JPS6123264B2 - - Google Patents
Info
- Publication number
- JPS6123264B2 JPS6123264B2 JP58157610A JP15761083A JPS6123264B2 JP S6123264 B2 JPS6123264 B2 JP S6123264B2 JP 58157610 A JP58157610 A JP 58157610A JP 15761083 A JP15761083 A JP 15761083A JP S6123264 B2 JPS6123264 B2 JP S6123264B2
- Authority
- JP
- Japan
- Prior art keywords
- tube
- nuclear fuel
- cladding tube
- zirconium alloy
- zirconium
- 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
Links
- 238000005253 cladding Methods 0.000 claims description 29
- 229910001093 Zr alloy Inorganic materials 0.000 claims description 17
- 239000003758 nuclear fuel Substances 0.000 claims description 17
- 238000000034 method Methods 0.000 claims description 15
- 238000010438 heat treatment Methods 0.000 claims description 14
- 238000001816 cooling Methods 0.000 claims description 7
- 238000004519 manufacturing process Methods 0.000 claims description 6
- 230000007797 corrosion Effects 0.000 description 20
- 238000005260 corrosion Methods 0.000 description 20
- 230000035882 stress Effects 0.000 description 15
- 238000000137 annealing Methods 0.000 description 6
- 239000013078 crystal Substances 0.000 description 6
- 239000008188 pellet Substances 0.000 description 6
- 238000010791 quenching Methods 0.000 description 6
- 230000000171 quenching effect Effects 0.000 description 6
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 description 4
- XEEYBQQBJWHFJM-UHFFFAOYSA-N Iron Chemical compound [Fe] XEEYBQQBJWHFJM-UHFFFAOYSA-N 0.000 description 4
- 229910045601 alloy Inorganic materials 0.000 description 4
- 239000000956 alloy Substances 0.000 description 4
- 239000001257 hydrogen Substances 0.000 description 4
- 229910052739 hydrogen Inorganic materials 0.000 description 4
- VYZAMTAEIAYCRO-UHFFFAOYSA-N Chromium Chemical compound [Cr] VYZAMTAEIAYCRO-UHFFFAOYSA-N 0.000 description 3
- 238000005452 bending Methods 0.000 description 3
- 229910052804 chromium Inorganic materials 0.000 description 3
- 239000011651 chromium Substances 0.000 description 3
- 230000000694 effects Effects 0.000 description 3
- 239000000463 material Substances 0.000 description 3
- 238000005482 strain hardening Methods 0.000 description 3
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 3
- PXHVJJICTQNCMI-UHFFFAOYSA-N Nickel Chemical compound [Ni] PXHVJJICTQNCMI-UHFFFAOYSA-N 0.000 description 2
- ATJFFYVFTNAWJD-UHFFFAOYSA-N Tin Chemical compound [Sn] ATJFFYVFTNAWJD-UHFFFAOYSA-N 0.000 description 2
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 description 2
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 description 2
- 238000007796 conventional method Methods 0.000 description 2
- 238000005242 forging Methods 0.000 description 2
- 229910052742 iron Inorganic materials 0.000 description 2
- 238000005259 measurement Methods 0.000 description 2
- 238000002844 melting Methods 0.000 description 2
- 230000008018 melting Effects 0.000 description 2
- 229910052758 niobium Inorganic materials 0.000 description 2
- 239000010955 niobium Substances 0.000 description 2
- GUCVJGMIXFAOAE-UHFFFAOYSA-N niobium atom Chemical compound [Nb] GUCVJGMIXFAOAE-UHFFFAOYSA-N 0.000 description 2
- 239000001301 oxygen Substances 0.000 description 2
