JPS6355494A - Composite coated tube for nuclear fuel - Google Patents
Composite coated tube for nuclear fuelInfo
- Publication number
- JPS6355494A JPS6355494A JP61198884A JP19888486A JPS6355494A JP S6355494 A JPS6355494 A JP S6355494A JP 61198884 A JP61198884 A JP 61198884A JP 19888486 A JP19888486 A JP 19888486A JP S6355494 A JPS6355494 A JP S6355494A
- Authority
- JP
- Japan
- Prior art keywords
- zirconium
- zircaloy
- tubular member
- cladding tube
- cladding
- 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
- 239000002131 composite material Substances 0.000 title claims description 12
- 239000003758 nuclear fuel Substances 0.000 title claims description 11
- 238000005253 cladding Methods 0.000 claims description 55
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 claims description 41
- 229910052726 zirconium Inorganic materials 0.000 claims description 41
- 229910045601 alloy Inorganic materials 0.000 claims description 25
- 239000000956 alloy Substances 0.000 claims description 25
- PXHVJJICTQNCMI-UHFFFAOYSA-N Nickel Chemical compound [Ni] PXHVJJICTQNCMI-UHFFFAOYSA-N 0.000 claims description 17
- 229910052787 antimony Inorganic materials 0.000 claims description 15
- WATWJIUSRGPENY-UHFFFAOYSA-N antimony atom Chemical compound [Sb] WATWJIUSRGPENY-UHFFFAOYSA-N 0.000 claims description 15
- 238000000137 annealing Methods 0.000 claims description 9
- VYZAMTAEIAYCRO-UHFFFAOYSA-N Chromium Chemical compound [Cr] VYZAMTAEIAYCRO-UHFFFAOYSA-N 0.000 claims description 8
- 229910052804 chromium Inorganic materials 0.000 claims description 8
- 239000011651 chromium Substances 0.000 claims description 8
- 229910052759 nickel Inorganic materials 0.000 claims description 8
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 claims description 7
- 239000001301 oxygen Substances 0.000 claims description 7
- 229910052760 oxygen Inorganic materials 0.000 claims description 7
- 229910001257 Nb alloy Inorganic materials 0.000 claims description 6
- 239000012535 impurity Substances 0.000 claims description 5
- 239000010955 niobium Substances 0.000 claims description 4
- GUCVJGMIXFAOAE-UHFFFAOYSA-N niobium atom Chemical compound [Nb] GUCVJGMIXFAOAE-UHFFFAOYSA-N 0.000 claims description 4
- 238000001953 recrystallisation Methods 0.000 claims description 4
- 229910052751 metal Inorganic materials 0.000 claims description 3
- 239000002184 metal Substances 0.000 claims description 3
- 229910052758 niobium Inorganic materials 0.000 claims description 3
- GFUGMBIZUXZOAF-UHFFFAOYSA-N niobium zirconium Chemical group [Zr].[Nb] GFUGMBIZUXZOAF-UHFFFAOYSA-N 0.000 claims description 2
- 239000004615 ingredient Substances 0.000 claims 1
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 description 19
- 239000001257 hydrogen Substances 0.000 description 19
- 229910052739 hydrogen Inorganic materials 0.000 description 19
- 229910001093 Zr alloy Inorganic materials 0.000 description 17
- XEEYBQQBJWHFJM-UHFFFAOYSA-N Iron Chemical compound [Fe] XEEYBQQBJWHFJM-UHFFFAOYSA-N 0.000 description 14
- 239000013078 crystal Substances 0.000 description 12
- 239000000446 fuel Substances 0.000 description 12
- 239000000498 cooling water Substances 0.000 description 11
- 229910001245 Sb alloy Inorganic materials 0.000 description 10
- 239000002140 antimony alloy Substances 0.000 description 10
- 239000008188 pellet Substances 0.000 description 8
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 8
- 230000007797 corrosion Effects 0.000 description 7
- 238000005260 corrosion Methods 0.000 description 7
- 230000003993 interaction Effects 0.000 description 7
- 229910052742 iron Inorganic materials 0.000 description 7
- 230000000694 effects Effects 0.000 description 6
- 238000004519 manufacturing process Methods 0.000 description 6
- 230000007423 decrease Effects 0.000 description 5
- 230000004992 fission Effects 0.000 description 5
- 239000000047 product Substances 0.000 description 5
- IJGRMHOSHXDMSA-UHFFFAOYSA-N Atomic nitrogen Chemical compound N#N IJGRMHOSHXDMSA-UHFFFAOYSA-N 0.000 description 4
- ATJFFYVFTNAWJD-UHFFFAOYSA-N Tin Chemical compound [Sn] ATJFFYVFTNAWJD-UHFFFAOYSA-N 0.000 description 4
- FSVFALHAIQMCRU-UHFFFAOYSA-N antimony;zirconium Chemical compound [Sb]#[Zr] FSVFALHAIQMCRU-UHFFFAOYSA-N 0.000 description 4
- 238000006243 chemical reaction Methods 0.000 description 4
- 238000005336 cracking Methods 0.000 description 4
- 238000000034 method Methods 0.000 description 4
- 239000002244 precipitate Substances 0.000 description 4
- 239000000126 substance Substances 0.000 description 4
- 238000005275 alloying Methods 0.000 description 3
- 238000010894 electron beam technology Methods 0.000 description 3
