JPH0881746A - Manufacture of zr-base alloy-made nuclear fuel cladding tube for pressurized water reactor - Google Patents

Manufacture of zr-base alloy-made nuclear fuel cladding tube for pressurized water reactor

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Publication number
JPH0881746A
JPH0881746A JP6247060A JP24706094A JPH0881746A JP H0881746 A JPH0881746 A JP H0881746A JP 6247060 A JP6247060 A JP 6247060A JP 24706094 A JP24706094 A JP 24706094A JP H0881746 A JPH0881746 A JP H0881746A
Authority
JP
Japan
Prior art keywords
heating
nuclear fuel
fuel cladding
corrosion
corrosion resistance
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.)
Pending
Application number
JP6247060A
Other languages
Japanese (ja)
Inventor
Hideaki Abe
秀明 阿部
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.)
Nippon Steel Corp
Original Assignee
Sumitomo Metal Industries Ltd
Priority date (The priority date 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 date listed.)
Filing date
Publication date
Application filed by Sumitomo Metal Industries Ltd filed Critical Sumitomo Metal Industries Ltd
Priority to JP6247060A priority Critical patent/JPH0881746A/en
Publication of JPH0881746A publication Critical patent/JPH0881746A/en
Pending legal-status Critical Current

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Classifications

    • 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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  • Heat Treatment Of Nonferrous Metals Or Alloys (AREA)

Abstract

PURPOSE: To enhance the uniform-corrosion resistance by executing a high temp. heating treatment for heating to a specific temp. in working and heat treatment process after hot-extrusion tube-making process. CONSTITUTION: The high temp. heating treatment at 750-850 deg.C is executed. Solid solution of elements of Sn, Fe, Cr, Ni, Nb, etc., in a Zr-base alloy is promoted with the high temp. heating treatment. In order to restrain the coarsening of the crystal particle, the high temp. heating treatment for serving to not combination with annealing but to independence may be executed. The heating time is desirable to 1sec-20min and the crystal particle diameter is desirable to be <=30μm. The solid solution effect by executing the high temp. heating only one time remains to the product to sufficiently improve the corrosion resistance. The heating method is desirable to use an induction heating or a laser beam heating, etc., which can heat the objective temp. in a short time.

Description

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

【０００１】[0001]

【産業上の利用分野】本発明は、加圧水型原子炉に使用
されるＺｒ基合金製核燃料被覆管、より具体的には、耐
一様腐食性に優れたジルカロイ−４（商品名）製の核燃
料被覆管の製造方法に関する。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a nuclear fuel cladding tube made of a Zr-based alloy used in a pressurized water reactor, and more specifically made of Zircaloy-4 (trade name) excellent in uniform corrosion resistance. The present invention relates to a method for manufacturing a nuclear fuel cladding tube.

【０００２】[0002]

【従来の技術】軽水型原子炉は加圧水型原子炉（ＰＷ
Ｒ）と沸騰水型原子炉（ＢＷＲ）に大別され、ＰＷＲに
はジルカロイ−４なるＺｒ基合金製の核燃料被覆管が、
またＢＷＲにはジルカロイ−２（商品名）なるＺｒ基合
金製の核燃料被覆管がそれぞれ使用される。2. Description of the Related Art A light water reactor is a pressurized water reactor (PW).
R) and a boiling water reactor (BWR), and the PWR is a Zr-based alloy nuclear fuel cladding tube called Zircaloy-4.
Further, a Zr-based alloy nuclear fuel cladding tube called Zircaloy-2 (trade name) is used for each BWR.

【０００３】これらの核燃料被覆管は、一般に次のよう
なプロセスで製造される。まず、真空溶解で作製したイ
ンゴットを熱間鍛造により丸棒状のビレットとなす。次
いで、この丸ビレットをβ焼入後に熱間押出加工して素
管となす。そして、素管に対して焼鈍を行った後、必要
な寸法が得られるまで冷間圧延−焼鈍を繰り返し、最後
の焼鈍の後に精整を行う。These nuclear fuel cladding tubes are generally manufactured by the following process. First, the ingot produced by vacuum melting is formed into a round bar-shaped billet by hot forging. Next, this round billet is β-quenched and then hot extruded to form a raw tube. Then, after annealing the blank tube, cold rolling-annealing is repeated until the required dimensions are obtained, and after the final annealing, refining is performed.

