JPH0415920B2 - - Google Patents
Info
- Publication number
- JPH0415920B2 JPH0415920B2 JP59209005A JP20900584A JPH0415920B2 JP H0415920 B2 JPH0415920 B2 JP H0415920B2 JP 59209005 A JP59209005 A JP 59209005A JP 20900584 A JP20900584 A JP 20900584A JP H0415920 B2 JPH0415920 B2 JP H0415920B2
- Authority
- JP
- Japan
- Prior art keywords
- weight
- powder
- starting powder
- rare earth
- ppm
- 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
- 239000000843 powder Substances 0.000 claims description 54
- 238000000034 method Methods 0.000 claims description 21
- 239000003758 nuclear fuel Substances 0.000 claims description 18
- XAGFODPZIPBFFR-UHFFFAOYSA-N aluminium Chemical compound [Al] XAGFODPZIPBFFR-UHFFFAOYSA-N 0.000 claims description 17
- 229910052782 aluminium Inorganic materials 0.000 claims description 17
- 229910001404 rare earth metal oxide Inorganic materials 0.000 claims description 17
- RTAQQCXQSZGOHL-UHFFFAOYSA-N Titanium Chemical compound [Ti] RTAQQCXQSZGOHL-UHFFFAOYSA-N 0.000 claims description 16
- 229910052719 titanium Inorganic materials 0.000 claims description 16
- 239000010936 titanium Substances 0.000 claims description 16
- 238000005245 sintering Methods 0.000 claims description 15
- 239000002245 particle Substances 0.000 claims description 10
- 239000000654 additive Substances 0.000 claims description 9
- 239000013078 crystal Substances 0.000 claims description 7
- 238000004519 manufacturing process Methods 0.000 claims description 5
- 238000005056 compaction Methods 0.000 claims description 4
- 239000000203 mixture Substances 0.000 claims description 4
- XLYOFNOQVPJJNP-UHFFFAOYSA-M hydroxide Chemical compound [OH-] XLYOFNOQVPJJNP-UHFFFAOYSA-M 0.000 claims description 3
- 239000000126 substance Substances 0.000 claims description 3
- 230000000996 additive effect Effects 0.000 claims 2
- 238000010438 heat treatment Methods 0.000 description 5
- FCTBKIHDJGHPPO-UHFFFAOYSA-N uranium dioxide Inorganic materials O=[U]=O FCTBKIHDJGHPPO-UHFFFAOYSA-N 0.000 description 5
- GWEVSGVZZGPLCZ-UHFFFAOYSA-N Titan oxide Chemical compound O=[Ti]=O GWEVSGVZZGPLCZ-UHFFFAOYSA-N 0.000 description 4
- 229910018626 Al(OH) Inorganic materials 0.000 description 3
- 229910010413 TiO 2 Inorganic materials 0.000 description 3
- 238000000354 decomposition reaction Methods 0.000 description 3