- 229910052760 oxygen Inorganic materials 0.000 description 2
- 229910052726 zirconium Inorganic materials 0.000 description 2
- WZECUPJJEIXUKY-UHFFFAOYSA-N [O-2].[O-2].[O-2].[U+6] Chemical compound [O-2].[O-2].[O-2].[U+6] WZECUPJJEIXUKY-UHFFFAOYSA-N 0.000 description 1
- 238000010521 absorption reaction Methods 0.000 description 1
- SHZGCJCMOBCMKK-KGJVWPDLSA-N beta-L-fucose Chemical compound C[C@@H]1O[C@H](O)[C@@H](O)[C@H](O)[C@@H]1O SHZGCJCMOBCMKK-KGJVWPDLSA-N 0.000 description 1
- 238000009835 boiling Methods 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 230000008602 contraction Effects 0.000 description 1
- 238000009792 diffusion process Methods 0.000 description 1
- 230000006355 external stress Effects 0.000 description 1
- 239000000446 fuel Substances 0.000 description 1
- 238000001192 hot extrusion Methods 0.000 description 1
- 150000004678 hydrides Chemical class 0.000 description 1
- 239000000203 mixture Substances 0.000 description 1
- 229910052759 nickel Inorganic materials 0.000 description 1
- 239000002245 particle Substances 0.000 description 1
- 239000002344 surface layer Substances 0.000 description 1
- 229910000439 uranium oxide Inorganic materials 0.000 description 1
- 229910000568 zirconium hydride Inorganic materials 0.000 description 1
- -1 zirconium hydrides Chemical class 0.000 description 1
Classifications
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22F—CHANGING THE PHYSICAL STRUCTURE OF NON-FERROUS METALS AND NON-FERROUS ALLOYS
- C22F1/00—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working
- C22F1/16—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working of other metals or alloys based thereon
- C22F1/18—High-melting or refractory metals or alloys based thereon
- C22F1/186—High-melting or refractory metals or alloys based thereon of zirconium or alloys based thereon
Description
【発明の詳細な説明】
〔発明の技術分野〕
本発明は核燃料ペレツトを装填する核燃料被覆
管の製造方法に係り、特に核燃料被覆管の耐食性
と機械的特性の向上を図つたものである。DETAILED DESCRIPTION OF THE INVENTION [Technical Field of the Invention] The present invention relates to a method for manufacturing a nuclear fuel cladding tube into which nuclear fuel pellets are loaded, and in particular aims to improve the corrosion resistance and mechanical properties of the nuclear fuel cladding tube.
一般にジルカロイ―２、ジルカロイ―４などの
ジルコニウム合金は、熱中性子吸収断面積が小さ
いこと、原子炉内環境に対する耐食性に優れてい
ること、構造材料として機械的性質を充分に備え
ていることなどの理由から原子炉の核燃料被覆管
として多く用いられている。
In general, zirconium alloys such as Zircaloy-2 and Zircaloy-4 have a small thermal neutron absorption cross section, excellent corrosion resistance in the reactor environment, and have sufficient mechanical properties as a structural material. For this reason, it is often used as nuclear fuel cladding for nuclear reactors.
この核燃料被覆管１は第１図および第２図に示
すように、ジルコニウム合金管で形成され、内部
にはペレツト状に形成された、例えば酸化ウラン
あるいは酸化プルトニウムなどの核燃料ペレツト
２が複数個積層充填され、更にこの核燃料ペレツ
ト２は、前記被覆管１の上部端栓３に一端が当接
したスプリング４により固定されている。 As shown in FIGS. 1 and 2, this nuclear fuel cladding tube 1 is formed of a zirconium alloy tube, and inside thereof, a plurality of nuclear fuel pellets 2 formed in the form of pellets, such as uranium oxide or plutonium oxide, are stacked. The nuclear fuel pellets 2 are then fixed by a spring 4 whose one end abuts against the upper end plug 3 of the cladding tube 1.