【０００４】ところで、いずれの核燃料被覆管にあって
も、その長寿命化は近年の大きな課題であり、そのため
にそれぞれの腐食対策が各方面で研究されている。By the way, in any of the nuclear fuel cladding tubes, the longevity of the nuclear fuel cladding tubes has been a big problem in recent years, and therefore, various corrosion countermeasures have been studied in various fields.

【０００５】ＰＷＲ用ジルカロイ−４核燃料被覆管で問
題となる腐食は、「一様腐食」である。一方、ＢＷＲ用
ジルカロイ−２核燃料被覆管で問題となる腐食は、「ノ
ジュラー腐食」である。腐食形態が異なるのは、合金成
分の違いよりも炉内使用環境が異なるためと言われてい
る。そのため、ＰＷＲ用ジルカロイ−４核燃料被覆管と
ＢＷＲ用ジルカロイ−２核燃料被覆管とでは、異なる腐
食対策が採用される。The problematic corrosion in Zircaloy-4 nuclear fuel cladding for PWRs is "uniform corrosion". On the other hand, the problematic corrosion in the Zircaloy-2 nuclear fuel cladding for BWRs is "nodular corrosion". It is said that the different corrosion forms are due to the different operating environments in the furnace rather than the different alloy components. Therefore, different corrosion countermeasures are adopted for the PWR Zircaloy-4 nuclear fuel cladding tube and the BWR Zircaloy-2 nuclear fuel cladding tube.

【０００６】それぞれの腐食対策は、例えばZIRCONIUM
IN THENUCLEAR INDUSTRY P202-212「Microstructure an
d Corrosion Studies for Optimized PWR and BWR Zirc
aloy Cladding」（文献１）、JAERI-M 92-162「軽水炉
燃料被覆の腐食研究の現状と今後の方向」（文献２）に
詳しく説明されている。[0006] Each corrosion countermeasure is, for example, ZIRCONIUM
IN THENUCLEAR INDUSTRY P202-212 `` Microstructure an
d Corrosion Studies for Optimized PWR and BWR Zirc
“Aloy Cladding” (reference 1) and JAERI-M 92-162 “Current status and future direction of corrosion research on LWR fuel cladding” (reference 2).

【０００７】これらによると、ＰＷＲ用ジルカロイ−４
核燃料被覆管およＢＷＲ用ジルカロイ−２核燃料被覆管
の腐食に有効な対策として熱処理のあることが分かる。
熱処理による一般的な腐食対策は、総入熱パラメータΣ
Ａｉによるβ焼入処理後の入熱管理である。 ΣＡｉ＝ｔｉ・ｅｘｐ（−40000 ／Ｔｉ） ｔｉ：焼鈍温度保持時間（ｈ） Ｔｉ：焼鈍温度（Ｋ）According to these, Zircaloy-4 for PWR
It can be seen that heat treatment is an effective measure for corrosion of nuclear fuel cladding and Zircaloy-2 nuclear fuel cladding for BWRs.
The general heat input parameter is the total heat input parameter Σ
This is heat input control after β quenching processing by Ai. ΣAi = ti · exp (−40,000 / Ti) ti: Annealing temperature holding time (h) Ti: Annealing temperature (K)

【０００８】これを示した代表的な文献が文献１であ
り、これによると、ＰＷＲ用ジルカロイ−４核燃料被覆
管で問題となる一様腐食に対しては、β焼入より後のΣ
Ａｉを２×１０-18 〜５×１０-17 にする必要があり、
ＢＷＲ用ジルカロイ−２核燃料被覆管で問題となるノジ
ュラー腐食に対しては、このΣＡｉを１０-18 以下にす
る必要があるとされている。[0008] A representative document showing this is Document 1, which shows that for uniform corrosion which is a problem in Zircaloy-4 nuclear fuel cladding for PWR, Σ after β quenching
Ai needs to be 2 × 10 −18 to 5 × 10 −17 ,
It is said that the ΣAi needs to be 10 −18 or less against nodular corrosion which is a problem in the Zircaloy-2 nuclear fuel cladding tube for BWRs.

