JP4458579B2 - Reactor material and manufacturing method thereof - Google Patents

Reactor material and manufacturing method thereof Download PDF

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JP4458579B2
JP4458579B2 JP20498399A JP20498399A JP4458579B2 JP 4458579 B2 JP4458579 B2 JP 4458579B2 JP 20498399 A JP20498399 A JP 20498399A JP 20498399 A JP20498399 A JP 20498399A JP 4458579 B2 JP4458579 B2 JP 4458579B2
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stainless steel
austenitic stainless
grain boundary
heat treatment
grain boundaries
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JP2001032045A (en
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文寿 鹿野
耕司 福谷
博司 坂本
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Toshiba Corp
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    • 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
    • Y02E30/30—Nuclear fission reactors
    • 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
    • Y02P—CLIMATE CHANGE MITIGATION TECHNOLOGIES IN THE PRODUCTION OR PROCESSING OF GOODS
    • Y02P10/00—Technologies related to metal processing
    • Y02P10/25—Process efficiency

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Description

【０００１】
【発明の属する技術分野】
本発明は、オーステナイト系ステンレス鋼からなる原子炉材料およびその製造方法に関する。
【０００２】
【従来の技術】
制御棒、炉内計装管のような原子炉の構造材料として、これまでは主にオーステナイト系ステンレスＳＵＳ３０４鋼が使用されてきた。オーステナイト系ステンレス鋼は、Ｆｅ−Ｃｒ−Ｎｉを基本とした合金であり、用途に応じてＭｏ等を微量に添加したＳＵＳ３１６鋼等がある。いずれの合金も、すべての成分元素が均一に分散していることを前提に設計されている。特に耐食性は、Ｃｒの酸化膜による保護性を利用することにより得られるものであり、Ｃｒの分布が均一であることが良好な耐食性を維持するための必要条件と考えられてきた。このＣｒの酸化膜が過酷な腐食環境において、たとえ何らかの原因で破損しても、周囲からＣｒ原子が供給されてただちに膜が再生されることで、耐食性は維持される。しかし、ＣｒはＣと結合しやすい特性を持つことから、熱処理条件さらには溶接施工等によってはクロム炭化物を生成し、その生成のためにＣｒが消費されることによって、Ｃｒの希薄または欠乏した領域が生ずる場合がある。これは熱鋭敏化と呼ばれる。この現象は、しばしば結晶粒界近傍で見られ、Ｃｒ欠乏部の耐食性を劣化させる。