しかしながら、これらジルコニウムの合金で形
成された核燃料被覆管は、その使用時間の経過と
ともに、いわゆるノジユラーコロージヨンと呼ば
れる腐食反応による白色腐食生成物が、その表面
に斑点状に生成してくることがある。これはジル
コニウム合金が高温水と反応し、この表面に酸化
膜が形成された状態で、生成された水素が合金基
材と表面の酸化膜との間に蓄積して腐食生成物を
形成するものである。この腐食生成物は、経時的
に表面に集積し、ついには表面から剥離して、被
覆管の強度低下を招くおそれがある。また生成さ
れた水素が合金内部に侵入するとジルコニウムの
水素化物が形成され、これが表面と垂直方向に形
成されると、連続した水素化物による、いわゆる
水素脆性の問題があつた。 However, as nuclear fuel cladding tubes made of these zirconium alloys are used, white corrosion products due to a corrosion reaction called so-called nodular corrosion can be formed in spots on the surface of the nuclear fuel cladding tubes. be. This occurs when the zirconium alloy reacts with high-temperature water, forming an oxide film on its surface, and the generated hydrogen accumulates between the alloy base material and the oxide film on the surface, forming corrosion products. It is. This corrosion product accumulates on the surface over time and may eventually peel off from the surface, leading to a decrease in the strength of the cladding. Furthermore, when the generated hydrogen penetrates into the alloy, zirconium hydrides are formed, and when these are formed in a direction perpendicular to the surface, there is a problem of so-called hydrogen embrittlement due to continuous hydrides.
このような問題を解決するため、従来は、仕上
りの被覆管寸法に管絞り工程を行つたジルコニウ
ム合金管を、最終の熱処理工程において、β焼入
する方法（特開昭55−100947）およびジルコニウ
ム合金管の表面を急熱溶融後急冷する方法（特開
昭55−50453）などが提案されている。これらの
方法は、何れもジルコニウム合金管の少なくとも
表面部の結晶構造を焼入により針状結晶粒のβ相
（体心立方格子）に変えることにより耐ノジユラ
ーコロージヨン性を向上させるものである。 In order to solve these problems, conventionally, zirconium alloy tubes are subjected to a tube drawing process to the finished cladding size, and then β-quenched in the final heat treatment process (Japanese Patent Application Laid-open No. 100947/1983) and zirconium alloy tubes are A method has been proposed in which the surface of an alloy tube is rapidly melted and then rapidly cooled (Japanese Unexamined Patent Publication No. 50453/1983). All of these methods improve the nodular corrosion resistance by changing the crystal structure of at least the surface of the zirconium alloy tube into a β phase (body-centered cubic lattice) of acicular crystal grains through quenching. .
しかしながら、これら焼入による方法は、耐ノ
ジユラーコロージヨン性を向上させる反面、被覆
管としての機械的特性が劣化する上、肉厚が0.8
mm、長さが４ｍもの細長い管であるため、焼入時
に曲りやねじれを生ずる問題があり、満足すべき
ものではなかつた。 However, although these quenching methods improve the nodular corrosion resistance, they deteriorate the mechanical properties of the cladding, and the wall thickness is 0.8
Since it is a long and thin tube with a length of 4 m, there is a problem that it bends or twists during quenching, which is not satisfactory.
本発明は、かかる従来の問題点に鑑みなされた
もので、優れた耐ノジユラーコロージヨン性を有
すると共に、機械的特性にも優れ、しかも曲りや
ねじれの少ない核燃料被覆管の製造方法を提供す
るものである。
The present invention has been made in view of such conventional problems, and provides a method for manufacturing a nuclear fuel cladding tube that has excellent nodular corrosion resistance, excellent mechanical properties, and less bending and twisting. It is something.
本発明はジルコニウム合金管を所定の内径およ
び肉厚まで縮少する管絞り工程を行つた後、最終
の熱処理として、ジルコニウム合金管をα領域の
高温に急加熱して、短時間保持した後、直ちに急
冷することを特徴とするものである。
In the present invention, after performing a tube drawing process to reduce the zirconium alloy tube to a predetermined inner diameter and wall thickness, as a final heat treatment, the zirconium alloy tube is rapidly heated to a high temperature in the α region and held for a short time. It is characterized by immediate quenching.
