JPH0441794B2 - - Google Patents
Info
- Publication number
- JPH0441794B2 JPH0441794B2 JP59046300A JP4630084A JPH0441794B2 JP H0441794 B2 JPH0441794 B2 JP H0441794B2 JP 59046300 A JP59046300 A JP 59046300A JP 4630084 A JP4630084 A JP 4630084A JP H0441794 B2 JPH0441794 B2 JP H0441794B2
- Authority
- JP
- Japan
- Prior art keywords
- zirconium
- cladding tube
- corrosion cracking
- precipitated particles
- stress corrosion
- 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 41
- 229910052726 zirconium Inorganic materials 0.000 claims description 35
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 claims description 34
- XEEYBQQBJWHFJM-UHFFFAOYSA-N Iron Chemical compound [Fe] XEEYBQQBJWHFJM-UHFFFAOYSA-N 0.000 claims description 25
- 239000002245 particle Substances 0.000 claims description 24
- 239000003758 nuclear fuel Substances 0.000 claims description 16
- 229910052742 iron Inorganic materials 0.000 claims description 13
- 239000000446 fuel Substances 0.000 claims description 10
- 239000008188 pellet Substances 0.000 claims description 10
- 229910001093 Zr alloy Inorganic materials 0.000 claims description 4
- 229910052804 chromium Inorganic materials 0.000 claims description 2
- 150000001875 compounds Chemical class 0.000 claims description 2
- 229910052759 nickel Inorganic materials 0.000 claims description 2
- 229910052718 tin Inorganic materials 0.000 claims description 2
- PXHVJJICTQNCMI-UHFFFAOYSA-N Nickel Chemical compound [Ni] PXHVJJICTQNCMI-UHFFFAOYSA-N 0.000 claims 2
- VYZAMTAEIAYCRO-UHFFFAOYSA-N Chromium Chemical compound [Cr] VYZAMTAEIAYCRO-UHFFFAOYSA-N 0.000 claims 1
- ATJFFYVFTNAWJD-UHFFFAOYSA-N Tin Chemical compound [Sn] ATJFFYVFTNAWJD-UHFFFAOYSA-N 0.000 claims 1
- 239000011651 chromium Substances 0.000 claims 1
- 239000011135 tin Substances 0.000 claims 1
- 238000005260 corrosion Methods 0.000 description 37
- 230000007797 corrosion Effects 0.000 description 37
- 238000005336 cracking Methods 0.000 description 26
- 239000012535 impurity Substances 0.000 description 18
- 239000002244 precipitate Substances 0.000 description 15
- 230000000694 effects Effects 0.000 description 12
- 239000010410 layer Substances 0.000 description 12
- 230000004992 fission Effects 0.000 description 7
- 239000007789 gas Substances 0.000 description 6
- 239000000463 material Substances 0.000 description 6
- ZCYVEMRRCGMTRW-UHFFFAOYSA-N 7553-56-2 Chemical compound [I] ZCYVEMRRCGMTRW-UHFFFAOYSA-N 0.000 description 4
- XKRFYHLGVUSROY-UHFFFAOYSA-N Argon Chemical compound [Ar] XKRFYHLGVUSROY-UHFFFAOYSA-N 0.000 description 4
- LFQSCWFLJHTTHZ-UHFFFAOYSA-N Ethanol Chemical compound CCO LFQSCWFLJHTTHZ-UHFFFAOYSA-N 0.000 description 4
- VSCWAEJMTAWNJL-UHFFFAOYSA-K aluminium trichloride Chemical compound Cl[Al](Cl)Cl VSCWAEJMTAWNJL-UHFFFAOYSA-K 0.000 description 4
- 238000006243 chemical reaction Methods 0.000 description 4
- 239000011630 iodine Substances 0.000 description 4
- 229910052740 iodine Inorganic materials 0.000 description 4
- 239000000047 product Substances 0.000 description 4
- 239000002344 surface layer Substances 0.000 description 4
- JIAARYAFYJHUJI-UHFFFAOYSA-L zinc dichloride Chemical compound [Cl-].[Cl-].[Zn+2] JIAARYAFYJHUJI-UHFFFAOYSA-L 0.000 description 4
- 239000002826 coolant Substances 0.000 description 3
- 238000005498 polishing Methods 0.000 description 3
- 239000000243 solution Substances 0.000 description 3