【０００３】
最近は、熱鋭敏化による腐食対策を目的として、低炭素化されたＳＵＳ３０４Ｌ鋼、３１６Ｌ鋼が使用されている。これはＣの量を減らすことで、クロム炭化物の生成を抑制し、粒界でのＣｒ欠乏を防ぐことをねらったものである。しかし、長期間の中性子照射環境下では、結晶粒界近傍でＣｒが欠乏することが知られている。これは、粒界応力腐食割れ（ＩＧＳＣＣ）と呼ばれ、溶接などの熱サイクルによる粒界炭化物の形成とそれに伴う粒界近傍におけるクロム欠乏層の形成、すなわち、溶接鋭敏化がその原因である。しかしながら近年、熱による鋭敏化が全く起こっていない溶体化オーステナイト系ステンレス鋼においても、照射を受けた場合応力腐食割れを示す可能性が報告されている。
【０００４】
上記の原因に着目して、不純物元素量を限定することにより、高照射を受けた場合の耐粒界腐食割れを改善することが考えられ、高純度オーステナイト系ステンレス鋼が開発されている。
【０００５】
また、例えば、溶体化処理した固溶状態のオーステナイト系ステンレス鋼は放射線損傷のない炉心外においては耐粒界型応力腐食割れを有するが、同じ材料が炉心内において高レベルの照射、特に中性子照射量で０.５×１０21ｎ／ｃｍ2程度以上の照射を受けた場合はそのような抵抗性が失われていく。このような割れは照射誘起応力腐食割れ（ＩＡＳＣＣ）と称して古い原子炉で問題にされつつある。この問題を解決する方法として、オーステナイト系ステンレスの構成元素、例えば、Ｎ、Ｐ、Ｓｉ、Ｓ、Ｃ、Ｍｎ、Ｃｒ、Ｎｉの含有量を調整するとともに、微量のＴｉ、Ｎｂを添加する方法が提案されている。また、一方向凝固法によりオーステナイト系ステンレス鋼のランダム結晶粒界を排除して単結晶とする方法も提案されている。
【０００６】
一方、粒界型応力腐食割れを防止する方法として、その発生源であり網目状に連結する粒界を排除する単結晶化法がある。
【０００７】
図６を用いて説明する。オーステナイト系ステンレス鋼１中の結晶粒界３でのＣｒ濃度が１２ｗｔ％程度以下になると、Ｃｒ酸化膜４が破損した時のＣｒ原子５の供給が円滑に行われず、表面２の粒界部３において腐食が進行する。これは、照射によって粒界での偏析が誘起されるためである。金属は照射されることによって、中性子がもとの構成原子をはじき出し、金属中に原子空孔６と格子間原子が生成される。これらは、対消滅するものや、拡散してシンクで消滅するもの、欠陥の集合体になるものがある。シンクとは欠陥の消滅場所のことで、結晶粒界３や、表面２、転位、析出物等がこれに含まれる。ステンレス鋼中での格子間原子の多くは、移動速度が早く、シンクに行く前に集まって転位ループを作る。この後、移動速度の遅い原子空孔６が過剰に残存し、粒界等のシンクに流れ込む。この時、空孔６とクロム原子５は互いに位置交換しながら拡散していく。そのため、空孔６が粒界３に流れ込むに伴い、粒界近傍のクロム原子５は粒界から遠くに拡散することになる。このような現象は照射誘起偏析と呼ばれ、他の構成元素でも同様の挙動を示す。これは、照射によって生ずる一般的な現象である。位置交換しながら拡散するのは、母相原子に比べてサイズが大きなものである。これは、隣に空孔が来ることでひずみが緩和されるため、隣に空孔が来る確率が高くなり、位置交換の頻度も増えることによる。逆にニッケル原子７のようなサイズが小さい原子では、空孔６と組んで移動するため、空孔６が粒界３に流れ込むとき、同時に移動するので粒界３近くに集まることになる。
【０００８】
以上のような原理でクロムの腐食が進むものと考えられている。
【０００９】
【発明が解決しようとする課題】
上述した通り、中性子照射環境下でオーステナイト系ステンレス鋼製原子炉内部機器及び構造物の耐食性が低下する原因として、このステンレスの構成元素の一つであり、均一に分布していたＣｒが、照射により結晶粒界で減少すること（照射誘起偏析現象）があげられる。
【００１０】
