JPH0137463B2 - - Google Patents
Info
- Publication number
- JPH0137463B2 JPH0137463B2 JP57068669A JP6866982A JPH0137463B2 JP H0137463 B2 JPH0137463 B2 JP H0137463B2 JP 57068669 A JP57068669 A JP 57068669A JP 6866982 A JP6866982 A JP 6866982A JP H0137463 B2 JPH0137463 B2 JP H0137463B2
- Authority
- JP
- Japan
- Prior art keywords
- nuclear fuel
- cladding tube
- zirconium
- iodine
- fuel element
- 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 42
- 239000003758 nuclear fuel Substances 0.000 claims description 34
- 239000008188 pellet Substances 0.000 claims description 19
- 229910052726 zirconium Inorganic materials 0.000 claims description 14
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 claims description 13
- 229910001093 Zr alloy Inorganic materials 0.000 claims description 12
- OYPRJOBELJOOCE-UHFFFAOYSA-N Calcium Chemical compound [Ca] OYPRJOBELJOOCE-UHFFFAOYSA-N 0.000 claims description 8
- FYYHWMGAXLPEAU-UHFFFAOYSA-N Magnesium Chemical compound [Mg] FYYHWMGAXLPEAU-UHFFFAOYSA-N 0.000 claims description 8
- 229910052791 calcium Inorganic materials 0.000 claims description 8
- 239000011575 calcium Substances 0.000 claims description 8
- 229910052749 magnesium Inorganic materials 0.000 claims description 8
- 239000011777 magnesium Substances 0.000 claims description 8
- 239000012535 impurity Substances 0.000 claims description 6
- 239000000463 material Substances 0.000 claims description 6
- PNDPGZBMCMUPRI-UHFFFAOYSA-N iodine Chemical compound II PNDPGZBMCMUPRI-UHFFFAOYSA-N 0.000 claims description 5
- 150000001875 compounds Chemical class 0.000 claims 1
- 150000004694 iodide salts Chemical class 0.000 claims 1
- 230000035882 stress Effects 0.000 description 14
- ZCYVEMRRCGMTRW-UHFFFAOYSA-N 7553-56-2 Chemical compound [I] ZCYVEMRRCGMTRW-UHFFFAOYSA-N 0.000 description 10
- 239000011630 iodine Substances 0.000 description 10
- 229910052740 iodine Inorganic materials 0.000 description 10
- 238000005336 cracking Methods 0.000 description 9
- 230000007797 corrosion Effects 0.000 description 8
- 238000005260 corrosion Methods 0.000 description 8
- 239000007789 gas Substances 0.000 description 7
- 230000004992 fission Effects 0.000 description 5
- 238000006243 chemical reaction Methods 0.000 description 4
- 239000000047 product Substances 0.000 description 4
- 229910045601 alloy Inorganic materials 0.000 description 3
- 239000000956 alloy Substances 0.000 description 3
- 239000002826 coolant Substances 0.000 description 3
- 238000000034 method Methods 0.000 description 3
- CGWDABYOHPEOAD-VIFPVBQESA-N (2r)-2-[(4-fluorophenoxy)methyl]oxirane Chemical compound C1=CC(F)=CC=C1OC[C@@H]1OC1 CGWDABYOHPEOAD-VIFPVBQESA-N 0.000 description 2
- XEEYBQQBJWHFJM-UHFFFAOYSA-N Iron Chemical compound [Fe] XEEYBQQBJWHFJM-UHFFFAOYSA-N 0.000 description 2
- PXHVJJICTQNCMI-UHFFFAOYSA-N Nickel Chemical compound [Ni] PXHVJJICTQNCMI-UHFFFAOYSA-N 0.000 description 2
- 238000005275 alloying Methods 0.000 description 2
- XAGFODPZIPBFFR-UHFFFAOYSA-N aluminium Chemical compound [Al] XAGFODPZIPBFFR-UHFFFAOYSA-N 0.000 description 2
- 229910052782 aluminium Inorganic materials 0.000 description 2
- TVFDJXOCXUVLDH-UHFFFAOYSA-N caesium atom Chemical compound [Cs] TVFDJXOCXUVLDH-UHFFFAOYSA-N 0.000 description 2
- 239000013078 crystal Substances 0.000 description 2
- 230000007423 decrease Effects 0.000 description 2
- 239000000446 fuel Substances 0.000 description 2
- 230000007246 mechanism Effects 0.000 description 2
- 239000002244 precipitate Substances 0.000 description 2
- VYZAMTAEIAYCRO-UHFFFAOYSA-N Chromium Chemical compound [Cr] VYZAMTAEIAYCRO-UHFFFAOYSA-N 0.000 description 1
- 238000010521 absorption reaction Methods 0.000 description 1
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 description 1