以下本発明を詳細に説明する。 The present invention will be explained in detail below.
本発明において用いるジルコニウム合金として
は、例えば重量比でスズ1.2〜1.7％、鉄0.07〜
0.20％、クロム0.05〜0.15％、ニツケル0.03〜
0.08％、残部ジルコニウムよりなるジルカロイ―
２と呼称されているもの、スズ1.2〜1.7％、鉄
0.18〜0.24％、クロム0.07〜0.13％、残部ジルコ
ニウムよりなるジルカロイ―４と呼称されている
もの、あるいはジルコニウム―2.5％ニオブ系、
ジルコニウム―１％ニオブ系、またはオーゼナイ
トなどのジルコニウム合金に適用することができ
る。 The zirconium alloy used in the present invention includes, for example, 1.2 to 1.7% tin and 0.07 to 1.7% iron by weight.
0.20%, chromium 0.05~0.15%, nickel 0.03~
Zircaloy consisting of 0.08%, balance zirconium
2, tin 1.2-1.7%, iron
Zircaloy-4, which consists of 0.18-0.24% chromium, 0.07-0.13% chromium, and the balance zirconium, or zirconium-2.5% niobium,
It can be applied to zirconium-1% niobium-based or zirconium alloys such as ausenite.
次に本発明被覆管の製造方法について説明す
る。 Next, a method for manufacturing the cladding tube of the present invention will be explained.
被覆管はジルコニウム合金を溶解、鍛造して中
空ビレツトを形成し、次いで熱間押出した後、冷
間加工による管絞り工程を経て、仕上りの内径お
よび肉厚まで縮少する。この冷間加工による管絞
り工程は、中間に焼鈍を組み合せて、３〜４回の
パスを経て、最終焼鈍を行なう。最終焼鈍の温度
は通常580℃近傍で約２時間半加熱して行うが、
このようにして得られた被覆管は残留歪がなく、
また結晶構造はα相（六方晶形）である。 The cladding tube is made by melting and forging zirconium alloy to form a hollow billet, then hot extruding and then cold working to reduce the tube to the finished inner diameter and wall thickness. This tube drawing process by cold working is combined with intermediate annealing, and final annealing is performed after three to four passes. The final annealing temperature is usually around 580℃ for about 2 and a half hours.
The cladding tube obtained in this way has no residual strain,
Moreover, the crystal structure is α phase (hexagonal crystal structure).
ここまでは従来の方法と同一であるが、本発明
においては、最終焼鈍後に、次の熱処理工程を付
加したものである。 The method up to this point is the same as the conventional method, but in the present invention, the following heat treatment step is added after the final annealing.
最終焼鈍したジルコニウム合金管を、例えば高
周波加熱により、表面部をα領域の780〜860℃ま
で急加熱して、数秒間保持した後、急冷して、表
面部に圧縮応力を残留させるものである。このよ
うに急加熱、急冷を行なうと、内部に大きな温度
差を生じ、表面部は内部の高温部を包んで冷却す
るため自由に収縮できない。この結果、表面部は
内部の高温部のために張力を受けながら冷却す
る。冷却初期においては、温度は比較的高いから
降伏点が低く、表面部は多少永久変形を起こす。
しかも中心部まで室温になつた時には、表面部は
中心部に比べて、変形分だけ伸び過ぎていること
になり、これが内部からの収縮力を受け、最終的
には、表面部は中心部によつて圧縮された状態と
なる。即ち表面部には圧縮応力が残留し、中心部
は引張応力が働いて、つり合つた状態となる。 The surface of a final annealed zirconium alloy tube is rapidly heated, for example, by high-frequency heating, to 780 to 860°C in the α region, held for several seconds, and then rapidly cooled to leave compressive stress on the surface. . When rapid heating and cooling are performed in this manner, a large temperature difference occurs inside the product, and the surface portion cannot contract freely because it envelops the high temperature portion inside and cools it. As a result, the surface part cools while being under tension due to the internal high temperature part. In the early stage of cooling, the temperature is relatively high, so the yield point is low, and the surface portion undergoes some permanent deformation.