- 239000000203 mixture Substances 0.000 description 3
- 238000010521 absorption reaction Methods 0.000 description 2
- 230000015572 biosynthetic process Effects 0.000 description 2
- 238000004140 cleaning Methods 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 238000001125 extrusion Methods 0.000 description 2
- 239000000463 material Substances 0.000 description 2
- 238000002844 melting Methods 0.000 description 2
- 230000008018 melting Effects 0.000 description 2
- 229910052757 nitrogen Inorganic materials 0.000 description 2
- 230000002285 radioactive effect Effects 0.000 description 2
- 238000005096 rolling process Methods 0.000 description 2
- -1 In addition Chemical compound 0.000 description 1
- ZCYVEMRRCGMTRW-UHFFFAOYSA-N 7553-56-2 Chemical compound [I] ZCYVEMRRCGMTRW-UHFFFAOYSA-N 0.000 description 1
- VYPSYNLAJGMNEJ-UHFFFAOYSA-N Silicium dioxide Chemical compound O=[Si]=O VYPSYNLAJGMNEJ-UHFFFAOYSA-N 0.000 description 1
- 229910052770 Uranium Inorganic materials 0.000 description 1
- 238000004458 analytical method Methods 0.000 description 1
- 239000003795 chemical substances by application Substances 0.000 description 1
- 238000011109 contamination Methods 0.000 description 1
- 238000005520 cutting process Methods 0.000 description 1
- 230000007547 defect Effects 0.000 description 1
- 230000006866 deterioration Effects 0.000 description 1
- 230000001627 detrimental effect Effects 0.000 description 1
- 229910001651 emery Inorganic materials 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- 238000007730 finishing process Methods 0.000 description 1
- 238000000227 grinding Methods 0.000 description 1
- 238000010438 heat treatment Methods 0.000 description 1
- 150000004678 hydrides Chemical class 0.000 description 1
- 238000010348 incorporation Methods 0.000 description 1
- 239000003317 industrial substance Substances 0.000 description 1
- 230000000977 initiatory effect Effects 0.000 description 1
- PNDPGZBMCMUPRI-UHFFFAOYSA-N iodine Chemical compound II PNDPGZBMCMUPRI-UHFFFAOYSA-N 0.000 description 1
- 239000011630 iodine Substances 0.000 description 1
- 229910052740 iodine Inorganic materials 0.000 description 1
- 238000003754 machining Methods 0.000 description 1
- 239000011159 matrix material Substances 0.000 description 1
- 230000035515 penetration Effects 0.000 description 1
- 230000005855 radiation Effects 0.000 description 1
- 239000002994 raw material Substances 0.000 description 1
- 238000009864 tensile test Methods 0.000 description 1
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 1
- 238000003466 welding Methods 0.000 description 1
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
- Rigid Pipes And Flexible Pipes (AREA)
Abstract
(57)【要約】本公報は電子出願前の出願データであるた
め要約のデータは記録されません。(57) [Abstract] This bulletin contains application data before electronic filing, so abstract data is not recorded.
Description
【発明の詳細な説明】
【産業上の利用分野〕
本発明は核燃料用複合被覆管に係り、特に水冷却型原子
炉核燃料要素の燃料ペレットと設料被覆管との相互作用
による好ましくない作用ならびに水素吸収を最小とする
のに好適なジルコニウム基合金を使用した核燃料用複合
被覆管に関するものである。DETAILED DESCRIPTION OF THE INVENTION [Field of Industrial Application] The present invention relates to a composite cladding tube for nuclear fuel, and in particular to a composite cladding tube for nuclear fuel, and particularly to a nuclear fuel element of a water-cooled nuclear reactor, the undesirable effects caused by the interaction between the fuel pellets and the facility cladding tube, and The present invention relates to a composite cladding tube for nuclear fuel using a zirconium-based alloy suitable for minimizing hydrogen absorption.
水冷却型原子炉核燃料要素には、高ジルコニウム合金か
らなる被覆管が使用されてきた。普通に使用されている
合金の例としては、ジルカロイ−２、ジルカロイ−４お
よびジルコニウム−２，５重量％ニオブ合金がある。こ
れらの合金は、耐放射線特性１機械的特性および高温水
あるいは高温の水蒸気に対する耐食性の面から採用され
ている。Cladding tubes made of high zirconium alloys have been used in water-cooled nuclear reactor nuclear fuel elements. Examples of commonly used alloys include Zircaloy-2, Zircaloy-4, and Zirconium-2.5% by weight niobium alloys. These alloys are selected from the viewpoints of radiation resistance, mechanical properties, and corrosion resistance against high-temperature water or high-temperature steam.
ジルカロイ−２およびジルカロイ−４の工業上の化学成
分に関する規格は、エイ・ニス・ティー・エム（ＡＳＴ
Ｍ）Ｂ２Ｓ３−８０ （合金ＵＳＮＮａＲ６０８０２お
よびＲ６０８０４）に公表されている要求に合致してい
る。さらに、これらの要求に加えて、これらの合金の酸
素含有量は９００ないし１６００ｐｐ＋＊、代表的には
１２００±Ｉ２００ｐｐ罵であることが要求されている
。The standards regarding the industrial chemical components of Zircaloy-2 and Zircaloy-4 are published by A.N.S.T.M. (AST).
M) Meets the requirements published in B2S3-80 (alloys USNNaR60802 and R60804). Further, in addition to these requirements, the oxygen content of these alloys is required to be between 900 and 1600 pp+*, typically 1200±I200 pp+.