そこで本発明の目的は、オーステナイト系ステンレス鋼製原子炉内部機器及び構造物の溶体化熱処理条件を最適化することによって、ステンレス鋼の粒界にＣｒ、またはＭｏのようなＣｒに類似した化学特性を持つ元素を偏析させ、Ｎｉのようにサイズ効果から粒界での濃度分布がＣｒと逆になる元素を制御し、Ｃｒの結晶粒界での欠乏を抑制することである。すなわち、本発明は、中性子照射環境下で、長期間の応力腐食割れの抑制効果を維持することのできるオーステナイト系ステンレス鋼製原子炉材料およびその製造方法を提供することを目的とする。
【００１１】
【課題を解決するための手段】
本発明の原子炉材料は、１１００℃以上の温度に加熱して粒界でのクロムとモリブデンを偏析させた後、８００℃から５００℃まで１℃／秒〜１０℃／秒の冷却速度で冷却されたオーステナイト系ステンレス鋼ＳＵＳ３１６Ｌからなり、粒界での偏析はクロムとモリブデンが濃縮する構成であることを特徴とする。
【００１５】
本発明において、前記加熱は、電気抵抗発熱体または高周波加熱装置によって行われる。
【００１６】
本発明の原子炉材料の製造方法は、オーステナイト系ステンレス鋼ＳＵＳ３１６Ｌを１１００℃以上の温度に加熱して粒界でのクロムとモリブデンを偏析させる工程と、この加熱する工程の後に前記オーステナイト系ステンレス鋼ＳＵＳ３１６Ｌを８００℃から５００℃まで１℃／秒〜１０℃／秒の冷却速度で冷却する工程とを具備し、粒界でクロムとモリブデンを濃縮偏析させることを特徴とする。
【００２０】
さらに、本発明の原子炉材料の製造方法において、前記加熱は、電気抵抗発熱体または高周波加熱装置によって行われる。
【００２３】
本発明によれば、オーステナイト系ステンレス鋼製の原子炉構造材料を通常の加熱温度よりも高い温度で加熱することにより、結晶粒界での偏析が促進される。また、オーステナイト系ステンレス鋼製の原子炉構造材料の加熱後の冷却速度を制御することで、結晶粒界での偏析が促進される。
【００２４】
また、熱処理を、本発明において限定した組成のオーステナイト系ステンレス鋼で実施することにより、さらに結晶粒界での偏析が促進される。さらに、熱処理を表面から０．１ｍｍ以上の深さで行うことにより、さらに結晶粒界での偏析が促進される。
【００２５】
本発明によれば、結晶粒界でＣｒ及びＣｒの拡散の抑制効果がある遷移元素を意図的に偏析させるための熱処理条件、表面処理条件、熱処理後の冷却時熱処理装置を規定することにより、オーステナイト系ステンレス鋼製原子炉構造材料に対する中性子照射環境下での結晶粒界でのＣｒ欠乏層の生成が抑制され、その結果、中性子照射環境下での耐食性を維持することができる。
【００２６】
【発明の実施の形態】
図１は、本実施の形態で用いたオーステナイト系ステンレス鋼ＳＵＳ３１６Ｌにおける結晶粒界近傍での偏析を示したものである。このステンレス鋼上で、０．５ｎｍ径の電子線プローブを利用し、結晶粒界でのＣｒおよびＭｏの元素濃度を０．４ｎｍ間隔で測定した。図１（ａ）、（ｂ）および（ｃ）に示すように、３種類のものを用いた。
【００２７】
図２にＳＵＳ３１６Ｌ鋼の熱処理時の温度を変化させた場合の粒界上でのＣｒ＋Ｍｏの濃度を示す。図２のグラフから明らかな通り、熱処理温度が高い方がＣｒ＋Ｍｏの濃度が高くなることがわかる。一般的なオーステナイト系ステンレス鋼では、１０５０℃程度で溶体化熱処理が行われるが、それより高い温度で熱処理を行うことで、粒界での偏析を生じさせることができる。これによりオーステナイト系ステンレス鋼製の原子炉構造材料の中性子照射環境下での耐食性を維持することができる。
【００２８】
図３に１０５０℃でのＳＵＳ３１６Ｌ鋼の熱処理時の冷却速度を変化させた場合の粒界上でのＣｒ＋Ｍｏの濃度を示す。この結果、８００℃から５００℃の間の冷却速度が１℃／秒から１０℃／秒の間でＣｒ＋Ｍｏの濃度が最高になることがわかった。これによりオーステナイト系ステンレス鋼製の原子炉構造材料の中性子照射環境下での耐食性を維持することができる。
【００２９】
本発明は、熱処理により粒界での偏析を生成させるが、オーステナイト相が安定である必要がある。また粒界での偏析は、クロムとモリブデンが濃縮し、ニッケルが欠乏することが望ましい。また熱鋭敏化によるクロム炭化物の生成を抑制するため、炭素量を低くする必要がある。
