracture surface rate: 27%)
  • test material m (TT600 alloy) (intergranular fracture surface rate: 32%)
  • Cr Cr
  • the workability in each of the test materials was evaluated in the process of performing extrusion molding by hot working and the process of manufacturing a small diameter tube by cold working according to the following conditions and procedures.
  • the plastic working in producing a heat transfer tube includes the production of a raw tube by hot working (extrusion molding), and subsequent final finishing by cold working.
  • the hot workability was evaluated based on the result of the greeble test, and the cold workability was evaluated based on the result of the tensile test, and it was determined whether or not a heat transfer tube of an actual machine size could be manufactured.
  • Hot ductility The size of the SG heat transfer tube of the actual machine in the hot working temperature range corresponds to the hot ductility, particularly the shape and the surface quality. That is, the hot workability is better as the drawing is larger.
  • Zero ductility temperature A temperature at which the elongation becomes 0 (practically defined as 20% or less). That is, as the zero ductility temperature is higher, partial melting at the grain boundary is less likely to occur, and hot workability is better.
  • Deformation resistance The magnitude of the tensile strength in the temperature range of hot working of the SG heat transfer tube of the actual machine corresponds to the deformation resistance during hot working. That is, the lower the deformation resistance, the better the hot workability.
  • a round bar tensile test piece having a parallel part ⁇ 10 mm was collected from each test material.
  • a tensile test was performed after a pattern in which the temperature was raised to the test temperature in 3 minutes and held for 3 minutes.
  • the test temperature 900 degreeC, 1000 degreeC, 1100 degreeC shown in following Table 4) ) Until the test temperature was reached, and a tensile test was immediately conducted.
  • the strain rate in the tensile test was 10 / s. Then, after the tensile test, the appearance of the test piece was observed and the outer diameter of the fractured portion was measured to determine the fracture drawing, the tensile strength was determined from the maximum load, and the deformation resistance (vs. TT690 alloy) was calculated.
  • the specimens a and b according to the present invention are both “A” for hot workability (hot ductility, zero ductility, deformation resistance) and cold workability (tensile strength).
  • the overall evaluation was “A (pipe can be made)”, and it was revealed that the plastic workability was excellent.
  • test materials h and j which are comparative examples manufactured under conditions outside the range defined in the present invention, are both inferior in hot ductility and cold workability, and the overall evaluation is “C ( It is understood that the plastic workability is inferior.
  • test materials c, d, and i which are comparative examples set as conditions outside the range specified in the present invention, are “A” in the comprehensive evaluation, but all are evaluated as “B” with respect to hot ductility.
  • the production of the heat transfer tube for the steam generator is sufficiently possible, but the material is relatively low in plastic workability.
  • test materials c, d, and e which are comparative examples set as conditions outside the range specified in the present invention
  • the comprehensive evaluation of workability is “A”, but the inclusion of either Cr or Ni Since the amount is outside the specified range of the present invention, it is inferior in terms of corrosion resistance and the like, and this is the same for the conventionally known test material k (TT690 alloy).
  • the nuclear equipment material according to the present invention is excellent in corrosion resistance and workability. Therefore, by applying the nuclear equipment material of the present invention to a steam generator heat transfer tube in a pressurized water light water reactor of a nuclear power plant, it is possible to realize a steam generator heat transfer tube having all the necessary characteristics in a pressurized water light water reactor. Is clear.
  • the nuclear equipment material according to one embodiment of the present invention is excellent in corrosion resistance and workability, and therefore can be applied to a heat transfer tube for a steam generator in a pressurized water reactor of a nuclear power plant.

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Abstract

L'invention concerne un matériau pour équipement nucléaire, lequel comprend, en masse %, 34 à 38 % de Cr, 44 à 56 % de Ni, 0,015 à 0,025 % de C, plus de 0 % à 0,5 % de Si, 0,05 à 0,5 % de Mn, au plus 0,003 % de S, au plus 0,015 % de P, au plus 0,05 % de N, au plus 0,5 % de Ti, 0,05 à 0,5 % de Al, le reste étant constitué de Fe ainsi que d'inévitables impuretés. L'invention concerne également un tuyau chauffant pour générateur de vapeur possédant ce matériau. L'invention concerne encore un générateur de vapeur comportant ce tuyau chauffant. Enfin, l'invention concerne une centrale nucléaire comportant ce générateur de vapeur.
PCT/JP2012/056178 2011-03-10 2012-03-09 Matériau pour équipement nucléaire, tuyau chauffant pour générateur de vapeur, générateur de vapeur et centrale nucléaire WO2012121390A1 (fr)

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WO2016208569A1 (fr) * 2015-06-26 2016-12-29 新日鐵住金株式会社 TUYAU EN ALLIAGE À BASE DE Ni POUR ÉNERGIE ATOMIQUE
EP3006589A4 (fr) * 2013-06-07 2017-03-15 Korea Atomic Energy Research Institute Procédé de production d'un alliage 690, alliage ordonné de conductivité thermique améliorée, et alliage 690 ordonné produit selon ledit procédé
CN110719964A (zh) * 2017-06-08 2020-01-21 日本制铁株式会社 原子能用Ni基合金管
US10760147B2 (en) 2013-06-07 2020-09-01 Korea Atomic Energy Research Insitute Ordered alloy 690 with improved thermal conductivity

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JPS58177444A (ja) * 1982-04-12 1983-10-18 Sumitomo Metal Ind Ltd Ｎｉ−Ｃｒ合金の熱処理法
JPS60245773A (ja) * 1984-05-18 1985-12-05 Sumitomo Metal Ind Ltd 高耐食性Ｎｉ基合金の製造方法
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Publication number Priority date Publication date Assignee Title
EP3006589A4 (fr) * 2013-06-07 2017-03-15 Korea Atomic Energy Research Institute Procédé de production d'un alliage 690, alliage ordonné de conductivité thermique améliorée, et alliage 690 ordonné produit selon ledit procédé
US10287664B2 (en) 2013-06-07 2019-05-14 Korea Atomic Energy Research Institute Production method for alloy 690 ordered alloy of improved thermal conductivity, and alloy 690 ordered alloy produced thereby
US10760147B2 (en) 2013-06-07 2020-09-01 Korea Atomic Energy Research Insitute Ordered alloy 690 with improved thermal conductivity
WO2016208569A1 (fr) * 2015-06-26 2016-12-29 新日鐵住金株式会社 TUYAU EN ALLIAGE À BASE DE Ni POUR ÉNERGIE ATOMIQUE
JPWO2016208569A1 (ja) * 2015-06-26 2018-02-08 新日鐵住金株式会社 原子力用Ｎｉ基合金管
CN110719964A (zh) * 2017-06-08 2020-01-21 日本制铁株式会社 原子能用Ni基合金管
KR20200016333A (ko) * 2017-06-08 2020-02-14 닛폰세이테츠 가부시키가이샤 원자력용 Ni기 합금관
EP3636785A4 (fr) * 2017-06-08 2020-10-28 Nippon Steel Corporation Tuyau d'alliage à base de ni, de qualité nucléaire
KR102256407B1 (ko) 2017-06-08 2021-05-26 닛폰세이테츠 가부시키가이샤 원자력용 Ni기 합금관
US11215356B2 (en) 2017-06-08 2022-01-04 Nippon Steel Corporation Ni-based alloy pipe for nuclear power
CN110719964B (zh) * 2017-06-08 2022-03-04 日本制铁株式会社 原子能用Ni基合金管

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