JPH0373832B2 - - Google Patents
Info
- Publication number
- JPH0373832B2 JPH0373832B2 JP57168385A JP16838582A JPH0373832B2 JP H0373832 B2 JPH0373832 B2 JP H0373832B2 JP 57168385 A JP57168385 A JP 57168385A JP 16838582 A JP16838582 A JP 16838582A JP H0373832 B2 JPH0373832 B2 JP H0373832B2
- Authority
- JP
- Japan
- Prior art keywords
- cladding tube
- nuclear fuel
- cladding
- zircaloy
- oxygen
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Expired - Lifetime
Links
- 238000005253 cladding Methods 0.000 claims description 48
- 239000003758 nuclear fuel Substances 0.000 claims description 31
- 229910001093 Zr alloy Inorganic materials 0.000 claims description 27
- 239000008188 pellet Substances 0.000 claims description 25
- 239000001301 oxygen Substances 0.000 claims description 19
- 229910052760 oxygen Inorganic materials 0.000 claims description 19
- 239000000446 fuel Substances 0.000 claims description 16
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 claims description 11
- 239000000956 alloy Substances 0.000 claims description 8
- 229910045601 alloy Inorganic materials 0.000 claims description 4
- 230000035882 stress Effects 0.000 description 20
- 239000000463 material Substances 0.000 description 15
- 230000007797 corrosion Effects 0.000 description 12
- 238000005260 corrosion Methods 0.000 description 12
- 238000005336 cracking Methods 0.000 description 10
- 239000007789 gas Substances 0.000 description 9
- 230000003993 interaction Effects 0.000 description 7
- 230000004888 barrier function Effects 0.000 description 5
- 229910052726 zirconium Inorganic materials 0.000 description 5
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 description 4
- 229910052782 aluminium Inorganic materials 0.000 description 4
- 230000004992 fission Effects 0.000 description 4
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Chemical compound O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 4
- 238000000576 coating method Methods 0.000 description 3
- 239000002826 coolant Substances 0.000 description 3
- 230000003247 decreasing effect Effects 0.000 description 3
- XEEYBQQBJWHFJM-UHFFFAOYSA-N iron Substances [Fe] XEEYBQQBJWHFJM-UHFFFAOYSA-N 0.000 description 3
- PXHVJJICTQNCMI-UHFFFAOYSA-N nickel Substances [Ni] PXHVJJICTQNCMI-UHFFFAOYSA-N 0.000 description 3
- 230000003647 oxidation Effects 0.000 description 3
- 238000007254 oxidation reaction Methods 0.000 description 3
- ZCYVEMRRCGMTRW-UHFFFAOYSA-N 7553-56-2 Chemical compound [I] ZCYVEMRRCGMTRW-UHFFFAOYSA-N 0.000 description 2
- 206010021143 Hypoxia Diseases 0.000 description 2
- 238000010521 absorption reaction Methods 0.000 description 2
- XAGFODPZIPBFFR-UHFFFAOYSA-N aluminium Chemical compound [Al] XAGFODPZIPBFFR-UHFFFAOYSA-N 0.000 description 2
- TVFDJXOCXUVLDH-UHFFFAOYSA-N caesium atom Chemical compound [Cs] TVFDJXOCXUVLDH-UHFFFAOYSA-N 0.000 description 2
- 238000006243 chemical reaction Methods 0.000 description 2
- 229910052804 chromium Inorganic materials 0.000 description 2
- 239000011651 chromium Substances 0.000 description 2
- 229910052802 copper Inorganic materials 0.000 description 2
- 239000010949 copper Substances 0.000 description 2
- 230000006378 damage Effects 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 230000000694 effects Effects 0.000 description 2
- 230000001146 hypoxic effect Effects 0.000 description 2
- 239000011630 iodine Substances 0.000 description 2
- 229910052740 iodine Inorganic materials 0.000 description 2
- 229910052742 iron Inorganic materials 0.000 description 2
- 239000000314 lubricant Substances 0.000 description 2
- 229910052749 magnesium Inorganic materials 0.000 description 2
- 238000000034 method Methods 0.000 description 2
- 229910052750 molybdenum Inorganic materials 0.000 description 2
- 229910052759 nickel Inorganic materials 0.000 description 2
- 229910052758 niobium Inorganic materials 0.000 description 2
- OKTJSMMVPCPJKN-UHFFFAOYSA-N Carbon Chemical compound [C] OKTJSMMVPCPJKN-UHFFFAOYSA-N 0.000 description 1
- VYZAMTAEIAYCRO-UHFFFAOYSA-N Chromium Chemical compound [Cr] VYZAMTAEIAYCRO-UHFFFAOYSA-N 0.000 description 1
- 229910018487 Ni—Cr Inorganic materials 0.000 description 1
- ATJFFYVFTNAWJD-UHFFFAOYSA-N Tin Chemical compound [Sn] ATJFFYVFTNAWJD-UHFFFAOYSA-N 0.000 description 1