- 230000015572 biosynthetic process Effects 0.000 description 1
- 229910052792 caesium Inorganic materials 0.000 description 1
- 230000008859 change Effects 0.000 description 1
- 229910052804 chromium Inorganic materials 0.000 description 1
- 239000011651 chromium Substances 0.000 description 1
- 238000001816 cooling Methods 0.000 description 1
- 230000006378 damage Effects 0.000 description 1
- 238000010586 diagram Methods 0.000 description 1
- 238000002474 experimental method Methods 0.000 description 1
- 238000010438 heat treatment Methods 0.000 description 1
- XMBWDFGMSWQBCA-UHFFFAOYSA-N hydrogen iodide Chemical compound I XMBWDFGMSWQBCA-UHFFFAOYSA-N 0.000 description 1
- 230000003993 interaction Effects 0.000 description 1
- 150000002497 iodine compounds Chemical class 0.000 description 1
- 229910052742 iron Inorganic materials 0.000 description 1
- 238000004519 manufacturing process Methods 0.000 description 1
- 239000013081 microcrystal Substances 0.000 description 1
- 239000000203 mixture Substances 0.000 description 1
- 238000000465 moulding Methods 0.000 description 1
- 229910052759 nickel Inorganic materials 0.000 description 1
- 229910052760 oxygen Inorganic materials 0.000 description 1
- 239000001301 oxygen Substances 0.000 description 1
- 239000000843 powder Substances 0.000 description 1
- 230000002285 radioactive effect Effects 0.000 description 1
- 230000009257 reactivity Effects 0.000 description 1
- 230000009467 reduction Effects 0.000 description 1
- 238000005245 sintering Methods 0.000 description 1
- 238000001179 sorption measurement Methods 0.000 description 1
- -1 that is Chemical compound 0.000 description 1
- 230000008646 thermal stress Effects 0.000 description 1
- WFKWXMTUELFFGS-UHFFFAOYSA-N tungsten Chemical compound [W] WFKWXMTUELFFGS-UHFFFAOYSA-N 0.000 description 1
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Chemical compound O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 1
- 150000003754 zirconium Chemical class 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
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Preventing Corrosion Or Incrustation Of Metals (AREA)
Description
【発明の詳細な説明】
本発明は核燃料要素に係り、特にジルコニウム
合金系被覆管の応力腐食割れ現象の発生を軽減化
できるようにした核燃料要素に関する。DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a nuclear fuel element, and more particularly to a nuclear fuel element that can reduce the occurrence of stress corrosion cracking in a zirconium alloy cladding tube.
核燃料要素は通常、第1図に示すように、被覆
管1内に複数個の核燃料ペレツト2を積層収納す
るとともに被覆管1の両端開口を端栓3a,3b
で密封したものとなつている。上記核燃料ペレツ
ト2は核***性の酸化物燃料粉末をたとえば長さ
と直径との比が約1の円柱状ペレツトに成型焼結
したものである。なお、第1図中4は被覆管1内
にガス溜め用プレナム5を形成する機能と、核燃
料ペレツト2を安定に支持する機能とを果すスプ
リングを示している。 As shown in FIG. 1, a nuclear fuel element usually stores a plurality of nuclear fuel pellets 2 in a stack in a cladding tube 1, and the openings at both ends of the cladding tube 1 are closed with end plugs 3a, 3b.
It is sealed. The nuclear fuel pellet 2 is formed by molding and sintering fissile oxide fuel powder into a cylindrical pellet having a length to diameter ratio of about 1, for example. In addition, numeral 4 in FIG. 1 indicates a spring which functions to form a gas reservoir plenum 5 within the cladding tube 1 and to stably support the nuclear fuel pellets 2.
ところで、上記のように構成された核燃料要素
において、被覆管1には、核燃料ペレツト2と冷
却材との接触および化学反応を阻止する機能と、
燃料から放出された放射性核***生成物が冷却材
中に混入するのを阻止する機能とが要求される。
したがつて、このような機能を満足しない被覆
管、つまり被覆管が破損したような場合には、冷
却系プラントの放射能レベルが上昇し、安全を確
保するために原子炉の運転を停止させなければな
らない事態となる。 By the way, in the nuclear fuel element configured as described above, the cladding tube 1 has the function of preventing contact and chemical reaction between the nuclear fuel pellets 2 and the coolant, and
A function is required to prevent radioactive fission products released from the fuel from entering the coolant.
