JP2000105289A - Corrosion-resistant nuclear fuel cladding pipe - Google Patents
Corrosion-resistant nuclear fuel cladding pipeInfo
- Publication number
- JP2000105289A JP2000105289A JP10274540A JP27454098A JP2000105289A JP 2000105289 A JP2000105289 A JP 2000105289A JP 10274540 A JP10274540 A JP 10274540A JP 27454098 A JP27454098 A JP 27454098A JP 2000105289 A JP2000105289 A JP 2000105289A
- Authority
- JP
- Japan
- Prior art keywords
- zirconium alloy
- cladding tube
- nuclear fuel
- surface layer
- crystal grain
- 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
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
- Butt Welding And Welding Of Specific Article (AREA)
Abstract
Description
【０００１】[0001]
【発明の属する技術分野】本発明は核燃料被覆管に関す
る。[0001] The present invention relates to a nuclear fuel cladding tube.
【０００２】[0002]
【従来の技術】現在、原子炉に使用されている核燃料被
覆管は概ね以下に示す構造を有している。図１に核燃料
被覆管の縦方向の断面概略図を示す。被覆管本体１はジ
ルコニウム合金より成り、内部に、例えば、酸化ウラ
ン、或いは、酸化プルトニウムから成る核燃料ペレット
２が複数個装填されている。被覆管本体１の端部には、
上部端栓３と下部端栓４が設けられており、被覆管内部
は密閉された状態となっている。核燃料ペレット２は、
上部端栓３に一端が当接したスプリング５により固定さ
れている。この時、被覆管本体１と上部端栓３、被覆管
本体１と下部端栓４は、溶接により接続されている。2. Description of the Related Art At present, a nuclear fuel cladding tube used in a nuclear reactor generally has the following structure. FIG. 1 is a schematic cross-sectional view of a nuclear fuel cladding tube in a vertical direction. The cladding tube main body 1 is made of a zirconium alloy, and contains a plurality of nuclear fuel pellets 2 made of, for example, uranium oxide or plutonium oxide. At the end of the cladding tube body 1,
An upper end plug 3 and a lower end plug 4 are provided, and the inside of the cladding tube is sealed. Nuclear fuel pellet 2
One end of the upper end plug 3 is fixed by a spring 5 in contact with the upper end plug 3. At this time, the cladding tube main body 1 and the upper end plug 3 and the cladding tube main body 1 and the lower end plug 4 are connected by welding.
【０００３】この様な被覆管本体は、高温高圧の水蒸気
にさらされるという特殊な条件下にあるため、この様な
条件下における強度や耐食性が要求される。被覆管本体
を構成する材料としては、ジルカロイ−２、ジルカロイ
−４等のジルコニウムを主成分とする合金が使用されて
いる。当該合金には、強度や耐食性を向上させるために
種々の元素、例えばＳｎ、Ｆｅ、Ｃｒ、Ｎｉ等が添加さ
れている。[0003] Since such a clad tube main body is under a special condition of being exposed to high-temperature and high-pressure steam, strength and corrosion resistance under such a condition are required. As a material forming the cladding tube main body, an alloy containing zirconium as a main component such as Zircaloy-2 or Zircaloy-4 is used. Various elements such as Sn, Fe, Cr, and Ni are added to the alloy in order to improve strength and corrosion resistance.
【０００４】また、耐食性を向上させるため、製造工程
中、被覆管本体に対しβ−急冷処理を施している。これ
は合金中の添加元素とジルコニウムによって形成される
析出物、例えばＺｒ（Ｃｒ，Ｆｅ）2 やＺｒ2 （Ｎｉ，
Ｆｅ）を均一微細に分布させるために行う処理である。
この処理を施す事により、耐食性、特に耐ノジュラーコ
ロージョン性を向上させる事が出来る。Further, in order to improve the corrosion resistance, a β-quenching treatment is applied to the cladding tube body during the manufacturing process. This is because precipitates formed by the additive element and zirconium in the alloy, for example, Zr (Cr, Fe) 2 and Zr 2 (Ni,
This is a process performed to distribute Fe) uniformly and finely.
By performing this treatment, the corrosion resistance, particularly the nodular corrosion resistance, can be improved.
