JPH0337594A - Nuclear fuel element and manufacture thereof - Google Patents
Nuclear fuel element and manufacture thereofInfo
- Publication number
- JPH0337594A JPH0337594A JP1171138A JP17113889A JPH0337594A JP H0337594 A JPH0337594 A JP H0337594A JP 1171138 A JP1171138 A JP 1171138A JP 17113889 A JP17113889 A JP 17113889A JP H0337594 A JPH0337594 A JP H0337594A
- Authority
- JP
- Japan
- Prior art keywords
- nuclear fuel
- cladding tube
- zirconium
- fuel element
- layer
- 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
- 239000003758 nuclear fuel Substances 0.000 title claims abstract description 39
- 238000004519 manufacturing process Methods 0.000 title claims description 10
- 238000005253 cladding Methods 0.000 claims abstract description 34
- 239000000446 fuel Substances 0.000 claims abstract description 15
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 claims abstract description 14
- 229910052726 zirconium Inorganic materials 0.000 claims abstract description 13
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 claims abstract description 7
- 229910052760 oxygen Inorganic materials 0.000 claims abstract description 7
- 239000001301 oxygen Substances 0.000 claims abstract description 7
- XEEYBQQBJWHFJM-UHFFFAOYSA-N Iron Chemical compound [Fe] XEEYBQQBJWHFJM-UHFFFAOYSA-N 0.000 claims abstract description 6
- 239000007789 gas Substances 0.000 claims abstract description 6
- 229910044991 metal oxide Inorganic materials 0.000 claims abstract description 6
- 150000004706 metal oxides Chemical class 0.000 claims abstract description 5
- VYZAMTAEIAYCRO-UHFFFAOYSA-N Chromium Chemical compound [Cr] VYZAMTAEIAYCRO-UHFFFAOYSA-N 0.000 claims abstract description 4
- 229910052804 chromium Inorganic materials 0.000 claims abstract description 3
- 239000011651 chromium Substances 0.000 claims abstract description 3
- 229910052742 iron Inorganic materials 0.000 claims abstract description 3
- 238000010438 heat treatment Methods 0.000 claims description 26
- 239000008188 pellet Substances 0.000 claims description 10
- 229910045601 alloy Inorganic materials 0.000 claims description 8
- 239000000956 alloy Substances 0.000 claims description 8
- 238000000034 method Methods 0.000 claims description 8
- PXHVJJICTQNCMI-UHFFFAOYSA-N Nickel Chemical compound [Ni] PXHVJJICTQNCMI-UHFFFAOYSA-N 0.000 claims description 6
- ATJFFYVFTNAWJD-UHFFFAOYSA-N Tin Chemical compound [Sn] ATJFFYVFTNAWJD-UHFFFAOYSA-N 0.000 claims description 3
- 229910052759 nickel Inorganic materials 0.000 claims description 3
- 239000011135 tin Substances 0.000 claims description 3
- 229910052718 tin Inorganic materials 0.000 claims description 3
- 229910001093 Zr alloy Inorganic materials 0.000 abstract description 12
- 239000000203 mixture Substances 0.000 abstract description 7
- 239000001307 helium Substances 0.000 abstract description 3
- 229910052734 helium Inorganic materials 0.000 abstract description 3
- SWQJXJOGLNCZEY-UHFFFAOYSA-N helium atom Chemical compound [He] SWQJXJOGLNCZEY-UHFFFAOYSA-N 0.000 abstract description 3
- RVTZCBVAJQQJTK-UHFFFAOYSA-N oxygen(2-);zirconium(4+) Chemical compound [O-2].[O-2].[Zr+4] RVTZCBVAJQQJTK-UHFFFAOYSA-N 0.000 abstract description 2
- 229910001928 zirconium oxide Inorganic materials 0.000 abstract description 2
- CLDVQCMGOSGNIW-UHFFFAOYSA-N nickel tin Chemical compound [Ni].[Sn] CLDVQCMGOSGNIW-UHFFFAOYSA-N 0.000 abstract 1
- 238000002485 combustion reaction Methods 0.000 description 5
- 238000010586 diagram Methods 0.000 description 3
- -1 ore Chemical compound 0.000 description 3
- MCMNRKCIXSYSNV-UHFFFAOYSA-N Zirconium dioxide Chemical compound O=[Zr]=O MCMNRKCIXSYSNV-UHFFFAOYSA-N 0.000 description 2