- 239000000126 substance Substances 0.000 description 3
- 238000009864 tensile test Methods 0.000 description 3
- KRHYYFGTRYWZRS-UHFFFAOYSA-N Fluorane Chemical compound F KRHYYFGTRYWZRS-UHFFFAOYSA-N 0.000 description 2
- 230000002411 adverse Effects 0.000 description 2
- 229910045601 alloy Inorganic materials 0.000 description 2
- 239000000956 alloy Substances 0.000 description 2
- 229910052786 argon Inorganic materials 0.000 description 2
- 235000019441 ethanol Nutrition 0.000 description 2
- 230000007774 longterm Effects 0.000 description 2
- 238000000034 method Methods 0.000 description 2
- 235000005074 zinc chloride Nutrition 0.000 description 2
- 239000011592 zinc chloride Substances 0.000 description 2
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 description 1
- 238000010521 absorption reaction Methods 0.000 description 1
- 239000002253 acid Substances 0.000 description 1
- 238000004458 analytical method Methods 0.000 description 1
- 239000011260 aqueous acid Substances 0.000 description 1
- 239000007864 aqueous solution Substances 0.000 description 1
- 230000015572 biosynthetic process Effects 0.000 description 1
- 230000002301 combined effect Effects 0.000 description 1
- 239000004020 conductor Substances 0.000 description 1
- 230000003247 decreasing effect Effects 0.000 description 1
- 238000010586 diagram Methods 0.000 description 1
- 238000010894 electron beam technology Methods 0.000 description 1
- 238000002003 electron diffraction Methods 0.000 description 1
- 229910000765 intermetallic Inorganic materials 0.000 description 1
- 238000004519 manufacturing process Methods 0.000 description 1
- 239000000203 mixture Substances 0.000 description 1
- 229910017604 nitric acid Inorganic materials 0.000 description 1
- 238000007254 oxidation reaction Methods 0.000 description 1
- 150000003061 plutonium compounds Chemical class 0.000 description 1
- 230000002265 prevention Effects 0.000 description 1
- 230000002285 radioactive effect Effects 0.000 description 1
- 238000007789 sealing Methods 0.000 description 1
- 238000004381 surface treatment Methods 0.000 description 1
- 150000003586 thorium compounds Chemical class 0.000 description 1
- 150000003671 uranium compounds Chemical class 0.000 description 1
- 238000005406 washing Methods 0.000 description 1
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances 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)
- Glass Compositions (AREA)
- Catalysts (AREA)
Description
【発明の詳細な説明】
〔発明の利用分野〕
本発明は核分裂原子炉の炉心に使用する核燃料
要素に係り、特にジルコニウムまたはジルコニウ
ムを主成分とする合金からなるジルコニウムライ
ナ被覆管を有する核燃料要素に関する。Detailed Description of the Invention [Field of Application of the Invention] The present invention relates to a nuclear fuel element used in the core of a nuclear fission reactor, and more particularly to a nuclear fuel element having a zirconium liner cladding made of zirconium or an alloy mainly composed of zirconium. .
核燃料要素は通常、第１図に示すように、被覆
管１内に多数の燃料ペレツトを収納し、被覆管１
の両端を端栓３ａ，３ｂで密封して構成される。
核燃料要素の上部には核分裂によつて生じる気体
状核分裂生成物を溜めるプレナム４及び燃料ペレ
ツト２を保持するスプリング５が設けられてい
る。被覆管１は、燃料ペレツトと冷却材または減
速材との接触および化学反応を防止するととも
に、核分裂によつて生じる放射性核分裂生成物が
核燃料要素から冷却材または減速材にもれ出るの
を防止する機能が要求される。このため、被覆管
１の材質は、機械的性質に優れ、原子炉内での使
用条件下で良好な耐食性を有し、更には中性子吸
収の少ない事が必要である。この様な観点から、
現在はジルコニウムを主成分とするジルカロイが
被覆管材料として多数用いられている。
As shown in FIG. 1, a nuclear fuel element usually stores a large number of fuel pellets in a cladding tube 1.