【００３０】
原子炉構造材料から作成される機器及び構造物には、一定の容積と熱容量があるため、均一に加熱後の冷却を進行させるのは容易ではない。一方、耐食性に対する影響を決定するのは表面状態であり、表面近傍の結晶粒界で十分なＣｒ濃度があれば耐食性を維持することができる。そこで、原子炉構造材料から作成される機器及び構造物の表面近傍の熱処理温度をモニターして温度制御を行うと好適であることを見出した。
【００３１】
図４に、本発明によるオーステナイト系ステンレス鋼からなる原子炉構造材料の加熱および冷却を行うための装置の概略構成図を示す。
【００３２】
オーステナイト系ステンレス鋼からなる原子炉構造材料８は、電気抵抗発熱体もしくは高周波加熱装置である加熱装置９により加熱され、冷却装置（図示せず）により冷却される。加熱および冷却は、温度制御装置１０によって処理温度がモニターされる。また、加熱装置９と温度制御装置１０にはそれぞれ接触型温度計１１と非接触温度計１２が具備されている。また、原子炉材料８の加熱および冷却は駆動装置１３によって、加熱装置９、冷却装置および温度制御装置１０と接続されており、これらの装置を任意の速度で移動させて原子炉構造材料８に所望の処理を施す。
【００３３】
図４（ａ）は熱処理開始時、（ｂ）は熱処理途中、（ｃ）熱処理終了時を示す。
【００３４】
原子炉構造材料から作成された内部機器及び構造物には、大型の一体化された部材もあり、均一な熱処理を行うためには、温度制御装置１０により熱処理時の温度をリアルタイムで感知しながら、加熱装置、冷却装置を制御、作動させることにより、中性子照射環境下での耐食性を維持できるものが作成できる。
【００３５】
図５は、熱処理により結晶粒界でＣｒ及びＭｏが濃縮した場合の中性子照射にともなう粒界での濃度変化を示す。これから、熱処理を最適化して、粒界でのＣｒ＋Ｍｏを濃縮することにより、１０２５ｎ／ｍ２においても、粒界でのＣｒの欠乏は生じていないことがわかる。
【００３６】
【発明の効果】
以上述べたように本発明によるオーステナイト系ステンレス鋼３１６Ｌ製原子炉構造材料によれば、長期の中性子照射環境下で照射による偏析は生ずるものの、照射誘起偏析による粒界でのＣｒ欠乏が抑制されることから、従来より長く結晶粒界での耐食性を持続することができる。さらに、本発明の原子炉構造材料の製造方法によれば、特に材料の組成を変化させずに加熱温度と冷却速度を制御するという簡便な方法を採用しているため、製造コストを増大させることがない。
【００３７】
これにより、原子炉の長寿命化に貢献する事が可能となり、さらには原子力プラントの製造コストの低減に寄与することができる。
【図面の簡単な説明】
【図１】 オーステナイト系ステンレス鋼の結晶粒界近傍でのＣｒとＭｏの組成濃度分布図。
【図２】 ＳＵＳ３１６Ｌステンレス鋼の結晶粒界におけるＣｒ＋Ｍｏ濃度と熱処理温度の関係を示すグラフ。
【図３】 ＳＵＳ３１６Ｌステンレス鋼の結晶粒界におけるＣｒ＋Ｍｏ濃度と１０５０℃での熱処理時の冷却速度との関係を示すグラフ。
【図４】 原子炉内部機器及び構造物の処理温度制御装置の模式図。
【図５】 結晶粒界に偏析がある場合の、結晶粒界でのＣｒ＋Ｍｏ濃度と照射量の関係を示すグラフ。
【図６】 従来の技術における照射誘起偏析の機構を示す模式図。
【符号の説明】
１…オーステナイト系ステンレス鋼
２…表面
３…結晶粒界
４…Ｃｒ酸化膜
５…Ｃｒ原子
６…原子空孔
７…Ｎｉ原子
８…原子炉構造材料
９…加熱装置
１０…冷却時温度制御装置
１１…接触型温度計
１２…非接触型温度計
１３…駆動装置[0001]
BACKGROUND OF THE INVENTION
The present invention relates to a nuclear reactor material made of austenitic stainless steel and a method for producing the same .
