JPH08961B2 - Method for manufacturing nuclear fuel cladding tube made of Zr-based alloy for pressurized water reactor - Google Patents
Method for manufacturing nuclear fuel cladding tube made of Zr-based alloy for pressurized water reactorInfo
- Publication number
- JPH08961B2 JPH08961B2 JP60001177A JP117785A JPH08961B2 JP H08961 B2 JPH08961 B2 JP H08961B2 JP 60001177 A JP60001177 A JP 60001177A JP 117785 A JP117785 A JP 117785A JP H08961 B2 JPH08961 B2 JP H08961B2
- Authority
- JP
- Japan
- Prior art keywords
- nuclear fuel
- based alloy
- fuel cladding
- cladding tube
- value
- 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.)
- Expired - Lifetime
Links
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
Landscapes
- Heat Treatment Of Nonferrous Metals Or Alloys (AREA)
- Pressure Welding/Diffusion-Bonding (AREA)
- Heat Treatment Of Steel (AREA)
Description
【発明の詳細な説明】 〈産業上の利用分野〉 この発明は、Zr基合金製核燃料被覆管の製造方法に係
り、特に、材料の塑性異方性に関して極めて好ましい性
状を備えていて、加圧水型原子炉用として優れた性能を
発揮するZr基合金製核燃料被覆管を安定・確実に製造す
る方法に関するものである。TECHNICAL FIELD The present invention relates to a method for producing a Zr-based alloy nuclear fuel clad tube, and in particular, it has extremely favorable properties with respect to plastic anisotropy of a material, and is a pressurized water type. The present invention relates to a method for stably and reliably manufacturing a nuclear fuel cladding tube made of a Zr-based alloy that exhibits excellent performance for a nuclear reactor.
〈従来技術並びにその問題点〉 現在、商業炉として稼動している軽水型原子炉は、更
に沸騰水型と加圧水型とに大別されるが、このうちの加
圧水型原子炉では、その炉心に第１図で示されるような
燃料要素が設置されてエネルギー源として使用されてい
る。<Prior art and its problems> Light water reactors currently operating as commercial reactors are further classified into boiling water reactors and pressurized water reactors. A fuel element as shown in FIG. 1 is installed and used as an energy source.
上述のように、第１図は加圧水型原子炉に使用される
燃料要素の概略構成図であり、この燃料要素は、両端を
端栓1,2で密封される核燃料被覆管３内に二酸化ウラン
（UO2）ペレツト４が装填され、かつプレナムスプリン
グ５により二酸化ウランペレツト４を押圧固定して成る
もので、符号６で示されるものは、核燃料被覆管内上部
に設けられた核***生成ガス蓄積用のプレナムである。As described above, FIG. 1 is a schematic configuration diagram of a fuel element used in a pressurized water reactor. This fuel element has uranium dioxide in a nuclear fuel cladding tube 3 whose both ends are sealed with end plugs 1 and 2. A (UO 2 ) pellet 4 is loaded and the uranium dioxide pellet 4 is pressed and fixed by a plenum spring 5. The reference numeral 6 is a plenum for storing fission product gas provided in an upper portion of a nuclear fuel cladding tube. Is.
そして、当然のことながら、核燃料被覆管が破損して
放射性物質の漏洩を来たすことは原子炉の安全上あつて
はならないこととされているので、該核燃料被覆管に
は、熱中性子吸収が少なく、しかも優れた耐食性や高温
強度を有しているZr基合合（特に、「ジルカロイ−
４」）が普通に使用されており、その製造に当つては第
２図に示されるような一連の加工工程が採用されてい
る。And, as a matter of course, it is considered that the nuclear fuel clad tube is damaged and causes leakage of radioactive material, so it is not appropriate for the safety of the nuclear reactor, so the nuclear fuel clad tube has a low thermal neutron absorption. In addition, Zr base compound (especially "Zircaloy-
4 ") is commonly used, and in its manufacture, a series of processing steps as shown in FIG. 2 is adopted.
ところで、加圧水型原子炉の核燃料被覆管に想定され
る破損の原因としては、UO2ペレツトからの大きな歪
と、高速中性子の照射による延性の低下があげられる。
つまり、UO2ペレツトの外周部と中心部との熱膨張差に
よりこの方向の割れが該ペレツトに生じると、この割れ
に面した被覆管の部分に歪が集中することとなるが、こ
れが高速中性子によつて低下せしめられた被覆管材料の
延性限界を越えてしまうほどに高いと核燃料被覆管の割
れを引き起こしてしまうのである。By the way, the possible causes of damage to the nuclear fuel cladding of a pressurized water reactor are a large strain from the UO 2 pellet and a decrease in ductility due to the irradiation of fast neutrons.
