JPS6127719B2 - - Google Patents
Info
- Publication number
- JPS6127719B2 JPS6127719B2 JP56117554A JP11755481A JPS6127719B2 JP S6127719 B2 JPS6127719 B2 JP S6127719B2 JP 56117554 A JP56117554 A JP 56117554A JP 11755481 A JP11755481 A JP 11755481A JP S6127719 B2 JPS6127719 B2 JP S6127719B2
- Authority
- JP
- Japan
- Prior art keywords
- powder
- nuclear fuel
- treatment
- nitric acid
- present
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Expired
Links
- 239000000843 powder Substances 0.000 claims description 40
- 238000000034 method Methods 0.000 claims description 20
- 239000003758 nuclear fuel Substances 0.000 claims description 15
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 claims description 13
- 229910017604 nitric acid Inorganic materials 0.000 claims description 13
- 238000010438 heat treatment Methods 0.000 claims description 12
- 239000000463 material Substances 0.000 claims description 9
- 238000005245 sintering Methods 0.000 claims description 9
- 238000000465 moulding Methods 0.000 claims description 4
- 238000001035 drying Methods 0.000 claims 1
- 239000002994 raw material Substances 0.000 claims 1
- JCMLRUNDSXARRW-UHFFFAOYSA-N trioxouranium Chemical compound O=[U](=O)=O JCMLRUNDSXARRW-UHFFFAOYSA-N 0.000 description 16
- 239000008188 pellet Substances 0.000 description 14
- 238000004519 manufacturing process Methods 0.000 description 7
- 239000000446 fuel Substances 0.000 description 5
- 230000000052 comparative effect Effects 0.000 description 4
- 230000000704 physical effect Effects 0.000 description 4
- 238000012545 processing Methods 0.000 description 3
- 238000007493 shaping process Methods 0.000 description 3
- 239000000243 solution Substances 0.000 description 3
- 239000002699 waste material Substances 0.000 description 3
- 229910052770 Uranium Inorganic materials 0.000 description 2
- 238000000227 grinding Methods 0.000 description 2
- 238000001556 precipitation Methods 0.000 description 2
- 238000002360 preparation method Methods 0.000 description 2
- 230000005855 radiation Effects 0.000 description 2
- 239000002344 surface layer Substances 0.000 description 2
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 2
- 238000009736 wetting Methods 0.000 description 2
- 229910052778 Plutonium Inorganic materials 0.000 description 1
- WZECUPJJEIXUKY-UHFFFAOYSA-N [O-2].[O-2].[O-2].[U+6] Chemical compound [O-2].[O-2].[O-2].[U+6] WZECUPJJEIXUKY-UHFFFAOYSA-N 0.000 description 1
- 230000002411 adverse Effects 0.000 description 1
- 230000032683 aging Effects 0.000 description 1
- 239000007864 aqueous solution Substances 0.000 description 1
- 238000000498 ball milling Methods 0.000 description 1
- SHZGCJCMOBCMKK-KGJVWPDLSA-N beta-L-fucose Chemical compound C[C@@H]1O[C@H](O)[C@@H](O)[C@H](O)[C@@H]1O SHZGCJCMOBCMKK-KGJVWPDLSA-N 0.000 description 1
- 239000011230 binding agent Substances 0.000 description 1
- 230000006835 compression Effects 0.000 description 1
- 238000007906 compression Methods 0.000 description 1
- 238000001739 density measurement Methods 0.000 description 1
- 238000010586 diagram Methods 0.000 description 1
- 238000007598 dipping method Methods 0.000 description 1
- 238000004090 dissolution Methods 0.000 description 1
- 230000002349 favourable effect Effects 0.000 description 1
- 230000008014 freezing Effects 0.000 description 1
- 238000007710 freezing Methods 0.000 description 1
- 239000007788 liquid Substances 0.000 description 1
- 238000002156 mixing Methods 0.000 description 1
- 239000000203 mixture Substances 0.000 description 1
- 230000003647 oxidation Effects 0.000 description 1