ジルカロイ被覆管の製造に際して、インゴットを熱間加
工して中間寸法のビレットまたは棒を造り、得られたビ
レットを切削加工した中空ビレットとし、この中空ビレ
ットを熱間押出成形して中空の素管とし、この素管を各
冷間ピルガー式圧延工程前にアルファ相再結晶化焼鈍工
程を備えた多数の冷間ピルガ−式圧延工程を通すことに
よって、実質上仕上がり寸法の被覆管に寸法を縮小させ
ることからなる製造工程が通常採用されている。こうし
て製造された冷間加工状態の実質上仕上がり寸法の被覆
管は、最終的に焼鈍される。この最終の焼鈍は、応力除
去焼鈍、部分的再結晶焼鈍または完全再結晶焼鈍である
。採用される最終焼鈍方式は、燃料被覆管の機械的性質
に関する設計者の仕様に基づいて選定される。When manufacturing Zircaloy cladding, an ingot is hot-processed to create a billet or bar of intermediate dimensions, the resulting billet is machined into a hollow billet, and this hollow billet is hot-extruded to form a hollow tube. , the raw tube is reduced in size to a cladding tube of substantially finished size by passing this blank tube through a number of cold pilger rolling steps with an alpha phase recrystallization annealing step before each cold pilger rolling step. A manufacturing process consisting of: The thus produced cold-worked cladding tube of substantially finished size is finally annealed. This final annealing is a stress relief annealing, a partial recrystallization annealing, or a complete recrystallization annealing. The final annealing method employed is selected based on the designer's specifications regarding the mechanical properties of the fuel cladding.
上記の被覆管を使用した燃料棒には、原子炉使用時に生
じる熱膨張した燃料ペレットと被覆管内面が強く接触し
て、被覆管の内側にき裂が生じる問題がある。以後、こ
のような作用で被覆管に生じるき裂をＰＣＩ（ペレット
・クラッド・インタラクションの略）き裂と呼ぶことに
する。これらのき裂は、時折被覆管壁を貫通して広がり
、燃料棒の一体性を破壊し、燃料内部から放射性核***
生成物が冷却水中に放出され、−次冷却水を汚染する。Fuel rods using the above-mentioned cladding tubes have the problem that the inner surface of the cladding tube comes into strong contact with the thermally expanded fuel pellets generated during use in a nuclear reactor, causing cracks to occur inside the cladding tube. Hereinafter, cracks that occur in the cladding due to such an action will be referred to as PCI (abbreviation for pellet clad interaction) cracks. These cracks occasionally propagate through the cladding walls, destroying the integrity of the fuel rods and releasing radioactive fission products from within the fuel into the cooling water, contaminating the secondary cooling water.
さらには被覆管壁を貫通したき裂を通って一次冷却水が
高温の燃料棒内部に侵入して被覆管を内側から腐食し、
貫通き裂を広げることがある。Furthermore, primary cooling water enters the high temperature fuel rods through cracks that penetrate the cladding wall, corroding the cladding from the inside.
May widen through cracks.
この場合には、−次冷却水中に放出される放出性核***
生成物の量が増し、−次冷却水の汚染度が増大すること
になる。ＰＣＩき裂現象は、被覆管の高速中性子照射に
よる脆化または硬化１機械的応力および腐食核***生成
物の相互作用によりジルコニウム合金中にき裂の発生お
よび伝播を誘発する環境を生ずることによって起こると
一般に信じられている。In this case, the amount of releasable fission products released into the secondary cooling water increases, and the degree of contamination of the secondary cooling water increases. The PCI cracking phenomenon occurs when the cladding is embrittled or hardened by fast neutron irradiation, and the interaction of mechanical stress and corrosion fission products creates an environment that induces crack initiation and propagation in the zirconium alloy. Generally believed.
被覆管の内側に実質上不純物を含む純ジルコニウム層を
結合させたジルカロイ被覆管は、水冷却型原子炉の運転
中に生じる燃料ペレットと被覆管の相互作用により発生
するＰＣＩき裂に対して抵抗性を発揮すると提晶されて
いる。これらの提晶に基づく考案としては特開昭５１−
６９７９５号公報。Zircaloy cladding, which has a substantially impure layer of pure zirconium bonded to the inside of the cladding, resists PCI cracking caused by interaction between fuel pellets and the cladding during operation of water-cooled reactors. It is said to be crystallized when it shows its sexuality. A device based on these crystals is JP-A-51-
Publication No. 69795.
特開昭５４−５９６００号公報、特公昭５５−３３０３
７号公報がある。Japanese Patent Publication No. 54-59600, Japanese Patent Publication No. 55-3303
There is Publication No. 7.
上記従来技術の内側に純ジルコニウム層を結合させたジ
ルカロイ被覆管は、ジルコニウムの内張りがＰＣＩき裂
を防止するために選択されたもので、高温水または高温
の水蒸気に耐する抵抗性ならびに水素吸収に対する抵抗
性が考慮されていない。したがって、もしジルコニウム
を内張りした被覆管に原子炉内で管壁貫通りラックが生
じて一次冷却水が被覆管の内側に侵入すると、内張りの
耐
高温水あるいは高温の水蒸気に対する低食性が被覆管母
材となる外側管状部材を構成する高ジルコニウム基合金
の耐食性より著しく劣ったものとなる。このような条件
下で内張りは完全に酸化されて比較的早く実質−ト役に
立たなくなり、被覆管内側に侵入した一次冷却水と高温
のペレットが反応して生じる水素あるいは内張りの酸化
にともなって発生する水素が内張りを通過してジルコニ
ウム合金からなる被覆管母材に侵入し、水素化物の生成
が増大し、これによって脆くなり、ジルコニウム合金の
一体構造を損傷する。被覆管のこのような品質低下は、
ウランおよび放射性核***生成物を一次冷却水中にかな
り多量に放出することになり、核燃料要素としての全体
の機能を損なうことになる。Zircaloy cladding with a pure zirconium layer bonded to the inside of the prior art is selected because the zirconium lining prevents PCI cracking, has high resistance to high temperature water or high temperature steam, and high hydrogen absorption. Resistance to is not taken into account. Therefore, if a rack penetrates the tube wall of a zirconium-lined cladding tube in a nuclear reactor and primary cooling water enters the inside of the cladding tube, the lining's resistance to high-temperature water or high-temperature water vapor corrosion resistance will decrease. The corrosion resistance is significantly inferior to that of the high zirconium-based alloy that makes up the outer tubular member. Under these conditions, the lining becomes completely oxidized and becomes virtually useless relatively quickly, and hydrogen is generated as a result of the reaction between the primary cooling water that has entered the inside of the cladding tube and the hot pellets, or as the lining oxidizes. Hydrogen penetrates through the lining and into the zirconium alloy cladding matrix, increasing hydride formation, which embrittles and damages the zirconium alloy integral structure. This deterioration in the quality of the cladding is caused by
Significant amounts of uranium and radioactive fission products would be released into the primary cooling water, compromising the overall functionality of the nuclear fuel element.