- 239000000446 fuel Substances 0.000 description 3
- 239000007789 gas Substances 0.000 description 3
- 230000002401 inhibitory effect Effects 0.000 description 3
- OOAWCECZEHPMBX-UHFFFAOYSA-N oxygen(2-);uranium(4+) Chemical compound [O-2].[O-2].[U+4] OOAWCECZEHPMBX-UHFFFAOYSA-N 0.000 description 3
- 230000002411 adverse Effects 0.000 description 2
- 230000000694 effects Effects 0.000 description 2
- 239000002574 poison Substances 0.000 description 2
- 231100000614 poison Toxicity 0.000 description 2
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 description 1
- AZDRQVAHHNSJOQ-UHFFFAOYSA-N alumane Chemical class [AlH3] AZDRQVAHHNSJOQ-UHFFFAOYSA-N 0.000 description 1
- 238000005253 cladding Methods 0.000 description 1
- 230000000052 comparative effect Effects 0.000 description 1
- 150000001875 compounds Chemical class 0.000 description 1
- 238000009792 diffusion process Methods 0.000 description 1
- 238000009826 distribution Methods 0.000 description 1
- 238000007730 finishing process Methods 0.000 description 1
- 230000004992 fission Effects 0.000 description 1
- 230000009969 flowable effect Effects 0.000 description 1
- 238000005469 granulation Methods 0.000 description 1
- 230000003179 granulation Effects 0.000 description 1
- 229910052739 hydrogen Inorganic materials 0.000 description 1
- 239000001257 hydrogen Substances 0.000 description 1
- 150000002823 nitrates Chemical class 0.000 description 1
- 230000001590 oxidative effect Effects 0.000 description 1
- TWNQGVIAIRXVLR-UHFFFAOYSA-N oxo(oxoalumanyloxy)alumane Chemical compound O=[Al]O[Al]=O TWNQGVIAIRXVLR-UHFFFAOYSA-N 0.000 description 1
- 239000011148 porous material Substances 0.000 description 1
- 229910052761 rare earth metal Inorganic materials 0.000 description 1
- 230000008961 swelling Effects 0.000 description 1
- -1 titanate compound Chemical class 0.000 description 1
- 239000004408 titanium dioxide Substances 0.000 description 1
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-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
Description
〔産業上の利用分野〕
本発明は、UO2又はUO2とPuO2との混合物か
ら成り、添加物として稀土類酸化物、特にGd2O3
を含む出発粉末を圧縮して加圧成形体とし、引続
き還元作用を有するガス雰囲気中でこの加圧成形
体を1500℃〜1750℃の範囲内の温度で焼結させて
固化することにより酸化物系核燃料焼結体を製造
する方法に関する。
〔従来の技術〕
この種の方法は西ドイツ特許第3144684号明細