- 238000007254 oxidation reaction Methods 0.000 description 1
- OOAWCECZEHPMBX-UHFFFAOYSA-N oxygen(2-);uranium(4+) Chemical compound [O-2].[O-2].[U+4] OOAWCECZEHPMBX-UHFFFAOYSA-N 0.000 description 1
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 description 1
- UTDLAEPMVCFGRJ-UHFFFAOYSA-N plutonium dihydrate Chemical compound O.O.[Pu] UTDLAEPMVCFGRJ-UHFFFAOYSA-N 0.000 description 1
- FLDALJIYKQCYHH-UHFFFAOYSA-N plutonium(IV) oxide Inorganic materials [O-2].[O-2].[Pu+4] FLDALJIYKQCYHH-UHFFFAOYSA-N 0.000 description 1
- 238000003672 processing method Methods 0.000 description 1
- 238000010298 pulverizing process Methods 0.000 description 1
- 238000012958 reprocessing Methods 0.000 description 1
- 238000002791 soaking Methods 0.000 description 1
- 239000002915 spent fuel radioactive waste Substances 0.000 description 1
- ZCUFMDLYAMJYST-UHFFFAOYSA-N thorium dioxide Chemical compound O=[Th]=O ZCUFMDLYAMJYST-UHFFFAOYSA-N 0.000 description 1
- FCTBKIHDJGHPPO-UHFFFAOYSA-N uranium dioxide Inorganic materials O=[U]=O FCTBKIHDJGHPPO-UHFFFAOYSA-N 0.000 description 1
- 229910000439 uranium oxide Inorganic materials 0.000 description 1
- MZFRHHGRNOIMLW-UHFFFAOYSA-J uranium(4+);tetrafluoride Chemical compound F[U](F)(F)F MZFRHHGRNOIMLW-UHFFFAOYSA-J 0.000 description 1
- XOOUIPVCVHRTMJ-UHFFFAOYSA-L zinc stearate Chemical compound [Zn+2].CCCCCCCCCCCCCCCCCC([O-])=O.CCCCCCCCCCCCCCCCCC([O-])=O XOOUIPVCVHRTMJ-UHFFFAOYSA-L 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
- Manufacture And Refinement Of Metals (AREA)
- Inorganic Compounds Of Heavy Metals (AREA)
Description
【発明の詳細な説明】
核燃料を製造する工程の一つとして、通常、核
燃料物質の粉末を圧縮成型した後、焼結してペレ
ツト化することが行なわれている。本発明は、か
かる核燃料ペレツト製造時における核燃料物質の
特性を改善する方法に関し、更に詳しくは、焼結
したときペレツトの密度特性が不良であるような
核燃料物質粉末の成型燃結性能を向上させ、良好
な密度特性の焼結ペレツトを得ることができるよ
うな方法に関するものである。DETAILED DESCRIPTION OF THE INVENTION As one of the steps for producing nuclear fuel, powder of nuclear fuel material is usually compression molded and then sintered to form pellets. The present invention relates to a method for improving the characteristics of nuclear fuel material during the production of nuclear fuel pellets, and more specifically, improving the molding and sintering performance of nuclear fuel material powder whose density characteristics are poor when sintered. The present invention relates to a method by which sintered pellets with good density characteristics can be obtained.
従来より、使用済み核燃料の再処理で分離して
得られたウラン溶液は、流動床による加熱処理に
より脱硝し、粉末化されているが、この方法によ
り得られる三酸化ウラン粉末は密度特性が悪いも
のが多く、酸化処理後、成型焼結してペレツトと
した時に、ペレツトの密度が必要な値まで上がら
ないという問題があつた。たとえば、通常の三酸
化ウラン粉末をもちいた混合酸化物燃料（酸化ウ
ランと酸化プルトニウムとの混合燃料）では成型
焼結後の密度が理論値の93〜97％となるのに対し
て、前記のような特性の悪い三酸化ウラン粉末を
用いた混合酸化物燃料では、成型焼結後の密度が
理論値の90％程度しか上がらず、そのままでは使
用できない。 Conventionally, uranium solutions obtained by separating spent nuclear fuel during reprocessing have been denitrified and powdered by heat treatment in a fluidized bed, but the uranium trioxide powder obtained by this method has poor density characteristics. There was a problem that the density of the pellets did not rise to the required value when the pellets were formed and sintered after oxidation treatment to form pellets. For example, in the case of mixed oxide fuel (mixed fuel of uranium oxide and plutonium oxide) using ordinary uranium trioxide powder, the density after shaping and sintering is 93 to 97% of the theoretical value, whereas Mixed oxide fuel using uranium trioxide powder with such poor properties has a density that increases by only about 90% of the theoretical value after shaping and sintering, and cannot be used as is.