Moreover, when the temperature reaches room temperature to the center, the surface will have expanded too much compared to the center by the amount of deformation, and this will receive a contraction force from within, and eventually the surface will expand to the center. This results in a compressed state. In other words, compressive stress remains on the surface, and tensile stress acts on the center, resulting in a balanced state.
このように表面部に圧縮応力が残留した場態
で、結晶構造がα相のままでも、耐ノジユラーコ
ロージヨン性が向上する理由について詳らかでは
ないが、圧縮応力が加つた状態では、結晶格子が
密につまつているため、表面から内部へ酸素が拡
散しにくくなるためであると考えられる。このよ
うに酸素の拡散が阻止されると酸化膜が形成され
にくくなり、酸化膜と合金表面との間の水素の蓄
積が防止され、耐ノジユラーコロージヨン性が向
上するものである。 It is not clear why the nodular corrosion resistance improves even if the crystal structure remains in the α phase when compressive stress remains on the surface, but when compressive stress is applied, the crystal lattice This is thought to be because oxygen is difficult to diffuse from the surface to the inside because the particles are tightly packed together. When oxygen diffusion is inhibited in this way, an oxide film is less likely to be formed, hydrogen is prevented from accumulating between the oxide film and the alloy surface, and the nodular corrosion resistance is improved.
なお本発明において加熱温度を780〜860℃のα
領域に限定した理由は、この温度範囲に加熱して
急冷することにより８〜46Kg／mm2の圧縮応力が残
留し、特に残留圧縮応力が20Kg／mm2以上で効果的
な耐ノジユラーコロージヨン性が得られる。この
場合780℃未満の加熱では、充分な圧縮応力が残
留せず、また860℃を越えるとβ領域となり、急
冷すると焼入が行われて、表面層がβ相（体心立
方格子）になる上、曲りやねじれが大きく、しか
も機械的特性が劣化するからである。 In addition, in the present invention, the heating temperature is set to α of 780 to 860°C.
The reason for limiting this range is that by heating to this temperature range and rapidly cooling it, a compressive stress of 8 to 46 kg/mm 2 remains, and the nodular corrosion resistance is particularly effective when the residual compressive stress is 20 kg/mm 2 or higher. You can get sex. In this case, if heated below 780℃, sufficient compressive stress will not remain, and if it exceeds 860℃, it will become a β region, and if it is rapidly cooled, quenching will occur and the surface layer will become a β phase (body-centered cubic lattice). First, the bending and twisting are large, and the mechanical properties are deteriorated.
また本発明では被覆管の表面に圧縮応力が残留
しているので、例えば引張強さが約45Kg／mm2のジ
ルコニウム合金からなる被覆管に、本発明の熱処
理を施して圧縮応力を20Kg／mm2残留させると、65
Kg／mm2までの外部応力で破断せずに耐えられるこ
とになる。更に圧縮応力が残留していることによ
り耐力が大きくなると共に伸びが小さなり、特に
沸騰高温水中に長時間曝らされる核燃料被覆管に
おいて、強度、クリープ特性など機械的特性の改
善効果が大きい。 In addition, in the present invention, since compressive stress remains on the surface of the cladding tube, for example, a cladding tube made of a zirconium alloy with a tensile strength of about 45 kg/mm 2 is subjected to the heat treatment of the present invention to reduce the compressive stress to 20 kg/mm 2. 2 remaining, 65
It can withstand external stress of up to Kg/ mm2 without breaking. Furthermore, residual compressive stress increases yield strength and reduces elongation, which has a significant effect on improving mechanical properties such as strength and creep properties, especially in nuclear fuel cladding tubes that are exposed to boiling, high-temperature water for long periods of time.