【０００５】ノジュラーコロージョンとは、白色の斑点
状の腐食であり、従来の核燃料被覆管においては、特に
核燃料被覆管の下部に発生しやすい腐食であった。現行
の核燃料被覆管においては、上記の様な材料や製造方法
を使用する事により耐ノジュラーコロージョン特性を向
上させており、現在の使用期限である４年間は十分な耐
食性を保持する事が出来、強度も十分である。[0005] The nodular corrosion is a white spot-like corrosion, and in a conventional nuclear fuel cladding tube, it is particularly likely to occur at a lower portion of the nuclear fuel cladding tube. In the current nuclear fuel cladding, the nodular corrosion resistance is improved by using the materials and manufacturing methods described above, and sufficient corrosion resistance can be maintained for the current expiration date of 4 years. , Strength is enough.
【０００６】しかしながら現在、核廃棄物減量化や経済
性の観点から、核燃料被覆管の使用可能な期間のさらな
る延長が望まれている。本発明者等は、核燃料被覆管を
現行より更に長期間使用した場合について試験・検討し
た結果、核燃料被覆管の下部に位置する被覆管本体と下
部端栓との溶接部及び熱影響部近傍、および、被覆管本
体に白色均一腐食が発生する事を見出した。この様な腐
食による生成物は、経時的には外表面に集積し、遂には
表面から剥離する。この剥離現象はブレイク・アウェイ
と呼ばれる。このブレイク・アウェイが起こると、核燃
料被覆管は水素吸収量の増加に起因する脆化で強度が低
下し、安全性が損なわれる。However, at present, from the viewpoint of reducing nuclear waste and economical efficiency, it is desired to further extend the usable period of the nuclear fuel cladding tube. The present inventors have conducted tests and studies on the case where the nuclear fuel cladding tube is used for a longer period of time than the current situation.As a result, the welded portion of the cladding body located at the lower part of the nuclear fuel cladding and the lower end plug and the vicinity of the heat affected zone, Also, it was found that white uniform corrosion occurred in the cladding tube main body. The products of such corrosion accumulate on the outer surface over time and eventually exfoliate from the surface. This peeling phenomenon is called break away. When this break away occurs, the nuclear fuel cladding becomes weaker due to embrittlement due to an increase in the amount of absorbed hydrogen, and the safety is impaired.
【０００７】更に、白色均一腐食は、基材変形強度に比
例した酸化被膜内の圧縮応力に起因すると予測し、白色
均一腐食抑制策として基材クリープによる適度な応力緩
和を考え、クリープ変形を起こしやすいように被覆管本
体のジルコニウム合金の平均結晶粒径を３μｍ以下にし
た核燃料被覆管を、現行より更に長期間使用した場合に
ついて試験・検討した結果、被覆管本体に白色均一腐食
は発生しないものの、クリープ伸びが著しくなり、核燃
料被覆管としての寸法精度が保てなくなる事が分かっ
た。Further, it is predicted that the uniform white corrosion is caused by the compressive stress in the oxide film in proportion to the deformation strength of the base material. Considering appropriate stress relaxation by creep of the base material as a measure for suppressing uniform white corrosion, creep deformation occurs. As a result of tests and examinations on the case where the average fuel grain diameter of the zirconium alloy of the cladding body was reduced to 3 μm or less for a longer time than the current case, the uniform cladding did not cause white uniform corrosion on the cladding body. It was found that the creep elongation became remarkable and the dimensional accuracy as a nuclear fuel cladding tube could not be maintained.
【０００８】[0008]
【発明が解決しようとする課題】上記した様に、従来の
核燃料被覆管においては、被覆管本体と下部端栓との溶
接部及び熱影響部近傍、および、被覆管本体に白色均一
腐食が発生し、長時間使用する事ができなかった。ま
た、白色均一腐食の抑制策として、被覆管本体のジルコ
ニウム合金の平均結晶粒径を小さくすると、クリープ伸
びで問題が生じた。本発明は以上の様な点に鑑み、白色
均一腐食を抑制し、更に、寸法精度を維持して、安全性
に優れ長期間使用可能な核燃料被覆管の提供を目的とす
るものである。As described above, in the conventional nuclear fuel cladding tube, white uniform corrosion occurs in the vicinity of the welded portion between the cladding tube main body and the lower end plug, the heat affected zone, and the cladding tube main body. And could not be used for a long time. In addition, when the average crystal grain size of the zirconium alloy in the cladding tube body is reduced as a measure for suppressing white uniform corrosion, a problem occurs in creep elongation. SUMMARY OF THE INVENTION In view of the foregoing, an object of the present invention is to provide a nuclear fuel cladding tube which suppresses white uniform corrosion, maintains dimensional accuracy, is excellent in safety, and can be used for a long time.