この問題を解決するため、特性不良の三酸化ウ
ラン粉末を再度硝酸に溶解し、沈澱法によつて再
び粉末化するという処理が必要であつた。ところ
が、このような処理を行なうと、核燃料製造工程
および製造時間が長くなり、作業員の被曝の機会
もそれだけ増加するばかりでなく、処理にともな
つて廃液などの二次廃棄物の発生量も増大する。 In order to solve this problem, it was necessary to dissolve the uranium trioxide powder with poor characteristics in nitric acid again and pulverize it again by a precipitation method. However, such treatment not only lengthens the nuclear fuel manufacturing process and production time, increasing the chances of workers being exposed to radiation, but also increases the amount of secondary waste such as waste liquid generated during treatment. increase
また、上記した流動床方式によりえられた三酸
化ウラン粉末以外にも、核燃料製造過程において
粉末物性の不良な粉末が発生した場合、これら粉
末についても再粉末化処理を行なつていたが、前
記と同様の問題点が生じるし、また再粉末化処理
によつても必ずしも所望の特性をもつ粉末が得ら
れるとは限らない。というのは、沈澱法におい
て、熟成時間、温度等の条件が十分管理されなけ
ればならず、混合時、酸化還元処理時の条件もそ
の数値によつては粉末特性に悪影響を及ぼす虞れ
がある。 In addition to the uranium trioxide powder obtained by the fluidized bed method described above, when powder with poor physical properties is generated during the nuclear fuel manufacturing process, these powders are also repulverized. Problems similar to those described above arise, and powder having desired properties cannot necessarily be obtained even by re-powdering treatment. This is because in the precipitation method, conditions such as aging time and temperature must be carefully controlled, and depending on the values of the conditions during mixing and redox treatment, there is a risk that the powder properties may be adversely affected. .
本発明の目的は、上記のような従来技術の欠点
を解消し、焼結したときのペレツトの密度特性が
不良であるような核燃料物質粉末の成型焼結性能
を、簡単な工程で、短時間で、しかも作業員の被
曝もほとんどなく、改善することのできる方法を
提供することにある。 The purpose of the present invention is to eliminate the above-mentioned drawbacks of the prior art, and to improve the molding and sintering performance of nuclear fuel material powder, which has poor density properties when sintered, in a simple process and in a short time. Our goal is to provide a method that can improve the situation, while minimizing the exposure of workers to radiation.
かかる目的を達成するため、本発明では、被処
理粉末に硝酸を加えて該粉末を湿潤または浸漬状
態となし、次いでマイクロ波加熱処理を施して再
び乾燥させるよう構成されており、まさにこのよ
うな処理方法に本発明の特徴がある。 In order to achieve such an object, the present invention is configured such that nitric acid is added to the powder to be treated to make the powder wet or immersed, and then the powder is subjected to microwave heating treatment and dried again. The processing method is a feature of the present invention.
ウランまたはプルトニウムの硝酸溶液をマイク
ロ波加熱することにより、好ましい物性を有する
核燃料物質粉末が得られることは既に公知の事実
であり、特性不良粉末を完全に溶解した後にマイ
クロ波加熱すれば物性の良好な粉末が得られるの
は当然のことであるが、特性不良粉末を湿潤また
は浸漬状態程度の処理をしただけでその特性を改
善することができるという事実は、本発明者等が
はじめて知得したものである。本発明は、かかる
知得に基づき完成されたものである。 It is already a well-known fact that nuclear fuel material powder with favorable physical properties can be obtained by heating a nitric acid solution of uranium or plutonium with microwaves. It is a matter of course that a powder with good properties can be obtained, but the present inventors were the first to know that the properties of powder with poor properties can be improved simply by processing it in a wet or immersed state. It is something. The present invention has been completed based on this knowledge.
以下、本発明について更に詳しく説明する。図
面は本発明方法の剥剥念を示すもので、バツチ式
処理の場合の一例である。要調製粉末１を容器２
に収納し、調製用のグローブボツクス３に移動さ
せる。その後、硝酸をシヤワ４等により適量添加
する。一定時間経過させて要調製粉末を湿潤もし
くは浸漬状態とした後、マイクロ波加熱炉５に収
容し、加熱処理する。マイクロ波は、シール部６
を介してグローブボツクス３の外のマイクロ波発
振器７より印加される。その際、オフガスはオフ
ガス処理系８で無害化処理する。調製終了した粉
末１′は、ペレツト製造工程に提供され粉末物性
良好の粉末と同様に取扱われることになる。な
お、この例は、バツチ式処理の例であるが、要調
整粉末が大量の場合は、連続式のマイクロ波加熱
炉を用い連続処理することも可能である。 The present invention will be explained in more detail below. The drawing shows the stripping process according to the present invention, and is an example of a batch type process. Powder to be prepared 1 in container 2
and move it to glove box 3 for preparation. Thereafter, an appropriate amount of nitric acid is added using a shower 4 or the like. After the powder to be prepared is brought into a wet or immersed state for a certain period of time, it is placed in a microwave heating furnace 5 and heat-treated. The microwave is applied to the seal part 6.