【０００９】また、ＢＷＲ用ジルカロイ−２核燃料被覆
管のノジュラー腐食対策については、製造工程中のいず
れかの段階で１３００°Ｆ（７０４℃）以上の高温域か
ら外面側のみを急冷する方法が、特開昭５８−２０７３
４９号公報（文献３）に示されている。As a measure against nodular corrosion of Zircaloy-2 nuclear fuel cladding for BWRs, a method of quenching only the outer surface side from a high temperature region of 1300 ° F (704 ° C) or higher at any stage in the manufacturing process is JP 58-2073
No. 49 (reference 3).

【００１０】[0010]

【発明が解決しようとする課題】ＰＷＲ用ジルカロイ−
４核燃料被覆管の一様腐食対策として、文献１に示され
たような入熱管理は非常に有効であり、現在、世界中の
この種の管はΣＡｉが２×１０-18 〜５×１０-17 の範
囲内で熱処理を受けたものになっている。しかし、入熱
管理だけでは最近のこの種の管に要求される高度の耐食
性能を満足させることが難しい。そのため文献２に示さ
れるように、主に合金成分の見直しに注力しているのが
現状である。Zircaloy for PWRs
As a countermeasure against uniform corrosion of 4 nuclear fuel cladding tubes, heat input management as shown in Reference 1 is very effective. Currently, this type of tubes in the world has ΣAi of 2 × 10 −18 to 5 × 10. It has been subjected to heat treatment within the range of -17 . However, it is difficult to satisfy the high level of corrosion resistance required for these types of tubes in recent years only by controlling heat input. Therefore, as shown in Reference 2, the current situation is that the focus is mainly on reviewing alloy components.

【００１１】文献３に示された高温域からの外面急冷処
理は、ＢＷＲ用ジルカロイ−２核燃料被覆管で問題とな
るノジュラー腐食に対しては有効であるが、ＰＷＲ用ジ
ルカロイ−４核燃料被覆管で問題となる一様腐食に対し
ては、必ずしも有効ではない。The outer surface quenching treatment from the high temperature range shown in Reference 3 is effective against nodular corrosion which is a problem in Zircaloy-2 nuclear fuel cladding tube for BWRs, but it is effective in Zircaloy-4 nuclear fuel cladding tube for PWRs. It is not always effective against the problem of uniform corrosion.

【００１２】なぜなら、文献３に示された実施例では、
１６５０°Ｆ（８９９℃）からの焼入処理が実施されて
いるが、このようなβ変態を既に生じている温度域から
の焼入処理では、ＰＷＲ用ジルカロイ−４核燃料被覆管
の耐一様腐食性に悪影響を及ぼすとされる析出物の微細
化が進むからである。また、その実施例では、途中熱処
理として１１５０°Ｆ（６２１℃）×２Ｈｒを３回行っ
ているが、このときの総入熱パラメータＡｉは2.２×１
０-19 と１０-18 を超え、この点からもＰＷＲ用ジルカ
ロイ−４核燃料被覆管の一様腐食対策としては不適当で
ある。Because, in the embodiment shown in Document 3,
Quenching treatment from 1650 ° F (899 ° C) has been carried out. However, in the quenching treatment from the temperature range in which such β transformation has already occurred, the uniform resistance of Zircaloy-4 nuclear fuel cladding tube for PWR This is because the refinement of precipitates, which are said to adversely affect the corrosivity, progresses. Further, in the embodiment, 1150 ° F. (621 ° C.) × 2 Hr is performed as the intermediate heat treatment three times, but the total heat input parameter Ai at this time is 2.2 × 1.
It exceeds 0 -19 and 10 -18, which is also unsuitable as a measure against uniform corrosion of Zircaloy-4 nuclear fuel cladding for PWR.