【０００９】[0009]
【課題を解決するための手段】本願第１の発明は、核燃
料ペレットを装填するジルコニウム合金からなる被覆管
本体と、この被覆管本体端部に溶接により接続され、前
記被覆管本体を密閉するジルコニウム合金からなる端栓
とを具備する核燃料被覆管において、前記被覆管本体と
端栓との溶接部及び熱影響部近傍の外表面層のジルコニ
ウム合金の平均結晶粒径が３μｍ以下であり、且つ、当
該外表面層の厚さが３〜５０μｍである核燃料被覆管で
ある。According to a first aspect of the present invention, there is provided a cladding tube body made of a zirconium alloy for loading nuclear fuel pellets, and a zirconium sealed by welding to an end of the cladding tube body to seal the cladding tube body. An end plug made of an alloy, the average crystal grain size of the zirconium alloy in the outer surface layer near the welded portion and the heat-affected zone between the clad tube main body and the end plug is 3 μm or less, and A nuclear fuel cladding tube having a thickness of the outer surface layer of 3 to 50 μm.
【００１０】本願第２の発明は、平均粒径５〜１０μｍ
のジルコニウム合金基体と、このジルコニウム合金基体
表面に形成された平均結晶粒径３μｍ以下、厚さ３〜５
０μｍのジルコニウム合金外表面とからなる核燃料被覆
管である。[0010] The second invention of the present application is characterized in that the average particle size is 5 to 10 µm
A zirconium alloy substrate having an average crystal grain size of 3 μm or less and a thickness of 3 to 5 μm formed on the surface of the zirconium alloy substrate.
A nuclear fuel cladding tube comprising a 0 μm zirconium alloy outer surface.
【００１１】[0011]
【発明の実施の形態】本発明に係わる被覆管を構成する
ジルコニウム合金としては、Ｚｒを主成分とする合金、
例えば、重量％でＳｎ：１．０〜１．７％、Ｆｅ：０．
０７〜０．２０％、Ｃｒ：０．０５〜０．１５％、Ｎ
ｉ：０．０３〜０．０８％、残部Ｚｒの組成を有するジ
ルカロイ−２、或いは、 Ｓｎ：１．０〜１．７％、Ｆ
ｅ：０．１８〜０．２４％、Ｃｒ：０．０５〜０．１５
％、 残部Ｚｒの組成を有するジルカロイ−４、上記ジ
ルカロイ−２またはジルカロイ−４に０．２％以上のＭ
ｏ及び０．１％以上のＮｂの少なくとも一種を合計量で
２．０％以下含有した合金、Ｚｒ−２．５％Ｎｂ系ジル
コニウム合金、 Ｚｒ−１％Ｎｂ系ジルコニウム合金、
または、オーゼナイト等のジルコニウム合金が挙げられ
る。特に、ジルカロイ−２またはジルカロイ−４が、強
度、及び、耐食性に優れており好ましい。BEST MODE FOR CARRYING OUT THE INVENTION As a zirconium alloy constituting a cladding tube according to the present invention, an alloy containing Zr as a main component,
For example, Sn: 1.0 to 1.7% and Fe: 0.
07 to 0.20%, Cr: 0.05 to 0.15%, N
i: 0.03 to 0.08%, Zircaloy-2 having a balance of Zr, or Sn: 1.0 to 1.7%, F
e: 0.18 to 0.24%, Cr: 0.05 to 0.15
%, Zircaloy-4 having a balance of Zr, 0.2% or more of M in the above Zircaloy-2 or Zircaloy-4.
an alloy containing at least one of o and 0.1% or more of Nb in a total amount of 2.0% or less, a Zr-2.5% Nb-based zirconium alloy, a Zr-1% Nb-based zirconium alloy,
Alternatively, a zirconium alloy such as ausenite may be used. In particular, Zircaloy-2 or Zircaloy-4 is preferable because of its excellent strength and corrosion resistance.
