JPH0961570A - Production of zirconium alloy-based reactor core structural material excellent in corrosion resistance, particularly in uniform corrosion resistance and hydrogen absorption resistance - Google Patents

Production of zirconium alloy-based reactor core structural material excellent in corrosion resistance, particularly in uniform corrosion resistance and hydrogen absorption resistance

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Publication number
JPH0961570A
JPH0961570A JP7235956A JP23595695A JPH0961570A JP H0961570 A JPH0961570 A JP H0961570A JP 7235956 A JP7235956 A JP 7235956A JP 23595695 A JP23595695 A JP 23595695A JP H0961570 A JPH0961570 A JP H0961570A
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JP
Japan
Prior art keywords
corrosion resistance
zirconium alloy
resistance
hydrogen absorption
content
Prior art date
Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
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Application number
JP7235956A
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Japanese (ja)
Inventor
Yasuhiro Mozumi
泰寛 茂住
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
Nuclear Fuel Industries Ltd
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Nuclear Fuel Industries Ltd
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Application filed by Nuclear Fuel Industries Ltd filed Critical Nuclear Fuel Industries Ltd
Priority to JP7235956A priority Critical patent/JPH0961570A/en
Publication of JPH0961570A publication Critical patent/JPH0961570A/en
Withdrawn legal-status Critical Current

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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

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  • Preventing Corrosion Or Incrustation Of Metals (AREA)

Abstract

PROBLEM TO BE SOLVED: To provide a method for producing zirconium alloy-based reactor core structural material having sufficient corrosion resistance, particularly excellent in uniform corrosion resistance and hydrogen absorption resistance in higher burn-up degree and for longer use than before. SOLUTION: Sn: 0.8-1.6wt.%, Fe: 0.17-0.25wt.%, Cr: 0.15-0.25wt.%, Ni: 0.01-0.08wt.% and impurity Si of concentration of 120ppm or less are contained, the residual is composed of Zr, and after zirconium alloy adjusted so as to satisfy Fe+Cr+Ni<=0.52wt.% is melted and made a piece of casting, Β hardening treatment is performed. Thereafter, rolling and annealing treatment are performed so that the total thermal input amount Σ A1 for the material may be in the range of 1×10<-19> -1×10<-17> .

Description

【発明の詳細な説明】Detailed Description of the Invention

【０００１】[0001]

【発明の属する技術分野】本発明は、原子炉炉心内構造
材の製造方法に関するものであり、特に、高燃焼度およ
び長期使用における耐食性、特に耐一様腐食性および耐
水素吸収性に優れたジルコニウム合金系原子炉炉心内構
造材の製造方法に関するものである。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for manufacturing a structural material in a reactor core, and particularly to a high burnup and corrosion resistance in long-term use, particularly excellent uniform corrosion resistance and hydrogen absorption resistance. The present invention relates to a method for manufacturing a zirconium alloy-based reactor core structural material.

【０００２】[0002]

【従来の技術】従来より、原子炉の燃料集合体など炉心
内構造材にはジルコニウム合金が用いられている。例え
ば、沸騰水型原子炉（以下、ＢＷＲとする）において
は、燃料被覆管材料やウォータロッドにジルカロイ−２
が、チャンネルボックスにジルカロイ−４が、またスペ
ーサ部材にジルカロイ−２およびジルカロイ−４が使用
されていた。一方、加圧水型原子炉（以下、ＰＷＲとい
う）においては、燃料被覆管材、制御棒案内管材として
ジルカロイ−４が用いられていた。2. Description of the Related Art Conventionally, a zirconium alloy has been used as a structural material in a core such as a fuel assembly of a nuclear reactor. For example, in a boiling water reactor (hereinafter referred to as BWR), Zircaloy-2 is used as a fuel cladding material or a water rod.
However, Zircaloy-4 was used for the channel box, and Zircaloy-2 and Zircaloy-4 were used for the spacer member. On the other hand, in a pressurized water nuclear reactor (hereinafter referred to as PWR), Zircaloy-4 was used as a fuel cladding tube material and a control rod guide tube material.

【０００３】歴史的には、ジルカロイ−２が先に開発さ
れ、そのジルカロイ−２の優れた耐食性を維持したまま
水素吸収性の低減を試みたジルカロイ−４が開発され
た。このようなジルカロイ−４が炉水水素濃度の高いＰ
ＷＲに、またＢＷＲにおける燃料集合体部材であるスペ
ーサおよびチャンネルボックスに使用されてきた。一
方、炉水水素濃度が比較的低いＢＷＲでは、燃料被覆管
用にジルカロイ−２が使用されている。Historically, Zircaloy-2 was first developed, and Zircaloy-4 was developed in an attempt to reduce hydrogen absorption while maintaining the excellent corrosion resistance of Zircaloy-2. Zircaloy-4 like this has a high P concentration in the reactor water hydrogen.
It has been used in WRs and in fuel assembly members in BWRs, such as spacers and channel boxes. On the other hand, in BWR having relatively low hydrogen concentration in reactor water, Zircaloy-2 is used for the fuel cladding tube.

【０００４】ジルカロイ−２の耐水素吸収性はジルカロ
イ−４に比べると劣っているが、４００℃以上の高温で
高圧の水蒸気中では、耐食性はジルカロイ−４に比べて
優れており、特に５３０℃の高温でも製造工程中の熱処
理を最適にした場合には、ノジュラー腐食（白いレンズ
状腐食物が生じる）は発生しない。Although hydrogen absorption resistance of Zircaloy-2 is inferior to that of Zircaloy-4, its corrosion resistance is superior to that of Zircaloy-4 at a high temperature of 400 ° C. or higher in high-pressure steam, especially at 530 ° C. Nodular corrosion (white lenticular corrosion products occur) does not occur when the heat treatment during the manufacturing process is optimized even at a high temperature.

【０００５】このようなジルカロイ−２の組成はＡＳＴ
Ｍ（American Society for Testingand Materials）お
よびＪＩＳ規格によって、Ｓｎ；１．２０〜１．７０ｗ
ｔ％，Ｆｅ；０．０７〜０．２０ｗｔ％，Ｃｒ；０．０
５〜０．１５ｗｔ％，Ｎｉ；０．０３〜０．０８ｗｔ
％，Ｆｅ＋Ｃｒ＋Ｎｉ；０．１８〜０．３８ｗｔ％に規
定されている。The composition of such Zircaloy-2 is AST
Sn: 1.20 to 1.70w according to M (American Society for Testing and Materials) and JIS standards
t%, Fe; 0.07 to 0.20 wt%, Cr; 0.0
5 to 0.15 wt%, Ni; 0.03 to 0.08 wt
%, Fe + Cr + Ni; 0.18 to 0.38 wt%.

【０００６】８０年代前半までは、ノミナル値でＳｎ
１．５ｗｔ％，Ｆｅ０．１５ｗｔ％，Ｃｒ０．１ｗｔ
％，Ｎｉ０．０５ｗｔ％程度で使用されていたが、８０
年代後半より現在まで、耐食性、主にノジュラー腐食性
の一層の改善を計って、Ｓｎが下限値１．２％，Ｆｅが
上限値０．２％，Ｎｉが上限値０．０８％，Ｃｒがノミ
ナル値０．１％とした合金組成をもつジルカロイ−２が
広く使用されている。Until the first half of the 1980s, Sn was the nominal value.
1.5wt%, Fe0.15wt%, Cr0.1wt
%, Ni was used in 0.05 wt%, but 80
From the latter half of the 1980s to the present, by further improving corrosion resistance, mainly nodular corrosion resistance, Sn has a lower limit of 1.2%, Fe has an upper limit of 0.2%, Ni has an upper limit of 0.08%, and Cr has an upper limit of 0.08%. Zircaloy-2 having an alloy composition with a nominal value of 0.1% is widely used.