- 150000001875 compounds Chemical class 0.000 description 2
- 230000000694 effects Effects 0.000 description 2
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 2
- VYPSYNLAJGMNEJ-UHFFFAOYSA-N Silicium dioxide Chemical compound O=[Si]=O VYPSYNLAJGMNEJ-UHFFFAOYSA-N 0.000 description 1
- 238000009835 boiling Methods 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- 238000002474 experimental method Methods 0.000 description 1
- 230000004992 fission Effects 0.000 description 1
- 230000001771 impaired effect Effects 0.000 description 1
- 230000003993 interaction Effects 0.000 description 1
- 210000002425 internal capsule Anatomy 0.000 description 1
- 229910052743 krypton Inorganic materials 0.000 description 1
- DNNSSWSSYDEUBZ-UHFFFAOYSA-N krypton atom Chemical compound [Kr] DNNSSWSSYDEUBZ-UHFFFAOYSA-N 0.000 description 1
- 230000007774 longterm Effects 0.000 description 1
- 229910052751 metal Inorganic materials 0.000 description 1
- 239000002184 metal Substances 0.000 description 1
- 230000000116 mitigating effect Effects 0.000 description 1
- 230000001590 oxidative effect Effects 0.000 description 1
- 230000002093 peripheral effect Effects 0.000 description 1
- 238000007789 sealing Methods 0.000 description 1
- 229910052814 silicon oxide Inorganic materials 0.000 description 1
- 239000000126 substance Substances 0.000 description 1
- 229910052724 xenon Inorganic materials 0.000 description 1
- FHNFHKCVQCLJFQ-UHFFFAOYSA-N xenon atom Chemical compound [Xe] FHNFHKCVQCLJFQ-UHFFFAOYSA-N 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)
Abstract
Description
【発明の詳細な説明】
［産業上の利用分野］
本発明は、核燃料要素およびその製造方法の改良に係り
、特にすぐれた機械的特性をもつ核燃料要素およびその
製造方法に関するものである。DETAILED DESCRIPTION OF THE INVENTION [Industrial Application Field] The present invention relates to improvements in nuclear fuel elements and methods for producing the same, and particularly to nuclear fuel elements with excellent mechanical properties and methods for producing the same.
［従来の技術］
従来の核燃料要素に関しては、「軽水炉燃料のふるまい
」 （原子力安全研究協会：昭和６０年８月、Ｐ６４〜
６９）に詳細に記載されている。[Conventional technology] Regarding conventional nuclear fuel elements, see "Behavior of Light Water Reactor Fuel" (Nuclear Safety Research Association: August 1985, p. 64~
69).
第３図に従来の沸騰水型原子炉で用いられている核燃料
要素の縦断面略示図を示した。同図において、核燃料要
素］には、被覆管２内に複数個の燃料ペレット３が装填
されており、被覆管２の上下両端の開口部には、夫々上
部端栓４および下部端栓５が溶接されている。また、被
覆管２内には、ヘリウムカスが到人されており、１１部
プレナｌ＼部には、プレナムスプリング７およびＺｒ合
金のチップ８を詰めたゲッタ９が設けられている。FIG. 3 shows a schematic vertical cross-sectional view of a nuclear fuel element used in a conventional boiling water reactor. In the figure, a plurality of fuel pellets 3 are loaded into a cladding tube 2 of a nuclear fuel element, and an upper end plug 4 and a lower end plug 5 are provided at the openings at both upper and lower ends of the cladding tube 2, respectively. Welded. Further, helium scum is introduced into the cladding tube 2, and a getter 9 filled with a plenum spring 7 and a Zr alloy chip 8 is provided in the 11th planar portion.
このように構成された核燃料要素１において、核燃料の
燃焼が進むに従って核分裂生成物（ＦＩ）という）は、
核燃料要素内に放出される。この放出されたＦＰの中で
、気体状のＸ、ｅ、Ｋｒさらに揮発性のＣｓ等は、核燃
料要素内の内圧上昇の要因となる。また、燃料ペレット
３は燃焼に件って変形、膨張するため内服の上昇と相乗
して被覆管には大きな応力が負荷されることとなる。In the nuclear fuel element 1 configured in this way, as the combustion of the nuclear fuel progresses, fission products (FI) are produced as follows:
Released into nuclear fuel elements. Among the released FP, gaseous X, e, Kr, volatile Cs, etc. become a factor in increasing the internal pressure within the nuclear fuel element. Moreover, since the fuel pellets 3 deform and expand during combustion, a large stress is applied to the cladding tube in combination with the rise of the internal capsule.