Therefore, if the cladding tube does not satisfy these functions, or if the cladding tube is damaged, the radioactivity level in the cooling system plant will increase, and reactor operation will have to be stopped to ensure safety. It becomes a situation where it has to happen.
軽水冷却型原子炉に用いられる核燃料要素の被
覆管は、一般にジルコニウムおよびその合金系で
形成されている。ジルコニウムおよびその合金
は、中性子吸収断面積が小さく、かつ、約400℃
以下の温度で強靭で延性が良く、しかも、冷却材
として用いられる水蒸気と反応しない安定した特
性を有している。 The cladding of nuclear fuel elements used in light water-cooled nuclear reactors is generally made of zirconium and its alloys. Zirconium and its alloys have a small neutron absorption cross section and a temperature of about 400℃.
It is strong and ductile at temperatures below, and has stable properties that do not react with water vapor used as a coolant.
しかしながら、現在までの運転経験によると、
ジルコニウムおよびその合金で形成された被覆管
にあつても、中性子照射を受けることによる材料
強度の低下および核***生成物との化学反応によ
る腐食などの相互作用に基づく脆性割れが発生す
る。 However, based on my driving experience to date,
Even in cladding tubes made of zirconium and zirconium alloys, brittle cracking occurs due to interaction such as reduction in material strength due to neutron irradiation and corrosion due to chemical reaction with nuclear fission products.
このような望ましくない現象は次のようにして
発生するものと考えられる。すなわち、核燃料ペ
レツト2で発生した熱を被覆管1の外表面に効率
よく伝えるには、被覆管1の内側面と核燃料ペレ
ツト2との間に形成されるギヤツプを約10ミクロ
ン以下に設定する必要があり、一方、運転時に
は、核燃料ペレツト2が発熱するのでペレツト自
身が熱応力で割れ、その破面のくい違いや、さら
には燃焼とともに核燃料ペレツト内に核***生成
物が蓄積して起こる体積膨張などが原因して第2
図に示すように被覆管1が核燃料ペレツト2によ
つて押し拡げられ応力を受ける。被覆管1が受け
る歪の周方向の平均値はさほど大きくはないが、
核燃料ペレツト2に生じたクラツク6近傍の壁に
は局部的に歪が集中し、この歪は降伏応力以上に
達する。さらに、核***に伴なつて核燃料ペレツ
ト2からヨウ素およびヨウ素化合物、セシウムお
よびセシウム化合物などの腐食性ガスが発生し、
この腐食性ガスは被覆管1内の自由空間、すなわ
ち、クラツク6などに集まる。つまり、被覆管1
の特に歪が集中している部分近傍に腐食性ガスが
集りやすい。 It is thought that such an undesirable phenomenon occurs as follows. That is, in order to efficiently transfer the heat generated by the nuclear fuel pellet 2 to the outer surface of the cladding tube 1, the gap formed between the inner surface of the cladding tube 1 and the nuclear fuel pellet 2 must be set to about 10 microns or less. On the other hand, during operation, the nuclear fuel pellet 2 generates heat, which causes the pellet itself to crack due to thermal stress, resulting in discrepancies in the fracture surfaces, as well as volumetric expansion caused by fission products accumulating within the nuclear fuel pellet as it burns. The second cause is
As shown in the figure, the cladding tube 1 is expanded by the nuclear fuel pellets 2 and is subjected to stress. Although the average value of the strain that the cladding tube 1 receives in the circumferential direction is not very large,
Strain is locally concentrated on the wall near the crack 6 formed in the nuclear fuel pellet 2, and this strain reaches more than the yield stress. Furthermore, corrosive gases such as iodine and iodine compounds, cesium and cesium compounds are generated from the nuclear fuel pellet 2 as a result of nuclear fission.
This corrosive gas collects in the free space within the cladding tube 1, ie in the cracks 6 and the like. In other words, cladding tube 1
Corrosive gas tends to collect especially near areas where strain is concentrated.
一般に、腐食性ガスの雰囲気中で応力(特に降
伏応力以上)が作用すると、材料の延性が低減
し、応力腐食割れと呼称される脆性破壊現象が発
生する。 Generally, when stress (particularly greater than yield stress) is applied in a corrosive gas atmosphere, the ductility of the material decreases and a brittle fracture phenomenon called stress corrosion cracking occurs.