【０００７】このような組成を持つジルカロイ−２の水
素吸収特性は、従来使用されている燃料集合体の燃焼度
範囲では、問題となることはなかった。また、一様腐食
（黒色皮膜状の腐食物が生じる）特性についても、ジル
カロイ−２の長期使用に際して一様腐食が加速される現
象は、８０年代のＢＷＲでは知られていなかった。The hydrogen absorption characteristics of Zircaloy-2 having such a composition have not been a problem in the burnup range of conventionally used fuel assemblies. Regarding the characteristic of uniform corrosion (a black film-like corrosive substance is generated), the phenomenon that uniform corrosion is accelerated during long-term use of Zircaloy-2 was not known in the BWRs of the 1980s.

【０００８】[0008]

【発明が解決しようとする課題】しかしながら、上記の
組成を持ち、且つ耐食性（耐ノジュラー腐食性、耐一様
腐食性）に対して最適化された熱処理を施されているジ
ルコニウム合金を使用したものであっても、大幅な高燃
焼度または長期使用に対しては、耐一様腐食加速現象お
よび耐水素吸収特性が十分でないことが判明した。However, a zirconium alloy having the above composition and having been subjected to a heat treatment optimized for corrosion resistance (nodular corrosion resistance, uniform corrosion resistance) is used. However, it has been found that the uniform corrosion acceleration accelerating phenomenon and the hydrogen absorption resistance are not sufficient for a significantly high burnup or long-term use.

【０００９】また、スペーサ、ウオーターロッド、チャ
ンネルボックス等については、温度勾配がないために部
材両面からの水素の侵入がスムーズとなってしまい、一
方構造部材としてはより薄肉化が望まれることから、水
素脆化に対して被覆管以上に改善が望まれる。Further, with respect to the spacer, water rod, channel box, etc., since there is no temperature gradient, the penetration of hydrogen from both sides of the member becomes smooth, while it is desired that the structural member be thinner. It is desired to improve hydrogen embrittlement more than cladding.

【００１０】またＢＷＲプラントにおいても、長寿命化
を図って炉水中に水素を添加する試みが始まり、炉心内
構造材に使用されるジルコニウム合金にとって、耐水素
吸収性などさらに厳しい条件となりつつある。Also in BWR plants, attempts have been made to add hydrogen to the reactor water in order to extend the service life thereof, and the zirconium alloy used for the structural material in the core has become more severe conditions such as hydrogen absorption resistance.

【００１１】本発明は、上記問題点に鑑み、従来より高
燃焼度および長期使用において十分な耐食性、特に耐一
様腐食性および耐水素吸収性に優れた原子炉炉心構造材
用ジルコニウム合金が得られる製造方法を提供すること
を目的とする。In view of the above problems, the present invention provides a zirconium alloy for a reactor core structural material, which has a higher burnup and a sufficient corrosion resistance in long-term use, particularly uniform corrosion resistance and hydrogen absorption resistance. It is an object of the present invention to provide a manufacturing method of the same.

【００１２】[0012]

【課題を解決するための手段】上記目的を達成するた
め、請求項１に記載の発明に係る耐食性、特に耐一様腐
食性と耐水素吸収性に優れたジルコニウム合金系原子炉
炉心内構造材の製造方法では、錫（Ｓｎ）；０．８〜
１．６ｗｔ％，鉄（Ｆｅ）；０．１７〜０．２５ｗｔ
％，クロム（Ｃｒ）；０．１５〜０．２５ｗｔ％，ニッ
ケル（Ｎｉ）；０．０１〜０．０８ｗｔ％および濃度１
２０ｐｐｍ以下の不純物けい素（Ｓｉ）を含み、残部は
ジルコニウム（Ｚｒ）からなり、且つ、 Ｆｅ＋Ｃｒ＋Ｎｉ≦０．５２ｗｔ％ を満足するように調節したジルコニウム合金を溶製して
鋳片としたのちβ焼入れ処理を施し、その後、下式で表
される材料への総入熱量ΣＡi が１×１０-19 〜１×１
０-17 の範囲内に納まるように圧延および焼鈍処理を施
すものである。 ΣＡi ＝Σｔi ・ｅｘｐ（−Ｑ／ＲＴi ） ｔi ：β焼入れ後のｉ番目熱処理工程における処理時間
（時間） Ｔi ：工程ｉの処理温度（Ｋ） Ｑ：活性化エネルギー Ｒ：気体定数 Ｑ／Ｒ：４００００In order to achieve the above object, a zirconium alloy-based reactor core structural material excellent in corrosion resistance, particularly uniform corrosion resistance and hydrogen absorption resistance according to the invention of claim 1 is provided. In the manufacturing method of, tin (Sn); 0.8 to
1.6 wt%, iron (Fe); 0.17 to 0.25 wt
%, Chromium (Cr); 0.15-0.25 wt%, nickel (Ni); 0.01-0.08 wt% and concentration 1
Beta-quenched after smelting a zirconium alloy containing 20 ppm or less of impurity silicon (Si), the balance being zirconium (Zr), and adjusting to satisfy Fe + Cr + Ni ≦ 0.52 wt% After the treatment, the total heat input ΣA i to the material represented by the following formula is 1 × 10 −19 to 1 × 1
The rolling and annealing treatments are performed so that it falls within the range of 0 -17 . ΣA i = Σt i · exp ( -Q / RT i) t i: β treatment time in the i-th heat treatment step after quenching (time) T i: step i treatment temperature (K) Q: activation energy R: gas Constant Q / R: 40000

【００１３】また、請求項２に記載の発明に係る耐食
性、特に耐一様腐食性と耐水素吸収性に優れたジルコニ
ウム合金系原子炉炉心内構造材の製造方法では、請求項
１に記載のジルコニウム合金系原子炉炉心内構造材の製
造方法において、錫（Ｓｎ）；０．８〜１．２ｗｔ％，
鉄（Ｆｅ）；０．１７〜０．２５ｗｔ％，クロム（Ｃ
ｒ）；０．１５〜０．２５ｗｔ％，ニッケル（Ｎｉ）；
０．０１〜０．０８ｗｔ％，酸素（Ｏ）１０００〜１５
００ｐｐｍおよび濃度１２０ｐｐｍ以下の不純物けい素
（Ｓｉ）を含み、残部はジルコニウム（Ｚｒ）からな
り、且つ、 Ｆｅ＋Ｃｒ＋Ｎｉ≦０．５２ｗｔ％ を満足するように調節したジルコニウム合金を溶製して
鋳片としたのちβ焼入れ処理を施し、その後、材料への
総入熱量ΣＡi が１×１０-19 〜１×１０-17 の範囲内
に納まるように圧延および焼鈍処理を施すものである。The method for producing a zirconium alloy-based reactor core structural material excellent in corrosion resistance, particularly uniform corrosion resistance and hydrogen absorption resistance, according to the invention described in claim 2, In the method for manufacturing a zirconium alloy-based reactor core structural material, tin (Sn); 0.8 to 1.2 wt%,
Iron (Fe); 0.17 to 0.25 wt%, chromium (C
r); 0.15-0.25 wt%, nickel (Ni);
0.01-0.08 wt%, oxygen (O) 1000-15
A zirconium alloy containing 00 ppm and a concentration of 120 ppm or less of impurity silicon (Si) and the balance of zirconium (Zr) and adjusted to satisfy Fe + Cr + Ni ≦ 0.52 wt% was melted to form a cast piece. After that, β quenching is performed, and then rolling and annealing are performed so that the total heat input ΣA i to the material falls within the range of 1 × 10 −19 to 1 × 10 −17 .