本発明の目的は、腐食性核***生成物が関与する燃料ペ
レットと被覆管との相互作用によって生じるｃｐｒｏ裂
の伝播を抑制し、かつ、何らかの原因で生じた被覆管の
管壁貫通き裂を通って被覆管内側に侵入した一次冷却水
の反応によって生じる水素の被覆管中への取り込みを低
減することができる核燃料用複合被覆管を提供すること
にある。The purpose of the present invention is to suppress the propagation of CPRO cracks caused by the interaction between fuel pellets and the cladding tube involving corrosive fission products, and to prevent the propagation of CPRO cracks that occur through the wall of the cladding tube due to any cause. An object of the present invention is to provide a composite cladding tube for nuclear fuel that can reduce the incorporation of hydrogen into the cladding tube caused by the reaction of primary cooling water that has entered the inside of the cladding tube.
上記目的は、被覆管が外側管状部材とこの外側管状部材
の内側に内張りした内側管状部材からなり、この内側管
状部材の外側周縁表面は全面に亘って上記外側管状部材
の内側周縁表面に内張すしてあり、かつ、上記外側管状
部材がジルカロイ−２、ジルカロイ−４，ニオブを１．
０〜３．０重量％含むジルコニウム−ニオブ合金および
ジルカロイ−２かジルカロイ−４に含まれる成分のほか
に０．１〜１重量％を含むジルコニウム基合金から選択
された第１ジルコニウム基合金からなり、上記内側管状
部材がアンチモンを０．１〜０．６重量％含み、かつ、
鉄、クロム、ニッケル等の金属不純物の総量が１１００
０ｐｐ以下で、酸素を１１０００ｐｐ以下含み、応力除
去焼鈍あるいは結晶焼鈍された第２ジルコニウム基合金
からなり、その厚さが少なくとも５０μｍあるものとし
て達成するようにした。The above object is such that the cladding tube consists of an outer tubular member and an inner tubular member lined inside the outer tubular member, and the outer circumferential surface of the inner tubular member is entirely lined with the inner circumferential surface of the outer tubular member. and the outer tubular member contains 1.2% Zircaloy, 4% Zircaloy, and 1.0% niobium.
Consists of a first zirconium-based alloy selected from a zirconium-niobium alloy containing 0 to 3.0% by weight and a zirconium-based alloy containing 0.1 to 1% by weight in addition to the components contained in Zircaloy-2 or Zircaloy-4. , the inner tubular member contains 0.1 to 0.6% by weight of antimony, and
The total amount of metal impurities such as iron, chromium, nickel, etc. is 1100
The second zirconium-based alloy is made of a stress-relief annealed or crystal annealed second zirconium-based alloy containing 0 pp or less, 11,000 pp or less of oxygen, and has a thickness of at least 50 μm.
純ジルコニウムに合金成分としてアンチモンを０．１〜
０．６重量％含有させると、水素取り込みに対する抵抗
性を純ジルコニウムよりも十分に増加することができ、
また、アンチモンを０．１〜０．６重量％含むジルコニ
ウム基合金は、被覆管母材を構成する市販のジルコニウ
ム基合金に比べてＰＣＩき裂の拡大に対して充分、がっ
、有効な ゛抵抗作用を発揮する。Pure zirconium with antimony as an alloying component of 0.1~
When the content is 0.6% by weight, the resistance to hydrogen uptake can be sufficiently increased compared to pure zirconium,
In addition, zirconium-based alloys containing 0.1 to 0.6% by weight of antimony are more effective against PCI crack expansion than commercially available zirconium-based alloys that constitute the cladding tube base material. exerts a resistance effect.
さらに、ジルカロイ−２あるいはジルカロイ−４から鉄
、クロム、ニッケルを排除すると、ＰＣＩき裂の発生と
伝播を抑制することができる。その理由は、鉄、クロム
、ニッケルがジルカロイ中でＺｒ （Ｆａ、Ｃｒ、Ｎｉ
）析出物として結晶粒内および結晶粒界に析出しやすく
、これらの析出物は応力集中源となり、ＰＣＩき裂の発
生と伝播を容易にする作用があることによる。なお、ス
ズおよび酸素はジルコニウム中に固溶していると強度を
付加するが、ＰＣＩき裂は延性が増すほど発生しやすく
なるので、その発生と伝播を抑制することから、これら
は少ないほどよい。Furthermore, by excluding iron, chromium, and nickel from Zircaloy-2 or Zircaloy-4, the occurrence and propagation of PCI cracks can be suppressed. The reason is that iron, chromium, and nickel are contained in Zircaloy.
) This is because precipitates tend to precipitate within grains and at grain boundaries, and these precipitates act as a source of stress concentration, facilitating the generation and propagation of PCI cracks. Incidentally, tin and oxygen add strength when dissolved in zirconium, but since PCI cracks are more likely to occur as the ductility increases, the less they are present, the better in order to suppress their occurrence and propagation. .
以下本発明を第１図〜第３図を用いて詳細に説明する。 The present invention will be explained in detail below using FIGS. 1 to 3.