[0002]
[Prior art]
Conventionally, austenitic stainless steel SUS304 has been mainly used as a structural material for nuclear reactors such as control rods and in-core instrument tubes. Austenitic stainless steel is an alloy based on Fe-Cr-Ni, and includes SUS316 steel to which a small amount of Mo or the like is added depending on the application. Both alloys are designed on the assumption that all the constituent elements are uniformly dispersed. In particular, the corrosion resistance is obtained by utilizing the protective property of the Cr oxide film, and it has been considered that the uniform distribution of Cr is a necessary condition for maintaining good corrosion resistance. Even if the Cr oxide film is damaged for some reason in a severe corrosive environment, the corrosion resistance is maintained by the regeneration of the film as soon as Cr atoms are supplied from the surroundings. However, since Cr has the property of easily combining with C, chromium carbide is generated depending on heat treatment conditions and welding work, and Cr is consumed for the generation thereof, so that the Cr is diluted or depleted region. May occur. This is called heat sensitization. This phenomenon is often seen in the vicinity of the crystal grain boundary and deteriorates the corrosion resistance of the Cr-deficient part.
[0003]
Recently, low-carbon SUS304L steel and 316L steel have been used for the purpose of countermeasures against corrosion by thermal sensitization. This is to reduce the amount of C to suppress the formation of chromium carbide and prevent Cr deficiency at the grain boundary. However, it is known that Cr is deficient in the vicinity of the grain boundary under a long-term neutron irradiation environment. This is called intergranular stress corrosion cracking (IGSCC) and is caused by the formation of intergranular carbide by thermal cycles such as welding and the formation of a chromium-deficient layer in the vicinity of the intergranularity, that is, welding sensitization. However, in recent years, even in solution austenitic stainless steel that has not been sensitized by heat, the possibility of stress corrosion cracking has been reported when irradiated.
[0004]
Focusing on the above causes, it is conceivable to improve the intergranular corrosion cracking when subjected to high irradiation by limiting the amount of impurity elements, and high purity austenitic stainless steel has been developed.
[0005]
In addition, for example, solution-treated austenitic stainless steel in a solid solution state has grain boundary type stress corrosion cracking outside the core without radiation damage, but the same material is irradiated at a high level in the core, particularly neutron irradiation. Such a resistance is lost when the irradiation is about 0.5 × 10 21 n / cm 2 or more. Such cracks are called radiation induced stress corrosion cracking (IASCC) and are becoming a problem in older reactors. As a method of solving this problem, there is a method of adjusting the content of constituent elements of austenitic stainless steel, for example, N, P, Si, S, C, Mn, Cr, and Ni, and adding a trace amount of Ti and Nb. Proposed. In addition, a method has been proposed in which random crystal grain boundaries of austenitic stainless steel are eliminated to form a single crystal by a unidirectional solidification method.
[0006]
On the other hand, as a method for preventing intergranular stress corrosion cracking, there is a single crystallization method that eliminates the grain boundary that is the source of the intergranular connection.