応力腐食割れは温度、応力、腐食性ガスの濃
度、溶存酸素、合金の組成、熱処理、加工度など
によつても左右されその発生メカニズムは単一で
はなく明らかでない。 Stress corrosion cracking is influenced by temperature, stress, concentration of corrosive gas, dissolved oxygen, alloy composition, heat treatment, degree of processing, etc., and the mechanism by which it occurs is not clear.
本発明の目的は、腐食性ガス中において応力が
加つた場合、ジルコニウム合金系被覆管が応力腐
食割れを起こす確率を大幅に低減でき、破損確率
が低い信頼性に富んだ核燃料要素を提供すること
にある。 An object of the present invention is to provide a highly reliable nuclear fuel element that can significantly reduce the probability that a zirconium alloy cladding tube will undergo stress corrosion cracking when stress is applied in a corrosive gas, and has a low probability of failure. It is in.
上記課題を解決するための本発明に係る核燃料
要素は、ジルコニウム合金系被覆管内に核燃料ペ
レツトが収納され前記被覆管の両端開口を密封し
てなる核燃料要素において、前記被覆管を構成す
る材料として、熱力学的にジルコニウムより安定
なヨウ化物を形成する不純物元素の含有率を、
1ppm以下に低減させたジルカロイを用いるよう
にしたものである。 A nuclear fuel element according to the present invention for solving the above problems is a nuclear fuel element in which nuclear fuel pellets are housed in a zirconium alloy-based cladding tube and both openings of the cladding tube are sealed, and the material constituting the cladding tube includes: The content of impurity elements that form iodide, which is thermodynamically more stable than zirconium, is
This uses Zircaloy reduced to 1 ppm or less.
まず、応力腐食割れの発生メカニズムについて
説明する。被覆管内面の結晶粒が特定の方向を向
いた粒界は、粒界析出物と核***生成物であるヨ
ウ素(I2)との反応が活発であり、いわゆるアク
テブパスとなる。この附近で歪集中を生じると、
表面酸化膜が破壊し、この破壊成分がピツトとな
つて割れの核が発生する。この初期クラツク先端
では、ヨウ素の表面吸着によつて界面エネルギー
が低下し、クラツクが伝播することが推測され
る。 First, the mechanism by which stress corrosion cracking occurs will be explained. Grain boundaries where the crystal grains on the inner surface of the cladding tube are oriented in a specific direction are active in the reaction between grain boundary precipitates and iodine (I 2 ), which is a fission product, resulting in a so-called active path. If strain concentration occurs near this area,
The surface oxide film is destroyed, and the components of this destruction form pits and generate crack nuclei. It is presumed that at the tip of this initial crack, the interfacial energy decreases due to surface adsorption of iodine, causing the crack to propagate.
そこで、何らかの手段でこの結晶粒界析出物と
ヨウ素との反応速度を低下させるか、または、停
止させることができれば、初期クラツクの核が発
生しないことになり、応力腐食割れが発生しなく
なる。 Therefore, if the reaction rate between grain boundary precipitates and iodine can be reduced or stopped by some means, the initial crack nucleus will not be generated, and stress corrosion cracking will no longer occur.
ジルカロイ被覆管の結晶粒界には、鉄、クロ
ム、ニツケルの合金元素のほかにアルミニウム、
マグネシウム、カルシウムなどが析出している。
被覆管内の各元素が350℃で1モルのヨウ素と反
応するときのギブスの生成自由エネルギー変化は
第3図に示してある。第3図からヨウ素とジルコ
ニウムとは反応しやすいが、結晶粒界に析出する
と思われるマグネシウム、カルシウムは、第3図
に31の符号をつけて示してあるように、ジルコ
ニウムよりさらにヨウ素と反応しやすいことがわ
かる。このため、ヨウ素によつて被覆管に粒界割
れが発生し、それが進展してゆくことが考えられ
る。 In addition to the alloying elements of iron, chromium, and nickel, the grain boundaries of Zircaloy-clad tubes contain aluminum,
Magnesium, calcium, etc. are precipitated.
Figure 3 shows the change in the Gibbs free energy of formation when each element in the cladding reacts with 1 mole of iodine at 350°C. As shown in Figure 3, iodine and zirconium react easily, but magnesium and calcium, which are thought to precipitate at grain boundaries, react with iodine even more than zirconium, as indicated by the symbol 31 in Figure 3. It turns out it's easy. For this reason, it is thought that intergranular cracks occur in the cladding tube due to iodine, and these cracks progress.