以下に説明する試験に先立ち、次に示す試験片を製作し
た。厚さ１■の市販のジルカロイ−２およびジルカロイ
−４の板を試験片に切断加工した後、表面を研磨して鏡
面とし、高真空中で５７７℃で２時間の再結晶焼鈍を施
した。また、クリスタルバー・ジルコニウムを高真空中
で電子ビームによりボタン溶解した後、厚さ１．２閣に
冷間圧延し、両面を研削して表面の欠陥や荒れを除去し
て厚さｌ閣の板とし、これを試験片に切断加工した後、
再び両面を鏡面仕上げして上記ジルカロイ試験片と同じ
条件で熱処理を施した。さらに、上記と同一のクリスタ
ルバー・ジルコニウムを純度９９．９９５％のアンチモ
ンと一緒に高真空中で電子ビームによりボタン溶解し、
最終的にジルコニウム−０，１重量％アンチモン合金、
ジルコニウム−０，５重量％アンチモン合金、ジルコニ
ウム−１，０重量％アンチモン合金、ジルコニウム−１
，５重量％アンチモン合金を製作した。これらのジルコ
ニウム基合金から上記クリスタルバー・ジルコニウムの
場合と同じ工程で試験片を製作した。クリスタルバー・
ジルコニウムおよび各種ジルコニウム−アンチモン合金
のボタン溶解後に分析した主な化学成分を第１表に示し
た。Prior to the tests described below, the following test pieces were manufactured. Commercially available Zircaloy-2 and Zircaloy-4 plates with a thickness of 1 inch were cut into test pieces, the surfaces of which were polished to a mirror finish and recrystallized annealed at 577°C for 2 hours in a high vacuum. In addition, crystal bar zirconium is button-melted using an electron beam in a high vacuum, then cold-rolled to a thickness of 1.2 mm, and both sides are ground to remove surface defects and roughness. After cutting this into a test piece,
Both sides were mirror-finished again and heat treated under the same conditions as the Zircaloy test piece. Furthermore, the same crystal bar zirconium as above was button melted together with antimony of 99.995% purity using an electron beam in a high vacuum.
Finally, zirconium-0.1% by weight antimony alloy,
Zirconium-0,5% by weight antimony alloy, zirconium-1,0% by weight antimony alloy, zirconium-1
, 5 wt% antimony alloy was fabricated. Test pieces were manufactured from these zirconium-based alloys using the same process as for the crystal bar zirconium described above. crystal bar・
Table 1 shows the main chemical components analyzed after button melting of zirconium and various zirconium-antimony alloys.
縦２０■、横２０閣の上記ジルカロイ、クリスタルバー
・ジルコニウム、ジルコニウム−アンチモン合金からな
る各試験片を４００℃、 １．５ｋｇ／ｍ”の水蒸気中
に５０日間保持し、その後各試験片表面を観察すると、
クリスタルバー・ジルコニウム試験片の全面が白色の粉
末状酸化物に覆われていたが、他のジルコニウム基合金
の試験片の酎
表面は黒色のち密な低置酸化膜で覆われていた。Each test piece made of Zircaloy, crystal bar zirconium, and zirconium-antimony alloy with a length of 20 cm and a width of 20 cm was held in water vapor of 1.5 kg/m'' at 400°C for 50 days, and then the surface of each test piece was When you observe,
The entire surface of the crystal bar zirconium specimen was covered with white powdery oxide, but the surface of the other zirconium-based alloy specimens was covered with a black, dense, low-lying oxide film.
各試験片の表面に形成された酸化膜をエメリー紙で研削
して除去した後、通常使用されている化学的方法によっ
て水素含有量を測定した。After removing the oxide film formed on the surface of each test piece by grinding with emery paper, the hydrogen content was measured by a commonly used chemical method.
各試験片の水素含有量ならびにこれとアンチモン含有量
との関係を第２図に示す。ジルカロイ−２およびジルカ
ロイ−４の水素含有量はほぼ１５０±１．０ＰＰ１１で
あった。クリスタルバー・ジルコニウムの水素含有量は
上記ジルカロイより著しく多い６５０±３０ＰｐＨであ
った。一方、ジルコニウム−アンチモン合金の水素含有
量は、上記ジルカロイのそれより少なく、かつ、アンチ
モン含有量が０．１〜０．５重量％において最大値を示
した。Figure 2 shows the hydrogen content of each test piece and the relationship between this and the antimony content. The hydrogen content of Zircaloy-2 and Zircaloy-4 was approximately 150±1.0 PP11. The hydrogen content of crystal bar zirconium was 650±30 PpH, which is significantly higher than the above Zircaloy. On the other hand, the hydrogen content of the zirconium-antimony alloy was lower than that of the above-mentioned zircaloy, and showed a maximum value when the antimony content was 0.1 to 0.5% by weight.
長さ３５■２幅５■の板の中央部を長さ１０ｎｍに亘り
幅５■、厚さ０．５−に加工した上記ジルカロイ、クリ
スタルバー・ジルコニウム、ジルコニウム−アンチモン
合金からなる試験片について。Regarding test pieces made of the above-mentioned Zircaloy, crystal bar zirconium, and zirconium-antimony alloy, in which the central part of a plate with a length of 35 cm and a width of 5 cm was processed to have a length of 10 nm, a width of 5 cm, and a thickness of 0.5 cm.