【００１４】[0014]

【発明の実施の形態】ジルカロイの耐食性は構造材とし
ての最終成形品に至るまでの製造工程中における熱処理
条件により影響される。ジルカロイによる構造材の製造
工程はビレットの段階でβ焼入れと呼ばれる溶体化処理
が行なわれ、次に熱間押出、進行焼鈍を行ない、その後
冷間圧延および真空焼鈍を３回繰返して行なう、という
ものである。BEST MODE FOR CARRYING OUT THE INVENTION The corrosion resistance of zircaloy is affected by the heat treatment conditions in the manufacturing process up to the final molded product as a structural material. In the manufacturing process of Zircaloy structural materials, solution treatment called β quenching is performed at the billet stage, followed by hot extrusion and progressive annealing, followed by cold rolling and vacuum annealing repeated three times. Is.

【００１５】β焼入れ後におけるα相領域での総入熱量
を表したパラメータとして、次式に示す累積入熱パラメ
ータΣＡi があり、このパラメータΣＡi がジルカロイ
の耐食性と良い相関を持つことが提案されている。ΣＡ
i が小さいと耐ノジュラー腐食特性が向上し、ΣＡi が
大きいと耐一様腐食特性が向上する。A cumulative heat input parameter ΣA i represented by the following equation is a parameter representing the total heat input in the α phase region after β quenching, and it is proposed that this parameter ΣA i has a good correlation with the corrosion resistance of zircaloy. Has been done. ΣA
When i is small, the nodular corrosion resistance is improved, and when ΣA i is large, the uniform corrosion resistance is improved.

【００１６】ΣＡi ＝Σｔi ・ｅｘｐ（−Ｑ／ＲＴi ） （但し、ｔi はβ焼入れ処理後のα相温度における熱処
理工程ｉの処理時間（単位：時間）を示し、Ｔi は工程
ｉの処理温度（単位：Ｋ）を示し、Ｑは活性化エネルギ
ー（単位：Ｊ／ｍｏｌ・Ｋ）を示し、Ｑ／Ｒ＝４０００
０Ｋである。）[0016] ΣA i = Σt i · exp ( -Q / RT i) ( where, t i is the heat treatment step i of processing time at α phase temperature after β quenching (unit: represents the time), T i the process i represents the treatment temperature (unit: K), Q represents the activation energy (unit: J / mol · K), and Q / R = 4000.
0K. )

【００１７】従来のジルカロイ−２についてこの総入熱
量ΣＡi を検討したところ、ΣＡiが１×１０-19 より
少ない場合、得られるジルコニウム合金は耐ノジュラー
腐食性に優れるものの耐一様腐食性が劣り、一方１×１
０-17 を越える場合、耐一様腐食性は優れるものの耐ノ
ジュラー腐食性が劣ることがわかっている。When the total heat input ΣA i of the conventional Zircaloy-2 was examined, when ΣA i was less than 1 × 10 −19 , the obtained zirconium alloy was excellent in nodular corrosion resistance but uniform corrosion resistance. Inferior, while 1 × 1
When it exceeds 0 -17 , it is known that the uniform corrosion resistance is excellent but the nodular corrosion resistance is poor.

【００１８】従って、従来のジルカロイ−２において
も、β焼入れ後の圧延および焼鈍処理における材料への
総入熱量ΣＡi を１×１０-19 〜１×１０-17 の範囲内
とすれば耐ノジュラー腐食性と耐一様腐食性との双方に
最適化させることができるが、この最適化範囲内であっ
ても、一様腐食性は１２００日の長期試験の結果から約
６００日程度で腐食速度が加速されることが判明した。Therefore, also in the conventional Zircaloy-2, if the total heat input ΣA i to the material in the rolling and annealing treatment after β quenching is within the range of 1 × 10 -19 to 1 × 10 -17 , nodular resistance is obtained. Although it is possible to optimize both the corrosion resistance and the uniform corrosion resistance, even within this optimization range, the uniform corrosion resistance is about 600 days from the result of the long-term test of 1200 days. Turned out to be accelerated.

【００１９】そこで本発明では、前記ΣＡi の最適化範
囲内で更に長期の耐一様腐食性を得るべく、まず、錫
（Ｓｎ）；０．８〜１．６ｗｔ％，鉄（Ｆｅ）；０．１
７〜０．２５ｗｔ％，クロム（Ｃｒ）；０．１５〜０．
２５ｗｔ％，ニッケル（Ｎｉ）；０．０１〜０．０８ｗ
ｔ％および濃度１２０ｐｐｍ以下の不純物けい素（Ｓ
ｉ）を含み、残部はジルコニウム（Ｚｒ）からなり、且
つ、Ｆｅ＋Ｃｒ＋Ｎｉ≦０．５２ｗｔ％を満足するよう
に調節したジルコニウム合金素材を溶製して鋳片とす
る。Therefore, in the present invention, in order to obtain a longer-term uniform corrosion resistance within the optimized range of ΣA i , first, tin (Sn); 0.8 to 1.6 wt%, iron (Fe); 0.1
7 to 0.25 wt%, chromium (Cr); 0.15 to 0.
25 wt%, nickel (Ni); 0.01 to 0.08 w
Impurity silicon (S
A slab is made by melting a zirconium alloy material containing i) and the balance being zirconium (Zr) and adjusted so as to satisfy Fe + Cr + Ni ≦ 0.52 wt%.

【００２０】ここで、Ｓｎは、その添加により合金の強
度を増加させ、水素吸収量を低く抑える効果を有し、ま
た合金表面に形成された酸化膜の剥離を防止する働きが
ある。但し、含有率が０．７ｗｔ％以下では充分な強度
が得られず、また、含有率が１．７ｗｔ％以上となる場
合には、一様腐食を加速させることがある。Here, Sn has the effect of increasing the strength of the alloy and suppressing the hydrogen absorption amount by the addition thereof, and also has the function of preventing the peeling of the oxide film formed on the alloy surface. However, if the content rate is 0.7 wt% or less, sufficient strength cannot be obtained, and if the content rate is 1.7 wt% or more, uniform corrosion may be accelerated.

【００２１】従って、Ｓｎは０．８〜１．６ｗｔ％の含
有率とし、高燃焼度および長期使用にあたって十分な強
度と耐水素吸収特性を発揮させるものとする。この含有
率の範囲は、従来のジルカロイにおけるＳｎ含有率範囲
の下限側を拡げるものである。特に、Ｓｎ０．８〜１．
２ｗｔ％と上記範囲内でも下限寄りの含有率とすると、
さらに耐食性が向上され、かつ耐食性に関して優れた性
能を発揮するΣＡi の領域を上端側、即ち耐一様腐食性
がより向上する側に広げる。Therefore, the content of Sn is set to 0.8 to 1.6 wt% so as to exhibit high burnup and sufficient strength and hydrogen absorption resistance during long-term use. This range of content extends to the lower limit side of the Sn content range in the conventional Zircaloy. In particular, Sn 0.8-1.
If the content rate is 2 wt%, which is closer to the lower limit even within the above range,
Further, the region of ΣA i where the corrosion resistance is improved and exhibits excellent performance in terms of the corrosion resistance is expanded to the upper end side, that is, the side where the uniform corrosion resistance is further improved.