It is constructed by sealing both ends with end plugs 3a and 3b.
A plenum 4 for storing gaseous fission products produced by nuclear fission and a spring 5 for holding fuel pellets 2 are provided in the upper part of the nuclear fuel element. The cladding tube 1 prevents contact and chemical reaction between the fuel pellets and the coolant or moderator, and also prevents radioactive fission products produced by nuclear fission from escaping from the nuclear fuel element into the coolant or moderator. functionality is required. For this reason, the material of the cladding tube 1 must have excellent mechanical properties, good corrosion resistance under the conditions of use in a nuclear reactor, and furthermore, must have low neutron absorption. From this perspective,
Currently, Zircaloy, whose main component is zirconium, is widely used as a cladding material.
しかしながら、現在までの経験によれば、ジル
カロイ被覆管の内表面は腐食性核分裂生成物との
化学反応および燃料ペレツト２の膨張によつて生
じる応力の重畳作用によつて応力腐食割れを起こ
すことがわかつてきた。また、被覆管１の外表面
は原子炉冷却材または減速材によつて局部的な酸
化反応を生じることがある。 However, experience to date has shown that the inner surface of the Zircaloy cladding is susceptible to stress corrosion cracking due to the combined effects of stress caused by chemical reactions with corrosive fission products and the expansion of the fuel pellets 2. I'm starting to feel strange. In addition, local oxidation reactions may occur on the outer surface of the cladding tube 1 due to reactor coolant or moderator.
上記の被覆管１の応力腐食割れを防ぐために、
第１図において被覆管１の内側部に例えば厚さ80
〜100μmの純ジルコニウム層６を冶金的に結合さ
せたいわゆるジルコニウムライナ被覆管が開発さ
れている。純ジルコニウム層６は、被覆管１と腐
食性核分裂生成物の接触を防止するとともに、燃
料ペレツト２の膨張によつて被覆管１に生じる局
所応力を緩和することにより応力腐食割れを防止
することが期待されている。 In order to prevent stress corrosion cracking of the above-mentioned cladding tube 1,
In Fig. 1, the inner part of the cladding tube 1 has a thickness of, for example, 80 mm.
A so-called zirconium liner cladding tube has been developed in which a pure zirconium layer 6 of ~100 μm is metallurgically bonded. The pure zirconium layer 6 prevents contact between the cladding tube 1 and corrosive fission products, and also prevents stress corrosion cracking by relieving local stress generated in the cladding tube 1 due to expansion of the fuel pellets 2. It is expected.
しかし、純ジルコニウム層６中には実質的に多
少の不純物が含有されており、不純物の種類およ
びそれらの含有量によつては応力腐食割れ防止効
果を低下することがある。このため、従来は高純
度に精製したジルコニウムライナや、ジルコニウ
ムの高純度の部分のみを選択的に使用して不純物
の含有量をできるだけ低くしたジルコニウムライ
ナ層を被覆管の内側に設けることが考えられた。
しかしこの様な高純度のジルコニウムを得るに
は、その製造が非常に面倒である。 However, the pure zirconium layer 6 substantially contains some impurities, and depending on the type of impurities and their content, the effect of preventing stress corrosion cracking may be reduced. For this reason, it has conventionally been considered to provide a highly purified zirconium liner or a zirconium liner layer inside the cladding tube that selectively uses only the high-purity portion of zirconium to keep the impurity content as low as possible. Ta.
However, in order to obtain such high-purity zirconium, the manufacturing process is extremely troublesome.
本発明の目的は、ジルコニウムライナ材に含有
する不純物による応力腐食割れ防止効果への悪影
響を低減させる事、更に本発明の他の目的は、被
覆管外表面と冷却材または減速材との反応による
局部腐食を防止し、核燃料の長期間使用時の信頼
性を向上させることにある。
An object of the present invention is to reduce the adverse effect of impurities contained in zirconium liner material on the effect of preventing stress corrosion cracking, and another object of the present invention is to reduce the adverse effect of impurities contained in zirconium liner material on the effect of preventing stress corrosion cracking. The purpose is to prevent local corrosion and improve the reliability of nuclear fuel during long-term use.