[0007]
This will be described with reference to FIG. When the Cr concentration at the crystal grain boundary 3 in the austenitic stainless steel 1 is about 12 wt% or less, the supply of Cr atoms 5 when the Cr oxide film 4 is broken is not smoothly performed, and the grain boundary part 3 on the surface 2 Corrosion proceeds at. This is because segregation at grain boundaries is induced by irradiation. When the metal is irradiated, neutrons eject the original constituent atoms, and atomic vacancies 6 and interstitial atoms are generated in the metal. Some of these disappear, some diffuse and disappear at the sink, and some become defects. A sink is a defect disappearance location, and includes crystal grain boundaries 3, surface 2, dislocations, precipitates, and the like. Many of the interstitial atoms in stainless steel move quickly and gather before they go to the sink to form a dislocation loop. Thereafter, the atomic vacancies 6 having a low moving speed remain and flow into sinks such as grain boundaries. At this time, the holes 6 and the chromium atoms 5 diffuse while exchanging positions with each other. Therefore, as the vacancies 6 flow into the grain boundaries 3, the chromium atoms 5 near the grain boundaries diffuse away from the grain boundaries. Such a phenomenon is called irradiation-induced segregation, and the same behavior is exhibited with other constituent elements. This is a general phenomenon caused by irradiation. Diffusion while exchanging positions is larger than that of the parent atom. This is because the distortion is eased by the adjoining vacancies, so that the probability that the vacancies are adjoining increases and the frequency of position exchange increases. On the other hand, since atoms having a small size such as nickel atoms 7 move in combination with the vacancies 6, when the vacancies 6 flow into the grain boundaries 3, they move at the same time and gather near the grain boundaries 3.
[0008]
It is thought that the corrosion of chromium progresses on the principle as described above.
[0009]
[Problems to be solved by the invention]
As described above, one of the constituent elements of this stainless steel, Cr, which was uniformly distributed, is the cause of the deterioration of the corrosion resistance of austenitic stainless steel reactor internal equipment and structures under the neutron irradiation environment. Is reduced at the grain boundary (irradiation-induced segregation phenomenon).
[0010]
Therefore, the object of the present invention is to optimize the chemical solution heat treatment conditions of the austenitic stainless steel reactor internal equipment and structures, and have chemical properties similar to Cr, such as Cr, or Cr such as Mo at the grain boundaries of stainless steel. And segregating elements having a concentration distribution such as Ni, and controlling the element whose concentration distribution at the grain boundary is opposite to that of Cr due to the size effect, thereby suppressing the deficiency at the crystal grain boundary of Cr. That is, an object of the present invention is to provide an austenitic stainless steel reactor material capable of maintaining the effect of suppressing stress corrosion cracking for a long time under a neutron irradiation environment, and a method for producing the same.
[0011]
[Means for Solving the Problems]
The reactor material of the present invention is heated to a temperature of 1 100 ° C. or higher to segregate chromium and molybdenum at the grain boundary, and then at a cooling rate of 1 ° C./second to 10 ° C./second from 800 ° C. to 500 ° C. It is made of cooled austenitic stainless steel SUS316L, and segregation at the grain boundary is characterized in that chromium and molybdenum are concentrated.
[0015]
In the present invention, the heating is performed by an electric resistance heating element or a high-frequency heating device.
[0016]
Manufacturing method of the reactor material of the present invention includes the steps of Ru is segregated chromium and molybdenum at the grain boundaries by heating the austenitic stainless steel SUS316L to 1 100 ° C. or higher, the austenitic after the step of heating A step of cooling stainless steel SUS316L from 800 ° C. to 500 ° C. at a cooling rate of 1 ° C./second to 10 ° C./second, and chromium and molybdenum are concentrated and segregated at grain boundaries.
[0020]
Furthermore, in the method for manufacturing a reactor material according to the present invention, the heating is performed by an electric resistance heating element or a high-frequency heating device.
[0023]
According to the present invention, segregation at a grain boundary is promoted by heating a reactor structural material made of austenitic stainless steel at a temperature higher than a normal heating temperature. Further, by controlling the cooling rate after heating the nuclear reactor structural material made of austenitic stainless steel, segregation at the grain boundaries is promoted.
[0024]
Further, by performing the heat treatment on the austenitic stainless steel having the composition limited in the present invention, segregation at the grain boundary is further promoted. Furthermore, by performing the heat treatment at a depth of 0.1 mm or more from the surface, segregation at the grain boundaries is further promoted.
[0025]
According to the present invention, by prescribing heat treatment conditions, surface treatment conditions, and heat treatment equipment during cooling after heat treatment for intentionally segregating transition elements that have an effect of suppressing the diffusion of Cr and Cr at grain boundaries, Formation of a Cr-deficient layer at the grain boundary in the neutron irradiation environment for the austenitic stainless steel reactor structural material is suppressed, and as a result, corrosion resistance in the neutron irradiation environment can be maintained.