【００２２】また、Ｆｅは、その添加によりジルコニウ
ム合金に対して強度に影響を与えることなく、添加量増
加に伴って耐一様腐食性および耐ノジュラー腐食性、耐
水素吸収性を向上させる。但し、Ｆｅの含有率が０．１
６ｗｔ％以下では一様腐食およびノジュラー腐食に対し
て十分な耐食性が得られず、含有率が０．２５ｗｔ％を
越えた場合はΣＡi の範囲を広げ得る反面、溶接熱影響
部の耐食性劣化が顕著になる。The addition of Fe improves the uniform corrosion resistance, the nodular corrosion resistance, and the hydrogen absorption resistance with an increase in the addition amount without affecting the strength of the zirconium alloy. However, the Fe content is 0.1
If it is 6 wt% or less, sufficient corrosion resistance against uniform corrosion and nodular corrosion cannot be obtained, and if the content exceeds 0.25 wt%, the range of ΣA i can be widened, but the corrosion resistance of the weld heat affected zone deteriorates. It will be noticeable.

【００２３】従って、Ｆｅは０．１７〜０．２５ｗｔ％
の含有率とし、耐一様腐食性、耐ノジュラー腐食性、耐
水素吸収性および溶接熱影響部の耐食性についての特性
を発揮させるものとする。Therefore, Fe is 0.17 to 0.25 wt%.
As the content ratio of, the properties of uniform corrosion resistance, nodular corrosion resistance, hydrogen absorption resistance, and corrosion resistance of the heat affected zone of welding are exhibited.

【００２４】また、Ｎｉは、その添加によりジルコニウ
ム合金の耐食性向上の効果を有する一方、水素吸収率を
高める性質がある。水素吸収率を高めないで耐食性の向
上が得られるＮｉの含有率の範囲として、Ｎｉは０．０
１〜０．０８ｗｔ％の含有率とした。Further, Ni has the effect of improving the corrosion resistance of the zirconium alloy by adding it, and has the property of increasing the hydrogen absorption rate. As the Ni content range in which the corrosion resistance is improved without increasing the hydrogen absorption rate, Ni is 0.0
The content rate was 1 to 0.08 wt%.

【００２５】ところで、前述したように従来にくらべて
Ｓｎの含有率が下限側に拡がり、Ｓｎを比較的低い含有
率とする場合、Ｎｉをその含有率範囲内の上限側の高い
含有率とした場合には、一様腐食の加速が問題となって
しまう。特に、Ｓｎの含有率が下限寄り、即ち０．８〜
１．２ｗｔ％という範囲にある時には、Ｎｉの含有率が
高くなると一様腐食の加速は顕著に現れる。By the way, as described above, when the content of Sn spreads to the lower limit side as compared with the conventional case and Sn is set to a relatively low content rate, Ni is set to a high content rate on the upper limit side within the content range. In this case, acceleration of uniform corrosion becomes a problem. In particular, the Sn content is closer to the lower limit, that is, 0.8 to
When it is in the range of 1.2 wt%, the acceleration of uniform corrosion becomes remarkable as the Ni content increases.

【００２６】通常、一様腐食は原子炉燃焼において時間
経過に伴って直線的に増加していくが、上記のようにＳ
ｎの含有率が比較的低くてＮｉの含有率が比較的高い場
合には、Ｎｉが持つ水素吸収率を高める性質の方がより
顕著にでてきてしまい、高燃焼度または長時間燃焼時に
炉心内構造材を構成するジルコニウム合金の水素吸収率
が高まり、また一様腐食の加速が生じてしまう。Normally, uniform corrosion increases linearly with the passage of time in reactor combustion.
When the content of n is relatively low and the content of Ni is relatively high, the property of increasing the hydrogen absorption rate of Ni becomes more prominent, and the core at high burnup or long burning The hydrogen absorption rate of the zirconium alloy constituting the inner structural material is increased, and uniform corrosion is accelerated.

【００２７】そこで、本発明では、Ｃｒの含有率を適当
な範囲とすることによって、この一様腐食の加速の問題
を解消した。具体的には、Ｃｒは、その添加によってジ
ルコニウム合金の強度に影響を与えることなく、Ｆｅの
効果に比べれば弱いながらも耐食性を向上させると共
に、水素吸収率を低減する働きを持つものである。従来
のジルカロイ−２ではこのようなＣｒの含有率を０．１
０ｗｔ％としていたが、本発明ではＣｒの含有率を０．
１５％ｗｔ以上と従来に比べて高く設定することによっ
て、水素吸収率が高まるのを抑え、高燃焼度または長時
間燃焼時における一様腐食の加速の発生を抑制すること
ができた。Therefore, in the present invention, the problem of acceleration of uniform corrosion is solved by setting the Cr content in an appropriate range. Specifically, Cr does not affect the strength of the zirconium alloy by the addition thereof, and although it is weaker than the effect of Fe, it has the function of improving the corrosion resistance and reducing the hydrogen absorption rate. In the conventional Zircaloy-2, such a Cr content is 0.1
Although it was set to 0 wt%, in the present invention, the content ratio of Cr is 0.
By setting it to 15% wt or more, which is higher than the conventional value, it was possible to suppress an increase in the hydrogen absorption rate, and to suppress the occurrence of uniform corrosion acceleration at high burnup or during long-term combustion.

【００２８】ただし、Ｃｒの含有率が０．２５ｗｔ％を
越えた場合には、Ｆｅと同様に溶接熱影響部の耐食性の
劣化が顕著となる。従って、本発明では、Ｃｒは０．１
５〜０．２５ｗｔ％の含有率とした。However, when the Cr content exceeds 0.25 wt%, the corrosion resistance of the weld heat affected zone is markedly deteriorated as in the case of Fe. Therefore, in the present invention, Cr is 0.1
The content rate was 5 to 0.25 wt%.

【００２９】また、本発明では、Ｆｅ＋Ｃｒ＋Ｎｉ≦
０．５２ｗｔ％とした。これは、ジルコニウム系合金で
はこれら３元素の合計の含有率が０．５３％以上の場合
に、溶接熱影響部の耐食性の劣化が顕著となるためであ
る。In the present invention, Fe + Cr + Ni≤
It was 0.52 wt%. This is because in the zirconium-based alloy, when the total content of these three elements is 0.53% or more, the corrosion resistance of the weld heat affected zone is significantly deteriorated.

【００３０】なお、不純物元素であるＳｉは、ジルコニ
ウム合金の耐ノジュラー腐食性に影響を及ぼすものであ
り、濃度が１２０ｐｐｍ以上の場合は、ノジュラー腐食
が生じることがあるため、一般的に１２０ｐｐｍ以下の
濃度に抑える必要があり、この数値はＡＳＴＭ規格でも
定められている。従って、本発明においてもＳｉの濃度
は１２０ｐｐｍ以下とし、もちろん、例えば９０ｐｐｍ
以下，６０ｐｐｍ以下とさらに減じることが望ましいこ
とは言うまでもない。Si, which is an impurity element, affects the nodular corrosion resistance of the zirconium alloy, and when the concentration is 120 ppm or more, nodular corrosion may occur. Therefore, it is generally 120 ppm or less. It is necessary to suppress the concentration, and this value is also defined in the ASTM standard. Therefore, also in the present invention, the Si concentration is 120 ppm or less, and of course, for example, 90 ppm.
It goes without saying that it is desirable to further reduce it to 60 ppm or less.