書（特開昭58−92987号公報参照）から公知であ
る。稀土類酸化物の添加量が10重量％までである
この公知方法では、比表面積２〜4.5m2／ｇ及
び／又は結晶体の平均直径80nm〜250nmを有す
るUO2出発粉末を圧縮のために使用する。加圧成
形体の焼結時における温度での保持時間は１〜10
時間であつてよい。この公知の方法は、中性子物
理学的に燃焼可能の中性子毒として稀土類元素を
含み、またその密度が理論的に可能な密度の93％
より多く、場合によつては95％以上である酸化物
系核燃料焼結体を生じる。この種の高密度核燃料
焼結体は、運転下にある原子炉内でガス状又は易
揮発性の核分裂生成物を極く僅かに遊離するにす
ぎない。従つてこの核燃料焼結体で満たされた燃
料棒は、燃料棒被覆管内で加圧を生じることはほ
とんどない。また運転中原子炉内で核燃料焼結体
が収縮又は局部的に過加熱（これは燃料棒を破損
するおそれがある）することはない。
この公知方法のための極めて小さな比表面積及
び極めて大きい結晶体直径を有するUO2出発粉末
は造粒不能であり、いわゆるADU法〔“Gmel−
in−Handbuch der anorganischen Chemie，
Uran”（別巻A3、第99頁〜第108頁、1981年）〕
により直接得ることができる。またこの種のU2O
出発粉末は水熱分解条件下における粉末の滞留時
間が十分に長く選択される限りにおいて、同様に
造粒不能であるが、いわゆるAUC法〔“Gmelin
−Handbuch der anorga−nischen Chemie，
Uran”（別巻A3、第101〜第104頁、1981年）〕に
より製造することができる。
ADU法により得られた造粒不能の二酸化ウラ
ン粉末は一般に非流動性であり、従つて取扱いが
極めて困難である。他方AUC法での水熱分解条
件下における粉末滞留時間の延長は、粉末製造装
置への通過量を減少させる。
〔発明が解決しようとする問題点〕
従つて本発明の目的は、公知方法を改良し、特
に中性子物理学的に燃焼可能の中性子毒を極めて
多量に含む高密度の酸化物系核燃料焼結体の製造
を可能にすることであり、その際ADU法により
製造した二酸化ウラン粉末を使用する必要はなく
またAUC法による製造で水熱分解条件下におけ
る粉末の長い滞留時間も必要としないようにする
ことである。
〔問題点を解決するための手段〕
この問題を解決するため本発明によれば、UO2
成分が比表面積４〜７m2／ｇ及び／又は結晶体の
平均直径0.5μm未満、有利には0.2μm〜0.01μmを
有し、またアルミニウム５〜500重量ppm又はチ
タン５〜50重量ppmを酸化物又は水酸化物の形
で、粉末の元の粒径を変化させない混合工程で加
えた出発粉末を圧縮のために使用し、焼結時の温
度での保持時間を１〜４時間の範囲内で選択す
る。
〔作用効果〕
ところでこの僅少量のアルミニウム又はチタン
が出発粉末に含まれることによつて焼結時の拡散
工程は、酸化物系の核燃料焼結体を製造するため
それ自体焼結抑制作用を有する稀土類酸化物、例
えばGd2O3粉末を極めて多量に添加するにもかか
わらず、通常の延長されない水熱分解時間で
AUC法により製造されたUO2粉末を使用でき、
またそれにもかかわらず場合によつては理論的に
可能な密度の95％よりも高い核燃料焼結体の高焼
結密度を得ることができるほど、促進されること
が判明した。焼結時の温度での１〜４時間の保持
時間（これを促進された仕上げ工程を意味する）
で、核燃料焼結体の最良の焼結密度が得られる。
これより長い熱処理は焼結温度はこれ以上改良せ
ず、場合によつては核燃料焼結体を好ましくなく
膨潤することになる。
更に酸化作用をするガス雰囲気中で焼結前又は
後に行う熱処理は、還元作用をするガス雰囲気中
での熱処理により、稀土類酸化物の添加されてい
るUO2出発粉末から得られた核燃料焼結体の焼結
密度に悪影響を及ぼすことなく省略することがで
きる粉末の元の粒径を変えることなく実施する混
合工程で出発粉末にアルミニウム又はチタン酸合
物又は水酸化物の粉末の形で加えることにより、
例えば各粉末を共通の粉破機よつて混合する場合
に避け得ない造粒工程を省くことができる。また
核燃料焼結体中のアルミニウム又はチタンは、該
物質含有量が少なければ少ないほど、その可塑性
（クリープ現象）及び熱伝導性に及ぼす影響は小
さい。実際にアルミニウム含有量が５〜200重量
ppm又はチタン含有量が５〜50重量ppmの場合、
その影響は認められない。
更に出発粉末中のアルミニウム又はチタンの最
良の作用にとつては、UO2成分の結晶格子中にお
けるアルミニウム及びチタン以外の異物質の量が
200重量ppmより小さいことが有利である。この
種の二酸化ウラン粉末はAUC法によつて得られ
る。
添加物として稀土類酸化物、特にGd2O3を10重
量％まで含む出発粉末を使用することが有利であ
る。更に好ましいのは添加物として稀土類酸化
物、特にGd2O3を２〜10重量％、有利には４〜10
重量％含有する出発粉末を使用することである
が、最良に仕上げられた核燃料焼結体にとつて
は、添加物として稀土類酸化物、特にGd2O3を10
〜20重量％以上含む出発粉末を使用するのが有利
である。
欧州特許出願第0076680号公開公報から、UO2