そこで、本発明においては、ジルコニウムより
ヨウ素と反応しやすい不純物、すなわち、マグネ
シウム、カルシウムの含有率が低いジルコニウム
合金で被覆管を構成するようにし、この粒界割れ
の発生の主要な原因を除去するようにした。 Therefore, in the present invention, the cladding tube is made of a zirconium alloy that has a lower content of impurities that are more likely to react with iodine than zirconium, that is, magnesium and calcium, thereby eliminating the main cause of intergranular cracking. I did it like that.
以下本発明の一実施例について詳細に説明す
る。 An embodiment of the present invention will be described in detail below.
まず、ジルコニウムの通常の製法であるクロー
ル法によつてスポンジジルコニウムを得る。この
ジルコニウム中にはマグネシウムおよびカルシウ
ムがそれぞれ約10ppm含まれているので、次に、
ジルコニウム中のマグネシウムとカルシウムの含
有率を極力低減化するため、スポンジジルコニウ
ムを一度ヨウ化させてヨウ化ジルコニウムとす
る。このヨウ化ジルコニウムを赤熱したタングス
テン線に触れさせて分解し、赤熱線を中心にジル
コニウムの微結晶を析出させる。この方法はヨー
ド法と呼ばれている。このようにして得られたク
リスタルバージルコニウム中のマグネシウムおよ
びカルシウムの含有率は、それぞれ1ppm以下で
あつた。そこで、この母材中に合金元素を添加し
てジルカロイを得て、このジルカロイを用いて通
常の被覆管の製造法にしたがつて外径12.52mm、
板厚0.86mmの被覆管に加工し、それを用いて核燃
料要素を構成した。 First, sponge zirconium is obtained by the Kroll method, which is a common method for producing zirconium. This zirconium contains about 10 ppm each of magnesium and calcium, so next,
In order to reduce the content of magnesium and calcium in zirconium as much as possible, sponge zirconium is once iodized to produce zirconium iodide. This zirconium iodide is brought into contact with a red-hot tungsten wire to decompose it, and zirconium microcrystals are precipitated around the red-hot wire. This method is called the iodine method. The contents of magnesium and calcium in the crystal verzirconium thus obtained were each 1 ppm or less. Therefore, alloying elements were added to this base material to obtain Zircaloy, and using this Zircaloy, an outer diameter of 12.52 mm,
It was processed into a cladding tube with a thickness of 0.86mm, and used to construct a nuclear fuel element.
このようにヨウ素と反応しやすいマグネシウ
ム、カルシウムなどの不純物の含有率を低減させ
た被覆管の特性を調べるために、被覆管内に中空
の核燃料ペレツトを挿入するとともに、核燃料ペ
レツトの中空部に円柱状の純アルミニウム棒を充
填し、ヨウ素濃度3mg/c.c.、被覆管温度350℃の
雰囲気下でアルミニウム棒を長手方向に圧縮し、
中空の核燃料ペレツトを介して被覆管に円周方向
応力を加えた。そしてこのときに被覆管に生じた
破断伸びを求めた。その結果、第4図に棒線42
で示す特性が得られた。一方、比較のために従来
の被覆管を用意し、同様の実験を行つた結果を第
4図に棒線41で示した。これより、ヨウ素と反
応しやすい不純物の含有率を低減させた被覆管
は、従来の被覆管より破断伸びが大きくなること
がわかる。なお、本発明に係る核燃料要素の構成
は第1図と同様であり、ここでは説明を省略す
る。 In order to investigate the characteristics of the cladding tube, which has a reduced content of impurities such as magnesium and calcium that easily react with iodine, a hollow nuclear fuel pellet was inserted into the cladding tube, and a cylindrical shape was inserted into the hollow part of the nuclear fuel pellet. filled with pure aluminum rods, and compressed in the longitudinal direction in an atmosphere with an iodine concentration of 3 mg/cc and a cladding temperature of 350°C.
Circumferential stress was applied to the cladding tube through a hollow nuclear fuel pellet. Then, the elongation at break that occurred in the cladding tube at this time was determined. As a result, the bar line 42 in Fig.