［発明が解決しようとする課題］
核燃料要素の高燃焼度化にイ゛１′ってＦＰの放出量は
、現行運転時のものと比較して益々増大する傾３
向にある１、また、燃焼にともなうペレット・の変形量
も増大し、被覆管の許容応力を超えてしまう可能性があ
る。[Problems to be Solved by the Invention] As the burn-up of nuclear fuel elements increases, the amount of FP released tends to increase compared to that during current operation. The amount of deformation of the pellets due to combustion also increases, potentially exceeding the allowable stress of the cladding.
本発明の「１的は、核燃料要素の高燃焼度化に伴ない、
被覆管の内圧が上昇することに対して、強度および靭性
を高めて、高応力条件に耐えうる核燃料要素およびその
製造方法を提（Ｊｌｉすることである。The first object of the present invention is that as the burn-up of nuclear fuel elements increases,
An object of the present invention is to provide a nuclear fuel element that can withstand high stress conditions with increased strength and toughness as the internal pressure of the cladding increases, and a method for manufacturing the same.
［課題を解決するための手段］
上記課題を解決するための本発明に係る核燃料要素の構
成は、ジルコニラ１１基合金からなる核燃料被覆管の内
周面に純ジルコニウムｚ ＪＭを設けたライナ被覆管内
に、複数個の燃料ペレットを積層して収納し、その上下
両端を端栓により密封してなる核燃料要素において、該
被覆管外表１ｆＩｉに、ジルコニウｌ＼、鉄、クロム、
錫、ニッケルと酸素とからなる非化学量論的組成の金属
酸化Ｎ （Ｍｏ２．ただし、０ 、０５ ＜ ｘ ＜
０ 、２ ）を形成するようにしたことである。[Means for Solving the Problems] In order to solve the above problems, the nuclear fuel element according to the present invention has a structure in which pure zirconium z JM is provided on the inner peripheral surface of the nuclear fuel cladding made of an 11-base alloy of zirconia. In a nuclear fuel element formed by stacking and storing a plurality of fuel pellets and sealing both upper and lower ends with end plugs, the outer surface 1fIi of the cladding tube contains zirconium, iron, chromium,
Metal oxide N (Mo2) with a non-stoichiometric composition consisting of tin, nickel and oxygen, where 0, 05 < x <
0, 2).
また、上記課題を解決するための本発明に係る核燃料要
素の製造方法の構成は、ジルコニウムライナ被覆管の内
部に、複数個の核燃料ペレツ１へを装填し、上下両端を
端栓溶接してなる核燃料要素の製造方法において、外表
面が、ジルコニウｌ＼、鉱、クロム、錫、ニッケルから
なる合金で形成された被覆管を、まず、酸素混合ガス中
で加熱（第１−段階の加熱）した後に、これらを真空中
で加熱（第２段階の加熱）するようにしたことである。Moreover, the structure of the method for manufacturing a nuclear fuel element according to the present invention for solving the above problems is such that a plurality of nuclear fuel pellets 1 are loaded into a zirconium liner cladding tube, and both upper and lower ends are welded with end plugs. In a method for manufacturing a nuclear fuel element, a cladding tube whose outer surface is formed of an alloy consisting of zirconium, ore, chromium, tin, and nickel is first heated in an oxygen mixed gas (first stage heating). Later, these were heated in vacuum (second stage heating).
［作用コ
燃料被覆管の内側に設けられた純ジルコニウム層は被覆
管のジルコニウム合金と比較して軟かいため、燃料の燃
焼によって起こる燃料ペレツ１へと被覆管との機械的相
互作用（ＰＣＭＩ）を緩和する作用が期待てきる。[Operation] The pure zirconium layer provided inside the fuel cladding is soft compared to the zirconium alloy of the cladding, so mechanical interaction (PCMI) with the cladding occurs in the fuel pellets 1 caused by combustion of the fuel. It is expected that it will have a mitigating effect.
