JPH0769467B2 - Ruthenium Separation Equipment from Radioactive Waste - Google Patents
Ruthenium Separation Equipment from Radioactive WasteInfo
- Publication number
- JPH0769467B2 JPH0769467B2 JP8116187A JP8116187A JPH0769467B2 JP H0769467 B2 JPH0769467 B2 JP H0769467B2 JP 8116187 A JP8116187 A JP 8116187A JP 8116187 A JP8116187 A JP 8116187A JP H0769467 B2 JPH0769467 B2 JP H0769467B2
- Authority
- JP
- Japan
- Prior art keywords
- ruthenium
- radioactive waste
- oxidation
- tank
- waste
- 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
Landscapes
- Treatment Of Water By Oxidation Or Reduction (AREA)
- Heat Treatment Of Water, Waste Water Or Sewage (AREA)
- Solid-Sorbent Or Filter-Aiding Compositions (AREA)
- Processing Of Solid Wastes (AREA)
Description
【発明の詳細な説明】 ［産業上の利用分野］ 本発明は、放射性廃棄物中に含まれるルテニウムを回収
するための装置に係り、特に放射性廃棄物中のルテニウ
ムを回収すべく四酸化ルテニウムとしてその廃棄物中か
ら分離する放射性廃棄物からのルテニウム分離装置に関
するものである。Description: TECHNICAL FIELD The present invention relates to an apparatus for recovering ruthenium contained in radioactive waste, and in particular as ruthenium tetroxide for recovering ruthenium in radioactive waste. The present invention relates to a device for separating ruthenium from radioactive waste that is separated from the waste.
［従来の技術］ 高レベルの放射性廃棄物の廃棄処理においては放射性廃
棄物をガラス原料と一緒に加熱させ、これを格納容器内
に入れてガラス固化させた状態で格納するようにしてい
る。[Prior Art] In the disposal of high-level radioactive waste, the radioactive waste is heated together with the glass raw material, and this is put in a storage container and stored in a vitrified state.
この放射性廃棄物中には、種々の重金属が含まれている
が、本出願人は先に放射性廃棄物中からルテニウムを回
収する装置（特願昭60−253649号）を提案した。Although various heavy metals are contained in this radioactive waste, the present applicant has previously proposed a device (Japanese Patent Application No. 60-253649) for recovering ruthenium from the radioactive waste.
この先願の発明においては、放射性廃棄物を加熱しなが
ら、オゾン等の酸化剤を吹き込み、放射性廃棄物中に含
まれるルテニウムを四酸化ルテニウムとして気化させ、
これを回収するようにしたものである。In the invention of this earlier application, while heating the radioactive waste, an oxidizing agent such as ozone is blown in to vaporize ruthenium contained in the radioactive waste as ruthenium tetraoxide,
It is designed to collect this.
［発明が解決しようとする問題点］ ところで、この先願の発明においては、放射性廃棄物を
反応容器内に入れ、その反応容器内で酸化と気化とを同
時に行なうようにしている。しかしながら、放射性廃棄
物中のルテニウムの酸化速度は気化速度よりも遅く、そ
のため、放射性廃棄物からルテニウムを四酸化ルテニウ
ムとして分離するには、酸化時間が支配し、その間放射
性廃棄物を無駄に加熱していることとなりやすい。[Problems to be Solved by the Invention] In the invention of this earlier application, radioactive waste is put in a reaction vessel, and oxidation and vaporization are simultaneously performed in the reaction vessel. However, the oxidation rate of ruthenium in radioactive waste is slower than the vaporization rate, so in order to separate ruthenium as ruthenium tetroxide from the radioactive waste, the oxidation time is dominated, during which the radioactive waste is wastefully heated. It is easy to become.
本発明は、上記事情を考慮してなされたもので、放射性
廃棄物中のルテニウムの酸化と、その酸化後の四酸化ル
テニウムの気化とを効率よく行なえる放射性廃棄物から
のルテニウム分離装置を提供することを目的とする。The present invention has been made in consideration of the above circumstances, and provides an apparatus for separating ruthenium from radioactive waste, which can efficiently oxidize ruthenium in radioactive waste and vaporize ruthenium tetroxide after the oxidation. The purpose is to do.