- RTAQQCXQSZGOHL-UHFFFAOYSA-N Titanium Chemical compound [Ti] RTAQQCXQSZGOHL-UHFFFAOYSA-N 0.000 description 1
- 238000009825 accumulation Methods 0.000 description 1
- 238000000137 annealing Methods 0.000 description 1
- 229910052790 beryllium Inorganic materials 0.000 description 1
- 229910052792 caesium Inorganic materials 0.000 description 1
- 238000001816 cooling Methods 0.000 description 1
- 230000007423 decrease Effects 0.000 description 1
- 238000002474 experimental method Methods 0.000 description 1
- 229910002804 graphite Inorganic materials 0.000 description 1
- 239000010439 graphite Substances 0.000 description 1
- 238000010438 heat treatment Methods 0.000 description 1
- 150000002497 iodine compounds Chemical class 0.000 description 1
- 239000011777 magnesium Substances 0.000 description 1
- 230000007246 mechanism Effects 0.000 description 1
- 239000012528 membrane Substances 0.000 description 1
- 239000007769 metal material Substances 0.000 description 1
- 239000000203 mixture Substances 0.000 description 1
- CWQXQMHSOZUFJS-UHFFFAOYSA-N molybdenum disulfide Chemical compound S=[Mo]=S CWQXQMHSOZUFJS-UHFFFAOYSA-N 0.000 description 1
- 229910052982 molybdenum disulfide Inorganic materials 0.000 description 1
- 239000010955 niobium Substances 0.000 description 1
- 239000000843 powder Substances 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 230000009467 reduction Effects 0.000 description 1
- 238000005096 rolling process Methods 0.000 description 1
- 239000010935 stainless steel Substances 0.000 description 1
- 229910001220 stainless steel Inorganic materials 0.000 description 1
- 238000005482 strain hardening Methods 0.000 description 1
- 229910052715 tantalum Inorganic materials 0.000 description 1
- 238000012360 testing method Methods 0.000 description 1
- 230000008646 thermal stress Effects 0.000 description 1
- 239000010936 titanium Substances 0.000 description 1
- 229910052719 titanium Inorganic materials 0.000 description 1
- 238000012546 transfer Methods 0.000 description 1
- 229910052721 tungsten Inorganic materials 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
Description
【発明の詳細な説明】
本発明は、核燃料要素の改良に関するものであ
る。DETAILED DESCRIPTION OF THE INVENTION The present invention relates to improvements in nuclear fuel elements.
一般に、核燃料要素は、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の両端開口が端栓３ａ，
３ｄにより密閉されている。上記燃料ペレツト２
は核***性の酸化物燃料粉末を、例えば長さと直
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。尚、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もたせたスプ
リングである。 Generally, in a nuclear fuel element, a plurality of nuclear fuel pellets 2 are stacked and stored in a cladding tube 1 as shown in FIG.
It is sealed by 3d. Above fuel pellet 2
is a fissile oxide fuel powder formed and sintered into a cylindrical pellet having a length to diameter ratio of about 1, for example. Reference numeral 4 in FIG. 1 is a spring having the function of forming a gas reservoir plenum 5 within the cladding tube 1 and the function of stably supporting 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との
間で冷却材が接触すること及び化学反応が生じる
ことを阻止する機能と、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。従つて、このような機能
を満足しない被覆管、即ち、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 a function of preventing the coolant from coming into contact with the nuclear fuel pellets 2 and preventing a chemical reaction from occurring, and a function of preventing the radioactivity released from the fuel. A function is required to prevent nuclear fission products from entering the coolant. Therefore, in the event that the cladding tube does not satisfy these functions, that is, the cladding tube is damaged, the radioactivity level in the cooling system plant will increase, and the reactor operation will have to be stopped to ensure safety. It becomes a situation where it has to happen.
一方、水冷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及びその合金系材
料で形成されている。ジルコニウム及びその合金
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よく、しかも、冷却材
として用いられる水蒸気とも反応しない特性を有
している。 On the other hand, cladding tubes of nuclear fuel elements used in water-cooled nuclear reactors are generally made of zirconium and zirconium alloy materials. 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 also does not react with water vapor, which is used as a coolant.