【００３１】また、酸素は従来よりジルコニウム合金へ
の添加に実績があり、これはジルコニウム合金の耐食性
に影響を与えることなく強度を向上させる効果がある。
本発明では、特にＳｎが０．８〜１．２ｗｔ％の比較的
低い含有率範囲にある時の強度低下を補うために、１０
００〜１５００ｐｐｍの範囲内で酸素を添加する。Oxygen has been added to zirconium alloys in the past, and this has the effect of improving the strength without affecting the corrosion resistance of the zirconium alloy.
In the present invention, in order to compensate for the strength decrease especially when Sn is in a relatively low content range of 0.8 to 1.2 wt%, 10
Oxygen is added within the range of 00 to 1500 ppm.

【００３２】上記の如き本発明の製造方法で得られたジ
ルコニウム系合金の原子炉炉心内構造材は、一様腐食性
の長期試験において１２００日経過しても腐食速度が加
速されることのないものであり、例えばＢＷＲ燃料集合
体取り出し燃焼度３９０００ＭＷＤ／ｔ以上の高燃焼度
条件下であっても、従来と同程度の良好な耐ノジュラー
腐食性を持ちながら、優れた耐一様腐食性および耐水素
吸収性を示すものである。The zirconium alloy structural material in the reactor core obtained by the manufacturing method of the present invention as described above does not accelerate the corrosion rate even after 1200 days in the long-term uniform corrosion test. For example, even under a high burn-up condition of, for example, BWR fuel assembly take-out burn-up of 39000 MWD / t or more, while maintaining good nodular corrosion resistance comparable to the conventional one, excellent uniform corrosion resistance and It exhibits hydrogen absorption resistance.

【００３３】[0033]

【実施例】本発明による製造方法で得られた種々の組成
のジルコニウム合金系原子炉炉心内構造材について、比
較例と共に、それぞれＢＷＲ高燃焼度運転の炉内環境相
当の条件下における耐ノジュラー腐食試験、耐一様腐食
試験、耐水素吸収性試験、耐溶接部腐食試験、機械特性
試験によって材料評価を行なった。EXAMPLES With respect to zirconium alloy-based nuclear reactor core structural materials of various compositions obtained by the manufacturing method according to the present invention, along with comparative examples, nodular corrosion resistance under conditions corresponding to the internal environment of BWR high burnup operation, respectively. Materials were evaluated by tests, uniform corrosion resistance test, hydrogen absorption resistance test, weld corrosion resistance test, and mechanical property test.

【００３４】まず製造工程は、図１のフローチャート図
で示すように、各組成の素材を溶製してそれぞれインゴ
ットを製造し、１０１５℃以上でβ焼入れを行なった
後、それぞれ総入熱量ΣＡi が１×１０-20 〜１×１０
-16 の範囲内の或る値におさまるように熱間圧延、焼
鈍、複数回の冷間圧延と焼鈍の繰返し（最終焼鈍５７７
℃）を行なって、原子炉炉心内構造材としての最終成形
品を得た。First, as shown in the flow chart of FIG. 1, in the manufacturing process, raw materials of each composition are melted to manufacture respective ingots, β-quenched at 1015 ° C. or more, and then the total heat input ΣA i. Is 1 × 10 −20 to 1 × 10
-16 hot rolling, annealing, multiple cold rolling and annealing (final annealing 577)
C.) to obtain a final molded product as a structural material in the reactor core.

【００３５】[0035]

【表１】 [Table 1]

【００３６】上記工程で得られた製品を試験片として、
表１に示すごとく、各種耐食試験を行なった。即ち、耐
ノジュラー腐食試験では試験片を循環式オートクレー
ブ内で５３０℃，１０５気圧の条件下に２４時間置いた
後、腐食増量を測定した。ここで、腐食増量が１００ｍ
ｇ／ｄｍ2 以下のものをランクＡとし、それ以上の腐食
増量の大きいものをランクＢと評価した。The product obtained in the above process is used as a test piece,
As shown in Table 1, various corrosion resistance tests were conducted. That is, in the nodular corrosion resistance test, the test piece was placed in a circulating autoclave under the conditions of 530 ° C. and 105 atmospheric pressure for 24 hours, and then the increase in corrosion was measured. Here, the increase in corrosion is 100 m
Those with g / dm 2 or less were evaluated as rank A, and those with a larger corrosion increase were evaluated as rank B.

【００３７】また、耐一様腐食試験では、試験片を循
環式オートクレーブ内で４００℃，１０５気圧の条件下
に置き、１２００日という長期の間、一様腐食増量を継
続的に測定し、腐食の加速が生じているかどうかを観測
した。観測結果の評価は、図２のグラフに示したように
Ａ，Ｂ1 ，Ｂ2 の３つにレベル別けした。In the uniform corrosion resistance test, the test piece was placed in a circulation type autoclave under the conditions of 400 ° C. and 105 atm, and the uniform corrosion amount was continuously measured for a long period of 1200 days to measure the corrosion. It was observed whether or not the acceleration was occurring. As shown in the graph of FIG. 2, the evaluation of the observation result was classified into three levels of A, B 1 and B 2 .

【００３８】即ち、加速が生じていないものをランクＡ
と評価し、また、加速が生じているもののうち、試験開
始６００日目ぐらいから加速し始めた加速程度が遅いも
のをランクＢ2 、試験開始４００日目ぐらいから加速し
始めた加速程度が早いものをランクＢ1 と評価した。That is, the one in which acceleration is not generated is rank A.
Also, of the accelerations that occurred, those that started to accelerate from the start of the test about 600 days and started slower were ranked B 2 , and those that started to accelerate from the start of the test about 400 days earlier. The item was evaluated as rank B 1 .

【００３９】また、耐水素吸収性試験では、上記耐一
様腐食試験における４００℃，１０５気圧条件下で２０
００時間（８３．３日）置いた試験片について、腐食測
定後に水素吸収率を測定した。水素吸収率４５％以下の
ものをランクＡ、４５％以上のものをランクＢと評価し
た。In addition, in the hydrogen absorption resistance test, 20 under the conditions of 400 ° C. and 105 atmospheric pressure in the uniform corrosion resistance test.
The hydrogen absorption rate of the test piece left for 00 hours (83.3 days) was measured after the corrosion measurement. Those having a hydrogen absorption rate of 45% or less were evaluated as rank A, and those having a hydrogen absorption rate of 45% or more were evaluated as rank B.

【００４０】また、耐溶接（熱影響）部腐食試験とし
て、上記耐一様腐食試験における４００℃，１０５気圧
条件下で４５０日置いた試験片について、腐食測定後に
熱影響部の酸化皮膜厚を測定し、熱影響部以外の母材の
酸化皮膜厚と同等のものをランクＡ、母材の酸化皮膜厚
より大きいものをランクＢと評価した。As a welding (heat-affected) zone corrosion test, the oxide film thickness of the heat-affected zone was measured after corrosion measurement for a test piece placed for 450 days under the conditions of 400 ° C. and 105 atm in the above uniform corrosion test. It was measured, and those having the same oxide film thickness as that of the base material other than the heat-affected zone were ranked as Rank A, and those having a larger oxide film thickness as the base material were ranked as Rank B.

【００４１】機械特性試験としては室温における引張
試験を行なった。この場合、引張強さが５０ｋｇ／ｍｍ
2 以上であると共に耐力が３５ｋｇ／ｍｍ2 以上、且つ
伸び３０％以上のものをランクＡとし、これらの条件を
満たさないものをランクＢと評価した。A tensile test at room temperature was carried out as a mechanical property test. In this case, the tensile strength is 50 kg / mm
Those having a proof stress of 2 or more and a proof stress of 35 kg / mm 2 or more and an elongation of 30% or more were evaluated as rank A, and those not satisfying these conditions were evaluated as rank B.