粉末に、その圧縮及び焼結の前に他の物質の他に
アルミニウム及びチタンをも硝酸塩又は酸化物の
形で混合することが公知である。しかし稀土類酸
化物はこのUO2粉末に添加しない。従つてアルミ
ニウム又はチタンの混合物によつては、焼結密度
を高めるための稀土類酸化物の焼結抑制特性を凌
駕することができず、原子炉内での後加熱に際し
て後凝縮せず、従つて大粒径及び比較的大きな孔
を有する減少した密度の焼結体が得られる。これ
を達成するには、UO2粉末にアルミニウム及びチ
タンのような添加物0.05〜1.7容量％、（従つて稀
土類酸化物の焼結抑制特性を克服する場合よりも
はやはるかに多くの量）を加える必要がある。
更にUO2粉末に圧縮及び焼結の前に稀土類酸化
物及びチタン又はアルミニウム化合物を添加する
ことは西ドイツ特許出願公開第2008855号公報か
ら公知である。稀土類酸化物の含有量はUO2粉末
中例えば0.6％にすぎず、従つて極めて僅かであ
るが、アルミニウム含有量は例えば酸化アルミニ
ウム0.1％又はアルミニウム約530重量ppm又はチ
タン含有量は例えば二酸化チタン0.01重量％又は
チタン約60重量ppmと極めて高い。UO2粉末への
アルミニウム又はチタン添加量が著しく高いこと
により、圧縮されたUO2粉末の焼結によつて得ら
れる焼結体中には稀土類の均一な配分が達成され
るべきである。しかしUO2粉末中の稀土類酸化物
の含有量は、実際には得られた核燃料焼結体の密
度を減少し得ないほど僅少である。これに対し
UO2粉末中のアルミニウム及びチタン含有量が多
量の場合には、圧縮されたUO2粉末から得られた
核燃料焼結体の可塑性（クリープ現象）及び熱伝
導性に悪影響が及ぶ。
〔実施例〕
次に本発明及びその利点を比較例及び実施例に
基づき、第１表及び第２表によつて詳述する。
AUC法〔前記Gmelin−Handbuch参照〕によ
り得られた粒径2μm〜15μm、比表面積4.5m2／ｇ
及び平均結晶体直径0.08μmを有しかつCl，Ｆ，
Ｎ，Ｃ，Ni，Ca，Fe、及びSiをUO2結晶格子中
に全部で85重量ppm含む、未粒化UO2出発粉末
を、粒径0.5μm〜5μmのGd2O3粉末及び場合によ
つては粒径2μm〜15μmのTiO2又はAl（OH）3粉末
と、元の粉末粒径を変化させない混合機中で混合
する。引続き混合物を密度5.6ｇ／cm3の加圧成形
体に圧縮する。焼結炉内でこの加圧成形体を引続
き還元作用をする純粋な水素雰囲気中で加熱速度
10℃／分で1750℃に加熱し、この温度で２時間保
つ。冷却後加圧成形体から得られた核燃料焼結体
は、出発粉末に添加したGd2O3及びTiO2又はAl
（OH）3粉末の量と関連して第１表に示した焼結
密度（理論密度の％）を有していた。
[Industrial field of application] The present invention consists of UO 2 or a mixture of UO 2 and PuO 2 , with rare earth oxides, especially Gd 2 O 3 as additives.
The starting powder containing the oxide is compressed to form a compact, and the compact is then solidified by sintering in a reducing gas atmosphere at a temperature within the range of 1500°C to 1750°C. The present invention relates to a method for producing a nuclear fuel sintered body. [Prior Art] A method of this type is known from West German Patent No. 3144684 (see Japanese Patent Laid-Open No. 58-92987). In this known process, in which the amount of rare earth oxide added is up to 10% by weight, a UO 2 starting powder with a specific surface area of 2 to 4.5 m 2 /g and/or an average crystalline diameter of 80 nm to 250 nm is compressed. use. The holding time at the temperature during sintering of the pressed compact is 1 to 10
It may be time. This known method contains rare earth elements as neutron poisons which are neutron-physically combustible and whose density is 93% of the theoretically possible density.