【００１２】また、本発明に係わる被覆管本体と溶接に
より接続される端栓の材料としても、上記ジルコニウム
合金を適用する事が望ましい。本発明に係わる核燃料被
覆管は、内面に純ジルコニウムからなるライナー層を有
していてもよい。Further, it is desirable to apply the above zirconium alloy also as a material of the end plug connected by welding to the cladding tube body according to the present invention. The nuclear fuel cladding tube according to the present invention may have a liner layer made of pure zirconium on the inner surface.
【００１３】本願の第１の発明に係わる、ジルコニウム
合金から成り核燃料ペレットを装填する被覆管本体と、
被覆管本体端部を密閉する端栓を備え、被覆管本体と端
栓とが溶接により接続された核燃料被覆管は、被覆管本
体と端栓との溶接部及び熱影響部近傍の外表面層のジル
コニウム合金の平均結晶粒径が３μｍ以下であり、且
つ、当該外表面層の厚さが３〜５０μｍである事を特徴
とする。[0013] A cladding tube body made of a zirconium alloy and loaded with nuclear fuel pellets according to the first invention of the present application,
A nuclear fuel cladding tube having an end plug for sealing the end of the cladding tube main body and the cladding tube main body and the end plug being connected by welding is provided on the outer surface layer near the welded portion between the cladding tube main body and the end plug and the heat-affected zone. Is characterized in that the average crystal grain size of the zirconium alloy is 3 μm or less, and the thickness of the outer surface layer is 3 to 50 μm.
【００１４】被覆管本体と端栓との溶接部及び熱影響部
近傍の外表面層のジルコニウム合金の平均結晶粒径が、
溶接後未処理で通常観察される約５〜３０μｍである
と、長時間の使用により白色均一腐食が発生する。溶接
部及び熱影響部近傍の外表面層のジルコニウム合金の平
均結晶粒径は３μｍ以下である事が望ましい。この範囲
であると酸化腐食に対する抵抗を高くする事が出来るた
めである。更に好ましくは、外表面層のジルコニウム合
金の平均結晶粒径は３μｍ以下である事が望ましい。こ
の範囲であると、より長時間の使用でも白色均一腐食の
問題は生じない。尚、溶接部及び熱影響部近傍の外表面
層のジルコニウム合金の平均結晶粒径は、実質的に０．
０１μｍ以上で良い。現在の技術水準では０．０１μｍ
より小さい平均結晶粒径を実現するのは困難である。The average crystal grain size of the zirconium alloy in the outer surface layer near the welded portion between the cladding tube main body and the end plug and the heat affected zone is as follows:
When it is about 5 to 30 μm which is usually observed after unwelding, uniform white corrosion occurs after prolonged use. It is desirable that the average crystal grain size of the zirconium alloy in the outer surface layer near the weld zone and the heat-affected zone be 3 μm or less. This is because the resistance to oxidation corrosion can be increased in this range. More preferably, the average crystal grain size of the zirconium alloy in the outer surface layer is desirably 3 μm or less. Within this range, the problem of white uniform corrosion does not occur even when used for a longer time. Incidentally, the average crystal grain size of the zirconium alloy in the outer surface layer in the vicinity of the weld zone and the heat-affected zone is substantially 0.1 mm.
It may be 01 μm or more. 0.01 μm at current technology level
It is difficult to achieve smaller average grain sizes.
【００１５】この様に、溶接部及び熱影響部近傍の外表
面層のジルコニウム合金の平均結晶粒径を、通常観察さ
れる５〜３０μｍの平均粒径と比べ小さい値に抑制する
事で、白色均一腐食の発生を通常より長時間側に遅延さ
せる事が出来る。As described above, by controlling the average crystal grain size of the zirconium alloy in the outer surface layer near the weld zone and the heat-affected zone to a value smaller than the average grain size of 5 to 30 μm normally observed, white The occurrence of uniform corrosion can be delayed to a longer time than usual.