さらに、燃料被覆管の外側に設けられた金属酸化層はジ
ルコニウム合金と比較して強度的にすぐれているため、
高燃焼度運転時に被覆管に負荷される応力に耐えること
が可能となる。ジルコニウム基合金（Ｍと略す）の酸化
物として最も安定な化学形態はＭｏ２である。しかしな
からＭｏ２は、脆いためにジルコニウム合金層から剥離
しやすく、長期使用には不適である。Furthermore, the metal oxide layer provided on the outside of the fuel cladding has superior strength compared to zirconium alloy.
It becomes possible to withstand the stress applied to the cladding tube during high burnup operation. The most stable chemical form of the oxide of zirconium-based alloy (abbreviated as M) is Mo2. However, Mo2 is brittle and easily peels off from the zirconium alloy layer, making it unsuitable for long-term use.
したかって、Ｍｏ２のような化学量論的な化合物ではな
く、例えば、Ｍ Ｏｏ、ｏ ｓ ２Ｍ Ｏｏ、ｚなどの
ような非化学量論的化合物を、被覆管の外側に設けるよ
うにして、機械的強度と柔軟性とを兼備した核燃料被覆
管を製造することである。Therefore, instead of a stoichiometric compound such as Mo2, a non-stoichiometric compound such as M Oo, o s 2M Oo, z, etc., is provided on the outside of the cladding tube, and the mechanical The objective is to manufacture nuclear fuel cladding tubes that have both physical strength and flexibility.
本発明では、Ｍｏ２に種々の熱処理を施して、非化学量
論的酸化金属ＭＯ，（０，０５＜Ｘ＜０．２）層の生成
に成功した。すなわち、実験的にＸの値を０．２より大
きな試料を作成し、本試料しこ応力を負荷した結果、Ｍ
ｏ２（ｘ＞０．２）層は、下地の金属層から剥離しやす
いことがわかった。一方、０．０５より過小になると、
従来のジルコニウム合金の強度とほとんど変化なく、好
適な強度特性が得られないため、実験的に最適値として
Ｘ−０゜１を選定した。In the present invention, a non-stoichiometric metal oxide MO layer (0,05<X<0.2) was successfully produced by subjecting Mo2 to various heat treatments. That is, as a result of experimentally creating a sample with a value of
It was found that the o2 (x>0.2) layer was easily peeled off from the underlying metal layer. On the other hand, if it becomes less than 0.05,
Since the strength is almost the same as that of conventional zirconium alloys and suitable strength characteristics cannot be obtained, X-0°1 was experimentally selected as the optimum value.
さらに、最適な厚さとして２００ｕｍを選定した。Furthermore, 200 um was selected as the optimal thickness.
すなわち、厚さの値は過大になるとジルコニウム合金の
延性かそこなわれ強度的に好ましくない可能性がある。That is, if the thickness is too large, the ductility of the zirconium alloy may be impaired, which may be unfavorable in terms of strength.
−・方、厚さの値が過小になると好適な強度特性か得ら
れない。- On the other hand, if the thickness value becomes too small, suitable strength characteristics cannot be obtained.
ＭＯｏ□を形１反するためには、２段階の熱処理を最適
条件で実施する必要かある３、第１段階の熱処理は、ジ
ルコニウム合金層の表面を酸化させるためのものである
。第２段階の熱処理は、酸化層（ＭＯ７に近い組成のも
の）から過剰の酸素を合金内部へ拡散させて、出来るだ
け、所望のＭＯｏ、に近つけるために実施する真空熱処
岬である。In order to convert MOo□ to form 1, it is necessary to carry out two-stage heat treatment under optimal conditions.3 The first stage heat treatment is for oxidizing the surface of the zirconium alloy layer. The second stage heat treatment is a vacuum heat treatment performed to diffuse excess oxygen from the oxide layer (which has a composition close to MO7) into the alloy to bring it as close to the desired MOo as possible.
これらの熱処理のための最適条件を得るためには、真空
度、加熱温座、加熱時間などのパラメータサーベイを行
なわねばならない。In order to obtain the optimum conditions for these heat treatments, it is necessary to conduct a survey of parameters such as the degree of vacuum, heating temperature, and heating time.
第１段階の熱処理については、２０％の酸素混合ガス中
で、加熱温度は４． ＯＯ・〜５００℃、加熱時間］０
０〜２００時間の範囲が最適であることかわかった。For the first stage heat treatment, the heating temperature was 4.5% in a 20% oxygen mixed gas. OO・~500℃, heating time] 0
A range of 0 to 200 hours was found to be optimal.