ジルカロイ―２を用い、通常の溶解、鍛造によ
り中空ビレツトを形成した後、熱間押出しを行
い、次いで４回の冷間加工による管絞絞りと、真
空焼鈍を繰り返して最終の仕上り形状とした。
Using Zircaloy-2, a hollow billet was formed by conventional melting and forging, followed by hot extrusion, followed by repeated tube drawing by cold working four times and vacuum annealing to obtain the final finished shape.
次にこの被覆管を、サイリスタ式の高周波加熱
炉を用いて、急速加熱して表面を800℃に加熱し
た。この場合炉内滞留時間（保持時間）は約５秒
であつた。この後、直ちに水冷して急速冷却を行
つた。 Next, this cladding tube was rapidly heated using a thyristor type high frequency heating furnace to heat the surface to 800°C. In this case, the residence time in the furnace (holding time) was about 5 seconds. Thereafter, the mixture was immediately cooled with water for rapid cooling.
次いで表面の酸化膜を研麿除去して被覆管を製
造した。 Next, the oxide film on the surface was removed to produce a cladding tube.
このようにして得られた被覆管を、500℃、105
気圧の高温高圧水蒸気中に放置して、加速腐食試
験を行なつて耐食性を調べた。この結果は第３図
のグラフに曲線ａで示すように腐食による増量は
48時間経過後も、僅かであつた。 The cladding tube thus obtained was heated at 500°C and 105°C.
The corrosion resistance was examined by leaving the specimen in high-temperature, high-pressure steam and conducting an accelerated corrosion test. This result shows that as shown by curve a in the graph of Figure 3, the weight increase due to corrosion is
Even after 48 hours, the amount was still small.
また機械的特性を調べるため、耐力と伸びを調
べたところ、耐力は42.8Kg／mm2、伸びは32.9％
で、優れたクリープ特性を有することが確認され
た。 In addition, in order to investigate the mechanical properties, we investigated the yield strength and elongation, and found that the yield strength was 42.8Kg/mm 2 and the elongation was 32.9%.
It was confirmed that the material had excellent creep properties.
更に本発明被覆管の表面応力状態を見るためＸ
線により残留圧縮応力を測定したところ、28Kg／
mm2であつた。なおこの場合のＸ線測定はCrKα
（クロムケアルフアー）線を用い（2022）面から
の回析線を用いて行なつた。 Furthermore, in order to observe the surface stress state of the cladding tube of the present invention,
When the residual compressive stress was measured using a wire, it was 28Kg/
It was warm in mm2 . In this case, the X-ray measurement is CrKα
(Chromkjärfer) line and the diffraction line from the (2022) plane.
次に本発明と比較するために、最終焼鈍後、何
ら熱処理を行わない従来の被覆管についても同様
に加速腐食試験を行つた。 Next, in order to compare with the present invention, accelerated corrosion tests were similarly conducted on conventional cladding tubes that were not subjected to any heat treatment after final annealing.
この結果、腐食による増量は第３図のグラフに
曲線ｂで示すように急激な増加カーブを画いた。
また同様に機械的特性を調べたところ耐力は42.8
Kg／mm2、伸びは32.9％であり、本発明の実施例品
と同等であつた。またＸ線による圧縮応力の測定
では、残留が認められなかつた。 As a result, the weight increase due to corrosion drew a sharp increasing curve as shown by curve b in the graph of FIG.
Similarly, when the mechanical properties were investigated, the yield strength was 42.8.
Kg/mm 2 and elongation were 32.9%, which were equivalent to the example products of the present invention. In addition, no residual was found in the measurement of compressive stress using X-rays.