【００１６】ジルコニウム合金の平均結晶粒径が抑制さ
れる溶接部及び熱影響部近傍の外表面層の厚さは、３〜
５０μｍである事が望ましい。３μｍより薄いと、酸化
腐食に対する抵抗が持続しない。５０μｍを超えると、
被覆管本体部分の溶接部及び熱影響部近傍の外表面層内
側の基体部分の厚さとの比率が増加し、０．９ｍｍの薄
肉被覆管本体のクリープ伸びに影響を与える。The thickness of the outer surface layer in the vicinity of the weld and the heat-affected zone where the average grain size of the zirconium alloy is suppressed is 3 to
Preferably, it is 50 μm. If the thickness is less than 3 μm, resistance to oxidative corrosion will not be maintained. If it exceeds 50 μm,
The ratio of the thickness of the clad tube main body portion to the thickness of the base portion inside the outer surface layer near the welded portion and the heat-affected zone increases, which affects the creep elongation of the thin clad tube main body of 0.9 mm.
【００１７】本願の第２の発明に係わるジルコニウム合
金から成る核燃料被覆管は、被覆管本体の外表面層のジ
ルコニウム合金の平均結晶粒径が３μｍ以下であり、且
つ、当該外表面層の厚さが３〜５０μｍであり、更に
は、被覆管本体の外表面層内側の基体部分のジルコニウ
ム合金の平均結晶粒径が５〜１０μｍである事を特徴と
する。In the nuclear fuel cladding tube made of a zirconium alloy according to the second invention of the present application, the average crystal grain size of the zirconium alloy in the outer surface layer of the cladding tube main body is 3 μm or less, and the thickness of the outer surface layer is Is 3 to 50 µm, and the average crystal grain size of the zirconium alloy in the base portion inside the outer surface layer of the cladding tube main body is 5 to 10 µm.
【００１８】被覆管本体の外表面層のジルコニウム合金
の平均結晶粒径が、通常観察される４〜５μｍである
と、長時間の使用により白色均一腐食が発生する。被覆
管本体の外表面層のジルコニウム合金の平均結晶粒径は
３μｍ以下である事が望ましい。この範囲であると酸化
腐食に対する抵抗を高くする事が出来るためである。更
に好ましくは、外表面層のジルコニウム合金の平均結晶
粒径は３μｍ以下である事が望ましい。この範囲である
と、より長時間の使用でも白色均一腐食の問題は生じな
い。尚、被覆管本体の外表面層のジルコニウム合金の平
均結晶粒径は、実質的に０．０１μｍ以上で良い。現在
の技術水準では０．０１μｍより小さい平均結晶粒径を
実現するのは困難である。If the average crystal grain size of the zirconium alloy in the outer surface layer of the cladding tube main body is 4 to 5 μm, which is usually observed, white uniform corrosion occurs after long use. The average crystal grain size of the zirconium alloy in the outer surface layer of the cladding tube main body is desirably 3 μm or less. This is because the resistance to oxidation corrosion can be increased in this range. More preferably, the average crystal grain size of the zirconium alloy in the outer surface layer is desirably 3 μm or less. Within this range, the problem of white uniform corrosion does not occur even when used for a longer time. The average crystal grain size of the zirconium alloy in the outer surface layer of the cladding tube body may be substantially 0.01 μm or more. With the current state of the art, it is difficult to achieve an average crystal grain size smaller than 0.01 μm.
【００１９】この様に、被覆管本体の外表面層のジルコ
ニウム合金の平均結晶粒径を、通常観察される４〜５μ
ｍの平均粒径と比べ小さい値に抑制する事で、白色均一
腐食の発生を通常より長時間側に遅延させる事が出来
る。As described above, the average crystal grain size of the zirconium alloy in the outer surface layer of the cladding tube main body is set to 4 to 5 μm, which is usually observed.
By suppressing the average particle size of m to a value smaller than the average particle size, the occurrence of white uniform corrosion can be delayed to a longer time than usual.
【００２０】ジルコニウム合金の平均結晶粒径が抑制さ
れる被覆管本体の外表面層の厚さは、３〜５０μｍであ
る事が望ましい。３μｍより薄いと、酸化腐食に対する
抵抗が持続しない。５０μｍを超えると、被覆管本体の
外表面層内側の基体部分の厚さとの比率が増加し、０．
９ｍｍの薄肉被覆管本体のクリープ伸びに影響を与え
る。The thickness of the outer surface layer of the cladding body in which the average crystal grain size of the zirconium alloy is suppressed is desirably 3 to 50 μm. If the thickness is less than 3 μm, resistance to oxidative corrosion will not be maintained. If it exceeds 50 μm, the ratio with respect to the thickness of the substrate portion inside the outer surface layer of the cladding tube body increases, and the ratio of the thickness to 0.