本発明は、ジルコニウムまたはジルコニウムを
主成分とする合金またはこれら両者を冶金的に結
合した材質から成る被覆管に、ウラン化合物、プ
ルトニウム化合物、トリウム化合物またはこれら
の混合物から成る燃料ペレツトを収納し、前記被
覆管の両端に固着された端栓を有する核燃料要素
において、本発明は前記被覆管の内表面または外
表面に分散しているFe、Cr、Ni、Snなどを含む
化合物の析出粒子を取り除いた事を特徴とする。
The present invention stores fuel pellets made of a uranium compound, a plutonium compound, a thorium compound, or a mixture thereof in a cladding tube made of zirconium, a zirconium-based alloy, or a metallurgical combination of the two; In a nuclear fuel element having end plugs fixed to both ends of a cladding tube, the present invention removes precipitated particles of compounds containing Fe, Cr, Ni, Sn, etc. dispersed on the inner or outer surface of the cladding tube. characterized by things.
一般に、被覆管の内外面に現われている粒子の
部分は、化学的活性度が高く、また荷重が加わつ
た場合に応力集中を生じ易い。発明者らによる
と、ライナ材であるジルコニウム層の内表面部に
おいて析出粒子の量が増すに従つて応力腐食割れ
の感受性が高くなる事が判明した。これは、腐食
性環境下で荷重を加えた場合、析出粒子の部分の
化学的活性度の高いことまたは応力集中域となる
ことによりき裂の発生及び進展が助長されるため
と考えられる。 Generally, the particulate portions appearing on the inner and outer surfaces of the cladding tube have a high degree of chemical activity and are likely to cause stress concentration when a load is applied. According to the inventors, it has been found that as the amount of precipitated particles increases on the inner surface of the zirconium layer, which is the liner material, the susceptibility to stress corrosion cracking increases. This is considered to be because, when a load is applied in a corrosive environment, the occurrence and propagation of cracks are promoted due to the high chemical activity of the precipitated particles or the formation of stress concentration areas.
一方、被覆管外面の水または水蒸気との反応に
よる局部腐食（ノジユラー腐食など）について
も、析出粒子が寄与していると考えられている。
すなわち、析出粒子が良好な電子伝導体となり、
その近傍の腐食反応を加速するという考えが提案
されている。 On the other hand, it is believed that precipitated particles also contribute to localized corrosion (such as nodular corrosion) caused by reaction with water or steam on the outer surface of the cladding tube.
In other words, the precipitated particles become good electron conductors,
The idea of accelerating corrosion reactions in the vicinity has been proposed.
上記の応力腐食割れおよび局部腐食はいずれも
被覆管の内面又は外面のごく表面層から生じて内
部へ進行するものであるため、表面層の応力腐食
割れあるいは局部腐食に対する初期段階の感受性
を低く保つておく必要がある。従つてジルコニウ
ムライナ被覆管表面部に析出した析出粒子を取り
除き、表面の応力腐食割れ、または局部腐食の感
受性を低くすることが本発明の意図するところで
ある。 Both the stress corrosion cracking and localized corrosion mentioned above originate from the very surface layer of the inner or outer surface of the cladding and progress to the inside, so the initial stage susceptibility to stress corrosion cracking or localized corrosion of the surface layer should be kept low. It is necessary to keep it. Therefore, it is the intention of the present invention to remove precipitated particles deposited on the surface of the zirconium liner cladding tube, thereby reducing the susceptibility of the surface to stress corrosion cracking or localized corrosion.