ＰＣＩき裂に関与していると信じられているヨウ素が蒸
気圧４０ｔｏｒｒで毎分１００ｃｃの流速で流れる石英
ガラス容器において３５０℃で引張り試験を実施し、試
験片が破断したときの中央減肉部の伸びを測定した。各
試験片の破断伸びを比較したものが第３図である。第３
図においては、ジルカロイ−２の破断伸びの平均値を１
とする相対的な破断伸びを示しである。破断伸びは、ク
リスタルバー・ジルコニウムが最も大きく、アンチモン
含有率の増加にともない減少し、特にアンチモン含有率
が０．５ 重量％を越えると比較的速く低下し、ジルカ
ロイの値に近づく傾向が認められる。破断伸びが大きい
ほどＰＣＩき裂の伝播に対する抵抗性が大きいことを意
味することから、実用的にはジルコニウム−０，１〜０
．５重量％アンチモン合金はクリスタルバー・ジルコニ
ウムにほぼ匹敵するＰＣＩき裂の伝播に対する抵抗性を
有するといえる。A tensile test was conducted at 350°C in a quartz glass container in which iodine, which is believed to be involved in PCI cracking, flows at a flow rate of 100 cc per minute at a vapor pressure of 40 torr. The elongation was measured. FIG. 3 shows a comparison of the elongation at break of each test piece. Third
In the figure, the average value of the elongation at break of Zircaloy-2 is 1
The relative elongation at break is shown below. The elongation at break is the largest for crystal bar zirconium, and decreases as the antimony content increases.Especially when the antimony content exceeds 0.5% by weight, it decreases relatively quickly and tends to approach the value of zircaloy. . Since the larger the elongation at break, the greater the resistance to PCI crack propagation, practical use of zirconium-0,1 to 0
．． It can be said that the 5 wt% antimony alloy has a resistance to PCI crack propagation that is nearly comparable to crystal bar zirconium.
−１−記２種類の試験から水素取り込みに対する抵抗性
およびＰＣＩき裂拡大に対する抵抗性の両面に対して総
合的に判断すると、ジルコニウムに０．１〜０．６重量
％のアンチモンを含む合金が好適であるといえる。これ
によりジルコニウム−０，１〜０．６重量％アンチモン
合金をジルカロイ−２，ジルカロイ−４あるいはジルコ
ニウム−１〜３重量％ニオブ合金からなる被覆管母材の
内側に結合させた複合被覆管とすることにより、ＰＩＣ
き裂拡大に対して十分な抵抗性を発揮し、かつ、何らか
の理由で被覆管に生じた貫通き裂を通って被覆管の内側
に侵入した一次冷却水が被覆管内部で高温のペレットな
どと反応して生じる水素を被覆管中に取り込むことに対
する抵抗性を発揮することができる。-1- Judging from the two types of tests described above in terms of both resistance to hydrogen uptake and resistance to PCI crack propagation, the alloy containing 0.1 to 0.6% by weight of antimony in zirconium It can be said that it is suitable. As a result, a composite cladding tube is obtained in which a zirconium-0.1 to 0.6% by weight antimony alloy is bonded to the inside of a cladding tube base material made of Zircaloy-2, Zircaloy-4, or zirconium-1 to 3% by weight niobium alloy. By this, PIC
It exhibits sufficient resistance to crack propagation, and primary cooling water that enters the inside of the cladding tube through a through crack that occurs in the cladding tube for some reason becomes hot pellets etc. inside the cladding tube. It can exhibit resistance to the hydrogen produced by the reaction being taken into the cladding tube.
本発明をよく明確に理解するために、第１図に示す実施
例を用いて本発明を説明する。第１図は本発明の核燃料
用複合被覆管の一実施例を示す横断面図で、第１図にお
いて、被覆管は、外側層、金からなり、原子炉−次冷却
水による腐食に対して優れた抵抗性ならびに高強度と低
クリープ速度を有している。この第１ジルコニウム基合
金は、ジルカロイ−２合金、ジルカロイー４合金または
ジルコニウムとニオブとの合金、例えば、ジルコニウム
−２，５重量％ニオブ合金であるのが好ましい、内側層
２の第２ジルコニウム基合金は、ペレットと被覆管の相
互作用により生ずるき裂、すなわち、ＰＣＩき裂の伝播
に対する抵抗性と水素取り込みに対する抵抗性とが増大
した金属であり、本発明の発明者が設計した合金である
。In order to have a clear understanding of the invention, the invention will be explained using the embodiment shown in FIG. FIG. 1 is a cross-sectional view showing an embodiment of the composite cladding tube for nuclear fuel of the present invention. In FIG. It has excellent resistance as well as high strength and low creep rate. The second zirconium-based alloy of the inner layer 2 is preferably a zirconium-2 alloy, a zircaloy-4 alloy or an alloy of zirconium and niobium, such as a zirconium-2.5 wt% niobium alloy. is a metal with increased resistance to propagation of cracks caused by pellet-cladding interaction, ie, PCI crack propagation, and resistance to hydrogen uptake, and is an alloy designed by the inventors of the present invention.
第２ジルコニウム基合金の組成範囲は、第２表に示す通
りであり、下記理由により決定しである。The composition range of the second zirconium-based alloy is as shown in Table 2, and was determined for the following reasons.
すなわち、鉄およびクロムとともにスズが存在し。That is, tin is present along with iron and chromium.
かつ、適宜ニッケルを添付すると、ジルコニウムの高温
水に対する腐食抵抗性が増大する。また、スズおよび酸
素はジルコニウムに固溶して強度を第 ２
表
高めるのに有効である。鉄、クロムおよびニッケルはＺ
ｒη （Ｆθ、Ｏｒ、Ｎｉ）析出物を形成することによ
って付加的な強度付与効果を呈する。上記の諸元素は、
ジルコニウムのクリープ速度を低下させ、ジルコニウム
の原子炉運転温度における中性子照射による照射クリー
プを相殺し、また、照射硬化を低減せしめる。しかしな
がら１強度が増し、クリープ速度が低下する反面延性が
低下し、かつ、上記析出物が応力集中の原因となってＰ
ＣＩき裂の拡大を容易にする。要約すると、スズ、鉄。In addition, when nickel is added as appropriate, the corrosion resistance of zirconium to high temperature water increases. In addition, tin and oxygen are dissolved in zirconium to improve its strength.