That is, when cracks in this direction occur in the pellet due to the difference in thermal expansion between the outer peripheral portion and the central portion of the UO 2 pellet, strain is concentrated in the cladding portion facing this crack, but this is the fast neutron. If it is so high as to exceed the ductility limit of the cladding material which has been reduced by the above, it will cause cracking of the nuclear fuel cladding.
このようなことから、加圧水型原子炉用核燃料被覆管
材料では、上記「割れ」に対する十分な特性が殊更に重
要視されてきた。Therefore, in the nuclear fuel clad tube material for pressurized water reactors, sufficient characteristics against the above-mentioned "crack" have been particularly emphasized.
しかし、核燃料被覆管材料として使用されているZr基
合金はその98％程度以上がZr成分であり、従つて該Zrの
稠密六方晶の結晶構造に起因する強い異方性が被覆管の
性質にそのまま現われるので「割れ」等を防止すると言
う観点からは上記材料は非常に厄介な代物だつたのであ
る。However, about 98% or more of the Zr-based alloy used as a nuclear fuel cladding material is a Zr component, and therefore strong anisotropy due to the crystal structure of the dense hexagonal crystal of Zr affects the properties of the cladding. Since it appears as it is, the above material is a very troublesome substitute from the viewpoint of preventing "cracking".
そこで、最近、加圧水型原子炉用Zr基合金製被覆管の
製造に際し、異方性の指標として“塑性異方性係数（C.
S.R.:収縮歪比）の観念が導入され、この「C.S.R.」を
調整することによつて適正な製品を確保しようとの試み
がなされるようになつた。この「C.S.R.」とは、管軸方
向での４〜５％の常温引張り試験時における円周方向と
半径方向の歪の比、即ち、式 で表わされるものである。Therefore, recently, in the production of Zr-based alloy cladding for pressurized water reactors, the "plastic anisotropy coefficient (C.
The concept of SR: shrinkage strain ratio) was introduced, and attempts were made to secure appropriate products by adjusting this "CSR". This “CSR” is the ratio of the strain in the circumferential direction to the strain in the radial direction during a room temperature tensile test of 4 to 5% in the pipe axis direction, that is, the formula Is represented by.
この「C.S.R」とZr基合金（ジルカロイ−４）製核燃
料被覆管の諸性質については種々の研究がなされ、 『「C.S.R.」の大きい場合には肉厚減少による変形が
起り難く、このような管は径方向の変形を拘束しない通
常の引張り試験では“径収縮”或いは“長さ収縮”が起
つて大きい変形能を示すが、多軸応力下での変形能は小
さいこと、即ち、上記加圧水型原子炉内でのUO2ペレツ
トとその被覆管との関係を模擬した長さ方向拘束内圧バ
ースト試験のような多軸応力下の変形試験では変形能が
小さく表われて実用上問題である』 ことが明らかとなつている。Various studies have been conducted on various properties of this "CSR" and a Zr-based alloy (Zircaloy-4) nuclear fuel cladding tube. "When" CSR "is large, deformation due to wall thickness reduction is unlikely to occur, and Shows a large deformability due to "radial contraction" or "length contraction" in a normal tensile test that does not restrain radial deformation, but the deformability under multiaxial stress is small, that is, the above pressurized water type In a deformation test under multiaxial stress, such as a longitudinally constrained internal pressure burst test that simulates the relationship between a UO 2 pellet and its cladding in a nuclear reactor, the deformability appears to be small, which is a practical problem. " Has become clear.