以上説明した如く、本発明による核燃料被覆管
の製造方法のよればα領域から急冷して、表面に
圧縮応力を残留させることにより、耐ノジユラー
コロージヨン性と機械的特性の向上が図れると共
に、β領域からの焼入に比べて低温度からの急冷
であるため曲りやねじれの発生が少なく、寸法精
度にも優れているなど顕著な効果を有するもので
ある。
As explained above, according to the method for manufacturing a nuclear fuel cladding according to the present invention, by rapidly cooling from the α region and leaving compressive stress on the surface, nodular corrosion resistance and mechanical properties can be improved, and Compared to quenching from the β region, this method has remarkable effects such as less bending and twisting due to rapid cooling from a lower temperature and superior dimensional accuracy.
第１図は核燃料被覆管の内部に核燃料ペレツト
を装填した燃料棒の縦断面図、第２図は第１図の
拡大水平断面図、第３図は本発明による被覆管と
従来方法による被覆管との腐食増量の時間変化を
示すグラフである。
１……被覆管、２……核燃料ペレツト、３……
上部端栓、４……スプリング、５……下部端栓。
Fig. 1 is a vertical cross-sectional view of a fuel rod in which nuclear fuel pellets are loaded inside a nuclear fuel cladding tube, Fig. 2 is an enlarged horizontal cross-sectional view of Fig. 1, and Fig. 3 is a cladding tube according to the present invention and a cladding tube according to a conventional method. It is a graph showing the change in corrosion weight increase over time. 1... Cladding tube, 2... Nuclear fuel pellets, 3...
Upper end plug, 4... Spring, 5... Lower end plug.
Claims (1)
ながら複数回のパスを順次経て、所定の内径およ
び肉厚まで縮小する管絞り工程を行つた後、最終
の熱処理として、ジルコニウム合金管の表面をα
領域の高温に急加熱して、短時間保持した後、直
ちに急冷することを特徴とする核燃料被覆管の製
造方法。 ２ α領域での加熱温度を780〜860℃としたこと
を特徴とする特許請求の範囲第１項記載の核燃料
被覆管の製造方法。[Claims] 1. A zirconium alloy tube is sequentially passed through multiple passes while performing intermediate heat treatment, and after a tube drawing process is performed to reduce the inner diameter and wall thickness to a predetermined value, the zirconium alloy tube is subjected to a final heat treatment. The surface of the tube is α
A method for producing a nuclear fuel cladding tube, which comprises rapidly heating the area to a high temperature, holding it for a short period of time, and then immediately cooling it rapidly. 2. The method for manufacturing a nuclear fuel cladding tube according to claim 1, characterized in that the heating temperature in the α region is 780 to 860°C.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP58157610A JPS6050155A (en) | 1983-08-29 | 1983-08-29 | Production of nuclear fuel cladding pipe |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP58157610A JPS6050155A (en) | 1983-08-29 | 1983-08-29 | Production of nuclear fuel cladding pipe |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS6050155A JPS6050155A (en) | 1985-03-19 |
| JPS6123264B2 true JPS6123264B2 (en) | 1986-06-05 |
Family
ID=15653485
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP58157610A Granted JPS6050155A (en) | 1983-08-29 | 1983-08-29 | Production of nuclear fuel cladding pipe |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS6050155A (en) |
Families Citing this family (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| FR2584097B1 (en) * | 1985-06-27 | 1987-12-11 | Cezus Co Europ Zirconium | METHOD FOR MANUFACTURING A BLIND CORROSIVE CLADDING TUBE BLANK IN ZIRCONIUM ALLOY |
| US6126762A (en) * | 1998-03-30 | 2000-10-03 | General Electric Company | Protective coarsening anneal for zirconium alloys |
-
1983
- 1983-08-29 JP JP58157610A patent/JPS6050155A/en active Granted
Also Published As
| Publication number | Publication date |
|---|---|
| JPS6050155A (en) | 1985-03-19 |
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