This results in more, in some cases more than 95%, oxide-based nuclear fuel sintered bodies. Dense nuclear fuel sintered bodies of this type liberate only a small amount of gaseous or easily volatile fission products in an operating nuclear reactor. Therefore, a fuel rod filled with this nuclear fuel sintered body hardly generates pressurization within the fuel rod cladding tube. Furthermore, the nuclear fuel sintered body does not shrink or locally overheat (which could potentially damage the fuel rods) within the nuclear reactor during operation. The UO 2 starting powder with a very small specific surface area and a very large crystalline diameter for this known process cannot be granulated and is not suitable for the so-called ADU process [“Gmel-
in−Handbuch der anorganischen Chemie,
Uran” (Special volume A3, pp. 99-108, 1981)]
can be obtained directly by Also this kind of U 2 O
The starting powder is likewise non-granulatable, provided that the residence time of the powder under hydrothermal decomposition conditions is selected to be long enough, but the so-called AUC method [“Gmelin
−Handbuch der anorga−nischen Chemie,
Uranium dioxide powder obtained by the ADU method is generally non-flowable and therefore extremely difficult to handle. On the other hand, prolonging the powder residence time under hydrothermal decomposition conditions in the AUC method reduces the amount of powder passing through the powder manufacturing equipment. [Problems to be Solved by the Invention] Therefore, the purpose of the present invention is to The object of the present invention is to improve the known method and to make it possible to produce a high-density oxide-based nuclear fuel sintered body containing an extremely large amount of neutron poison, which is particularly neutron-physically combustible. It is not necessary to use uranium dioxide powder, nor is the long residence time of the powder under hydrothermal decomposition conditions produced by the AUC method. According to the present invention, UO 2
The component has a specific surface area of 4 to 7 m 2 /g and/or an average diameter of the crystals of less than 0.5 μm, advantageously 0.2 μm to 0.01 μm, and oxidizes 5 to 500 ppm by weight of aluminum or 5 to 50 ppm by weight of titanium. The starting powder, in the form of a compound or hydroxide, added during the mixing process without changing the original particle size of the powder, is used for compaction, and the holding time at the temperature during sintering is within the range of 1 to 4 hours. Select with . [Function and Effect] By the way, because this small amount of aluminum or titanium is included in the starting powder, the diffusion process during sintering has a sintering inhibiting effect in order to produce an oxide-based nuclear fuel sintered body. Despite the addition of very large amounts of rare earth oxides, e.g. Gd 2 O 3 powder, the
UO 2 powder produced by AUC method can be used,
It has also been found that it is nevertheless possible to obtain high sintered densities of nuclear fuel sintered bodies, in some cases even higher than 95% of the theoretically possible density. Holding time of 1 to 4 hours at sintering temperature (this represents an accelerated finishing process)
, the best sintered density of the nuclear fuel sintered body can be obtained.
A longer heat treatment will not further improve the sintering temperature and may lead to undesirable swelling of the nuclear fuel sintered body. Further, the heat treatment performed before or after sintering in an oxidizing gas atmosphere is a method for producing nuclear fuel sintered from UO2 starting powder to which rare earth oxides have been added, by heat treatment in a reducing gas atmosphere. Addition in the form of aluminum or titanate compound or hydroxide powder to the starting powder in a mixing step carried out without changing the original particle size of the powder, which can be omitted without adversely affecting the sintered density of the body By this,
For example, the granulation step that is unavoidable when the powders are mixed using a common crusher can be omitted. Furthermore, the smaller the content of aluminum or titanium in the nuclear fuel sintered body, the smaller the influence on its plasticity (creep phenomenon) and thermal conductivity. Actually the aluminum content is 5~200wt
ppm or titanium content is 5 to 50 ppm by weight,
Its influence is not recognized. Furthermore, for the best effect of aluminum or titanium in the starting powder, the amount of foreign substances other than aluminum and titanium in the crystal lattice of the UO binary components is
Advantageously less than 200 ppm by weight. This type of uranium dioxide powder is obtained by the AUC method. It is advantageous to use starting powders containing up to 10% by weight of rare earth oxides, in particular Gd 2 O 3 as additives. More preferably, 2 to 10% by weight, preferably 4 to 10% by weight of rare earth oxides, especially Gd 2 O 3 are added as additives.
For the best finished nuclear fuel sintered bodies, rare earth oxides, especially Gd 2 O 3 are added as additives to the starting powder containing 10% by weight.