また、このほかにも、「C.S.R.」と照射成長（応力の
かかつていない状態下で高速中性子照射によつて起きる
変形）、クリープ特性、又は応力腐食割れ等の材料特性
との関係についても研究がなされ、 『「C.S.R.」が大き過ぎる場合には、照射成長によつ
て被覆管が長手方向に伸び変形を起すこととなり、一
方、「C.S.R.」が小さ過ぎる場合にはクリープ特性や耐
応力腐食割れに悪影響が及ぶこととなる』 との知見も得られている。In addition, research on the relationship between “CSR” and irradiation growth (deformation caused by fast neutron irradiation under unstressed conditions), creep properties, or stress corrosion cracking and other material properties is also conducted. , "If" CSR "is too large, the cladding will stretch and deform in the longitudinal direction due to irradiation growth. On the other hand, if" CSR "is too small, the creep characteristics and stress corrosion cracking resistance will be adversely affected. It has been obtained. ”
そして、これらの知見事項の総合検討によつて １≦C.S.R.≦２ が「C.S.R.」の最適範囲であるとの結論が得られたこと
より、Zr基合金（ジルカロイ−４）製核燃料被覆管の製
造にあたつては、その異方性が上記適正範囲内に入るよ
うに制御する努力が払われてきた。From the comprehensive examination of these findings, it was concluded that 1 ≦ CSR ≦ 2 is the optimum range of “CSR”. Therefore, the production of Zr-based alloy (Zircaloy-4) nuclear fuel cladding For this reason, efforts have been made to control the anisotropy so that it falls within the appropriate range.
一方、これまでの研究では、この異方性はZr基合金製
核燃料被覆管製造工程中の仕上げ圧延条件に大きく影響
され、特に で表わされる加工パラメータ〔QE値〕によつて決定され
ると言われていた。On the other hand, in previous studies, this anisotropy was greatly affected by the finish rolling conditions during the manufacturing process of Zr-based alloy nuclear fuel cladding, It is said that it is determined by the processing parameter [Q E value] represented by.
従つて、これまで、Zr基合金製核燃料被覆管の異方性
の制御は〔QE値〕の調整を図ることのみに重点が置かれ
て実施されてきたのである。そして、それも〔QE値〕と
異方性の関係についての定量的な把握に基づくものでは
なく、単なる経験的な事実に基づいての〔QE値〕設定が
なされていたに過ぎないのが現状であつた。Therefore, until now, the control of the anisotropy of the Zr-based alloy nuclear fuel cladding tube has been carried out with emphasis on only adjusting the [Q E value]. And that is not based on a quantitative grasp of the relationship between [Q E value] and anisotropy, and the setting of [Q E value] was based only on empirical facts. Was the current situation.
ところが、様々な観点に立つてなされた加圧水型原子
炉の核燃料被覆管に関する本発明者等の研究結果は、仕
上げ圧延条件によつて異方性が決まることを確認はした
が、他方で、Zr基合金製核燃料被覆管の「C.S.R.」は、
仕上げ圧延加工時の「QE値」を制御するだけでは最適範
囲に調整することができず、仕上げ圧延の際の加工度を
も考慮する必要のあることを明らかにしたのである。現
に、Zr基合金製核燃料被覆管の実生産工程において、仕
上げ圧延の「QE値」を同程度に設定したとしても、その
加工度が変われば「C.S.R.」も変化すると言う思わぬ事
態が発生し、製造上大きな問題となることがあつた。However, the research results of the present inventors regarding the nuclear fuel cladding of the pressurized water reactor made from various viewpoints confirmed that the anisotropy is determined by the finish rolling conditions, but on the other hand, Zr "CSR" of nuclear fuel cladding tube made of base alloy is
It was clarified that it is not possible to adjust to the optimum range simply by controlling the "Q E value" during finish rolling, and it is also necessary to consider the workability during finish rolling. Actually, in the actual production process of Zr-based alloy nuclear fuel clad tubes, even if the "Q E value" of finish rolling was set to the same level, the unexpected change in "CSR" would occur if the workability changed. However, this may cause a big problem in manufacturing.
〈問題点を解決するための手段〉 そこで本発明者等は、上述のような研究結果を踏えた
上で、従来品に懸念される前記問題点が払拭された核燃
料被覆管を実現すべく、特に加工スケジユールと製品被
覆管の異方性との相関関係の解明を目指して更に研究を
重ねたところ、次の（ａ）〜（ｆ）に示す如き事項が確
認されたのである。<Means for Solving Problems> Therefore, the inventors of the present invention, on the basis of the above research results, to realize a nuclear fuel clad tube in which the above-mentioned problems of concern in conventional products are eliminated. In particular, further research was conducted with the aim of clarifying the correlation between the processing schedule and the anisotropy of the product coating tube, and the following matters (a) to (f) were confirmed.