(1) 内側に純ジルコニウム層、その外側にジルコ
ニウム合金（例えばジルカロイ−２）層を有す
る被覆管の表面の析出物を取り除くことは、電
解研磨によつて実施できる。例えば、約1500〜
2000ppmの不純物を含有する第１図に示す純ジ
ルコニウム層６の内表面を約20000倍の倍率で
観察すると、約0.5μm以下の析出粒子が多数観
察される。この部分を塩化アルミニウム及び塩
化亜鉛を含むエチルアルコール溶液中で電流密
度約10mA／mm2で10秒間電解研磨することによ
り、約20000倍で観察しても析出物が確認でき
ない程度まで析出物を取り除くことができた。
このような被覆管内に多数の燃料ペレツトを装
填し、被覆管の両端を密封して第１図のような
核燃料要素を得る。
(1) Precipitates on the surface of a cladding tube having a pure zirconium layer on the inside and a zirconium alloy (for example, Zircaloy-2) layer on the outside can be removed by electrolytic polishing. For example, about 1500 ~
When the inner surface of the pure zirconium layer 6 shown in FIG. 1 containing 2000 ppm of impurities is observed at a magnification of about 20000 times, many precipitated particles of about 0.5 μm or less are observed. By electrolytically polishing this area in an ethyl alcohol solution containing aluminum chloride and zinc chloride at a current density of approximately 10 mA/mm 2 for 10 seconds, the precipitates are removed to the extent that they cannot be seen even when observed at approximately 20,000x magnification. I was able to do that.
A large number of fuel pellets are loaded into such a cladding tube, and both ends of the cladding tube are sealed to obtain a nuclear fuel element as shown in FIG.
(2) また、上記二重層からなるジルコニウムライ
ナ被覆管表面の析出物を除く他の例として、表
面を酸で洗浄する酸洗いを適当な時間行う事に
より析出物を取り除く事ができる。第２図は第
１図に示すジルカロイ−２被覆管１の外表面
を、容積率５％のフツ化水素酸、45％の硝酸の
水溶液中で20秒間酸洗したものの外表面の走査
型電子顕微鏡写真である。表面積１mm2あたりの
析出粒子の平均個数は約２×105個であるが、
第３図に示す様に、上記酸の水溶液中で約２分
間酸洗したものは析出粒子が非常に少なく、平
均析出粒子数は約1/10にまで減少した。(2) As another example of removing the precipitates on the surface of the double-layered zirconium liner cladding tube, the precipitates can be removed by washing the surface with acid for an appropriate period of time. Figure 2 shows the scanning electron beam of the outer surface of the Zircaloy-2 cladding tube 1 shown in Figure 1 which was pickled for 20 seconds in an aqueous solution of hydrofluoric acid with a volume ratio of 5% and nitric acid with a volume ratio of 45%. This is a microscopic photograph. The average number of precipitated particles per 1 mm 2 of surface area is approximately 2 × 10 5 , but
As shown in FIG. 3, the particles pickled in the aqueous acid solution for about 2 minutes had very few precipitated particles, and the average number of precipitated particles was reduced to about 1/10.
得られた被覆管内に多数の燃料ペレツト、ス
プリングを装填し、被覆管の両端を密封する。
このようにして核燃料要素が得られる。 A large number of fuel pellets and springs are loaded into the resulting cladding tube, and both ends of the cladding tube are sealed.
In this way a nuclear fuel element is obtained.
(3) 更に他の実施例として、第１図に示す二重層
のジルコニウムライナ被覆管において不純物と
して約800ppmの鉄を。含む純ジルコニウム層
６の内表面を約20000倍の倍率で観察すると、
約2μm以下の析出粒子が多数観察され、この析
出粒子をＸ線マイクロアナライザー（XMA）
及び電子線回析により分析したところ、ジルコ
ニウムと鉄の金属間化合物であると判つた。こ
の純ジルコニウム層６の内表面を上記実施例(1)
と同様に塩化アルミニウム及び塩化亜鉛を含む
エチルアルコール溶液中で電流密度約20mA／
mm2で10秒間電解研磨することにより、約20000
倍で観察しても析出物が確認できなかつた。す
なわち、純ジルコニウム層６の内表面部に析出
粒子として存在した不純物の鉄を、電解研磨に
よつて取り除くことができた。(3) As yet another example, about 800 ppm iron as an impurity in the double layer zirconium liner cladding shown in FIG. When observing the inner surface of the pure zirconium layer 6 at a magnification of about 20,000 times,
Many precipitated particles of approximately 2 μm or less were observed, and these precipitated particles were analyzed using an X-ray microanalyzer (XMA).
Analysis by electron diffraction revealed that it was an intermetallic compound of zirconium and iron. The inner surface of this pure zirconium layer 6 was
Similarly, in an ethyl alcohol solution containing aluminum chloride and zinc chloride, the current density is approximately 20 mA/
Approximately 20000 by electropolishing for 10 seconds at mm2
No precipitate could be observed even when observed under magnification. That is, the impurity iron present as precipitated particles on the inner surface of the pure zirconium layer 6 could be removed by electrolytic polishing.