[0026]
DETAILED DESCRIPTION OF THE INVENTION
FIG. 1 shows segregation near the grain boundary in the austenitic stainless steel SUS316L used in the present embodiment. On this stainless steel, using an electron beam probe with a diameter of 0.5 nm, the elemental concentrations of Cr and Mo at the grain boundaries were measured at intervals of 0.4 nm. As shown in FIGS. 1 (a), (b) and (c), three types were used.
[0027]
FIG. 2 shows the Cr + Mo concentration on the grain boundary when the temperature during the heat treatment of SUS316L steel is changed. As is apparent from the graph of FIG. 2, it can be seen that the higher the heat treatment temperature, the higher the Cr + Mo concentration. In general austenitic stainless steel, solution heat treatment is performed at about 1050 ° C., but by performing heat treatment at a temperature higher than that, segregation at grain boundaries can be caused. As a result, the corrosion resistance of the austenitic stainless steel reactor structural material in a neutron irradiation environment can be maintained.
[0028]
FIG. 3 shows the Cr + Mo concentration on the grain boundary when the cooling rate during heat treatment of SUS316L steel at 1050 ° C. is changed. As a result, it was found that the Cr + Mo concentration was highest when the cooling rate between 800 ° C. and 500 ° C. was between 1 ° C./second and 10 ° C./second. As a result, the corrosion resistance of the austenitic stainless steel reactor structural material in a neutron irradiation environment can be maintained.
[0029]
In the present invention, segregation at grain boundaries is generated by heat treatment, but the austenite phase needs to be stable. In addition, segregation at the grain boundaries is preferably concentrated in chromium and molybdenum and deficient in nickel. Moreover, in order to suppress the production | generation of chromium carbide by heat sensitization, it is necessary to make carbon content low .
[0030]
Since equipment and structures made from nuclear reactor structural materials have a certain volume and heat capacity, it is not easy to proceed with cooling after heating uniformly. On the other hand, the influence on the corrosion resistance is determined by the surface state, and the corrosion resistance can be maintained if there is a sufficient Cr concentration at the grain boundary near the surface. Therefore, it has been found that it is preferable to control the temperature by monitoring the heat treatment temperature in the vicinity of the surface of the equipment and structure made from the reactor structural material.
[0031]
In FIG. 4, the schematic block diagram of the apparatus for heating and cooling the nuclear reactor structural material which consists of austenitic stainless steel by this invention is shown.
[0032]
The nuclear reactor structural material 8 made of austenitic stainless steel is heated by a heating device 9 which is an electric resistance heating element or a high-frequency heating device, and is cooled by a cooling device (not shown). In the heating and cooling, the processing temperature is monitored by the temperature control device 10. The heating device 9 and the temperature control device 10 are provided with a contact-type thermometer 11 and a non-contact thermometer 12, respectively. In addition, the heating and cooling of the reactor material 8 are connected to the heating device 9, the cooling device, and the temperature control device 10 by the driving device 13, and these devices are moved at an arbitrary speed to the reactor structural material 8. Apply the desired treatment.
[0033]
FIG. 4A shows the start of heat treatment, FIG. 4B shows the middle of the heat treatment, and FIG. 4C shows the end of the heat treatment.
[0034]
Internal equipment and structures made from nuclear reactor structural materials also have large integrated members. In order to perform uniform heat treatment, the temperature controller 10 senses the temperature during heat treatment in real time. By controlling and operating the heating device and the cooling device, a device capable of maintaining the corrosion resistance under the neutron irradiation environment can be created.
[0035]
FIG. 5 shows the change in concentration at the grain boundary accompanying neutron irradiation when Cr and Mo are concentrated at the grain boundary by heat treatment. From this, it is understood that Cr depletion at the grain boundary does not occur even at 10 25 n / m 2 by optimizing the heat treatment and concentrating Cr + Mo at the grain boundary.