なお、加工スケジユールと異方性との関連性調査に
は、まず第３図で示すような種々の加工スケジユール
（押出し素管から種々の外径及び肉厚の途中管を製造
し、次いで仕上げ冷間圧延を実施）でZr基合金（ジルカ
ロイ−４）管を製造し、これに応力除去焼鈍を施したも
のについて、各圧延条件が異方性にどのような影響を及
ぼしたかを調べると言う方法を主として採用した。そし
て、このときの冷間圧延加工のパラメータとしては、
〔QE値〕並びに式 で定義される加工度（Rd）を採用し、Zr基合金管の異方
性の指標としては「C.S.R.」を用いた。In order to investigate the relationship between the processing schedule and anisotropy, first, various processing schedules (manufacturing intermediate pipes of various outer diameters and wall thicknesses from extruded raw pipes as shown in FIG. Hot-rolling) to produce a Zr-based alloy (Zircaloy-4) tube, which was stress-relieved and annealed, and how the rolling conditions affect the anisotropy. Was mainly adopted. And as the parameters of the cold rolling at this time,
[Q E value] and formula The workability (Rd) defined by the above was adopted, and "CSR" was used as an index of the anisotropy of the Zr-based alloy tube.
確認事項 （ａ）以前からの報告通り、Zr基合金製核燃料被覆管製
品の異方性は、加工工程中の途中管の異方性には何ら左
右されることなく、仕上げ圧延の条件のみによつて影響
を受けるものであること。Items to be confirmed (a) As previously reported, the anisotropy of the Zr-based alloy nuclear fuel clad tube product is not affected by the anisotropy of the intermediate tube during the working process, only the finish rolling conditions. Being affected.
第４図は、２種の圧延スケジユールを持つ最終管につ
いて、圧延スケジユールと〔C.S.R.〕の値を比較したも
のであるが、第４図からも、途中工程が異なるために20
φの時点では異方性に差がある２種の母管であつても、
仕上げ圧延条件を同じにすると、仕上げ管の異方性には
差がなくなることが明らかである。Fig. 4 compares the values of [CSR] with the rolling schedule for the final pipe with two types of rolling schedules.
Even if there are two types of mother tubes that differ in anisotropy at the time of φ,
It is clear that if the finish rolling conditions are the same, there will be no difference in the anisotropy of the finished tubes.
（ｂ）「QE値」は、確かにZr基合金製核燃料被覆管の異
方性に大きな影響を及ぼす要素であり、その増加は「C.
S.R.」の値を増加することにつながるものであるが、両
者間には、特に第５図で示されるような関係が存在して
いること。(B) The "Q E value" is certainly a factor that has a great influence on the anisotropy of Zr-based alloy nuclear fuel cladding, and its increase is "C.
This will lead to an increase in the value of "SR", but there must be a relationship between the two, especially as shown in Fig. 5.
第５図は、仕上げ圧延加工度が70％程度のZr基合金
（ジルカロイ−４）製核燃料被覆管について、仕上げ圧
延時の「QE値」と仕上げ管の「C.R.S.」との関係を示す
グラフであるが、第５図からも、「QE値」が増加するに
つれて「C.R.S.」の値も増加することが明らかである。
そして、これらの関係より、先に述べたような １≦C.S.R.≦２ の条件を満たし得る「QE値」は、加工度の影響を加味し
たとしても〔1.5〜3.5〕に抑えておく必要のあることが
わかる。Fig. 5 is a graph showing the relationship between the "Q E value" during finish rolling and the "CRS" of the finish tube for a Zr-based alloy (Zircaloy-4) nuclear fuel cladding tube with a finish rolling degree of about 70%. However, also from FIG. 5, it is clear that the value of “CRS” increases as the “Q E value” increases.
From these relationships, the “Q E value” that can satisfy the condition of 1 ≦ CSR ≦ 2 as described above needs to be suppressed to [1.5 to 3.5] even if the influence of the workability is taken into consideration. I know there is.