【００１３】更に言えば、文献３に示された実施例で
は、９３２°Ｆ（５００℃）の２４時間スチーム腐食試
験でその耐食性が評価されているが、ＰＷＲ用ジルカロ
イ−４核燃料被覆管の腐食については、試験温度が４１
５℃以上では試験結果が炉内の腐食結果と矛盾する（In
ternational Symposium on Environmental Degradation
of Materials in Nuclear Power Systems PAGE. 10.15-
10.24 1990) 。そのため、文献３に示された腐食試験で
は、ＰＷＲ用ジルカロイ−４核燃料被覆管の耐食性を評
価することができない。Furthermore, in the example shown in Document 3, its corrosion resistance was evaluated by a 24-hour steam corrosion test at 932 ° F. (500 ° C.), but the corrosion of Zircaloy-4 nuclear fuel clad pipe for PWR was corroded. For, the test temperature is 41
At 5 ℃ or higher, the test results conflict with the results of corrosion in the furnace (In
ternational Symposium on Environmental Degradation
of Materials in Nuclear Power Systems PAGE. 10.15-
10.24 1990). Therefore, the corrosion test shown in Document 3 cannot evaluate the corrosion resistance of the PWR Zircaloy-4 nuclear fuel cladding tube.

【００１４】以上のようにＰＷＲ用ジルカロイ−４核燃
料被覆管の腐食対策、特に熱処理による腐食対策として
は、総入熱パラメータΣＡｉによる入熱管理を超えるも
のが存在しないのが現状であり、そのために最初に記し
た通り、最近の厳しい要求に対してはもっぱら合金成分
の見直しの点から対策が講じられているわけである。As described above, as a countermeasure against the corrosion of the PWR Zircaloy-4 nuclear fuel clad tube, in particular, as a countermeasure against the corrosion by the heat treatment, there is no one exceeding the heat input control by the total heat input parameter ΣAi. As mentioned at the beginning, measures have been taken to meet the recent strict requirements, mainly from the viewpoint of reviewing alloy components.

【００１５】本発明はかかる事情に鑑みなされたもので
あって、ＰＷＲ用ジルカロイ−４核燃料被覆管の耐食性
を熱処理により高める核燃料被覆管製造方法を提供する
ことを目的とする。The present invention has been made in view of the above circumstances, and an object of the present invention is to provide a method for producing a nuclear fuel clad tube for enhancing the corrosion resistance of the PWR Zircaloy-4 nuclear fuel clad tube by heat treatment.

【００１６】[0016]

【課題を解決するための手段】ＰＷＲ用ジルカロイ−４
核燃料被覆管の耐一様腐食性を高めるために、β焼入後
の総入熱パラメータΣＡｉを２×１０-18 ×５〜１０
-17 にするのが有効なことは事実であり、従来は複数回
繰り返される熱処理および各熱処理での温度・時間が均
等に耐食性に影響するとの観点から、各熱処理での入熱
および条件が設定されてきた。[Means for Solving the Problems] Zircaloy-4 for PWR
In order to improve the uniform corrosion resistance of the nuclear fuel cladding tube, the total heat input parameter ΣAi after β quenching is set to 2 × 10 -18 × 5 to 10
It is true that setting -17 is effective, and in the past, heat input and conditions for each heat treatment were set from the viewpoint that heat treatment repeated multiple times and temperature / time at each heat treatment evenly affect corrosion resistance. It has been.

【００１７】しかし、本発明者の調査によれば、複数回
の熱処理および処理条件が全て均等に耐食性に影響する
のではなく、高温処理での加熱温度が特に強く耐食性に
影響することが判明した。However, according to the research conducted by the present inventor, it was found that the heat treatment and the treatment conditions of a plurality of times do not uniformly affect the corrosion resistance, but that the heating temperature in the high temperature treatment particularly strongly affects the corrosion resistance. .