Affects the creep elongation of the 9 mm thin cladding body.
【００２１】被覆管本体の外表面層内側の基体部分のジ
ルコニウム合金の平均結晶粒径は、通常観察される４〜
５μｍであっても良い。しかし、被覆管本体の外表面層
内側の基体部分のジルコニウム合金の平均結晶粒径は、
５〜１０μｍである事が望ましい。基体部分のジルコニ
ウム合金の平均結晶粒径が５μｍより小さいと、クリー
プ強度が低下する傾向を示し、更に、圧延、焼鈍の繰り
返し数が増加し、核燃料被覆管としての製造コストが上
昇する。被覆管本体の外表面層内側の基体部分のジルコ
ニウム合金の平均結晶粒径が１０μｍより大きいと、靭
性が劣化し、核燃料被覆管としての信頼性が維持出来な
くなる。The average crystal grain size of the zirconium alloy in the base portion inside the outer surface layer of the cladding tube main body is usually 4 to 4.
It may be 5 μm. However, the average crystal grain size of the zirconium alloy in the base portion inside the outer surface layer of the cladding tube main body is:
It is desirable that it is 5 to 10 μm. If the average crystal grain size of the zirconium alloy in the base portion is smaller than 5 μm, the creep strength tends to decrease, the number of rolling and annealing cycles increases, and the production cost as a nuclear fuel cladding tube increases. If the average crystal grain size of the zirconium alloy in the base portion inside the outer surface layer of the clad tube main body is larger than 10 μm, the toughness is deteriorated and the reliability as a nuclear fuel clad tube cannot be maintained.
【００２２】被覆管本体の外表面層に耐食性の機能を集
約する事で、被覆管本体の外表面層内側の基体部分のジ
ルコニウム合金の平均結晶粒径を、通常観察される４〜
５μｍの平均粒径と比べ、この様に、よりクリープ強度
が強い範囲から選択出来るようになる。By integrating the function of corrosion resistance into the outer surface layer of the cladding tube main body, the average crystal grain size of the zirconium alloy in the base portion inside the outer surface layer of the cladding tube main body can be measured in the range of 4 to 4 which is usually observed.
As compared with the average particle diameter of 5 μm, it becomes possible to select from the range in which the creep strength is stronger.
【００２３】[0023]
【実施例】実施例１〜６、比較例１〜６ 工業的に通常に施工されている冷間加工と焼鈍の条件を
調節し、ジルカロイ−２から成る肉厚９００μｍの被覆
管本体の基体部分について所望の平均結晶粒径を得た事
を、ＳＥＭ（Scanning Electron Microscope）による断
面観察で確認した。次いで、上記被覆管本体とジルカロ
イ−２から成る端栓とを溶接し、溶接部及び熱影響部近
傍の平均結晶粒径を同様にして確認した。結果をまとめ
て表１に示す。更に、ショットピーニングと熱処理を組
み合わせた条件により、被覆管本体、溶接部、及び、熱
影響部近傍の各外表面層の平均結晶粒径を同様にして確
認した。同じく結果を表１に示す。各条件で作製した被
覆管試験片を高温高圧水蒸気試験装置であるオートクレ
ーブに装荷し、７７３Ｋ、１０．３ＭＰａの条件で２０
日間腐食試験した後、外観観察により色を判定すると共
に、 ＳＥＭによる断面観察から酸化被膜の厚さを求め
た。同じく結果を表１に示す。また、上記被覆管本体の
みを、５６１Ｋにて、１００ＭＰａの引っ張り荷重の
下、１００日後のクリープ変形量を測定した。結果を表
１に併せて示す。Examples 1 to 6 and Comparative Examples 1 to 6 The base portion of a 900 μm-thick cladding tube main body made of Zircaloy-2 by adjusting the conditions of cold working and annealing conventionally used in industry. That a desired average crystal grain size was obtained was confirmed by cross-sectional observation using a scanning electron microscope (SEM). Next, the cladding tube main body and the end plug made of Zircaloy-2 were welded, and the average crystal grain size in the vicinity of the welded portion and the heat-affected zone was similarly confirmed. The results are summarized in Table 1. Furthermore, the average crystal grain size of the cladding tube main body, the welded portion, and each outer surface layer in the vicinity of the heat-affected zone was similarly confirmed under the conditions in which shot peening and heat treatment were combined. Table 1 also shows the results. The cladding tube test piece prepared under each condition was loaded into an autoclave, which is a high-temperature and high-pressure steam tester, and subjected to 773K and 10.3 MPa conditions for 20 minutes.