（ｃ）Zr基合金（ジルカロイ−４）製核燃料被覆管の異
方性には明瞭な加工度依存性があり、例えば仕上げ圧延
の加工度と「C.S.R.」との関係を示す第６図からも明ら
かなように、仕上げ加工度の増加は「C.S.R.」に大きく
影響してその値を大幅に減少する。従つて、Zr基合金製
核燃料被覆管の「C.S.R.」を制御する場合には、「Q
E値」だけの考慮では極めて不十分であり、「QE値」と
「仕上げ圧延加工度」の両方を共に調整する必要がある
こと。(C) The anisotropy of the Zr-based alloy (Zircaloy-4) nuclear fuel cladding tube has a clear dependence on the workability. For example, from FIG. 6 showing the relationship between the workability of finish rolling and “CSR”. As is clear, an increase in the finishing degree has a great influence on "CSR", and the value is greatly reduced. Therefore, when controlling "CSR" of nuclear fuel cladding tube made of Zr-based alloy,
Extremely insufficient for E value "only consideration, it is necessary to both adjust both the" Q E value "" finish rolling working ratio. "
（ｄ）更に、Zr基合金（ジルカロイ−４）製核燃料被覆
管の「仕上げ加工度〔Rd〕（％）」，「QE値」」及び
「C.S.R.」との間には、 C.S.R.＝5.3−4.8×10-2・Rd＋0.16・lnQEなる関係式
が成立すること。(D) Furthermore, between the “finishing degree [Rd] (%)”, “Q E value” and “CSR” of the Zr-based alloy (Zircaloy-4) nuclear fuel cladding tube, CSR = 5.3− The relational expression of 4.8 × 10 -2 · Rd + 0.16 · lnQ E must hold.
従つて、この式が与えられれば、好適な異方性（１≦
C.S.R.≦２）を得るための「仕上げ加工度」と「QE値」
の適正値が定まることとなる。Therefore, given this equation, the preferred anisotropy (1 ≦
“Finishing degree” and “Q E value” to obtain CSR ≦ 2)
The appropriate value of is determined.
特に「C.S.R」の上限値に注目すれば、好適異方性を備
えた被覆管を製造するには、 と適正加工度が限定される。In particular, paying attention to the upper limit of "CSR", in order to manufacture a cladding tube with suitable anisotropy, And the proper processing degree is limited.
（ｅ）Zr基合金（ジルカロイ−４）製核燃料被覆管の信
頼性には、前述したような材料の異方性が大きな影響を
与えるばかりでなく、その機械的性質も重要な要素とな
つているが、必要とされる高温引き張り伸び率:15％程
度以上を確保するためには、仕上げ圧延の加工度を90％
以下に設定する必要があること。(E) Not only the anisotropy of the material as described above has a great influence on the reliability of the Zr-based alloy (Zircaloy-4) nuclear fuel cladding tube, but also its mechanical properties are important factors. However, in order to secure the required high temperature tensile elongation of 15% or more, the workability of finish rolling is 90%.
Must be set to:
第７図は、Zr基合金（ジルカロイ−４）製核燃料被覆
管の仕上げ加工度と400℃高温引張り伸びとの関係を示
したグラフであるが、第７図からも、仕上げ加工度を90
％以下に設定すれば15％以上の高温引張り伸び率を十分
に達成できることが明らかである。FIG. 7 is a graph showing the relationship between the finish workability of the Zr-based alloy (Zircaloy-4) nuclear fuel cladding tube and the 400 ° C. high temperature tensile elongation. From FIG. 7 as well, the finish workability is 90%.
It is clear that a high temperature tensile elongation of 15% or more can be sufficiently achieved by setting it to be less than or equal to%.
（ｅ）従つて、Zr基合金（ジルカロイ−４）製核燃料被
覆管の製造に当つて、その仕上げ圧延の加工スケジユー
ルを第８図におけるハツチング部内の条件に設定すれ
ば、実用上最適な異方性を備えた製品が得られること。(E) Therefore, in manufacturing a nuclear fuel cladding tube made of a Zr-based alloy (Zircaloy-4), if the processing schedule of finish rolling is set to the conditions in the hatched portion in FIG. To obtain a product with the characteristics.
（ｆ）また、燃料被覆管にあつては、材料の「異方性」
や「高温引張り伸び」のほか、常温における「引張り強
さ」や「耐力」、並びに「伸び」も信頼性の重要な要素
となつているが、仕上げ圧延後の最終焼鈍を、焼鈍温
度:430〜500℃の応力除去焼鈍とすれば、材料の異方性
に影響することなく、必要とされる13％以上の伸び、75
Kgf/mm2以上の引張り強さ、及び55Kgf/mm2以上の耐力が
確保されること。(F) Also, for fuel cladding, the "anisotropic" nature of the material
In addition to "and high temperature tensile elongation", "tensile strength", "proof stress" at room temperature, and "elongation" are also important factors for reliability, but the final annealing after finish rolling is performed at an annealing temperature of 430. With stress relief annealing at ~ 500 ℃, the required elongation of 13% or more, 75% without affecting the anisotropy of the material.