第２段階の熱処理については、ます、真空度はｌＸｌ０
−”ｒｏｒｒ−ＩＸＩＯ”Ｔ”ｏｒｒてあればよい１、
加熱温度は、／１． Ｏ○〜５００ ’Ｃ１加熱時間は
５００時間（４００〜６００時間）であれは最適である
ことが確認できた。For the second stage of heat treatment, the degree of vacuum is lXl0.
-"rorr-IXIO"T"orr is enough1,
The heating temperature is /1. It was confirmed that 500 hours (400 to 600 hours) was the optimum heating time for O○~500' C1.
以−にのようにして、ＭＯ２Ｍから所定の厚さのＭｏ
ｏ、１１を生成することにより、強度のすぐれた核燃料
要素の製造が可能となり、その［」的を達成することか
できた。As described above, a predetermined thickness of Mo is obtained from MO2M.
By producing O, 11, it became possible to manufacture a nuclear fuel element with excellent strength, and the objective was achieved.
［実施例］
以下本発明の１実施例について、第王図を用いて説明す
る。[Example] Hereinafter, one example of the present invention will be described using a king diagram.
第１図（ａ）は、本発明に係る実施例の核燃料要素の縦
１１７１而略示図、同図（ｂ）は横断面鴫示図である。FIG. 1(a) is a vertical 1171 schematic diagram of a nuclear fuel element according to an embodiment of the present invention, and FIG. 1(b) is a horizontal cross-sectional diagram.
本実施例の構成は、核燃料要素１には、ジルコニウム合
金製の被覆管２内に複数個の燃料ベーン１〜３を積層し
て充填し、上下両端部は−に部端栓４および下部端栓５
で密封溶接されている。また、核燃料要素上内には、ヘ
リウムガスが封入されており、」二部プレナム部６には
、ブレナムスプリング７およびＺｒ合金チップ８を詰め
たゲッタ９が設けられている。さらに、被覆管２の内層
に、厚さ約８０μｍの純ジルコニウム層１０を設け、外
層はジルコニウム合金（ジルカロイ−２）からなり、特
に、その表面から２００μｍを組成がＭＯ。The structure of this embodiment is such that a nuclear fuel element 1 is filled with a plurality of fuel vanes 1 to 3 stacked in a zirconium alloy cladding tube 2, and both upper and lower ends have a negative end plug 4 and a lower end. Stopper 5
It is hermetically welded. Further, helium gas is sealed inside the nuclear fuel element, and the two-part plenum portion 6 is provided with a getter 9 filled with a blemish spring 7 and a Zr alloy chip 8. Further, a pure zirconium layer 10 with a thickness of about 80 μm is provided on the inner layer of the cladding tube 2, and the outer layer is made of a zirconium alloy (Zircaloy-2), and in particular, the composition is MO in the 200 μm from the surface.
１なる非化学量論的組成の酸化シルコニウス１合金層Ｉ
Ｊを設けるようにしたものである。Silicon oxide 1 alloy layer I with a non-stoichiometric composition of 1
J is provided.
つぎに、本実施例の核燃料要素を原子炉炉心に装荷して
使用した場合の動作について説明する。Next, the operation when the nuclear fuel element of this embodiment is loaded into a nuclear reactor core and used will be explained.
原子炉運転が進行し、装荷燃料の燃焼につれて燃料ベー
ン１〜から気体状のＦＰ（クリプトン及びキセノン）お
よび揮発性のセシウ１１等が放出されるため、しだいに
燃利要素内の内圧か上昇することとなる。さらに、燃焼
に伴なうペレットの変形管により、被覆管には応力が負
荷され、核燃料要素は機械強度的に厳しい環境下に置か
れることとなる。１
ｊ″Ｊ、上の実施例に−）いてその効果を検討するため
に行トＩ・、った実験結果を第２図に示す。すなＪ）ち
、第２図１．Ｊ： （１１，）従来の被覆管および（２
）本実施例の熱処理を施した場合の被覆管の各々を室温
において内圧力ｏ ｎ；による破裂試験の結果を総括し
た＃）のである、。As the reactor operation progresses and as the loaded fuel burns, gaseous FP (krypton and xenon) and volatile cesium-11 are released from the fuel vanes 1~, so the internal pressure within the fuel element gradually increases. It happens. Furthermore, due to the deformation of the pellet tube during combustion, stress is applied to the cladding tube, and the nuclear fuel element is placed under a harsh environment in terms of mechanical strength. Figure 2 shows the results of an experiment carried out in order to examine the effect of the above example. 11,) Conventional cladding tube and (2
) This is a summary of the results of a rupture test at room temperature under internal pressure on each of the cladding tubes subjected to the heat treatment of this example.