【００４２】まず、Ｓｎが１．７ｗｔ％と本発明におけ
る範囲上限を越えた高い含有率で、Ｃｒが０．１ｗｔ％
と本発明における範囲下限を越えた従来と同様の低い含
有率の場合および本発明における範囲内の含有率０．１
５〜０．１７ｗｔ％である場合とで、Ｆｅ；０．１９ｗ
ｔ％，Ｎｉ；０．０１〜０．０８ｗｔ％，残りＺｒとい
う各組成の素材について、総入熱量ΣＡi を従来の最適
化領域である１×１０ -19 〜１×１０-17 で製造した試
験片No.41 〜50について、上記〜の評価試験を行な
った。結果を表２に示す。First, in the present invention, Sn is 1.7 wt%.
Cr content of 0.1 wt%
And a low content similar to the conventional one, which exceeds the lower limit of the range in the present invention.
In the case of a ratio and the content ratio within the range of the present invention 0.1
In the case of 5 to 0.17 wt%, Fe; 0.19w
t%, Ni; 0.01 to 0.08 wt%, remaining Zr
Total heat input ΣA for each compositioni The conventional optimum
1 × 10 which is the digitized area -19 ~ 1 × 10-17 Trial manufactured in
Perform the above-mentioned evaluation tests for Specimen Nos. 41-50.
Was. Table 2 shows the results.

【００４３】[0043]

【表２】 [Table 2]

【００４４】表２からもわかるように、Ｓｎの含有率が
１．７ｗｔ％と高い場合には、Ｃｒの含有率が低め（N
o.41 〜46）であっても高め（No.47 〜50）であって
も、またＮｉの含有率が上限寄り（No.41 ，No.49 ，5
0）であっても下限寄り（No42〜48）であっても、いず
れにしても耐一様腐食試験でランクＢ1 の加速腐食が
見られた。Ｓｎの含有率が１．７ｗｔ％以上であると一
様腐食は加速されてしまうと考えられ、このような組成
では、長期使用される耐食性原子炉炉内構造材としては
不十分である。本発明では、まずＳｎの含有率を０．８
〜１．６ｗｔ％とすることによって優れた耐一様腐食性
を有する製品を得ようとするものである。As can be seen from Table 2, when the Sn content is as high as 1.7 wt%, the Cr content is low (N
o.41 to 46) or higher (No.47 to 50), the Ni content is closer to the upper limit (No.41, No.49, 5).
Accelerated corrosion of rank B 1 was observed in the uniform corrosion resistance test regardless of whether it was 0) or near the lower limit (No 42 to 48). It is considered that if the Sn content is 1.7 wt% or more, uniform corrosion will be accelerated, and such a composition is insufficient as a corrosion-resistant nuclear reactor internal structural material that is used for a long period of time. In the present invention, first, the Sn content is set to 0.8.
By setting the content to ˜1.6 wt%, it is intended to obtain a product having excellent uniform corrosion resistance.

【００４５】以下、Ｓｎの含有率が０．８〜１．６ｗｔ
％という本発明の範囲内にある場合に得られた製品（試
験片）の耐食性の評価を行なった。まず、Ｓｎの含有率
が１．３〜１．５ｗｔ％と前記範囲内の上限側にある際
の各組成の場合における上記〜の評価試験を行な
い、その結果を表３（No.1〜19）に示す。Below, the Sn content is 0.8 to 1.6 wt.
The corrosion resistance of the product (test piece) obtained when it was within the range of the present invention of% was evaluated. First, the above-mentioned evaluation tests were performed for each composition when the Sn content was 1.3 to 1.5 wt% and were on the upper limit side within the above range, and the results are shown in Table 3 (No. 1 to 19). ).

【００４６】このとき、Ｃｒについては０．１ｗｔ％の
本発明における範囲下限を越えた従来同様に低い含有率
の場合（No.1〜13) と、０．１５〜０．２５ｗｔ％の本
発明の範囲内の含有率である場合（No.14 〜19）とで、
それぞれＮｉが０．０１〜０．０８ｗｔ％の間の各含有
率について、Ｆｅ；０．１９ｗｔ％，残りＺｒという各
組成の素材において、総入熱量ΣＡi を従来の最適化領
域である１×１０-19〜１×１０-17 およびその下限を
越えた１×１０-20 と上限を越えた１×１０-1 6 で製造
した試験片をそれぞれ評価した。At this time, in the case where the Cr content is as low as 0.1 wt%, which is below the lower limit of the present invention, as in the conventional case (No. 1 to 13), the present invention is 0.15 to 0.25 wt%. When the content rate is within the range (No.14 to 19),
For each content of Ni between 0.01 and 0.08 wt%, the total heat input ΣA i is 1 ×, which is the conventional optimization range, in the material of each composition of Fe: 0.19 wt% and the remaining Zr. 10 -19 ~1 × 10 -17 and the lower limit exceeds 1 × 10 -20 and specimens prepared in 1 × 10 -1 6 exceeding the upper limit was evaluated respectively.

【００４７】[0047]

【表３】 [Table 3]

【００４８】まず、Ｃｒの含有率が０．１ｗｔ％のもの
で、Ｎｉの含有率が０．０６〜０．０８ｗｔ％と高い
（上限寄り）もの（No.1〜 5）では、総入熱量ΣＡi が
いずれのものであっても、全て耐一様腐食試験ではラ
ンクＢ1 あるいはＢ2 で一様腐食に加速が生じていただ
けでなく、耐水素吸収試験でもランクＢであった。First, when the Cr content is 0.1 wt% and the Ni content is as high as 0.06 to 0.08 wt% (close to the upper limit) (No. 1 to 5), the total heat input is Whatever ΣA i , not only accelerated uniform corrosion at rank B 1 or B 2 in the uniform corrosion resistance test, but also rank B in the hydrogen absorption resistance test.

【００４９】また、Ｃｒの含有率が０．１ｗｔ％のもの
で、Ｎｉの含有率が０．０１〜０．０５ｗｔ％と低い
（下限寄り）もの（No.6〜13）でも、総入熱量ΣＡi が
１×１０-19 〜１×１０-17 以外としたものでは耐食性
に問題があった。即ち、ΣＡiを１×１０-20 としたも
の（No.6，No.11 ）では耐一様腐食試験でランクＢ1
を示し、ΣＡi を１×１０-16 としたもの（No.8，No.1
3 ）では耐ノジュラー腐食試験でランクＢを示した。Even if the Cr content is 0.1 wt% and the Ni content is as low as 0.01 to 0.05 wt% (close to the lower limit) (No. 6 to 13), the total heat input is If ΣA i is other than 1 × 10 −19 to 1 × 10 −17 , there is a problem in corrosion resistance. That is, in the case where ΣA i is set to 1 × 10 -20 (No.6, No.11), the rank B 1 in uniform corrosion resistance test
With ΣA i of 1 × 10 -16 (No.8, No.1
In No. 3), Rank B was shown in the nodular corrosion resistance test.

【００５０】これらに対して、Ｃｒの含有率が０．１５
〜０．２５ｗｔ％で総入熱量ΣＡiを１×１０-19 〜１
×１０-17 としたもの（No.14 〜19）では、Ｎｉの含有
率が０．０１〜０．０８ｗｔ％の範囲内において上限寄
りであろうと下限寄りであろうと、いずれの試験片も評
価試験〜全てでランクＡを示した。On the other hand, the Cr content is 0.15
~ 0.25 wt% total heat input ΣA i is 1 × 10 -19 -1
In the case of x10 -17 (Nos. 14 to 19), all the test pieces were evaluated regardless of whether the Ni content was in the range of 0.01 to 0.08 wt%, near the upper limit or near the lower limit. A rank A was shown in all the tests.