Kgf / mm 2 or more tensile strength, and 55 kgf / mm 2 or more that yield strength is ensured.
この発明は、上記知見に基づいてなされたもので、 「ジルカロイ−４」として知られるところの、 Sn:1.20〜1.70％（以下、成分割合を表す％は重量％と
する）、 Fe:0.18〜0.24％、 Cr:0.07〜0.13％ を含有するとともに、〔Fe（％）〜Cr（％）〕の値が0.
28〜0.37の範囲内で、かつ、 残部:Zr及び不可避不純物 から成る成分組成の加圧水型原子炉用Zr基合金製核燃料
被覆管の製造方法であって、仕上げ冷間圧延加工を 加工パラメータ〔QE値〕:1.5〜3.5, 加工度〔Rd〕： で実施するとともに、これに続く焼鈍を、焼鈍温度:430
〜500℃の応力除去焼鈍とすることにより、信頼性の極
めて高い核燃料被覆管、特に材料特性の異方性に関して
極めて好ましい加圧水型原子炉用Zr基合金製核燃料被覆
管を安定・確実に製造する点、に特徴を有するものであ
る。The present invention was made based on the above findings, and is known as "Zircaloy-4", Sn: 1.20 to 1.70% (hereinafter,% representing the component ratio shall be% by weight), Fe: 0.18 to 0.24%, Cr: 0.07 to 0.13%, and the value of [Fe (%) to Cr (%)] is 0.
A method for producing a Zr-based alloy nuclear fuel clad tube for a pressurized water reactor with a composition within the range of 28 to 0.37 and the balance: Zr and unavoidable impurities. E value]: 1.5 to 3.5, processing degree [Rd]: And the subsequent annealing at an annealing temperature of 430.
Stable and reliable production of highly reliable nuclear fuel clad tubes, especially Zr-based alloy nuclear fuel clad tubes for pressurized water reactors, which are extremely favorable in terms of anisotropy of material properties, by stress relief annealing at ~ 500 ° C It is characterized by points.
なお、この発明の核燃料被覆管の製造方法において仕
上げ冷間圧延加工の〔QE値〕を1.5〜3.5と限定したの
は、前述の第３図からも明らかであるが、〔QE値〕が1.
5未満であつても、また3.5を越えても〔C.S.R.〕が適正
な値（１〜２）とならず、いずれにしても被覆管の信頼
性が不十分となるからである。Incidentally, the finish cold rolling process the [Q E value] is limited with 1.5 to 3.5 in the method for manufacturing a nuclear fuel cladding tube of the present invention, it will be apparent from FIG. 3 described above, [Q E value] Is 1.
Even if it is less than 5, or exceeds 3.5, the [CSR] does not become an appropriate value (1 to 2), and in any case, the reliability of the cladding tube becomes insufficient.
そして、仕上げ冷間圧延加工における加工度〔Rd〕が 未満であつても〔C.S.R.〕の値が２を越えてしまつて適
正範囲外となり、一方、前記加工度が90％を越えると要
求される高温引張り伸び率:15％以上を確保できなくな
ることから、仕上げ冷間圧延加工の加工度〔Rd〕を と定めた。And the workability [Rd] in finish cold rolling is Even if it is less than the above, the value of [CSR] will exceed 2 and will be out of the proper range. On the other hand, if the working degree exceeds 90%, the required high temperature tensile elongation: 15% or more cannot be secured. , Finish cold rolling processability [Rd] I decided.
更に、本発明のZr基合金製核燃料被覆管の製造方法に
おいては、必要な強度を確保しつつ常温における「伸
び」を十分に回復させて被覆管の信頼性を一層向上させ
るため、430〜500℃の焼鈍温度で応力除去焼鈍を施すこ
とは極めて重要なことである。そして、この焼鈍温度が
430℃未満では必要とされる常温伸び:13％以上が達成で
きず、一方500℃を越える温度で焼鈍すると、要求され
る強張り強さ:75Kgf/mm2以上及び耐力:55Kgf/mm2を確保
できなくなることから、焼鈍温度を430〜500℃と定め
た。Further, in the method for producing a Zr-based alloy nuclear fuel cladding tube of the present invention, in order to further improve the reliability of the cladding tube by sufficiently recovering the "elongation" at room temperature while ensuring the required strength, 430-500 It is extremely important to perform stress relief annealing at an annealing temperature of ° C. And this annealing temperature
If it is less than 430 ℃, the required room temperature elongation: 13% or more cannot be achieved, while if it is annealed at a temperature over 500 ℃, the required tensile strength: 75Kgf / mm 2 or more and yield strength: 55Kgf / mm 2 are obtained. The annealing temperature was set at 430-500 ° C because it could not be ensured.