［問題点を解決するための手段及び作用］ 本発明は上記の目的を達成するために、放射性廃棄物中
に酸化剤を供給してその廃棄物中のルテニウムを四酸化
ルテニウムに酸化する酸化槽と、酸化後の放射性廃棄物
を導入し、その廃棄物中に溶け込んだ四酸化ルテニウム
を加熱蒸留して四酸化ルテニウムを気化させる蒸留槽と
を備えたもので、酸化槽と蒸留槽とに分けることで、酸
化処理と蒸留処理とを個々に最適に操作でき、ルテニウ
ムの分離回収率を向上できる。[Means and Actions for Solving Problems] In order to achieve the above object, the present invention provides an oxidizing tank for supplying an oxidizing agent to radioactive waste to oxidize ruthenium in the waste to ruthenium tetraoxide. And a distillation tank for introducing radioactive waste after oxidation, and heating and distilling ruthenium tetroxide dissolved in the waste to vaporize ruthenium tetroxide, which is divided into an oxidation tank and a distillation tank. As a result, the oxidation treatment and the distillation treatment can be optimally operated individually, and the separation and recovery rate of ruthenium can be improved.
［実施例］ 以下に本発明に係る放射性廃棄物からのルテニウム分離
装置の好適−実施例を添付図面に基づいて説明する。[Embodiment] A preferred-embodiment of the apparatus for separating ruthenium from radioactive waste according to the present invention will be described below with reference to the accompanying drawings.
第１図において、１は高レベル放射性廃棄物２を酸化処
理する酸化槽で、その酸化槽１に液供給ライン３及びそ
の供給ポンプ４を介して蒸留槽５が接続され、その蒸留
槽５の頂部より、四酸化ルテニウム回収装置６と、オフ
ガス処理装置７とが順に接続される。In FIG. 1, reference numeral 1 denotes an oxidation tank for oxidizing high-level radioactive waste 2. A distillation tank 5 is connected to the oxidation tank 1 via a liquid supply line 3 and a supply pump 4 thereof. From the top, the ruthenium tetroxide recovery device 6 and the off-gas processing device 7 are connected in order.
酸化槽１は、放射性廃棄物２を槽１内に供給する処理液
供給管８が接続され、また槽１内の放射性廃棄物２中
に、オゾン、過マンガン酸カリ、セリウム（IV）化合物
などの酸化剤を吹き込む酸化剤供給管９が設けられる。The oxidation tank 1 is connected to a treatment liquid supply pipe 8 for supplying the radioactive waste 2 into the tank 1, and the radioactive waste 2 in the tank 1 contains ozone, potassium permanganate, cerium (IV) compounds, etc. An oxidant supply pipe 9 for blowing the oxidant is provided.
酸化槽１の頂部には、酸化後のオフガスを、蒸留槽５内
の処理液（酸化後の放射性廃棄物）2aに流すオフガス排
出管10が接続される。An off-gas discharge pipe 10 is connected to the top of the oxidation tank 1 for flowing the off-gas after oxidation to the treatment liquid (radioactive waste after oxidation) 2a in the distillation tank 5.
蒸留槽５は、その外周にスチームジャケット、電気ヒー
タなどの加熱手段11が設けられ、また処理液2a中に空気
などのキャリアガスを吹き込むキャリアガス供給管12が
設けられ、さらに下部には処理液2aを排出してガラス固
化させるための排出管13が接続される。The distillation tank 5 is provided with a heating means 11 such as a steam jacket or an electric heater on the outer periphery thereof, a carrier gas supply pipe 12 for blowing a carrier gas such as air into the treatment liquid 2a, and the treatment liquid at a lower portion thereof. A discharge pipe 13 for discharging 2a and vitrifying it is connected.
次に放射性廃棄物２からルテニウムの分離回収を説明す
る。Next, separation and recovery of ruthenium from the radioactive waste 2 will be described.
先ず、処理液供給管８より酸化槽１内に放射性廃棄物２
が所定量供給される。この放射性廃棄物２中には、酸化
剤供給管９よりオゾン等の酸化剤が吹き込まれ、廃棄物
２中のルテニウムが酸化されて四酸化ルテニウムとされ
る。First, the radioactive waste 2 is introduced into the oxidation tank 1 through the treatment liquid supply pipe 8.
Is supplied in a predetermined amount. An oxidant such as ozone is blown into the radioactive waste 2 through the oxidant supply pipe 9, and ruthenium in the waste 2 is oxidized to ruthenium tetroxide.
この酸化処理は常温で行なわれ、また廃棄物２は硝酸酸
性の状態に保たれ、その硝酸濃度が2.5規定以上、好ま
しくは10規定（濃度40％）以上に保たれている。廃棄物
２中のルテニウムは酸化剤により酸化され四酸化ルテニ
ウムとなる（Ru＋2O2→RuO4）。この四酸化ルテニウム
は常温では液状態であり、廃棄物２中に溶け込んでい
る。This oxidation treatment is performed at room temperature, the waste 2 is kept in a nitric acid acidic state, and its nitric acid concentration is kept at 2.5 normal or higher, preferably 10 normal (concentration 40%) or higher. Ruthenium in waste 2 is oxidized by an oxidant to become ruthenium tetroxide (Ru + 2O 2 → RuO 4 ). This ruthenium tetroxide is in a liquid state at room temperature and is dissolved in the waste 2.