After a corrosion test for one day, the color was determined by observing the appearance, and the thickness of the oxide film was determined by observing the cross section by SEM. Table 1 also shows the results. In addition, the creep deformation of only the cladding tube main body was measured at 561 K under a tensile load of 100 MPa after 100 days. The results are shown in Table 1.
【００２４】[0024]
【表１】 [Table 1]
【００２５】[0025]
【発明の効果】以上説明したように、本発明の核燃料被
覆管によれば、被覆管表面の結晶粒径に起因する白色均
一腐食が抑制でき、長期間に亘って安全に使用できる核
燃料被覆管を提供する事が出来る。As described above, according to the nuclear fuel cladding of the present invention, uniform white corrosion caused by the crystal grain size on the cladding surface can be suppressed, and the nuclear fuel cladding can be used safely for a long period of time. Can be provided.
【図１】核燃料被覆管の断面概略図。FIG. 1 is a schematic sectional view of a nuclear fuel cladding tube.
１・・・被覆管本体 ２・・・核燃料ペレット ３・・・上部端栓 ４・・・下部端栓 ５・・・スプリング DESCRIPTION OF SYMBOLS 1 ... Cladding pipe main body 2 ... Nuclear fuel pellet 3 ... Upper end plug 4 ... Lower end plug 5 ... Spring
Claims (2)
合金からなる被覆管本体と、この被覆管本体端部に溶接
により接続され、前記被覆管本体を密閉するジルコニウ
ム合金からなる端栓とを具備する核燃料被覆管におい
て、 前記被覆管本体と端栓との溶接部及び熱影響部近傍の外
表面層のジルコニウム合金の平均結晶粒径が３μｍ以下
であり、且つ、当該外表面層の厚さが３〜５０μｍであ
ることを特徴とする核燃料被覆管。1. A nuclear fuel cladding comprising: a cladding tube main body made of a zirconium alloy for loading nuclear fuel pellets; and an end plug made of a zirconium alloy sealed by welding to an end of the cladding tube main body to seal the cladding tube main body. In the pipe, the average crystal grain size of the zirconium alloy in the outer surface layer near the welded portion and the heat-affected zone between the cladding tube main body and the end plug is 3 μm or less, and the thickness of the outer surface layer is 3 to 50 μm A nuclear fuel cladding tube, characterized in that:
金基体と、このジルコニウム合金基体表面に形成された
平均結晶粒径３μｍ以下、厚さ３〜５０μｍのジルコニ
ウム合金外表面とからなる核燃料被覆管。2. A nuclear fuel cladding tube comprising a zirconium alloy substrate having an average particle size of 5 to 10 μm and an outer surface of a zirconium alloy having an average crystal grain size of 3 μm or less and a thickness of 3 to 50 μm formed on the surface of the zirconium alloy substrate.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP10274540A JP2000105289A (en) | 1998-09-29 | 1998-09-29 | Corrosion-resistant nuclear fuel cladding pipe |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP10274540A JP2000105289A (en) | 1998-09-29 | 1998-09-29 | Corrosion-resistant nuclear fuel cladding pipe |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JP2000105289A true JP2000105289A (en) | 2000-04-11 |
Family
ID=17543143
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP10274540A Pending JP2000105289A (en) | 1998-09-29 | 1998-09-29 | Corrosion-resistant nuclear fuel cladding pipe |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JP2000105289A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US8989339B2 (en) | 2010-11-08 | 2015-03-24 | Hitachi, Ltd. | Zirconium alloy material |
-
1998
- 1998-09-29 JP JP10274540A patent/JP2000105289A/en active Pending
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US8989339B2 (en) | 2010-11-08 | 2015-03-24 | Hitachi, Ltd. | Zirconium alloy material |
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