しかしながら、現在までの運転経験によると、
ジルコニウム及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
度の低下及び核***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してい
る。このような望ましくない現象は次のようにし
て発生するものとして考えられる。即ち、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数十ミク
ロン以下に設定する必要がある。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、その破面の喰い違いや、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***に伴つて核燃料ペレツト
２からよう素及びよう素化合物、セシウム及びセ
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、即ち、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まり易
い。 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. Such an undesirable phenomenon is thought to occur 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, it is necessary to set the gap formed between the inner surface of the cladding tube 1 and the nuclear fuel pellet 2 to several tens of microns or less. There is. 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 and volumetric expansion caused by the accumulation of fission products within the nuclear fuel pellet as it burns. second
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, and these corrosive gases collect in the free space within the cladding tube 1, that is, in the crack 6, etc. . In other words, corrosive gas tends to collect near the portion of the cladding tube 1 where strain is particularly 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。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
は単一でない。 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. 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 unique.
これらの好ましくない破壊を防止する目的で、
従来例として、例えば燃料ペレツト２と被覆管１
との間に潤滑剤を挿入する方法が米国特許
3018238号明細書に示されており、また、被覆管
１と燃料ペレツト２間に障壁を設けた例として
DAS1238115ではチタン層を述べている。さら
に、Nb，Ta，Mo，Zrの金網状膜で燃料を包囲
した核燃料棒も知られている。その他の障壁材と
して、ステンレス鋼、ガラス質物質、Al，Be，
Mg，Cu等が米国特許3080893号、同3085059号、
同3212788号、同3291700号、同3230150号明細書
及び特開昭50−109397号公報で公知になつてい
る。 In order to prevent these undesirable destructions,
As a conventional example, for example, fuel pellets 2 and cladding tube 1
U.S. patent for method of inserting lubricant between
3018238, and as an example in which a barrier is provided between the cladding tube 1 and the fuel pellets 2.
DAS1238115 mentions a titanium layer. Furthermore, nuclear fuel rods in which the fuel is surrounded by a wire mesh membrane of Nb, Ta, Mo, and Zr are also known. Other barrier materials include stainless steel, glassy materials, Al, Be,
Mg, Cu, etc. are U.S. Patent No. 3080893, U.S. Patent No. 3085059,
This method is known from the specifications of 3212788, 3291700, 3230150, and Japanese Patent Application Laid-open No. 109397/1983.
同様に被覆管を内張りする概念は周知であり、
米国特許3502549号、同3625821号明細書、特開昭
51−69792号、同51−69795号、同51−69796号及
び同51−71497号公報において、内張り材として
Mo，Ｗ，Nb，Cr，Ni，Fe，Mg，Cu，純Zr，
Al，Ni−Cr合金、アルミ化コーテイング、珪素
化コーテイング等が示されている。 Similarly, the concept of lining cladding is well known;
U.S. Patent No. 3502549, Specification No. 3625821, JP-A-Sho
No. 51-69792, No. 51-69795, No. 51-69796, and No. 51-71497, as a lining material.
Mo, W, Nb, Cr, Ni, Fe, Mg, Cu, pure Zr,
Al, Ni-Cr alloys, aluminized coatings, silicided coatings, etc. are shown.
さらに、上記摩擦力に原因する応力集中を緩和
させる目的で、例えば、グラフアイト、二硫化モ
リブデン等の高潤滑剤を単体で、もしくは障壁と
呼ばれる他の金属材とともに燃料ペレツト２と被
覆管１との中間に挿入する提案が、それぞれ米国
特許3018238号明細書及び特開昭50−109396号公
報に記載されている。 Furthermore, for the purpose of alleviating the stress concentration caused by the above-mentioned frictional force, a highly lubricant such as graphite or molybdenum disulfide is added to the fuel pellets 2 and the cladding tube 1, either alone or together with another metal material called a barrier. Proposals for inserting it in the middle are described in the specification of US Pat.