このように酸化処理を終えたのち、供給ポンプ４にて液
供給ライン３を介して酸化処理後の廃棄物を処理液2aと
して蒸留槽５内に供給する。After the oxidation treatment is completed in this way, the waste material after the oxidation treatment is supplied into the distillation tank 5 as the treatment liquid 2a through the liquid supply line 3 by the supply pump 4.
蒸留槽５内の処理液2aは、加熱手段11により四酸化ルテ
ニウムの分離温度又は沸点（93℃）以上すなわち50〜12
0℃或いはこれ以上に加熱され、またこの処理液2a中に
はキャリアガス供給管12からキャリアガスが供給され
る。The treatment liquid 2a in the distillation tank 5 is heated by the heating means 11 to a temperature above the separation temperature or boiling point (93 ° C.) of ruthenium tetroxide, that is, 50 to 12
The treatment liquid 2a is heated to 0 ° C. or higher, and a carrier gas is supplied into the treatment liquid 2a through a carrier gas supply pipe 12.
この状態で処理液2a中の四酸化ルテニウムは気化し、処
理液2aから分離し、キャリアガスと共に蒸留槽５の頂部
から排出され、四酸化ルテニウム回収装置６に導入され
る。この四酸化ルテニウム回収装置６は詳細は図示して
いないが、例えば四酸化ルテニウムをNaOH水溶液などの
吸収剤で吸収する吸収塔からなり、ガス中の四酸化ルテ
ニウムを吸収して回収する。また四酸化ルテニウムが回
収されたのちオフガスは、オフガス処理装置７で排気処
理される。In this state, the ruthenium tetroxide in the treatment liquid 2a is vaporized, separated from the treatment liquid 2a, discharged together with the carrier gas from the top of the distillation tank 5, and introduced into the ruthenium tetroxide recovery device 6. Although not shown in detail, the ruthenium tetroxide recovery device 6 is composed of an absorption tower that absorbs ruthenium tetroxide with an absorbent such as an aqueous NaOH solution, and absorbs and recovers ruthenium tetroxide in the gas. After the ruthenium tetroxide is recovered, the off gas is exhausted by the off gas processing device 7.
このように酸化槽１と蒸留槽５とに分けることで、例え
ば酸化槽１内では略一日かけて酸化処理したのち、蒸留
槽５に移し、次の日、再度新たな放射性廃棄物２を酸化
槽１で酸化処理している間に、任意の時間に蒸留槽５で
蒸留処理することができる。By dividing the oxidation tank 1 and the distillation tank 5 in this manner, for example, after the oxidation treatment is performed in the oxidation tank 1 for about one day, the oxidation tank 1 is transferred to the distillation tank 5 and the next day, a new radioactive waste 2 is again collected. Distillation can be carried out in the distillation tank 5 at an arbitrary time while the oxidation is carried out in the oxidation tank 1.
この場合、酸化槽１で酸化処理中に生じる酸化剤のオフ
ガスは、オフガス排出管10を介して、酸化の完了した蒸
留槽５内の処理液5a中に吹き込むようにしているが、こ
れを直接オフガス処理装置７に供給することにしてもよ
い。In this case, the off-gas of the oxidizing agent generated during the oxidation treatment in the oxidation tank 1 is blown into the treatment liquid 5a in the distillation tank 5 in which oxidation is completed through the off-gas discharge pipe 10. It may be supplied to the off-gas processing device 7.
第２図は、本発明の他の実施例を示すものである。本例
は基本的には、第１図の実施例て同様であるが、酸化槽
１の酸化処理中生じる未反応酸化剤であるオフガスをオ
フガスライン14を介してオフガス処理装置７に供給する
ようになし、また蒸留槽５内で処理液2a中の四酸化ルテ
ニウムを気化、分離させるにおいて、その処理液2aに蒸
気発生源15から吹込管16を介して水蒸気を吹き込み、水
蒸気蒸留を行なうようにした点で相違する。FIG. 2 shows another embodiment of the present invention. This example is basically the same as the example of FIG. 1, but the off gas which is the unreacted oxidant generated during the oxidation treatment of the oxidation tank 1 is supplied to the off gas treatment device 7 through the off gas line 14. In addition, when vaporizing and separating ruthenium tetroxide in the treatment liquid 2a in the distillation tank 5, steam is blown into the treatment liquid 2a from the vapor generation source 15 through the blowing pipe 16 to perform steam distillation. The difference is that you did.