【００１８】本発明はかかる知見に基づくものであっ
て、加圧水型原子炉に使用されるＺｒ基合金製核燃料被
覆管の製造方法であって、熱間押出製管工程より後の加
工・熱処理工程において、７５０〜８５０℃の温度域に
加熱する少なくとも１回の高温加熱処理を行うことを特
徴とする加圧水型原子炉用Ｚｒ基合金製核燃料被覆管の
製造方法を要旨とする。The present invention is based on such knowledge, and is a method for producing a Zr-based alloy nuclear fuel clad tube used in a pressurized water reactor, which comprises a working / heat treatment step after the hot extrusion tube making step. The method of manufacturing a nuclear fuel clad tube made of a Zr-based alloy for a pressurized water reactor is characterized in that at least one high-temperature heat treatment for heating in a temperature range of 750 to 850 ° C. is performed.

【００１９】高温加熱処理としては、高温加熱処理が、
加工後の焼鈍から分離独立した１秒〜２０分の短時間加
熱処理が望ましい。As the high temperature heat treatment, a high temperature heat treatment is
Short-time heat treatment for 1 second to 20 minutes, which is independent of annealing after working, is desirable.

【００２０】[0020]

【作用】一般に、ＰＷＲ用ジルカロイ−４核燃料被覆管
の製造において、β焼入以降の工程で処理温度が高いの
は、３〜５回繰り返される冷間圧延後の焼鈍工程の１回
目か２回目であるが、高温とは言ってもその温度は７０
０〜７４０℃程度である。これは、焼鈍温度が７００℃
より低くなると先の総入熱パラメータΣＡｉが小さくな
って、耐一様腐食性が低下し、逆に７４０℃を超える場
合は、通常行われる真空焼鈍だと加熱時間が数時間と長
いため、焼鈍中に結晶粒の粗大化を招き、引き続いて行
われる冷間圧延時に管に疵が発生しやすくなるからであ
る。In general, in the production of Zircaloy-4 nuclear fuel clad tube for PWR, the treatment temperature is high in the process after β quenching, which is the first or second annealing process after cold rolling repeated 3 to 5 times. However, the temperature is 70
It is about 0 to 740 ° C. This is an annealing temperature of 700 ℃
When it becomes lower, the total heat input parameter ΣAi becomes smaller and uniform corrosion resistance deteriorates. On the contrary, when it exceeds 740 ° C, the vacuum annealing that is usually performed requires a long heating time of several hours. This is because the crystal grains are coarsened therein, and the pipe is likely to be flawed during the subsequent cold rolling.

【００２１】しかし、本発明者の調査によれば、耐一様
腐食性向上のためには、７５０℃以上の高温加熱が有効
である。その理由はジルカロイ−４に耐食性改善のため
に通常添加されるＳｎ，Ｆｅ，Ｃｒ，Ｎｉ，Ｎｂ等の元
素の固溶化が、この高温加熱より促進されるからであ
る。そして好都合なことに、この耐食性改善のための高
温処理は、数秒〜数１０分の短時間処理でよく、そのた
め、焼鈍処理を兼ねずに独立して行うことにより、結晶
粒の粗大化を回避でき、また総入熱パラメータの増大に
よる耐食性低下を回避できる。However, according to the investigation by the present inventor, heating at a high temperature of 750 ° C. or higher is effective for improving the uniform corrosion resistance. The reason is that the solid solution of elements such as Sn, Fe, Cr, Ni and Nb, which are usually added to Zircaloy-4 for improving the corrosion resistance, is promoted by the high temperature heating. And, advantageously, the high temperature treatment for improving the corrosion resistance may be a short time treatment of several seconds to several tens of minutes, and therefore, by independently performing without annealing treatment, coarsening of crystal grains is avoided. It is also possible to avoid deterioration of corrosion resistance due to an increase in total heat input parameter.

【００２２】加熱温度の上限については、ＢＷＲ用ジル
カロイ−２核燃料被覆管では文献３に示すように高α領
域からβ領域での熱処理が耐ノジュラー腐食性の改善に
有効であるが、ＰＷＲ用ジルカロイ−４核燃料被覆管で
の一様腐食に対しては、むしろβ領域からの急冷処理が
悪影響を及ぼすので、加熱温度の上限をα＋β変態点で
ある８５０℃以下にしてβ変態発生領域からの急冷処理
をまぬがれる必要がある。As for the upper limit of the heating temperature, in the case of Zircaloy-2 nuclear fuel cladding tube for BWR, heat treatment in the high α region to β region is effective for improving nodular corrosion resistance as shown in Reference 3, but it is effective for ZWR for PWR. -4 Since the rapid cooling process from the β region rather adversely affects the uniform corrosion in the nuclear fuel cladding tube, the upper limit of the heating temperature is set to 850 ° C or lower, which is the α + β transformation point, and the rapid cooling from the β transformation generation region is performed. It is necessary to skip processing.