しかしながら、以上の従来技術に述べてある障
壁材及び内張り材のあるものは中性子吸収断面積
が大きく炉の経済性を低下させるなどの欠点があ
る。また、上記引用した提案の幾つかは障壁とし
使用する物質が核燃料ペレツトと両立し難い物質
であるが、被覆管と両立し難い物質である場合が
あり、上記引用した提案はいずれも最近問題とな
つている核燃料と被覆管との間の局部的な化学的
−機械的相互作用に対する根本的な解決法まで達
しているとは云えない。 However, some of the barrier materials and lining materials described in the above prior art have drawbacks such as a large neutron absorption cross section that reduces the economic efficiency of the reactor. Additionally, in some of the proposals cited above, the material used as a barrier is a material that is incompatible with nuclear fuel pellets, but in some cases it is a material that is incompatible with cladding, and all of the proposals cited above have recently been problematic. It cannot be said that a fundamental solution to the local chemical-mechanical interaction between the nuclear fuel and the cladding has yet been found.
本発明は上記の状況に鑑みなされたものであ
り、腐食性ガス中において燃料との相互作用によ
り被覆管に応力が作用した場合、応力腐食割れが
起り難く、被覆管破損時の管内表面の急激な酸化
を防止でき信頼性を向上できる核燃料要素を提供
することを目的としたものである。 The present invention has been developed in view of the above circumstances, and is designed to prevent stress corrosion cracking from occurring when stress is applied to the cladding tube due to interaction with fuel in corrosive gases, and to prevent sudden damage to the inner surface of the tube when the cladding tube breaks. The objective is to provide a nuclear fuel element that can prevent severe oxidation and improve reliability.
本発明の核燃料要素は、ジルコニウム合金系の
材料からなる被覆管内に燃料ペレツトが充填さ
れ、上記被覆管の両端開口が端栓を介し密閉され
てなり、上記被覆管の内表面の少なくとも一部に
低酸素ジルコニウム合金系の材料から形成された
ライナー層が内張りされてなるものである。本発
明者らは、水に対する耐腐食性にすぐれたジルコ
ニウム合金のうち、低酸素濃度のジルコニウム合
金が、燃料の相互作用による応力腐食割れに対し
てもすぐれていることを見い出し、ジルコニウム
合金系被覆管内表面上にこの低酸素ジルコニウム
合金を内張りにしたものである。 In the nuclear fuel element of the present invention, a cladding tube made of a zirconium alloy material is filled with fuel pellets, both openings of the cladding tube are sealed via end plugs, and at least a portion of the inner surface of the cladding tube is filled with fuel pellets. It is lined with a liner layer made of a low-oxygen zirconium alloy material. The present inventors discovered that among zirconium alloys that have excellent corrosion resistance against water, zirconium alloys with low oxygen concentrations also have excellent resistance to stress corrosion cracking caused by interaction with fuel, and found that zirconium alloy-based coatings The inner surface of the tube is lined with this low-oxygen zirconium alloy.
以下本発明の核燃料要素の一実施例を従来と同
構造の説明は省略し第３図により説明する。第３
図は横軸に酸素濃度をとり縦軸にビツカース硬度
をとつてジルカロイー２の含有酸素濃度とビツカ
ース硬度との関係を示すグラフである。低酸素ス
ポンジジルコニウムにそれぞれ重量％で、錫1.59
％、鉄0.16％、クロム0.11％、ニツケル0.06％を
添加し、酸素濃度が500ppm以下の低酸素ジルカ
ロイー２のインゴツトを得た後、厚肉円筒状に形
成加工した。次に、酸素を約1300ppm含んだジル
カロイー２厚肉管の内面に上記の低酸素ジルカロ
イー２厚肉管を挿入し、ピルガー圧延機による冷
間加工と、焼鈍とを組み合せながら、外径12.52
mm、肉厚0.86mmの被覆管に仕上げた。このように
加工されたジルカロイー２被覆管においては、そ
の内表面に低酸素ジルカロイー２のライナーが形
成されている。上記低酸素ジルカロイー２ライナ
ー部のオートクレープ試験結果から、高温水に対
する耐腐食性は、従来の1200ppm程度の酸素を含
んだジルカロイー２と同等であることが判つた。 Hereinafter, one embodiment of the nuclear fuel element of the present invention will be explained with reference to FIG. 3, omitting the explanation of the same structure as the conventional one. Third
The figure is a graph showing the relationship between the oxygen concentration contained in Zircaloy 2 and the Vickers hardness, with the horizontal axis representing the oxygen concentration and the vertical axis representing the Vickers hardness. 1.59% by weight of tin to hypoxic sponge zirconium, respectively
%, 0.16% iron, 0.11% chromium, and 0.06% nickel to obtain a low-oxygen Zircaloy 2 ingot with an oxygen concentration of 500 ppm or less, which was then formed into a thick-walled cylindrical shape. Next, the above-mentioned low-oxygen Zircaloy 2 thick-walled tube was inserted into the inner surface of the Zircaloy 2 thick-walled tube containing about 1300 ppm of oxygen, and while cold working with a Pilger rolling mill and annealing were combined, the outer diameter was 12.52.