It is advantageous to use starting powders containing ~20% by weight or more. From European Patent Application Publication No. 0076680, UO 2
It is known to mix aluminum and titanium, in addition to other substances, in the form of nitrates or oxides into the powder before its compaction and sintering. However, rare earth oxides are not added to this UO 2 powder. Therefore, some mixtures of aluminum or titanium cannot overcome the sintering-inhibiting properties of rare earth oxides to increase the sintered density, do not post-condense during post-heating in the reactor, and do not A reduced density sintered body with large grain size and relatively large pores is thus obtained. To achieve this, add additives such as aluminum and titanium to the UO2 powder 0.05-1.7% by volume (no longer much higher amounts than when overcoming the sintering-inhibiting properties of rare earth oxides). need to be added. Furthermore, it is known from DE 2008 855 A1 to add rare earth oxides and titanium or aluminum compounds to UO 2 powder before compaction and sintering. The content of rare earth oxides is, for example, only 0.6% in the UO 2 powder and is therefore very small, while the aluminum content is, for example, 0.1% aluminum oxide or approximately 530 ppm by weight of aluminum or the titanium content is, for example, titanium dioxide. Extremely high at 0.01% by weight or approximately 60 ppm by weight of titanium. Due to the significantly high aluminum or titanium addition to the UO 2 powder, a homogeneous distribution of the rare earths in the sintered body obtained by sintering the compacted UO 2 powder should be achieved. However, the content of rare earth oxides in the UO 2 powder is actually so small that it cannot reduce the density of the obtained nuclear fuel sintered body. In contrast to this
A high content of aluminum and titanium in the UO 2 powder adversely affects the plasticity (creep phenomenon) and thermal conductivity of the nuclear fuel sintered body obtained from the compressed UO 2 powder. [Example] Next, the present invention and its advantages will be explained in detail in Tables 1 and 2 based on comparative examples and examples. Particle size 2 μm to 15 μm, specific surface area 4.5 m 2 /g obtained by AUC method [see Gmelin-Handbuch above]
and has an average crystal diameter of 0.08 μm and contains Cl, F,
Ungranulated UO 2 starting powder containing a total of 85 wt ppm of N, C, Ni, Ca, Fe, and Si in the UO 2 crystal lattice was mixed with Gd 2 O 3 powder with a particle size of 0.5 μm to 5 μm and optionally TiO 2 or Al(OH) 3 powder with a particle size of 2 μm to 15 μm is then mixed in a mixer without changing the original powder particle size. The mixture is then pressed into compacts having a density of 5.6 g/cm 3 . This pressed body is then heated in a reducing pure hydrogen atmosphere in a sintering furnace at a heating rate.
Heat to 1750°C at 10°C/min and keep at this temperature for 2 hours. The nuclear fuel sintered body obtained from the cooled and pressed body contains Gd 2 O 3 and TiO 2 or Al added to the starting powder.
They had the sintered densities (% of theoretical density) shown in Table 1 in relation to the amount of (OH) 3 powder.
【表】
更にAUC法（前記Gmelin−Handbuch参照）
により得られた粒径2μm〜15μm、比表面積6.0
m2／ｇ及び平均結晶体直径0.08μmを有しかつCl，
Ｎ，Ｃ，Ni，Ca，Fe及びSiをUO2結晶格子中に
全部で85重量ppm含む未粒化UO2出発粉末を、粒
径0.5μm〜5μmのGd2O3粉末及び場合によつては
粒径2μm〜15μmのTiO2又はAl（OH）3粉末（第２
表に示した量）と同じ混合機中で混合し、次いで
第１表による数値を確認するために同じ方法で圧
縮し焼結した。出発粉末に添加したGd2O3及び
TiO2又はAl（OH）3粉末量との関連において第２
表に示した焼結密度を理論密度の％で示した。[Table] Furthermore, the AUC method (see Gmelin-Handbuch above)
Particle size 2 μm ~ 15 μm, specific surface area 6.0 obtained by
m 2 /g and an average crystal diameter of 0.08 μm, and Cl,
Ungranulated UO 2 starting powder containing a total of 85 ppm by weight of N, C, Ni, Ca, Fe and Si in the UO 2 crystal lattice was mixed with Gd 2 O 3 powder with a particle size of 0.5 μm to 5 μm and optionally is TiO 2 or Al(OH) 3 powder with a particle size of 2 μm to 15 μm (second
(in the amounts indicated in the table) in the same mixer and then compacted and sintered in the same way to confirm the values according to Table 1. Gd 2 O 3 and
The second in relation to the amount of TiO2 or Al(OH) 3 powder
The sintered densities shown in the table are expressed as % of the theoretical density.