【００２３】特に望ましい加熱温度は８００〜８５０℃
である。A particularly desirable heating temperature is 800 to 850 ° C.
Is.

【００２４】実際、文献３に示された実施例と同様のプ
ロセスで製造されたジルカロイ−２管について、その実
施例で採用された腐食試験（９３２°Ｆ×２４Ｈｒスチ
ーム中）を行ったところ良好な耐食性が得られたが、Ｐ
ＷＲ用ジルカロイ−４核燃料被覆管で問題となる耐一様
腐食性を評価する４００℃×２４Ｈｒスチーム試験では
むしろ腐食増量が増加し、耐食性の低下が認められた。In fact, a Zircaloy-2 tube manufactured by the same process as the example shown in Document 3 was subjected to the corrosion test (in 932 ° F. × 24 Hr steam) adopted in the example, and was good. Excellent corrosion resistance was obtained, but P
In the 400 ° C. × 24 Hr steam test for evaluating uniform corrosion resistance, which is a problem in Zircaloy-4 nuclear fuel cladding tube for WR, the amount of corrosion increase was rather increased and the corrosion resistance was decreased.

【００２５】７５０〜８５０℃の高温加熱は、結晶粒の
粗大化を抑制するために、焼鈍を兼ねずに独立して行う
のが良く、加熱時間は１秒〜２０分が望ましい。１秒未
満では固溶化が不足し、耐食性改善効果が小さい。２０
分を超えると結晶粒の粗大化による冷間加工性の悪化が
問題になり、場合によっては総入熱パラメータΣＡｉの
増大による耐食性劣化も問題になる。結晶粒径は３０μ
ｍ以下が適当である。特に望ましい加熱時間は、下限に
ついては３秒以上であり、上限については３分以下であ
る。It is preferable that the high temperature heating at 750 to 850 ° C. is independently performed without also performing annealing in order to suppress the coarsening of crystal grains, and the heating time is preferably 1 second to 20 minutes. If it is less than 1 second, solid solution is insufficient and the effect of improving corrosion resistance is small. 20
If the amount exceeds the limit, deterioration of cold workability due to coarsening of crystal grains becomes a problem, and deterioration of corrosion resistance due to increase in total heat input parameter ΣAi becomes a problem in some cases. Crystal grain size is 30μ
m or less is suitable. Particularly desirable heating time is 3 seconds or more as the lower limit and 3 minutes or less as the upper limit.

【００２６】高温加熱の実施回数については、１回実施
すれば固溶化効果は製品まで残り耐食性向上には十分で
ある。実施時期については、製造コストのかかる特殊作
業であるので、できうる限り上工程で行うことが望まし
い。With respect to the number of times of high-temperature heating, if it is performed once, the solution-solubilizing effect remains in the product and it is sufficient for improving the corrosion resistance. As for the implementation time, it is a special operation that requires manufacturing cost, so it is desirable to perform it in the upper process as much as possible.

【００２７】加熱方法としては、短時間で目標温度まで
加熱できる誘導加熱、レーザー加熱等が望ましい。As a heating method, it is desirable to use induction heating, laser heating, or the like, which can heat to a target temperature in a short time.

【００２８】ＰＷＲ用ジルカロイ−４核燃料被覆管は、
外面のみの耐食性が優れていればよいので、外面加熱処
理でも十分である。Zircaloy-4 nuclear fuel cladding for PWR is
Since it suffices that the corrosion resistance of only the outer surface is excellent, the outer surface heat treatment is also sufficient.