Finished in a cladding tube with a wall thickness of 0.86 mm. In the thus processed Zircaloy 2 cladding tube, a liner of low oxygen Zircaloy 2 is formed on its inner surface. The results of the autoclave test of the above-mentioned low-oxygen Zircaloy 2 liner section revealed that the corrosion resistance against high-temperature water was equivalent to that of the conventional Zircaloy 2 containing about 1200 ppm of oxygen.
更に、低酸素ジルカロイー２ライナー部の硬度
は、第３図の曲線Ａに示したように、酸素濃度の
減少とともに低下することがわかつた。硬度は、
材料の変形のし易さ、例えば、降伏応力と比例関
係にあることはよく知られている。従つて、ライ
ナー部の低酸素ジルカロイー２は被覆管母材より
低い降伏応力を持つことになり、被覆管と燃料ペ
レツトの力学的相互作用による燃料被覆管の応力
腐食割れに対してすぐれた材料であると云える。 Furthermore, it was found that the hardness of the hypoxic Zircaloy 2 liner portion decreased as the oxygen concentration decreased, as shown by curve A in FIG. The hardness is
It is well known that the ease with which a material deforms, for example, is proportional to its yield stress. Therefore, the low-oxygen Zircaloy 2 in the liner has a lower yield stress than the cladding base material, making it an excellent material against stress corrosion cracking of the fuel cladding due to mechanical interaction between the cladding and fuel pellets. I can say that there is.
このように、低酸素ジルカロイー２を内張りし
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共に、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、よう素濃度３mg／１c.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、このときに得られた被覆
管に生じた破断伸びを求めた。その結果、第４図
の棒グラフの棒線Ｂに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、第４図の棒線Ｃで示すよう
に破断伸びが減少していることが認められた。 In order to investigate the characteristics of a cladding tube lined with low-oxygen Zircaloy 2, a hollow nuclear fuel pellet was inserted into the cladding tube, a cylindrical pure aluminum rod was filled in the hollow part of the nuclear fuel pellet, and iodine was added. An aluminum rod was compressed in the longitudinal direction in an atmosphere with a concentration of 3 mg/1 c.c. 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 obtained at this time was determined. As a result, the characteristics shown in bar line B of the bar graph in FIG. 4 were obtained. On the other hand, for comparison, a conventional cladding tube was prepared and a similar experiment was conducted, and as a result, it was observed that the elongation at break was decreased as shown by bar C in FIG. 4.
第４図の結果から明らかなように本実施例の核
燃料要素は、腐食性ガス中において燃料との相互
作用により被覆管に応力が作用した場合、被覆管
の応力腐食割れが起り難く、即ち、低酸素ジルカ
ロイー２のライナーが応力腐食割れ防止に有効に
働き、抵抗力が強く大きな伸びを許容している。
また、被覆管内に水が浸入した場合の管内表面の
急激な酸化を防止でき、信頼性を向上できる。 As is clear from the results shown in FIG. 4, in the nuclear fuel element of this example, stress corrosion cracking of the cladding tube is less likely to occur when stress is applied to the cladding tube due to interaction with fuel in corrosive gas. The low-oxygen Zircaloy 2 liner works effectively to prevent stress corrosion cracking, has strong resistance, and allows for large elongation.
Further, rapid oxidation of the inner surface of the tube when water enters the cladding tube can be prevented, and reliability can be improved.
上記実施例は、ジルカロイー２からなる被覆管
と、ジルカロイー２からなるライナー材の場合に
ついて説明したが、他のジルカロイー４などのジ
ルコニウム合金系で構成された被覆管の場合も同
様の作用効果を有する。また、ライナー層は、被
覆管の軸方向の中間部分が最も腐食割れが生じ易
いのでこの中間部分だけに配置しても効果があ
る。 In the above embodiment, a cladding tube made of Zircaloy 2 and a liner material made of Zircaloy 2 were described, but similar effects can be obtained in the case of a cladding tube made of other zirconium alloys such as Zircaloy 4. . Further, since corrosion cracking is most likely to occur in the axially intermediate portion of the cladding tube, it is effective to arrange the liner layer only in this intermediate portion.