第２図によれは、（１）の場合には７０ Ｍ Ｐ ａで
破断したのに対して、（２）の場合には９０ＭＰａまで
破談しなかった。すなわち、強度は約３０％増大したこ
とになる。According to FIG. 2, the case (1) broke at 70 MPa, while the case (2) did not break at 90 MPa. That is, the strength increased by about 30%.
本実施例による熱処理方法を採用することにより、従来
の被覆管より機械的強度のすぐれた燃料被覆管を得られ
ることが実証された。It has been demonstrated that by employing the heat treatment method according to the present example, a fuel cladding tube with superior mechanical strength than conventional cladding tubes can be obtained.
［発明の効果］
上述のように、本発明によれば、高燃焼度時の被覆管の
強度性能を高め、高応力条件に耐えうる核燃料要素およ
びその製造方法を提供することができる。[Effects of the Invention] As described above, according to the present invention, it is possible to provide a nuclear fuel element that can enhance the strength performance of a cladding tube at high burn-up and withstand high stress conditions, and a method for manufacturing the same.
第１図（ａ）は、本発明の１実施例に係る核燃料要素の
縦断面略示図、第１図（ｂ ）は同上横断面略示図、第
２図は、内圧破裂試験結果の比較図、第３図は、従来の
核燃料要素の縦断面略示図である。
〈符Ｘの説明〉
１・・核燃料要素、２ 被覆管、３ 燃料ペレッＩ・、
４ 」一部端栓、５ 下部端栓、６ 」二部ブレナム、
９ ゲッター、」０ 純ジルコニウム屑、１１ 酸化ジ
ルコニウム合金層。FIG. 1(a) is a schematic vertical cross-sectional view of a nuclear fuel element according to an embodiment of the present invention, FIG. 1(b) is a schematic cross-sectional view of the same, and FIG. 2 is a comparison of internal pressure burst test results. FIG. 3 is a schematic longitudinal cross-sectional view of a conventional nuclear fuel element. <Explanation of mark X> 1...Nuclear fuel element, 2: Cladding tube, 3: Fuel pellet I...
4” part end plug, 5 lower end plug, 6” two part blenheim,
9 Getter, 0 Pure zirconium scrap, 11 Zirconium oxide alloy layer.
Claims (1)
に純ジルコニウム層を設けたライナ被覆管内に、複数個
の燃料ペレットを積層して収納し、その上下両端を端栓
により密封してなる核燃料要素において、該被覆管外表
面に、ジルコニウム、鉄、クロム、錫、ニッケルと酸素
からなる非化学量論的組成の金属酸化層（ＭＯ＿ｘ、た
だし、０．０５＜ｘ０．２）を形成したことを特徴とす
る核燃料要素。 ２、金属酸化層の厚さを、１００μｍ〜５００μｍとし
たことを特徴とする請求項１記載の核燃料要素。 ３、ジルコニウムライナ被覆管の内部に、複数個の核燃
料ペレットを装填し、上下両端を端栓溶接してなる核燃
料要素の製造方法において、外表面が、ジルコニウム、
鉄、クロム、錫、ニッケルからなる合金で形成された被
覆管を、まず、酸素混合ガス中で加熱（第１段階の加熱
）した後に、これらを真空中で加熱（第２段階の加熱）
することを特徴とする核燃料要素の製造方法。 ４、第１段階の熱処理は、酸素混合ガス中で、加熱温度
を４００℃〜５００℃、加熱時間を１００〜２００時間
とし、第２段階の熱処理は、真空度が１×１０＾−＾６
〜１×１０＾−＾７Ｔｏｒｒ中で、加熱温度を４００℃
〜５００℃、加熱時間を４００〜６００時間とすること
を特徴とする請求項３記載の核燃料要素の製造方法。[Claims] 1. A plurality of fuel pellets are stacked and stored in a liner cladding tube made of a zirconium-based alloy with a pure zirconium layer on the inner circumferential surface, and both upper and lower ends are connected with end plugs. In a nuclear fuel element sealed by a non-stoichiometric metal oxide layer (MO_x, where 0.05<x0. 2) A nuclear fuel element characterized by forming: 2. The nuclear fuel element according to claim 1, wherein the metal oxide layer has a thickness of 100 μm to 500 μm. 3. A method for manufacturing a nuclear fuel element in which a plurality of nuclear fuel pellets are loaded inside a zirconium liner cladding tube, and the upper and lower ends are welded with end plugs, in which the outer surface is made of zirconium,
A cladding tube made of an alloy consisting of iron, chromium, tin, and nickel is first heated in an oxygen mixed gas (first stage heating), and then heated in a vacuum (second stage heating).