[0036]
【The invention's effect】
As described above, according to the reactor structural material made of austenitic stainless steel 316L according to the present invention, although segregation due to irradiation occurs in a long-term neutron irradiation environment, Cr deficiency at grain boundaries due to irradiation-induced segregation is suppressed. For this reason, the corrosion resistance at the grain boundaries can be maintained longer than before. Furthermore, according to the method for manufacturing a nuclear reactor structure material of the present invention, since a simple method of controlling the heating temperature and the cooling rate without changing the composition of the material is adopted, the manufacturing cost is increased. There is no.
[0037]
As a result, it is possible to contribute to extending the life of the nuclear reactor, and further contribute to reducing the manufacturing cost of the nuclear power plant.
[Brief description of the drawings]
FIG. 1 is a composition concentration distribution diagram of Cr and Mo in the vicinity of a grain boundary of austenitic stainless steel.
FIG. 2 is a graph showing the relationship between the Cr + Mo concentration at the grain boundary of SUS316L stainless steel and the heat treatment temperature.
FIG. 3 is a graph showing the relationship between the Cr + Mo concentration at the grain boundary of SUS316L stainless steel and the cooling rate during heat treatment at 1050 ° C.
FIG. 4 is a schematic diagram of a reactor temperature control device for reactor internal equipment and structures.
FIG. 5 is a graph showing the relationship between the Cr + Mo concentration at the crystal grain boundary and the irradiation dose when there is segregation at the crystal grain boundary.
FIG. 6 is a schematic diagram showing a mechanism of irradiation-induced segregation in the prior art.
[Explanation of symbols]
DESCRIPTION OF SYMBOLS 1 ... Austenitic stainless steel 2 ... Surface 3 ... Grain boundary 4 ... Cr oxide film 5 ... Cr atom 6 ... Atomic vacancy 7 ... Ni atom 8 ... Reactor structural material 9 ... Heating device 10 ... Cooling temperature control device 11 ... Contact-type thermometer 12 ... Non-contact-type thermometer 13 ... Drive device

Claims (2)

１１００℃以上の温度に加熱して粒界でのクロムとモリブデンを偏析させた後、８００℃から５００℃まで１℃／秒〜１０℃／秒の冷却速度で冷却されたオーステナイト系ステンレス鋼ＳＵＳ３１６Ｌからなり、粒界での偏析はクロムとモリブデンが濃縮する構成であることを特徴とする原子炉材料。1 Austenitic stainless steel SUS316L heated to a temperature of 100 ° C. or higher to segregate chromium and molybdenum at grain boundaries and then cooled from 800 ° C. to 500 ° C. at a cooling rate of 1 ° C./second to 10 ° C./second. Reactor material characterized in that the segregation at the grain boundary is composed of chromium and molybdenum. オーステナイト系ステンレス鋼ＳＵＳ３１６Ｌを１１００℃以上の温度に加熱して粒界でのクロムとモリブデンを偏析させる工程と、この加熱する工程の後に前記オーステナイト系ステンレス鋼ＳＵＳ３１６Ｌを８００℃から５００℃まで１℃／秒〜１０℃／秒の冷却速度で冷却する工程とを具備し、粒界でクロムとモリブデンを濃縮偏析させることを特徴とする原子炉材料の製造方法。A step of Ru is segregated chromium and molybdenum at the grain boundaries by heating the austenitic stainless steel SUS316L to 1 100 ° C. or higher, 1 the austenitic stainless steel SUS316L after the step of heating from 800 ° C. to 500 ° C. And a step of cooling at a cooling rate of 10 ° C./second to 10 ° C./second, wherein chromium and molybdenum are concentrated and segregated at grain boundaries.
JP20498399A 1999-07-19 1999-07-19 Reactor material and manufacturing method thereof Expired - Fee Related JP4458579B2 (en)

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