以上記述した如く本発明の核燃料要素は、腐食
性ガス中において燃料との相互作用により被覆管
に応力が作用した場合に、応力腐食割れが起り難
く、被覆管破損時の管内表面の急激な酸化も防止
でき、信頼性を向上できる効果を有するものであ
る。 As described above, the nuclear fuel element of the present invention is resistant to stress corrosion cracking when stress is applied to the cladding tube due to interaction with fuel in corrosive gas, and rapid oxidation of the inner surface of the tube when the cladding tube breaks. This has the effect of preventing this and improving reliability.
第１図は通常の核燃料要素の断面図、第２図は
第１図の核燃料要素に起り易い問題点の説明図、
第３図は本発明の核燃料要素の実施例のライナー
材のジルカロイー２の含有酸素濃度とビツカース
硬度との関係説明図、第４図は本発明の核燃料要
素の実施例の被覆管の特性と従来の被覆管の特性
比較用の説明図である。
１…被覆管、２…核燃料ペレツト、３…端栓。
Figure 1 is a cross-sectional view of a normal nuclear fuel element, Figure 2 is an explanatory diagram of problems that tend to occur in the nuclear fuel element shown in Figure 1,
FIG. 3 is an explanatory diagram of the relationship between the oxygen content of Zircaloy 2 and the Vickers hardness of the liner material of the embodiment of the nuclear fuel element of the present invention, and FIG. 4 shows the characteristics of the cladding tube of the embodiment of the nuclear fuel element of the present invention and the conventional FIG. 1... Cladding tube, 2... Nuclear fuel pellet, 3... End plug.
Claims (1)
に燃料ペレツトが充填され、上記被覆管の両端開
口が端栓を介し密閉されてなるものにおいて、上
記被覆管の内表面の少なくとも１部分に、ジルカ
ロイ（Zr−Sn系合金）中の酸素濃度が500ppm以
下のライナ層を内張りにしたことを特徴とする核
燃料要素。1. A cladding tube made of a zirconium alloy material is filled with fuel pellets, and both openings of the cladding tube are sealed via end plugs, and at least a portion of the inner surface of the cladding tube is coated with Zircaloy (Zr). -A nuclear fuel element characterized by being lined with a liner layer in which the oxygen concentration in the Sn-based alloy is 500 ppm or less.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP57168385A JPS5958389A (en) | 1982-09-29 | 1982-09-29 | Nuclear fuel element |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP57168385A JPS5958389A (en) | 1982-09-29 | 1982-09-29 | Nuclear fuel element |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS5958389A JPS5958389A (en) | 1984-04-04 |
| JPH0373832B2 true JPH0373832B2 (en) | 1991-11-25 |
Family
ID=15867121
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP57168385A Granted JPS5958389A (en) | 1982-09-29 | 1982-09-29 | Nuclear fuel element |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS5958389A (en) |
Families Citing this family (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US4775508A (en) * | 1985-03-08 | 1988-10-04 | Westinghouse Electric Corp. | Zirconium alloy fuel cladding resistant to PCI crack propagation |
| US4933136A (en) * | 1985-03-08 | 1990-06-12 | Westinghouse Electric Corp. | Water reactor fuel cladding |
| US4751045A (en) * | 1985-10-22 | 1988-06-14 | Westinghouse Electric Corp. | PCI resistant light water reactor fuel cladding |
| EP0223291B1 (en) | 1985-11-07 | 1991-07-31 | Akzo N.V. | Reinforcing element of synthetic material for use in reinforced concrete, more particularly prestressed concrete, reinforced concrete provided with such reinforcing elements, and processes of manufacturing reinforcing elements, and reinforced and prestressed concrete |
Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5744886A (en) * | 1980-07-04 | 1982-03-13 | Asea Atom Ab | Nuclear fuel rod |
| JPS58195185A (en) * | 1982-03-31 | 1983-11-14 | ゼネラル・エレクトリツク・カンパニイ | Zirconium alloy membrane having improved corrosion resistance |
-
1982
- 1982-09-29 JP JP57168385A patent/JPS5958389A/en active Granted
Patent Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5744886A (en) * | 1980-07-04 | 1982-03-13 | Asea Atom Ab | Nuclear fuel rod |
| JPS58195185A (en) * | 1982-03-31 | 1983-11-14 | ゼネラル・エレクトリツク・カンパニイ | Zirconium alloy membrane having improved corrosion resistance |
Also Published As
| Publication number | Publication date |
|---|---|
| JPS5958389A (en) | 1984-04-04 |
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