【００２９】対象管は、ＰＷＲ用核燃料被覆管に用いら
れるジルカロイ−４またはその相当品である。The target tube is Zircaloy-4 used for the nuclear fuel cladding tube for PWR or its equivalent.

【００３０】[0030]

【実施例】以下に本発明の実施例を示し、比較例と対比
することにより、本発明の効果を明らかにする。EXAMPLES Examples of the present invention will be shown below, and the effects of the present invention will be clarified by comparison with Comparative Examples.

【００３１】真空溶解、熱間鍛造を経て製造したジルカ
ロイ−４からなる外径１９０ｍｍの中実ビレットをβ焼
入した。β焼入条件は、１０５０℃×３０分加熱−水冷
（１０℃／Ｓ）とした。β焼入後のビレットを孔繰りし
た後、熱間押出加工により外径８６ｍｍ、肉厚１４ｍｍ
の素管とした。素管に対して６４０℃×１Ｈｒの焼鈍を
行った後、表１に示すスケジユールで５回の冷間圧延お
よび焼鈍を行って、外径9.５ｍｍ、肉厚0.６ｍｍの最終
製品を得た。また、冷間圧延工程途中、すなわち、１回
目の圧延を終えた外径６3.５ｍｍ、肉厚１0.９ｍｍの管
（焼鈍後）に高温処理を誘導加熱法によって施した。A solid billet made of Zircaloy-4 manufactured by vacuum melting and hot forging and having an outer diameter of 190 mm was β-quenched. The β-quenching condition was 1050 ° C. × 30 minutes heating-water cooling (10 ° C./S). Beta-quenched billet is bored and then hot extruded to 86 mm outer diameter and 14 mm wall thickness
It was used as a tube. After annealing the blank tube at 640 ° C for 1 hour, cold rolling and annealing were performed 5 times with the schedule shown in Table 1 to obtain a final product with an outer diameter of 9.5 mm and a wall thickness of 0.6 mm. It was In the middle of the cold rolling process, that is, a tube having an outer diameter of 63.5 mm and a wall thickness of 10.9 mm (after annealing) after the first rolling was subjected to a high temperature treatment by an induction heating method.

【００３２】冷間圧延工程途中で高温処理を受けた最終
製品に対し、４００℃×１０５気圧の水蒸気中で２７日
間の腐食試験を行い、高温処理条件と腐食増量との関係
を調査した。なお、腐食増量は、高温処理を行わなかっ
たものの腐食増量を１としたときの比率で表わした。腐
食増量と合わせて高温処理後の結晶粒径、各圧延後の疵
発生の有無を調査すると共に、β焼入後の総入熱パラメ
ータΣＡｉを計算した結果を表２に示す。The final product subjected to the high temperature treatment during the cold rolling process was subjected to a corrosion test for 27 days in steam at 400 ° C. × 105 atm to investigate the relationship between the high temperature treatment condition and the corrosion amount increase. Note that the corrosion weight increase was expressed as a ratio when the corrosion weight increase was 1 without the high temperature treatment. Table 2 shows the results of the calculation of the total heat input parameter ΣAi after β-quenching, together with the increase in corrosion amount, the crystal grain size after high-temperature treatment, the presence or absence of defects after each rolling.

【００３３】熱間圧延にて製造された素管に７５０〜８
５０℃の高温加熱を実施することにより、成分組成の変
更なしに耐一様腐食性が改善される。これには、高温加
熱が焼鈍から独立した短時間加熱のために、総入熱パラ
メータΣＡｉを大きく増加させないことが寄与してい
る。加熱時間が長くなると、結晶粒径が増大し、高温加
熱が１回の場合は、総入熱パラメータΣＡｉの増大によ
る耐食性低下が問題となる前に、疵が発生する。短時間
の高温加熱を複数回実施した場合は、結晶粒の粗大化は
抑制されるが、総入熱パラメータΣＡｉの増大による耐
食性低下が問題になる。加熱温度が８５０℃を超える
と、β領域からの急冷となり、耐食性が悪化する。750 to 8 is applied to the raw pipe manufactured by hot rolling.
By carrying out high temperature heating at 50 ° C., uniform corrosion resistance is improved without changing the composition of components. This contributes to the fact that the total heat input parameter ΣAi is not significantly increased because the high temperature heating is a short time heating independent of the annealing. When the heating time becomes longer, the crystal grain size increases, and when the high temperature heating is performed once, a flaw occurs before the deterioration of the corrosion resistance due to the increase of the total heat input parameter ΣAi becomes a problem. When high-temperature heating for a short time is performed a plurality of times, coarsening of crystal grains is suppressed, but deterioration of corrosion resistance due to increase in total heat input parameter ΣAi becomes a problem. When the heating temperature exceeds 850 ° C., the β region is rapidly cooled and the corrosion resistance deteriorates.