The voltage is applied from a microwave oscillator 7 outside the glove box 3 via a microwave oscillator 7 outside the glove box 3. At this time, the off-gas is rendered harmless by the off-gas treatment system 8. The prepared powder 1' is provided to the pellet manufacturing process and handled in the same manner as powder having good physical properties. Although this example is an example of batch processing, if a large amount of powder needs to be adjusted, it is also possible to carry out continuous processing using a continuous microwave heating furnace.
さて、本発明方法で処理できる核燃料物質粉末
には、前記した三酸化ウランのほか、二酸化プル
トニウム、二酸化ウラン、四フツ化ウラン、二酸
化トリウムの単体またはそれらの２種以上の混合
物からなり、成型焼結性能の悪い粉末をあげるこ
とができる。本発明方法で使用する硝酸として
は、通常、数規定程度の水溶液が用いられるが、
濃硝酸から非常に薄い希硝酸の範囲まで適宜選択
しうる。要は、硝酸により被処理粉末の表面層の
一部を溶解させることができればよいのである。
つまり、表面層の一部を溶解させる目的をもつ
て、被処理粉末を硝酸によつて湿潤または浸漬処
理するのである。かかる処理を行なつた後、前記
のようにマイクロ波加熱処理を施して再び乾燥粉
末とするのであるが、マイクロ波加熱処理の条件
は、従来の硝酸溶液を処理する場合と同様のもの
でもよい。また、場合によつては、マイクロ波加
熱処理と同様にボールミル方式などの粉砕処理を
行なつて、粉末特性の調製を行なうこともでき
る。 Now, the nuclear fuel material powder that can be treated by the method of the present invention includes, in addition to the above-mentioned uranium trioxide, plutonium dioxide, uranium dioxide, uranium tetrafluoride, and thorium dioxide alone or in a mixture of two or more thereof. Powders with poor freezing performance can be used. As the nitric acid used in the method of the present invention, an aqueous solution of about several normals is usually used.
It can be selected as appropriate from a range of concentrated nitric acid to very dilute diluted nitric acid. In short, it is sufficient if a part of the surface layer of the powder to be treated can be dissolved by nitric acid.
That is, the powder to be treated is wetted or immersed in nitric acid for the purpose of dissolving a portion of the surface layer. After performing such treatment, it is subjected to microwave heat treatment as described above to form a dry powder again, but the conditions for the microwave heat treatment may be the same as those for conventional treatment of nitric acid solution. . Further, depending on the case, the powder characteristics may be adjusted by performing a pulverization process such as a ball mill method similar to the microwave heat treatment.
被処理粉末の量に対して使用する硝酸の量は、
完全に溶解させる従来の場合よりも、湿潤または
浸漬させる場合の方がはるかに少なくてすみ、こ
のことは、その後のマイクロ波加熱処理時間は、
本発明方法では極めて短かくてよいことを意味す
る。しかも、湿潤または浸漬処理は、粉末を溶解
する必要がないので短時間で完了させることがで
きる。さらに、この処理により発生するガスの量
も本発明方法では極めて少なくて済む。 The amount of nitric acid used for the amount of powder to be treated is:
Wetting or soaking requires much less time than the traditional case of complete dissolution, which means that the subsequent microwave heating time is
This means that the method of the present invention can be extremely short. Moreover, the wetting or dipping process can be completed in a short time because it is not necessary to dissolve the powder. Furthermore, the amount of gas generated by this treatment is also extremely small in the method of the present invention.
本発明方法の有効性を比較例と対比しながら説
明すると、次の通りである。 The effectiveness of the method of the present invention will be explained in comparison with comparative examples as follows.
比較例
流動床で脱硝して得た三酸化ウラン（UO3）で
以下のフローによりペレツト製造を行なつた。そ
の結果、得られた焼結ペレツトの密度は、理論密
度の88〜90％であり、平均的な核燃料ペレツトに
対して低い密度であつた。Comparative Example Pellet production was carried out using uranium trioxide (UO 3 ) obtained by denitrification in a fluidized bed according to the following flow. As a result, the density of the sintered pellets obtained was 88-90% of the theoretical density, which was lower than that of an average nuclear fuel pellet.