Effective for increasing the table. Z for iron, chromium and nickel
The formation of rη (Fθ, Or, Ni) precipitates provides an additional strength-imparting effect. The above elements are
It reduces the creep rate of zirconium, offsets irradiation creep due to neutron irradiation of zirconium at reactor operating temperatures, and reduces irradiation hardening. However, while the strength increases and the creep rate decreases, the ductility decreases, and the precipitates cause stress concentration, causing P
Facilitates CI crack expansion. In summary: tin, iron.
クロムおよびニッケルはジルコニウムの高温水に対する
腐食抵抗性を改善する傾向があるが、ＰＣＩき裂の拡大
を阻止する能力には有害である。Although chromium and nickel tend to improve zirconium's corrosion resistance to hot water, they are detrimental to its ability to inhibit PCI crack propagation.
純ジルコニウムには、実質的に鉄、クロム、ニッケルな
どが含まれるが、これらの総量を１１０００ｐｐ以下、
酸素を５００ｐｐｉｌ以下（１０００ｐｐｍ以下であっ
てもよい場合あり）、窒素を４０ＰＰ■以下におさえれ
ば、ＰＣＩき裂の拡大に対する抵抗性が十分に発揮され
る。しかしながら、純ジルコニウムは高温水に対する腐
食抵抗性と水素取り込みに対する抵抗性が劣る。上記不
純物を含む純ジルコニウムに０．１〜０．６重量％のア
ンチモンを添加した合金は、ＰＣＩき裂の拡大に対する
抵抗性と水素取り込みに対する抵抗性の両面において優
れた能力を発揮することができる。不純物の総量は１５
００ｐｐｍ以下が好ましいが、最も好ましくは、１１０
００ｐｐ以下に保つのがよい。Pure zirconium substantially contains iron, chromium, nickel, etc., but the total amount of these is 11,000 pp or less,
If oxygen is kept to 500 ppil or less (1000 ppm or less may be acceptable in some cases) and nitrogen is kept to 40 PP■ or less, sufficient resistance to PCI crack expansion can be achieved. However, pure zirconium has poor corrosion resistance to high temperature water and poor resistance to hydrogen uptake. An alloy made by adding 0.1 to 0.6% by weight of antimony to pure zirconium containing the above impurities can exhibit excellent ability in terms of both resistance to PCI crack expansion and resistance to hydrogen uptake. . The total amount of impurities is 15
00 ppm or less is preferred, most preferably 110 ppm or less.
It is best to keep it below 00pp.
ここに記載した被覆管の化学組成に対する条件は、全合
金元素および不純物についてインゴット製造段階での化
学分析を行い、次いで、例えば、同時押し出し段階近く
のような被覆管製造の中間段階において間隙に入る元素
である酸素、窒素および水素について分析を行うことに
よって満足させることができる。最終寸法の被覆管の化
学分析は必要でない。The conditions for the chemical composition of the cladding described here are such that all alloying elements and impurities are chemically analyzed at the ingot manufacturing stage and then entered into the interstitial space at an intermediate stage of cladding manufacturing, e.g. near the co-extrusion stage. This can be achieved by analyzing the elements oxygen, nitrogen and hydrogen. Chemical analysis of the final size cladding is not required.
内側層用の原料となる市販のジルコニウムとアンチモン
を必要な合金生成条件下で、例えば、アーク溶解して第
２表に記載の公称組成をもつ合金を造る。こうして製造
したインゴットを慣用の穴鼠きビレット製造方法によっ
て内側層２用の穴あきビレットを製造する。外側層１用
の穴あきビレットも慣用の製造方法によって製造する。Commercially available zirconium and antimony raw materials for the inner layer are melted under the necessary alloying conditions, for example by arc melting, to produce an alloy having the nominal composition listed in Table 2. A perforated billet for the inner layer 2 is manufactured from the thus manufactured ingot by a conventional perforated billet manufacturing method. The perforated billet for the outer layer 1 is also produced by conventional manufacturing methods.
これらの内側層２用および外側層１用穴あきビレットは
。These perforated billets for inner layer 2 and outer layer 1 are.
冷間加工、熱間加工、アルファ焼鈍組織またはベータ焼
入れ組織を有する。It has cold worked, hot worked, alpha annealed structure or beta quenched structure.
外側層１用穴あきビレットの内壁ならびに内側層２用穴
あきビレットの外壁は、次いでこれらの穴あきビレット
を嵌合したときにそれらの穴あきビレット間の隙間が最
小となる寸法に機械加工する１機械加工後、各穴あきビ
レットを清浄し、結合しようとする表面からできるだけ
全部の表面の汚れを除去する。清浄化後、穴あきビレッ
トの結合しようとする表面をこの穴あきビレットが１つ
に溶接されるまで清浄に保つのが好ましい。次いで各穴
あきビレットを嵌合し、隣接する各部材の間隙に形成さ
れる環状空間が両穴あきビレット両端を真空電子ビーム
溶接によって溶接した後も真空に保たれるように封止す
る。この段階で結合していない環状合わせビレットは、
既知の押し出し法、冷間ピルガ−式圧延工程および焼鈍
工程により完全に複合被覆管に加工できる。仕上がり寸
法の複合被覆管において、内側層２の厚さは、少なくと
も５０μｍが好ましく、さらには５０〜１００μｍの範
囲が好ましい。The inner wall of the perforated billet for outer layer 1 and the outer wall of the perforated billet for inner layer 2 are then machined to dimensions that minimize the gap between the perforated billets when they are mated. 1 After machining, each perforated billet is cleaned to remove as much surface dirt as possible from the surfaces to be joined. After cleaning, the surfaces of the perforated billets to be joined are preferably kept clean until the perforated billets are welded together. Next, the perforated billets are fitted together and sealed so that the annular space formed in the gap between adjacent members is kept in a vacuum even after both ends of the perforated billets are welded by vacuum electron beam welding. The annular bonded billet that is not bonded at this stage is
It can be completely processed into composite cladding by known extrusion, cold pilgering and annealing processes. In the finished size composite cladding tube, the thickness of the inner layer 2 is preferably at least 50 μm, more preferably in the range of 50 to 100 μm.