このような被覆管内に燃料ペレツトを装填し、
被覆管の両端を密封して、核燃料要素を得ること
ができる。 Loading fuel pellets into such a cladding tube,
Both ends of the cladding tube can be sealed to obtain a nuclear fuel element.
ここで、応力腐食割れに対する感受性の指標と
して、ヨウ素ガス雰囲気中とアルゴンガス雰囲気
中でそれぞれの被覆管の単軸引張試験を行い、試
料破断部の断面減少率の比を求めた。すなわち、
アルゴンガス雰囲気中で延性破断したときの断面
減少率をE1、ヨウ素ガス雰囲気中で破断したと
きの断面減少率をE2として、応力腐食割れ感受
性指標を（E1−E2）／E1とする。この値が小さ
い方が、応力腐食割れ感受性が低い。すなわち耐
応力割れ性に優れている。第４図は、不純物量が
約2000ppm及び1500ppmのジルコニウムについ
て、それぞれ実施例(1)で示した電解研磨によつて
析出物を取り除いたものと取り除かないものの応
力腐食割れ感受性指標を示したものである。すな
わち、Ａ及びＢは不純物含有量2000ppmの場合で
あつてＡが析出物を取除く前、Ｂがそれを取除い
た後のものである。また、Ｃ及びＤは不純物含有
量1500ppmの場合であつてＣが析出物を取除く
前、Ｄがそれを取除いた後のものである。不純物
含有量約2000ppmの場合は、析出物を取り除かな
いものが応力腐食割れ感受性指標が0.25であるの
に対し、析出物を取り除くと0.10まで減少した。
同様に、不純物量約1500ppmの場合は、析出物を
取り除くことにより0.15から0.05に減少した。こ
の様に、表面層に存在する析出粒子を取り除くこ
とにより応力腐食割れに対する感受性を著しく低
くすることができた。 Here, as an index of susceptibility to stress corrosion cracking, a uniaxial tensile test was performed on each cladding tube in an iodine gas atmosphere and an argon gas atmosphere, and the ratio of the cross-sectional reduction rate of the sample fractured part was determined. That is,
The stress corrosion cracking susceptibility index is (E 1 − E 2 )/E 1 , where E 1 is the area reduction rate when ductile fracture occurs in an argon gas atmosphere, and E 2 is the area reduction rate when fracture occurs in an iodine gas atmosphere. shall be. The smaller this value is, the lower the stress corrosion cracking susceptibility is. In other words, it has excellent stress cracking resistance. Figure 4 shows the stress corrosion cracking susceptibility index for zirconium containing about 2000 ppm and 1500 ppm of impurities, with and without precipitates removed by electropolishing as shown in Example (1), respectively. be. That is, A and B are for the case where the impurity content is 2000 ppm, A is before removing the precipitate, and B is after removing the precipitate. Further, C and D are for the case where the impurity content is 1500 ppm, C is before removing the precipitate, and D is after removing the precipitate. When the impurity content was approximately 2000 ppm, the stress corrosion cracking susceptibility index was 0.25 without removing the precipitates, but it decreased to 0.10 when the precipitates were removed.
Similarly, when the amount of impurities was about 1500 ppm, it was reduced from 0.15 to 0.05 by removing the precipitates. In this way, by removing the precipitated particles present in the surface layer, the susceptibility to stress corrosion cracking could be significantly lowered.