UO3
↓
焙焼還元 700℃焙焼２時間，還元４時間
↓
粉砕（ボールミル）４時間
↓
バインダ添加 ステアリン酸亜鉛 0.3wt％
↓
プレス ダイス径 6.4mmφ 2ton／cm2
↓
予焼 800℃２時間，200℃／時
↓
焼結 1700℃２時間，400℃／時
↓
密度測定
実施例
前記と同じ三酸化ウラン粉末を下記の条件で調
整し、その後、前記比較例と同様のフローでペレ
ツト製造を行なつたところ、得られた焼結ペレツ
トの密度は理論密度の93〜95％まで向上させるこ
とができた。UO 3 ↓ Roasting reduction 700℃ roasting 2 hours, reduction 4 hours ↓ Grinding (ball mill) 4 hours ↓ Addition of binder Zinc stearate 0.3wt% ↓ Press Die diameter 6.4mmφ 2ton/cm 2 ↓ Prebaking 800℃ 2 hours, 200℃/hour ↓ Sintering 1700℃ for 2 hours, 400℃/hour ↓ Density measurement example The same uranium trioxide powder as above was prepared under the following conditions, and then pellets were manufactured using the same flow as in the comparative example. As a result, the density of the obtained sintered pellets could be increased to 93-95% of the theoretical density.
UO3（100ｇ）
↓
硝酸添加 IN HNO3 20ml
↓
10分放置
↓
マイクロ波加熱処理 2450MHz，2KW，３
分印加
↓
以降は比較例と同じフロー
この結果からみて、本発明による前処理により
核燃料物質粉末の成型焼結性能が著しく改善され
ることが判る。UO 3 (100g) ↓ Nitric acid addition IN HNO 3 20ml ↓ Leave for 10 minutes ↓ Microwave heat treatment 2450MHz, 2KW, 3
Minute application ↓ From then on, the flow is the same as that of the comparative example. From these results, it can be seen that the pretreatment according to the present invention significantly improves the shaping and sintering performance of nuclear fuel material powder.
以上詳記したことから明らかなように、本発明
によれば、そのままでは燃料ペレツトの製造に使
用しえないような粉末物性（成型焼結性能）をも
つ核燃料物質粉末が、極めて簡単な処理を行なう
だけで燃料ペレツトの製造に使用できることにな
り、また、処理は短時間内で行なえるから作業員
の被曝を低減化でき、二次廃棄物の発生量も少な
くてすむなど、その実施効果は極めて大きいもの
である。 As is clear from the detailed description above, according to the present invention, nuclear fuel material powder having powder physical properties (molding and sintering performance) that cannot be used as it is for producing fuel pellets can be processed in an extremely simple manner. The benefits of implementing this method are that it can be used for the production of fuel pellets, and that the treatment can be carried out within a short period of time, reducing radiation exposure for workers and reducing the amount of secondary waste generated. It is extremely large.
図面は本発明方法の概念図である。
１……要調製粉末、３……調製用グローブボツ
クス、４……シヤワ、５……マイクロ波加熱炉、
８……オフガス処理系、１′……調製終了粉末。
The drawing is a conceptual diagram of the method of the present invention. 1... Powder to be prepared, 3... Glove box for preparation, 4... Shower, 5... Microwave heating furnace,
8...off gas treatment system, 1'...prepared powder.
Claims (1)
るような核燃料物質原料粉末に硝酸を加えて該粉
末を湿潤または浸漬状態となし、次いでマイクロ
波加熱処理を施して乾燥させることを特徴とする
核燃料粉末の成型焼結性能改善方法。1. Nuclear fuel characterized by adding nitric acid to a nuclear fuel material raw material powder whose density characteristics become poor after sintering treatment to make the powder wet or immersed, and then drying it by subjecting it to microwave heating treatment. Method for improving powder molding and sintering performance.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP56117554A JPS5819593A (en) | 1981-07-27 | 1981-07-27 | Method of improving mold sintering performance of nuclear fuel powder |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP56117554A JPS5819593A (en) | 1981-07-27 | 1981-07-27 | Method of improving mold sintering performance of nuclear fuel powder |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS5819593A JPS5819593A (en) | 1983-02-04 |
| JPS6127719B2 true JPS6127719B2 (en) | 1986-06-26 |
Family
ID=14714684
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP56117554A Granted JPS5819593A (en) | 1981-07-27 | 1981-07-27 | Method of improving mold sintering performance of nuclear fuel powder |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS5819593A (en) |
Families Citing this family (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP4890515B2 (en) * | 2008-08-08 | 2012-03-07 | ウエタックス株式会社 | Speaker |
| JP5555869B2 (en) * | 2009-10-20 | 2014-07-23 | 独立行政法人日本原子力研究開発機構 | Manufacturing method of nickel oxide using microwave absorption and heat generation effect by product addition |
-
1981
- 1981-07-27 JP JP56117554A patent/JPS5819593A/en active Granted
Also Published As
| Publication number | Publication date |
|---|---|
| JPS5819593A (en) | 1983-02-04 |
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