The characteristics shown are obtained. On the other hand, for comparison, a conventional cladding tube was prepared and a similar experiment was conducted, and the results are shown by the bar line 41 in FIG. This shows that the cladding tube with a reduced content of impurities that easily react with iodine has a larger elongation at break than the conventional cladding tube. Note that the configuration of the nuclear fuel element according to the present invention is the same as that shown in FIG. 1, and the description thereof will be omitted here.
上記した本発明に係る核燃料要素によれば、第
4図の結果から明らかなように、被覆管の応力腐
食割れに対する抵抗力が強く、大きな伸びまで許
容し得て、応力腐食割れを起こす確率を大幅に抵
減できる。 According to the nuclear fuel element according to the present invention described above, as is clear from the results shown in FIG. You can reduce the cost significantly.
なお、本発明はジルカロイ−2製の被覆管に限
らず、ジルカロイ−4などの他のジルコニウム合
金系で形成された被覆管にも適用できる。 The present invention is not limited to cladding tubes made of Zircaloy-2, but can also be applied to cladding tubes made of other zirconium alloys such as Zircaloy-4.
以上説明したように、本発明によれば、ジルコ
ニウム合金系被覆管が応力腐食割れを起こす確率
を大幅に低減でき、破損確率が低い信頼性に富ん
だ核燃料要素とすることができるという効果があ
る。 As explained above, according to the present invention, the probability of stress corrosion cracking occurring in the zirconium alloy cladding tube can be significantly reduced, and a highly reliable nuclear fuel element with a low probability of breakage can be obtained. .
図面は本発明を説明するためのもので、第1図
は核燃料要素の縦断面図、第2図は核燃料要素に
起こりやすい問題点を説明するための一部切開し
た斜視図、第3図は被覆管材料内部の元素とヨウ
素との反応性を示す相対図、第4図は本発明で用
いる被覆管と従来の被覆管の円周方向破断伸の比
較を示すグラフである。
1……被覆管、2……核燃料ペレツト、3a,
3b……端栓。
The drawings are for explaining the present invention; FIG. 1 is a longitudinal cross-sectional view of a nuclear fuel element, FIG. 2 is a partially cutaway perspective view for explaining problems that tend to occur in nuclear fuel elements, and FIG. FIG. 4 is a relative diagram showing the reactivity between elements inside the cladding material and iodine, and is a graph showing a comparison of elongation at break in the circumferential direction of the cladding tube used in the present invention and a conventional cladding tube. 1... Cladding tube, 2... Nuclear fuel pellet, 3a,
3b... end plug.
Claims (1)
が収納され前記被覆管の両端開口を密封してなる
核燃料要素において、前記被覆管を構成する材料
として、熱力学的にジルコニウムより安定なヨウ
化物を形成する不純物元素の含有率を、1ppm以
下に低減させたジルカロイを用いてあることを特
徴とする核燃料要素。 2 安定なヨウ化物を形成する不純物元素が、マ
グネシウムおよびカルシウムであることを特徴と
する特許請求の範囲第1項記載の核燃料要素。[Scope of Claims] 1. A nuclear fuel element in which nuclear fuel pellets are housed in a zirconium alloy cladding tube and openings at both ends of the cladding tube are sealed, in which iodine, which is thermodynamically more stable than zirconium, is used as the material constituting the cladding tube. A nuclear fuel element characterized by using Zircaloy in which the content of impurity elements that form compounds is reduced to 1 ppm or less. 2. The nuclear fuel element according to claim 1, wherein the impurity elements that form stable iodides are magnesium and calcium.
Priority Applications (1)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
JP57068669A JPS58185740A (en) | 1982-04-26 | 1982-04-26 | Nuclear fuel element |
Applications Claiming Priority (1)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
JP57068669A JPS58185740A (en) | 1982-04-26 | 1982-04-26 | Nuclear fuel element |
Publications (2)
Publication Number | Publication Date |
---|---|
JPS58185740A JPS58185740A (en) | 1983-10-29 |
JPH0137463B2 true JPH0137463B2 (en) | 1989-08-07 |
Family
ID=13380345
Family Applications (1)
Application Number | Title | Priority Date | Filing Date |
---|---|---|---|
JP57068669A Granted JPS58185740A (en) | 1982-04-26 | 1982-04-26 | Nuclear fuel element |
Country Status (1)
Country | Link |
---|---|
JP (1) | JPS58185740A (en) |
-
1982
- 1982-04-26 JP JP57068669A patent/JPS58185740A/en active Granted
Also Published As
Publication number | Publication date |
---|---|
JPS58185740A (en) | 1983-10-29 |
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