また、不純物として約800ppmの鉄を含むジル
コニウムと約600ppmの鉄を含むジルコニウムを
ヨウ素ガス雰囲気中で単軸引張試験を行つたとこ
ろ、鉄の含有量が多い試料の方が少ない伸びで応
力腐食割れを生じ、不純物の鉄が応力腐食割れ感
受性を高めている事が判つた。この約800ppmの
鉄不純物を含むジルコニウムを前記実施例(3)の方
法で電解研磨を行い、表面に析出する不純物の鉄
を含む析出粒子を取り除いた後にヨウ素ガス雰囲
気中で単軸引張試験を行つたところ、破断するま
での試料延びが著しく大きくなり、より不純物鉄
の濃度の低い600ppmの試料よりも延びが大きく
なつた。すなわち、試料表面に析出した鉄を含む
析出粒子を取り除く事により応力腐食割れに対す
る感受性を著しく低くする事ができた。 In addition, when a uniaxial tensile test was conducted on zirconium containing about 800 ppm iron as an impurity and zirconium containing about 600 ppm iron in an iodine gas atmosphere, the sample with a higher iron content showed stress corrosion cracking with less elongation. It was found that impurity iron increased the susceptibility to stress corrosion cracking. This zirconium containing about 800 ppm of iron impurities was electrolytically polished using the method described in Example (3) above, and after removing precipitated particles containing impurity iron on the surface, a uniaxial tensile test was conducted in an iodine gas atmosphere. As a result, the elongation of the sample before breaking was significantly larger than that of the sample with a lower concentration of impurity iron at 600 ppm. That is, by removing the precipitated particles containing iron deposited on the sample surface, the susceptibility to stress corrosion cracking could be significantly lowered.
上述の様に、被覆管内表面の析出粒子を取り除
くあるいは減少させる事により、応力腐食割れに
対する感受性を著しく低下させる事が可能とな
り、安価で、多少純度の悪いジルコニウム層６を
用いた場合でも、応力腐食割れ防止効果を十分に
発揮できる。また、被覆管外表面についても局部
腐食に大きな影響を及ぼすと考えられている析出
粒子を取り除くことで、長期間使用時の局部腐食
の増大を低減し、核燃料要素の信頼性を向上させ
る事ができる。 As mentioned above, by removing or reducing the precipitated particles on the inner surface of the cladding tube, it is possible to significantly reduce the susceptibility to stress corrosion cracking. Can fully demonstrate corrosion cracking prevention effect. In addition, by removing precipitated particles from the outer surface of the cladding tube, which are thought to have a large effect on local corrosion, it is possible to reduce the increase in local corrosion during long-term use and improve the reliability of nuclear fuel elements. can.
なお、本発明は、被覆管表面の析出粒子を取り
除く事を意図するところであり、その方法は上述
の実施例に限るものでは無く。化学的あるいは電
気化学的表面処理を適宜行う事で本発明を実施す
ることができる。また、析出粒子は完全に取り除
かれなくても、その量を減少させる事によつて、
表面層を確率的に生じる初期段階の応力腐食割れ
あるいは局部腐食に対する発生確率を下げる事が
可能となり、被覆管の改良効果が発揮される。 Note that the present invention is intended to remove precipitated particles on the surface of the cladding tube, and the method is not limited to the above-mentioned embodiments. The present invention can be carried out by appropriately performing chemical or electrochemical surface treatment. In addition, even if the precipitated particles are not completely removed, by reducing their amount,
It becomes possible to lower the probability of occurrence of early-stage stress corrosion cracking or local corrosion that occurs stochastically in the surface layer, and the effect of improving the cladding is exhibited.
また、上述の応力腐食割れおよび局部腐食は、
析出物の存在する部分を核として確率的に生じる
現象であるため、析出物を完全に取り除かなくて
も減少させる事により効果があり、実質的に観測
できる析出粒子のうち、体積率で50％以下または
析出粒子の数で50％以下になるまで取り除くこと
により充分な改善効果が得られる。 In addition, the stress corrosion cracking and local corrosion mentioned above are
Since this is a phenomenon that occurs stochastically with the area where precipitates exist as a nucleus, it is effective to reduce the precipitates even if they are not completely removed. A sufficient improvement effect can be obtained by removing the particles until the number of precipitated particles becomes 50% or less.
本発明によれば、応力腐食割れの危険性が著し
く少なく、寿命の長い核燃料要素を得ることがで
きる。
According to the present invention, a nuclear fuel element with a significantly reduced risk of stress corrosion cracking and a long life can be obtained.