A method for producing a nuclear fuel element, characterized in that: 4. The first stage heat treatment is performed in an oxygen mixed gas at a heating temperature of 400°C to 500°C and a heating time of 100 to 200 hours, and the second stage heat treatment is performed at a vacuum degree of 1x10^-^6.
~1×10^-^7 Torr, heating temperature 400℃
4. The method for producing a nuclear fuel element according to claim 3, wherein the heating time is 400 to 600 hours.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP1171138A JPH0337594A (en) | 1989-07-04 | 1989-07-04 | Nuclear fuel element and manufacture thereof |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP1171138A JPH0337594A (en) | 1989-07-04 | 1989-07-04 | Nuclear fuel element and manufacture thereof |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPH0337594A true JPH0337594A (en) | 1991-02-18 |
Family
ID=15917692
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP1171138A Pending JPH0337594A (en) | 1989-07-04 | 1989-07-04 | Nuclear fuel element and manufacture thereof |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH0337594A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US5715290A (en) * | 1993-07-01 | 1998-02-03 | Hitachi, Ltd. | Reactor water control method in BWR power plant, BWR power plant having low radioactivity concentration reactor water and fuel clad tube for BWR |
-
1989
- 1989-07-04 JP JP1171138A patent/JPH0337594A/en active Pending
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US5715290A (en) * | 1993-07-01 | 1998-02-03 | Hitachi, Ltd. | Reactor water control method in BWR power plant, BWR power plant having low radioactivity concentration reactor water and fuel clad tube for BWR |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| JP2846266B2 (en) | Cladding tube | |
| JPS63499B2 (en) | ||
| US3647623A (en) | Fuel element for a nuclear reactor | |
| JPH0337594A (en) | Nuclear fuel element and manufacture thereof | |
| JPS63284490A (en) | Fuel element for pressurized water type reactor | |
| US3898125A (en) | Nuclear fuel element containing strips of an alloyed Zr, Ti and Ni getter material | |
| KR101779128B1 (en) | Alumina-forming duplex stainless steels as accident resistant fuel cladding materials for light water reactors | |
| JP4999270B2 (en) | Methods, applications, and apparatus for boiling water nuclear fuel and fuel assembly cladding. | |
| US10217533B2 (en) | Fuel rod cladding and methods for making and using same | |
| JPH01232289A (en) | Fuel element and fuel assembly | |
| JP2010145234A (en) | Nuclear fuel rod | |
| JPH02167494A (en) | Nuclear fuel element and its production | |
| KR100982665B1 (en) | Super-plasticity uranium oxide nuclear fuel pellet and method of manufacturing the same | |
| JPH06281766A (en) | Nuclear reactor fuel rod | |
| JP4708032B2 (en) | Nuclear fuel rod and method of manufacturing nuclear fuel rod | |
| US3154845A (en) | Method of fabricating a fuel element | |
| JPH01318993A (en) | Nuclear fuel element | |
| JPH0472594A (en) | Oxide nuclear fuel rod | |
| JPS6026991B2 (en) | Nuclear fuel elements and their manufacturing methods | |
| JPS58186081A (en) | Nuclear fuel assembly | |
| Clayton | Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program) | |
| JPH06222174A (en) | Fuel rod for light water reactor | |
| JPH0777589A (en) | Nuclear fuel element | |
| JPS64675B2 (en) | ||
| JPH04236395A (en) | Production method of nuclear fuel pellet |