【００３４】[0034]

【表１】 [Table 1]

【００３５】[0035]

【表２】 ※１回目の焼鈍の後の結晶粒径 ※※高温処理後の２回目加工で疵発生[Table 2] * Crystal grain size after the first annealing * * Scratches occur in the second processing after high temperature treatment

【００３６】[0036]

【発明の効果】以上に説明した通り、本発明の加圧水型
原子炉用Ｚｒ基合金製核燃料被覆管の製造方法は、総入
熱パラメータを適正範囲内に維持した高温加熱処理によ
り、性能に大きな影響を及ぼす合金成分の変更なしに、
耐一様腐食性を高めることができる。従って、その被覆
管の高品質化に大きな効果を発揮する。As described above, the method for producing a Zr-based alloy nuclear fuel clad tube for a pressurized water reactor according to the present invention has a large performance due to the high temperature heat treatment in which the total heat input parameter is maintained within an appropriate range. Without changing the alloying composition
Uniform corrosion resistance can be improved. Therefore, it exerts a great effect in improving the quality of the cladding tube.

Claims (2)

【特許請求の範囲】[Claims] 【請求項１】 加圧水型原子炉に使用されるＺｒ基合金
製核燃料被覆管の製造方法であって、熱間押出製管工程
より後の加工・熱処理工程において、７５０〜８５０℃
の温度域に加熱する少なくとも１回の高温加熱処理を行
うことを特徴とする加圧水型原子炉用Ｚｒ基合金製核燃
料被覆管の製造方法。1. A method of manufacturing a nuclear fuel cladding tube made of a Zr-based alloy used in a pressurized water nuclear reactor, comprising 750 to 850 ° C. in a working / heat treating step after a hot extrusion tube making step.
The method for producing a nuclear fuel clad tube made of a Zr-based alloy for a pressurized water reactor, comprising performing at least one high-temperature heat treatment for heating to a temperature range of 1. 【請求項２】 高温加熱処理が、加工後の焼鈍から分離
独立した１秒〜２０分の短時間加熱処理であることを特
徴とする請求項１に記載の加圧水型原子炉用Ｚｒ基合金
製核燃料被覆管の製造方法。2. The Zr-based alloy for a pressurized water reactor according to claim 1, wherein the high-temperature heat treatment is a short-time heat treatment for 1 second to 20 minutes, which is independent of annealing after working. Manufacturing method of nuclear fuel cladding.
JP6247060A 1994-09-13 1994-09-13 Manufacture of zr-base alloy-made nuclear fuel cladding tube for pressurized water reactor Pending JPH0881746A (en)

Priority Applications (1)

Application Number Priority Date Filing Date Title
JP6247060A JPH0881746A (en) 1994-09-13 1994-09-13 Manufacture of zr-base alloy-made nuclear fuel cladding tube for pressurized water reactor

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
JP6247060A JPH0881746A (en) 1994-09-13 1994-09-13 Manufacture of zr-base alloy-made nuclear fuel cladding tube for pressurized water reactor

Publications (1)

Publication Number Publication Date
JPH0881746A true JPH0881746A (en) 1996-03-26

Family

ID=17157829

Family Applications (1)

Application Number Title Priority Date Filing Date
JP6247060A Pending JPH0881746A (en) 1994-09-13 1994-09-13 Manufacture of zr-base alloy-made nuclear fuel cladding tube for pressurized water reactor

Country Status (1)

Country Link
JP (1) JPH0881746A (en)

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