【００５１】ここで、表３中には示していないが、表３
で示したうち耐ノジュラー腐食試験でランクＡを示し
たのものの組成および入熱量と同一条件で、但し合金素
材のＳｉの濃度が１２０ｐｐｍ以上となるよう調製した
ものについて、耐ノジュラー腐食試験を別途に行なった
ところ、評価はランクＢと耐食性の劣るものであった。
また、Ｓｉの濃度を１２０ｐｐｍ以下とすると、耐ノジ
ュラー腐食特性は向上し、９０ｐｐｍ以下でランクＡと
なる。Although not shown in Table 3, Table 3
In the Nodular Corrosion Resistance Test, a Nodular Corrosion Resistance Test is separately conducted for the composition and heat input under the same conditions as those of Rank A in the Nodular Corrosion Resistance Test, except that the alloy material has a Si concentration of 120 ppm or more. As a result, the evaluation was rank B and the corrosion resistance was inferior.
Further, when the Si concentration is 120 ppm or less, the nodular corrosion resistance is improved, and it becomes rank A at 90 ppm or less.

【００５２】さらに、Ｓｎが１．００ｗｔ％と前記範囲
内の下限寄りにある際の各組成の場合における上記〜
の評価試験を行なった。その結果は以下の表４（No.2
1 〜32）に示す。Further, in the case of each composition when Sn is near the lower limit of 1.00 wt%, the above-mentioned
Was evaluated. The results are shown in Table 4 (No. 2 below).
1 to 32).

【００５３】なお、このとき、Ｃｒについては０．１ｗ
ｔ％の本発明における範囲下限を越えた従来同様に低い
含有率の場合（No.21 〜26) と、０．１５〜０．２５ｗ
ｔ％の本発明の範囲内である場合（No.27 〜32）とで、
それぞれＮｉが０．０１〜０．０８ｗｔ％の間の各含有
率について、Ｆｅ；０．１９ｗｔ％，残りＺｒという各
組成の素材について、総入熱量ΣＡi を最適化領域であ
る１×１０-19 〜１×１０-17 で製造した試験片をそれ
ぞれ評価した。At this time, Cr is 0.1 w
When the content is as low as the conventional one, which exceeds the lower limit of t% in the present invention (No. 21 to 26), 0.15 to 0.25w
When t% is within the range of the present invention (No. 27 to 32),
For each content of Ni between 0.01 and 0.08 wt%, Fe: 0.19 wt%, and the remaining Zr, the total heat input ΣA i is 1 × 10 − Each of the test pieces manufactured at 19 to 1 × 10 -17 was evaluated.

【００５４】[0054]

【表４】 [Table 4]

【００５５】まず、表４から明らかなように、Ｃｒの含
有率が０．１ｗｔ％のものでも、Ｎｉの含有率が０．０
１〜０．０５ｗｔ％とあまり高くない場合（No.22 〜2
6）では全ての評価試験（〜）でランクＡを示した
が、Ｎｉの含有率が０．０８ｗｔ％と高い場合（No.21
）には、やはり耐一様腐食試験でランクＢ1 を示す
と共に耐水素吸収性試験でもランクＢであった。First, as is clear from Table 4, even if the Cr content is 0.1 wt%, the Ni content is 0.0
When it is not so high as 1 to 0.05 wt% (No.22 to 2
In 6), all evaluation tests (-) showed rank A, but when the Ni content was as high as 0.08 wt% (No. 21).
) Also shows rank B 1 in the uniform corrosion resistance test and rank B in the hydrogen absorption resistance test.

【００５６】一方、Ｓｎの含有率が比較的低い場合で
も、従来に比べて高い本発明のＣｒの０．１５〜０．２
５ｗｔ％という範囲内の含有率であれば、Ｎｉの含有率
が０．０１〜０．０８ｗｔ％の範囲内で下限寄りである
場合はもちろん、０．０８ｗｔ％を含む上限寄り側の高
い含有率の場合であっても、いずれも（No.27 〜32）全
ての評価試験（〜）でランクＡを示した。On the other hand, even when the Sn content is relatively low, the content of Cr of the present invention is higher than that of the conventional one of 0.15 to 0.2.
If the content ratio is within the range of 5 wt%, not only when the Ni content ratio is within the range of 0.01 to 0.08 wt% near the lower limit, but also the higher content ratio near the upper limit including 0.08 wt%. In each case (No. 27 to 32), all evaluation tests (to) showed rank A.

【００５７】なお、表４中には示していないが、試料N
o.22 〜32について溶製時に酸素を１０００〜１５００
ｐｐｍの濃度で添加したものについて同様の試験を行な
ったところ、全ての耐食性に影響なく機械的強度がジル
カロイ−２と同等以上に向上したことが確認された。Although not shown in Table 4, sample N
O.22-32 Oxygen 1000-1500 during melting
When the same test was performed for the one added at the concentration of ppm, it was confirmed that the mechanical strength was improved to the same level or more as Zircaloy-2 without affecting all the corrosion resistance.

【００５８】以上の結果から、本発明では、合金素材の
組成において、Ｓｎの含有率が０．８〜１．６ｗｔ％と
比較的低く、Ｃｒの含有率を０．１５〜０．２５ｗｔ％
と従来より高くし、総入熱量ΣＡi を１×１０-19 〜１
×１０-17 としてジルコニウム合金系構造材を製造する
ことによって、高燃焼度または長期使用の原子炉炉心内
構造材として、特に耐一様腐食性と耐水素吸収性が従来
のジルコニウム系合金に比べて改善されたものが得られ
ることがわかる。なお、第３冷間圧延後の最終焼鈍処理
によって内部歪や応力を除去した構造材とすることもで
きるのは述べるまでもない。From the above results, in the present invention, in the composition of the alloy material, the Sn content is relatively low at 0.8 to 1.6 wt%, and the Cr content is 0.15 to 0.25 wt%.
And the total heat input ΣA i is 1 × 10 -19 -1
By producing a zirconium alloy-based structural material of × 10 -17 , it is possible to obtain high burnup or long-term use as a structural material in a nuclear reactor core, especially in terms of uniform corrosion resistance and hydrogen absorption resistance compared to conventional zirconium-based alloys. It can be seen that an improved product can be obtained. Needless to say, a structural material from which internal strain and stress have been removed by the final annealing treatment after the third cold rolling can be made.

【００５９】従って、上記の如き本発明の組成および入
熱量で製造されたジルコニウム合金系原子炉炉心内構造
材は、例えばＢＷＲ燃料集合体取り出し燃焼度３９００
０ＭＷＤ／ｔ以上の高燃焼度条件下においても、優れた
耐水素吸収性のために水素吸収による水素脆化および照
射脆化が抑えられると共に、優れた耐一様腐食性のため
に長期使用における一様腐食の加速も抑えられる。従っ
て、従来のものに比べその使用寿命が伸び、より長期間
使用することが可能となる。Therefore, the zirconium alloy-based reactor core structural material produced with the composition and heat input of the present invention as described above is, for example, a BWR fuel assembly take-out burnup of 3900.
Even under conditions of high burn-up of 0 MWD / t or higher, hydrogen embrittlement and irradiation embrittlement due to hydrogen absorption are suppressed due to excellent hydrogen absorption resistance, and excellent uniform corrosion resistance results in long-term use. Acceleration of uniform corrosion can also be suppressed. Therefore, it has a longer service life than conventional ones and can be used for a longer period of time.