このことは、第９図で示したところの、Zr基合金（ジ
ルカロイ−４）製核燃料被覆管について調査した焼鈍軟
化曲線からも明らかである。This is also clear from the annealing softening curve investigated for the Zr-based alloy (Zircaloy-4) nuclear fuel cladding tube shown in FIG.
次に、この発明を実施例により具体的に説明する。 Next, the present invention will be specifically described with reference to examples.
〈実施例〉 まず、第１表に示される化学成分組成のZr基合金製熱
間押出し素管を準備した。<Example> First, a Zr-based alloy hot extruded raw tube having the chemical composition shown in Table 1 was prepared.
次いで、常法通り、これに冷間圧延及び歪取り焼鈍を
繰り返して、第２表に示される寸法の母管を製造した。 Then, as usual, this was repeatedly subjected to cold rolling and strain relief annealing to manufacture a mother tube having the dimensions shown in Table 2.
続いて、同じく第２表に示される条件の仕上げ冷間圧
延と最終焼鈍を実施して、Zr基合金製核燃料被覆管製品
を得た。Subsequently, finish cold rolling and final annealing were also performed under the conditions shown in Table 2 to obtain a Zr-based alloy nuclear fuel clad tube product.
これらの製品について、その異方性並びに軸方向にお
ける機械的性質を調査し、その結果を同じく第２表に示
した。The anisotropy and mechanical properties in the axial direction of these products were investigated, and the results are also shown in Table 2.
第２表に示される結果からも、仕上げ冷間圧延の条件
が第８図のハツチングで示した適正加工スケジユールに
適合し、かつ430〜500℃の焼鈍温度で応力除去焼鈍を施
すと言う本発明方法によれば、良好な「C.S.R.」や「伸
び」等を備えた信頼性の高いZr基合金製核燃料被覆管が
得られるのに対して、本発明の製造条件を満たしていな
い比較法によると、例えば「C.S.R.」の値が２を越えて
しまうなど、得られる製品の信頼性に難点のあることが
明白である。 The results shown in Table 2 also indicate that the conditions of finish cold rolling conform to the proper working schedule shown by hatching in FIG. 8 and that stress relief annealing is performed at an annealing temperature of 430 to 500 ° C. According to the method, while a highly reliable Zr-based alloy nuclear fuel clad tube having good “CSR” and “elongation” can be obtained, according to the comparison method which does not satisfy the manufacturing conditions of the present invention, It is obvious that there is a problem in the reliability of the obtained product, for example, the value of "CSR" exceeds 2.
〈総括的な効果〉 以上に説明した通り、この発明によれば、材料特性の
異方性や機械的性質が極めて優れたZr基合金製核燃料被
覆管を安定・確実に製造することができ、核燃料被覆管
の信頼性向上、ひいては原子炉の安全性を一層確実なも
のとなし得るなど、産業上極めて優れた効果がもたらさ
れるのである。<Overall Effect> As described above, according to the present invention, it is possible to stably and reliably produce a nuclear fuel cladding tube made of a Zr-based alloy having excellent anisotropy of material properties and mechanical properties. The industrially excellent effects are brought about, such as improving the reliability of the nuclear fuel cladding tube and further ensuring the safety of the nuclear reactor.
第１図は、加圧水型原子炉用燃料要素の概略構成図、 第２図は、Zr基合金製核燃料被覆管の製造工程図、 第３図は、Zr基合金製核燃料被覆管製造試験における加
工テストスケジユールを示す説明図、 第４図は、Zr基合金製核燃料被覆管製造試験における２
種の加工スケジユールと製品の「C.S.R.」を示す説明
図、 第５図は、Zr基合金製核燃料被覆管における「QE値」と
「C.S.R.」との関係を示すグラフ、 第６図は、Zr基合金製核燃料被覆管における「加工度
（Rd）」と「C.S.R.」との関係を示すグラフ、 第７図は、Zr基合金製核燃料被覆管における「加工度
（Rd）」と「400℃高温引張り伸び」との関係を示すグ
ラフ、 第８図は、Zr基合金製核燃料被覆管の適正加工スケジユ
ールを示すグラフ、 第９図は、仕上げ冷間圧延後のZr基合金製核燃料被覆管
の焼鈍軟化曲線である。 図面において、 1,2……端栓、３……核燃料被覆管、４……UO2ペレツ
ト、５……プレナムスプリング、６……プレナム。FIG. 1 is a schematic configuration diagram of a fuel element for a pressurized water nuclear reactor, FIG. 2 is a manufacturing process diagram of a Zr-based alloy nuclear fuel cladding, and FIG. 3 is a process in a Zr-based alloy nuclear fuel cladding production test. Fig. 4 is an explanatory diagram showing the test schedule, and Fig. 4 shows 2 in the manufacturing test of the nuclear fuel cladding tube made of Zr-based alloy.