Claims (1)
添加物として稀土類酸化物を含む出発粉末を圧縮
して加圧成形体とし、引続き還元作用を有するガ
ス雰囲気中でこの加圧成形体を1500℃〜1750℃の
範囲内の温度で焼結させて固化することにより酸
化物系の核燃料焼結体を製造する方法において、
UO2成分が比表面積４〜７m2／ｇ及び／又は結晶
体の平均直径0.5μm未満を有し、またアルミニウ
ム５〜500重量ppm又はチタン５〜50重量ppmを
酸化物又は水酸化物の形で、粉末の元の粒径を変
化させない混合工程で加えた出発粉末を圧縮のた
めに使用し、焼結時の温度での保持時間を１〜４
時間の範囲で選択することを特徴とする酸化物系
核燃料焼結体の製法。 ２ アルミニウムを５〜200重量ppm含有する出
発粉末を圧縮するために使用することを特徴とす
る特許請求の範囲第１項記載の方法。 ３ UO2成分の結晶格子中に含まれるアルミニウ
ム及びチタン以外の異物質含有量が200重量ppm
より少ない出発粉末を使用することを特徴とする
特許請求の範囲第１項記載の方法。 ４ 添加物として稀土類酸化物を10重量％まで含
む出発粉末を使用することを特徴とする特許請求
の範囲第１項記載の方法。 ５ 添加物として稀土類酸化物を２〜10重量％含
有する出発粉末を使用することを特徴とする特許
請求の範囲第４項記載の方法。 ６ 添加物として稀土類酸化物を10〜20重量％よ
りも多く含む出発粉末を使用することを特徴とす
る特許請求の範囲第１項記載の方法。[Claims] 1 Consisting of UO 2 or a mixture of UO 2 and PuO 2 ,
A starting powder containing a rare earth oxide as an additive is compressed into a pressed body, and the pressed body is subsequently sintered at a temperature within the range of 1500°C to 1750°C in a reducing gas atmosphere. In a method for producing an oxide-based nuclear fuel sintered body by solidifying it with
The two UO components have a specific surface area of 4 to 7 m 2 /g and/or an average crystalline diameter of less than 0.5 μm, and 5 to 500 ppm by weight of aluminum or 5 to 50 ppm of titanium in the form of oxide or hydroxide. The starting powder added during the mixing process without changing the original particle size of the powder is used for compaction, and the holding time at the sintering temperature is 1 to 4.
A method for producing an oxide-based nuclear fuel sintered body, which is characterized by selection within a time range. 2. Process according to claim 1, characterized in that it is used for compacting a starting powder containing from 5 to 200 ppm by weight of aluminum. 3 The content of foreign substances other than aluminum and titanium in the crystal lattice of the two UO components is 200 ppm by weight
2. Process according to claim 1, characterized in that less starting powder is used. 4. Process according to claim 1, characterized in that a starting powder is used which contains up to 10% by weight of rare earth oxides as additives. 5. Process according to claim 4, characterized in that a starting powder containing 2 to 10% by weight of rare earth oxides as additive is used. 6. Process according to claim 1, characterized in that a starting powder is used which contains more than 10 to 20% by weight of rare earth oxides as additives.