【００６０】また、一様腐食が現れ易いＰＷＲ用の炉心
内構造材としても、上記の如き本発明の製造方法によっ
て得られたジルコニウム合金系構造材であれば、その優
れた耐一様腐食性および耐水素吸収性は有効であり、よ
り長期間の使用が可能となる。Further, as a structural material in the core for a PWR in which uniform corrosion is likely to occur, if the zirconium alloy structural material obtained by the manufacturing method of the present invention as described above has excellent uniform corrosion resistance. And the hydrogen absorption resistance is effective, and it becomes possible to use it for a longer period of time.

【００６１】[0061]

【発明の効果】本発明は以上説明したとおり、耐食性、
特に長期の耐一様腐食性および耐水素吸収性に優れたジ
ルコニウム合金系原子炉炉心内構造材が得られるという
効果がある。従って、このような耐一様腐食性および耐
水素吸収性に優れたジルコニウム合金系原子炉炉心内構
造材は例えばＢＷＲ燃料集合体取り出し燃焼度３９００
０ＭＷＤ／ｔ以上の高燃焼度条件下において、或はＰＷ
Ｒにおいてもより長期間用いることが可能となる。As described above, the present invention provides corrosion resistance,
In particular, there is an effect that a zirconium alloy-based reactor core structural material having excellent long-term uniform corrosion resistance and hydrogen absorption resistance can be obtained. Accordingly, such a zirconium alloy-based nuclear reactor core structural material having excellent uniform corrosion resistance and hydrogen absorption resistance has, for example, a BWR fuel assembly removal burnup of 3900.
Under high burn-up condition of 0 MWD / t or more, or PW
It is possible to use R for a longer period of time.

【図面の簡単な説明】[Brief description of drawings]

【図１】本発明の一実施例によるジルコニウム合金系原
子炉炉心内構造材の製造工程を示すフローチャート図で
ある。FIG. 1 is a flow chart showing a manufacturing process of a zirconium alloy-based nuclear reactor core structural material according to an embodiment of the present invention.

【図２】耐一様腐食試験における腐食加速レベルに応じ
た評価ランク別けを説明するためのグラフ図であり、横
軸に試験時間；日，縦軸に（一様）腐食増量；ｍｇ／ｄ
ｍ2 を示したものである。FIG. 2 is a graph for explaining evaluation ranks according to a corrosion acceleration level in a uniform corrosion resistance test, in which the horizontal axis represents test time; the vertical axis represents (uniform) corrosion increase amount; mg / d.
It shows m 2 .

Claims (2)

【特許請求の範囲】[Claims] 【請求項１】 錫（Ｓｎ）；０．８〜１．６ｗｔ％，鉄
（Ｆｅ）；０．１７〜０．２５ｗｔ％，クロム（Ｃ
ｒ）；０．１５〜０．２５ｗｔ％，ニッケル（Ｎｉ）；
０．０１〜０．０８ｗｔ％および濃度１２０ｐｐｍ以下
の不純物けい素（Ｓｉ）を含み、残部はジルコニウム
（Ｚｒ）からなり、且つ、 Ｆｅ＋Ｃｒ＋Ｎｉ≦０．５２ｗｔ％ を満足するように調節したジルコニウム合金を溶製して
鋳片としたのちβ焼入れ処理を施し、その後、下式で表
される材料への総入熱量ΣＡi が１×１０-19 〜１×１
０-17 の範囲内に納まるように圧延および焼鈍処理を施
すことを特徴とする耐食性、特に耐一様腐食性と耐水素
吸収性に優れたジルコニウム合金系原子炉炉心内構造材
の製造方法。 ΣＡi ＝Σｔi ・ｅｘｐ（−Ｑ／ＲＴi ） ｔi ：β焼入れ後のｉ番目熱処理工程における処理時間
（時間） Ｔi ：工程ｉの処理温度（Ｋ） Ｑ：活性化エネルギー Ｒ：気体定数 Ｑ／Ｒ：４００００1. Tin (Sn); 0.8 to 1.6 wt%, iron (Fe); 0.17 to 0.25 wt%, chromium (C)
r); 0.15-0.25 wt%, nickel (Ni);
A zirconium alloy containing 0.01 to 0.08 wt% and a concentration of 120 ppm or less of impurity silicon (Si), the balance being zirconium (Zr), and adjusted to satisfy Fe + Cr + Ni ≦ 0.52 wt% is melted. After being made into a slab, it is subjected to β quenching treatment, and then the total heat input ΣA i to the material represented by the following formula is 1 × 10 -19 to 1 × 1
A method for producing a zirconium alloy-based reactor core structural material having excellent corrosion resistance, particularly uniform corrosion resistance and hydrogen absorption resistance, which is characterized by performing rolling and annealing treatment so as to fall within a range of 0 -17 . ΣA i = Σt i · exp ( -Q / RT i) t i: β treatment time in the i-th heat treatment step after quenching (time) T i: step i treatment temperature (K) Q: activation energy R: gas Constant Q / R: 40000 【請求項２】 錫（Ｓｎ）；０．８〜１．２ｗｔ％，鉄
（Ｆｅ）；０．１７〜０．２５ｗｔ％，クロム（Ｃ
ｒ）；０．１５〜０．２５ｗｔ％，ニッケル（Ｎｉ）；
０．０１〜０．０８ｗｔ％，酸素（Ｏ）１０００〜１５
００ｐｐｍおよび濃度１２０ｐｐｍ以下の不純物けい素
（Ｓｉ）を含み、残部はジルコニウム（Ｚｒ）からな
り、且つ、 Ｆｅ＋Ｃｒ＋Ｎｉ≦０．５２ｗｔ％ を満足するように調節したジルコニウム合金を溶製して
鋳片としたのちβ焼入れ処理を施し、その後、材料への
総入熱量ΣＡi が１×１０-19 〜１×１０-17 の範囲内
に納まるように圧延および焼鈍処理を施すことを特徴と
する請求項１に記載の耐食性、特に耐一様腐食性と耐水
素吸収性に優れたジルコニウム合金系原子炉炉心内構造
材の製造方法。2. Tin (Sn); 0.8 to 1.2 wt%, iron (Fe); 0.17 to 0.25 wt%, chromium (C)
r); 0.15-0.25 wt%, nickel (Ni);
0.01-0.08 wt%, oxygen (O) 1000-15
A zirconium alloy containing 00 ppm and a concentration of 120 ppm or less of impurity silicon (Si) and the balance of zirconium (Zr) and adjusted to satisfy Fe + Cr + Ni ≦ 0.52 wt% was melted to form a slab. After that, β-quenching is performed, and then rolling and annealing are performed so that the total heat input ΣA i to the material falls within the range of 1 × 10 −19 to 1 × 10 −17. 2. A method for producing a zirconium alloy-based reactor core structural material having excellent corrosion resistance, particularly uniform corrosion resistance and hydrogen absorption resistance.
JP7235956A 1995-08-23 1995-08-23 Production of zirconium alloy-based reactor core structural material excellent in corrosion resistance, particularly in uniform corrosion resistance and hydrogen absorption resistance Withdrawn JPH0961570A (en)

Priority Applications (1)

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JP7235956A JPH0961570A (en) 1995-08-23 1995-08-23 Production of zirconium alloy-based reactor core structural material excellent in corrosion resistance, particularly in uniform corrosion resistance and hydrogen absorption resistance

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JPH0961570A true JPH0961570A (en) 1997-03-07

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Cited By (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JP2008045980A (en) * 2006-08-15 2008-02-28 Toshihisa Shirakawa Square-shaped nuclear fuel assembly for bwr type using pwr nuclear fuel rod

Cited By (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JP2008045980A (en) * 2006-08-15 2008-02-28 Toshihisa Shirakawa Square-shaped nuclear fuel assembly for bwr type using pwr nuclear fuel rod

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