仕上がり寸法に加工された複合被覆管は、最終的に少な
くとも内側層２および外側層１が完全に応力緩和するま
で焼鈍する必要があり、再結晶化するまで焼鈍してもよ
い。最終熱処理後、慣用の被覆管の清浄、直線化処理、
最終寸法調整および仕上げ工程を実施する。The composite cladding tube processed to the finished size must finally be annealed until at least the inner layer 2 and the outer layer 1 are completely stress-relaxed, and may be annealed until recrystallized. After final heat treatment, conventional cladding cleaning, straightening treatment,
Perform final dimensional adjustments and finishing processes.
以ト説明した本発明によれば、燃料ペレットと被覆管の
相互作用に基づ（ＰＣＩき裂の拡大を抑制することがで
き、がっ、何らかの原因で被覆管に生じた貫通き裂を通
って被覆管の内側に侵入した一次冷却水が燃料棒で生じ
る反応によって発生する水素を被覆管中に取り込み脆化
して燃料棒破損を拡大することを低減できるという効果
がある。According to the present invention described above, it is possible to suppress the expansion of PCI cracks based on the interaction between the fuel pellets and the cladding tube, and it is possible to suppress the expansion of PCI cracks through the penetration cracks that have occurred in the cladding tube for some reason. This has the effect of reducing the possibility that the primary cooling water that has entered the inside of the cladding tube will absorb hydrogen generated by reactions occurring in the fuel rods into the cladding tubes, causing embrittlement and increasing damage to the fuel rods.
第１図は本発明の核燃料用複合被覆管の一実施例を示す
横断面図、第２図は試験後に各試験片に取り込まれた水
素含有量とアンチモン含有率の関係線図、第３図は各種
試験片のヨウ素中での破断歪とアンチモン含有量との関
係線図である。
１・・・外側層、２・・・内側層。
代理人 弁理士 小川轡′男・ン：
又（
￥３図
アンチモン含有率（ｉｔ、ｘ）Fig. 1 is a cross-sectional view showing one embodiment of the composite cladding tube for nuclear fuel of the present invention, Fig. 2 is a relationship diagram between the hydrogen content and antimony content taken into each test piece after the test, and Fig. 3 is a relationship diagram between the breaking strain in iodine and the antimony content of various test pieces. 1...outer layer, 2...inner layer. Agent: Patent attorney Tsuyoshi Ogawa/N: Also ( ¥3 figure antimony content (it, x)
Claims (1)
張りした内側管状部材からなり、該内側管状部材の外側
周縁表面は全面に亘つて前記外側管状部材の内側周縁表
面に内張りしてあり、かつ、前記外側管状部材がジルカ
ロイ−２、ジルカロイ−４、ニオブを１．０〜３．０重
量％含むジルコニウム−ニオブ合金およびジルカロイ−
２かジルカロイ−４に含まれる成分のほかに０．１〜１
重量％のニオブを含むジルコニウム基合金から選択され
た第１ジルコニウム基合金からなり、前記内側管状部材
がアンチモンを０．１〜０．６重量％含み、かつ、鉄、
クロム、ニツケル等の金属不純物の総量が１０００ｐｐ
ｍ以下で、酸素を１０００ｐｐｍ以下含み、応力除去焼
鈍あるいは再結晶焼鈍された第２ジルコニウム基合金か
らなり、その厚さが少なくとも５０μｍあることを特徴
とする核燃料用複合被覆管。1. The cladding tube consists of an outer tubular member and an inner tubular member lined inside the outer tubular member, and the outer circumferential surface of the inner tubular member is entirely lined with the inner circumferential surface of the outer tubular member. , and the outer tubular member is a zirconium-niobium alloy containing 1.0 to 3.0% by weight of Zircaloy-2, Zircaloy-4, and niobium, and Zircaloy-2.
In addition to the ingredients contained in 2 or Zircaloy-4, 0.1 to 1
a first zirconium-based alloy selected from zirconium-based alloys containing niobium by weight %, said inner tubular member containing 0.1 to 0.6 weight % antimony;
Total amount of metal impurities such as chromium and nickel is 1000pp
A composite cladding tube for nuclear fuel, characterized in that it is made of a second zirconium-based alloy which contains 1,000 ppm or less of oxygen, is subjected to stress relief annealing or recrystallization annealing, and has a thickness of at least 50 μm.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61198884A JPS6355494A (en) | 1986-08-27 | 1986-08-27 | Composite coated tube for nuclear fuel |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61198884A JPS6355494A (en) | 1986-08-27 | 1986-08-27 | Composite coated tube for nuclear fuel |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPS6355494A true JPS6355494A (en) | 1988-03-09 |
Family
ID=16398530
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP61198884A Pending JPS6355494A (en) | 1986-08-27 | 1986-08-27 | Composite coated tube for nuclear fuel |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS6355494A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US6419772B1 (en) | 1998-02-10 | 2002-07-16 | Otsuka Chemical Co., Ltd. | Method for attaching radio wave absorber and structure for attaching the same |
-
1986
- 1986-08-27 JP JP61198884A patent/JPS6355494A/en active Pending
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US6419772B1 (en) | 1998-02-10 | 2002-07-16 | Otsuka Chemical Co., Ltd. | Method for attaching radio wave absorber and structure for attaching the same |
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