Applications Claiming Priority (3)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| DE3336387 | 1983-10-06 | ||
| DE3336387.0 | 1983-10-06 | ||
| DE3425581.8 | 1984-07-11 |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS6097294A JPS6097294A (en) | 1985-05-31 |
| JPH0415920B2 true JPH0415920B2 (en) | 1992-03-19 |
Family
ID=6211174
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP59209005A Granted JPS6097294A (en) | 1983-10-06 | 1984-10-04 | Manufacture of oxide group nuclear fuel sintered body |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS6097294A (en) |
Families Citing this family (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS6236589A (en) * | 1985-08-12 | 1987-02-17 | 日本ニユクリア・フユエル株式会社 | Manufacture of nuclear-fuel sintered body containing gadolinium oxide |
| JP2655908B2 (en) * | 1989-03-10 | 1997-09-24 | 三菱原子燃料株式会社 | Method for producing nuclear fuel pellet containing gatolinium oxide having large crystal grain size |
Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| DE2008855A1 (en) * | 1969-02-25 | 1970-09-03 | Centre d'Etude de !.'Energie Nucleaire, Brüssel | Nuclear fuel |
| FR2104136A5 (en) * | 1970-08-10 | 1972-04-14 | Gen Electric |
-
1984
- 1984-10-04 JP JP59209005A patent/JPS6097294A/en active Granted
Patent Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| DE2008855A1 (en) * | 1969-02-25 | 1970-09-03 | Centre d'Etude de !.'Energie Nucleaire, Brüssel | Nuclear fuel |
| FR2104136A5 (en) * | 1970-08-10 | 1972-04-14 | Gen Electric |
Also Published As
| Publication number | Publication date |
|---|---|
| JPS6097294A (en) | 1985-05-31 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| US6251310B1 (en) | Method of manufacturing a nuclear fuel pellet by recycling an irradiated oxide fuel pellet | |
| KR100287326B1 (en) | Method for recycling inferior goods of oxide nuclear fuel pellet | |
| US4774051A (en) | Sintered nuclear fuel compact and method for its production | |
| US4512939A (en) | Method for manufacturing oxidic sintered nuclear fuel bodies | |
| US5999585A (en) | Nuclear fuel having improved fission product retention properties | |
| US3995000A (en) | Ceramic nuclear fuel pellets | |
| JP2004538475A (en) | Method for producing mixed oxide nuclear fuel powder and mixed oxide nuclear fuel sintered body | |
| KR101302695B1 (en) | Fabrication method of burnable absorber nuclear fuel pellet, and the burnable absorber nuclear fuel pellet thereby | |
| US3715273A (en) | Nuclear fuel element containing sintered uranium dioxide fuel with a fine particulate dispersion of an oxide additive therein,and method of making same | |
| US4671904A (en) | Method for producing oxidic sintered nuclear fuel bodies | |
| JPH0120399B2 (en) | ||
| JPH0415920B2 (en) | ||
| US3342562A (en) | High density urania fuel elements | |
| US3761546A (en) | Method of making uranium dioxide bodies | |
| US3168479A (en) | Process for producing high density nuclear fuel particles | |
| KR100331483B1 (en) | Method of manufacturing oxide fuel pellets containing neutron-absorbing materials | |
| KR920000286B1 (en) | Manufacture of oxide group nuclear fuel sintered body | |
| JP2003509659A (en) | Manufacturing method of MOX type nuclear fuel pellets | |
| JPH01248092A (en) | Manufacture of nuclear fuel pellet | |
| KR101564744B1 (en) | Boron-bearing burnable absorber nuclear fuel pellet and the method for manufacturing thereof | |
| US3327027A (en) | Process for the production of plutonium oxide-containing nuclear fuel powders | |
| DE60120863T2 (en) | Nuclear fuel with burnt poison and process for its preparation | |
| Butler et al. | Ceramic nuclear fuel pellets | |
| JPS6129678B2 (en) | ||
| JP2002181975A (en) | Fuel pellet, its manufacturing method, and its fuel element and fuel assembly |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| R250 | Receipt of annual fees |
Free format text: JAPANESE INTERMEDIATE CODE: R250 |
|
| EXPY | Cancellation because of completion of term |