Explanatory drawing showing "CSR" of various processing schedules and products, Fig. 5 is a graph showing the relationship between "Q E value" and "CSR" in nuclear fuel cladding tube made of Zr-based alloy, and Fig. 6 is Zr Graph showing the relationship between "working degree (Rd)" and "CSR" in base alloy nuclear fuel cladding, Fig. 7 shows "working degree (Rd)" and "400 ℃ high temperature" in Zr-based alloy nuclear fuel cladding Fig. 8 is a graph showing the relationship with "tensile elongation", Fig. 8 is a graph showing the proper processing schedule of the Zr-based alloy nuclear fuel cladding, and Fig. 9 is the annealing of the Zr-based alloy nuclear fuel cladding after finish cold rolling. It is a softening curve. In the drawing, 1,2 ... end plug, 3 ... nuclear fuel cladding tube, 4 ... UO 2 pellet, 5 ... plenum spring, 6 ... plenum.
Claims (1)
基合金製核燃料被覆管の製造方法であって、仕上げ冷間
延加工を 加工パラメータ〔QE値〕:1.5〜3.5, 加工度〔Rd〕： で実施するとともに、これに続く焼鈍を、焼鈍温度:430
〜500℃の応力除去焼鈍とする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。 記 重量割合で、 Sn:1.20〜1.70％,Fe:0.18〜0.24％,Cr:0.07〜0.13％ を含有するとともに、〔Fe（％）＋Cr（％）〕の値が0.
28〜0.37の範囲内で、かつ、 残部:Zr及び不可避不純物。1. A Zr for pressurized water reactor having the following composition:
A method for manufacturing a base alloy nuclear fuel cladding tube, in which finish cold-rolling is performed. Processing parameters [Q E value]: 1.5 to 3.5, Degree of processing [Rd]: And the subsequent annealing at an annealing temperature of 430.
A method for producing a Zr-based alloy nuclear fuel cladding tube for a pressurized water reactor, which is characterized in that the stress relief annealing is performed at 〜 500 ℃. In the weight ratio, Sn: 1.20 to 1.70%, Fe: 0.18 to 0.24%, Cr: 0.07 to 0.13%, and the value of [Fe (%) + Cr (%)] is 0.
Within the range of 28 to 0.37, and the balance: Zr and unavoidable impurities.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP60001177A JPH08961B2 (en) | 1985-01-08 | 1985-01-08 | Method for manufacturing nuclear fuel cladding tube made of Zr-based alloy for pressurized water reactor |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP60001177A JPH08961B2 (en) | 1985-01-08 | 1985-01-08 | Method for manufacturing nuclear fuel cladding tube made of Zr-based alloy for pressurized water reactor |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS61179860A JPS61179860A (en) | 1986-08-12 |
| JPH08961B2 true JPH08961B2 (en) | 1996-01-10 |
Family
ID=11494157
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP60001177A Expired - Lifetime JPH08961B2 (en) | 1985-01-08 | 1985-01-08 | Method for manufacturing nuclear fuel cladding tube made of Zr-based alloy for pressurized water reactor |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH08961B2 (en) |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2521328B2 (en) * | 1988-06-20 | 1996-08-07 | 三菱重工業株式会社 | Zirconium-based alloy nuclear fuel cladding |
Family Cites Families (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS6049539B2 (en) * | 1981-09-30 | 1985-11-02 | 日本電気ホームエレクトロニクス株式会社 | Conveyance device |
-
1985
- 1985-01-08 JP JP60001177A patent/JPH08961B2/en not_active Expired - Lifetime
Also Published As
| Publication number | Publication date |
|---|---|
| JPS61179860A (en) | 1986-08-12 |
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