本例においては、各槽1,5をオフガス系が完全に独立し
ているため、個々に酸化処理と蒸留処理が行なえる。ま
た処理液2a中に吹き込む水蒸気は、約120℃前後のもの
を用いる。この水蒸気の吹き込みにより四酸化ルテニウ
ムの気化・分離が効率よくできる。この場合、キャリア
ガス供給管12は水蒸気がキャリアガスとなるため、必ず
しも設ける必要はない。In this example, the off-gas system of each of the tanks 1 and 5 is completely independent, so that the oxidation treatment and the distillation treatment can be individually performed. Further, the steam blown into the treatment liquid 2a is about 120 ° C. By injecting this water vapor, ruthenium tetroxide can be vaporized and separated efficiently. In this case, the carrier gas supply pipe 12 does not necessarily have to be provided because water vapor serves as the carrier gas.
［発明の効果］ 以上説明してきたことから明らかなように本発明によれ
ば次のごとき優れた効果を発揮する。[Effects of the Invention] As is apparent from the above description, the present invention exhibits the following excellent effects.
（１） 放射性廃棄物中のルテニウムを酸化し、気化さ
せるにおいて、酸化槽と蒸留槽とに分けて設けたので、
個々の処理を別個に最適に操作できルテニウムの回収効
率を上げることができる。(1) Since the ruthenium in the radioactive waste is oxidized and vaporized, it is provided separately in the oxidation tank and the distillation tank.
The individual treatments can be optimally operated separately and the recovery efficiency of ruthenium can be increased.
（２） 酸化処理と蒸留処理とが別個に行なえるため処
理液を無駄に加熱したりすることがなくなる。(2) Since the oxidation treatment and the distillation treatment can be performed separately, the treatment liquid is not wastefully heated.
第１図は本発明の一実施例を示す図、第２図は本発明の
他の実施例を示す図である。 図中、１は酸化槽、２は放射性廃棄物、2aは酸化後の処
理液、５は蒸留槽、６は四酸化ルテニウム回収装置であ
る。FIG. 1 is a diagram showing an embodiment of the present invention, and FIG. 2 is a diagram showing another embodiment of the present invention. In the figure, 1 is an oxidation tank, 2 is a radioactive waste, 2a is a treatment liquid after oxidation, 5 is a distillation tank, and 6 is a ruthenium tetroxide recovery device.
Claims (1)
棄物中のルテニウムを四酸化ルテニウムに酸化する酸化
槽と、酸化後の放射性廃棄物を導入し、その廃棄物中に
溶け込んだ四酸化ルテニウムを加熱蒸留して四酸化ルテ
ニウムを気化させる蒸留槽とを備えたことを特徴とする
放射性廃棄物からのルテニウム分離装置。1. An oxidizing tank for supplying an oxidant to radioactive waste to oxidize ruthenium in the waste to ruthenium tetroxide and a radioactive waste after oxidation are introduced and dissolved in the waste. An apparatus for separating ruthenium from radioactive waste, comprising: a distillation tank for vaporizing ruthenium tetraoxide by heating and distilling ruthenium tetraoxide.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP8116187A JPH0769467B2 (en) | 1987-04-03 | 1987-04-03 | Ruthenium Separation Equipment from Radioactive Waste |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP8116187A JPH0769467B2 (en) | 1987-04-03 | 1987-04-03 | Ruthenium Separation Equipment from Radioactive Waste |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS63247698A JPS63247698A (en) | 1988-10-14 |
| JPH0769467B2 true JPH0769467B2 (en) | 1995-07-31 |
Family
ID=13738725
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP8116187A Expired - Lifetime JPH0769467B2 (en) | 1987-04-03 | 1987-04-03 | Ruthenium Separation Equipment from Radioactive Waste |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH0769467B2 (en) |
Families Citing this family (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2530025B2 (en) * | 1989-05-12 | 1996-09-04 | 日本碍子株式会社 | Generation method of ruthenium tetroxide |
| JP6343929B2 (en) * | 2013-12-26 | 2018-06-20 | 株式会社Ihi | Method and apparatus for removing ruthenium in high-level radioactive liquid waste |
-
1987
- 1987-04-03 JP JP8116187A patent/JPH0769467B2/en not_active Expired - Lifetime
Also Published As
| Publication number | Publication date |
|---|---|
| JPS63247698A (en) | 1988-10-14 |
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