第１図は核燃料要素の縦断面図、第２図は約20
秒間酸洗した被覆管のジルカロイ−２の表面の金
属組織を示す走査型電子顕微鏡写真、第３図は約
２分間酸洗した被覆管のジルカロイ−２の表面の
金属組織を示す走査型電子顕微鏡写真、第４図は
ジルコニウムの応力腐食割れ感受性指標をそれぞ
れ示す特性図である。
１…被覆管、２…核燃料物質、３…端栓、４…
プレナム、５…保持部材（スプリング）、６…ジ
ルコニウムライナ。
Figure 1 is a vertical cross-sectional view of a nuclear fuel element, Figure 2 is approximately 20 mm
A scanning electron microscope photograph showing the metallographic structure of the surface of Zircaloy-2 of the cladding tube that was pickled for about 2 minutes. Fig. 3 is a scanning electron microscope photograph showing the metallographic structure of the surface of Zircaloy-2 of the cladding tube that was pickled for about 2 minutes. The photographs and FIG. 4 are characteristic diagrams showing the stress corrosion cracking susceptibility index of zirconium. 1... Cladding tube, 2... Nuclear fuel material, 3... End plug, 4...
Plenum, 5... Holding member (spring), 6... Zirconium liner.
Claims (1)
合金層を前記ジルコニウム層の外側に配置して成
り、両端が密封されたジルコニウムライナ被覆管
と、前記被覆管内に装填された複数の燃料ペレツ
トとを有する核燃料要素において、前記被覆管の
内表面または内表面と外表面両面に徴視的な大き
さで分散している鉄またはニツケルまたはクロム
またはスズを含む化合物の析出粒子を取り除いた
事を特徴とする核燃料要素。1. A nuclear fuel element having a zirconium liner cladding tube, which has a zirconium layer on the inside and a zirconium alloy layer on the outside of the zirconium layer, and which is sealed at both ends, and a plurality of fuel pellets loaded into the cladding tube. . A nuclear fuel element, characterized in that precipitated particles of iron, nickel, chromium, or tin-containing compounds dispersed in noticeable sizes on the inner surface or both the inner and outer surfaces of the cladding tube are removed.
Priority Applications (4)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP59046300A JPS60190889A (en) | 1984-03-09 | 1984-03-09 | Nuclear fuel element |
| EP85102649A EP0155603B1 (en) | 1984-03-09 | 1985-03-08 | Cladding tube for nuclear fuel and nuclear fuel element having this cladding tube |
| DE8585102649T DE3571096D1 (en) | 1984-03-09 | 1985-03-08 | Cladding tube for nuclear fuel and nuclear fuel element having this cladding tube |
| US07/059,175 US4863679A (en) | 1984-03-09 | 1987-06-12 | Cladding tube for nuclear fuel and nuclear fuel element having this cladding tube |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP59046300A JPS60190889A (en) | 1984-03-09 | 1984-03-09 | Nuclear fuel element |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS60190889A JPS60190889A (en) | 1985-09-28 |
| JPH0441794B2 true JPH0441794B2 (en) | 1992-07-09 |
Family
ID=12743352
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP59046300A Granted JPS60190889A (en) | 1984-03-09 | 1984-03-09 | Nuclear fuel element |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS60190889A (en) |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO2018062670A1 (en) * | 2016-09-29 | 2018-04-05 | Korea Atomic Energy Research Institute | Method for preparing nuclear fuel cladding for reducing crud deposition and method for reducing crud deposition on the nuclear fuel cladding |
Citations (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS507540A (en) * | 1973-05-18 | 1975-01-25 | ||
| JPS5368399A (en) * | 1976-12-01 | 1978-06-17 | Hitachi Ltd | Manufacturing method of nuclear fuel rod |
| JPS5636870A (en) * | 1979-08-27 | 1981-04-10 | Duracell Int | Chemical battery |
-
1984
- 1984-03-09 JP JP59046300A patent/JPS60190889A/en active Granted
Patent Citations (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS507540A (en) * | 1973-05-18 | 1975-01-25 | ||
| JPS5368399A (en) * | 1976-12-01 | 1978-06-17 | Hitachi Ltd | Manufacturing method of nuclear fuel rod |
| JPS5636870A (en) * | 1979-08-27 | 1981-04-10 | Duracell Int | Chemical battery |
Also Published As
| Publication number | Publication date |
|---|---|
| JPS60190889A (en) | 1985-09-28 |
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