JPS6227699A - Manufacture of radioactive waste intermediate storage body - Google Patents
Manufacture of radioactive waste intermediate storage bodyInfo
- Publication number
- JPS6227699A JPS6227699A JP16727585A JP16727585A JPS6227699A JP S6227699 A JPS6227699 A JP S6227699A JP 16727585 A JP16727585 A JP 16727585A JP 16727585 A JP16727585 A JP 16727585A JP S6227699 A JPS6227699 A JP S6227699A
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- Japan
- Prior art keywords
- radioactive waste
- intermediate storage
- storage body
- body according
- manufacturing
- Prior art date
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Abstract
(57)【要約】本公報は電子出願前の出願データであるた
め要約のデータは記録されません。(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.
Description
【発明の詳細な説明】
［発明の技術分野］
本発明は、原子力発電所や核燃料再処理１Ｍ段等の放射
性物質取り扱い施設から発生する中レベルないし低レベ
ルの放射性廃棄物中間貯蔵体の製造方法に関する。[Detailed Description of the Invention] [Technical Field of the Invention] The present invention relates to a method for producing an intermediate storage medium for intermediate to low level radioactive waste generated from facilities handling radioactive materials such as nuclear power plants and 1M stages of nuclear fuel reprocessing. Regarding.
［発明の技術的背景とその問題点コ
従来から、原子力発電所等の放射性物質取り扱い施設で
発生する放射性廃液の減容、同化処理法としでは、以下
のような、セメント固化法、アスファルト同化法、プラ
スチック同化法、ベレット化法、ガラス同化法等が知ら
れているが、それぞれ固有の欠点を有している。[Technical background of the invention and its problems] Conventionally, methods for volume reduction and assimilation of radioactive waste fluid generated in facilities handling radioactive materials such as nuclear power plants include the following cement solidification method and asphalt assimilation method. , plastic assimilation method, pelletization method, glass assimilation method, etc. are known, but each has its own drawbacks.
すなわち、
（イ）セメント同化法
この方法は、溶液状あるいはスラリー状の液状放射性廃
棄物を濃縮した後、セメントを固化剤として同化処理す
る方法である。(a) Cement assimilation method This method is a method in which liquid radioactive waste in the form of a solution or slurry is concentrated and then assimilated using cement as a solidifying agent.
このセメント固化法では、耐火性に優れ、機械的強度の
大きい同化体が得られる利点があるが、同化体の重量が
大きくなって取り扱いが困難となるうえに、同化体の発
生量が多くなるという欠点がある。This cement solidification method has the advantage of producing an assimilate with excellent fire resistance and high mechanical strength, but the weight of the assimilate increases, making it difficult to handle, and a large amount of assimilate is generated. There is a drawback.
（ロ）アスファルト同化法
この方法は、放射性廃棄物を濃縮した後、アスファルト
を同化剤として同化処理する方法である。(b) Asphalt assimilation method This method is a method in which radioactive waste is concentrated and then assimilated using asphalt as an assimilation agent.
この方法では、安価に同化体を形成し得る利点がおるが
、固化剤が可燃性であるために得られる同化体が耐火性
に乏しく、かつ機械的衝撃に弱いという欠点がある。This method has the advantage that the assimilate can be formed at low cost, but has the disadvantage that the assimilate obtained has poor fire resistance and is susceptible to mechanical shock because the solidifying agent is flammable.
（ハ）プラスチック固化法
この方法は、放射性廃棄物を濃縮乾燥して硫酸ナトリウ
ムを含む粉体とした後、熱硬化性のプラスチックを同化
剤として同化処理する方法である。(c) Plastic solidification method This method is a method in which radioactive waste is concentrated and dried to form a powder containing sodium sulfate, and then assimilated using thermosetting plastic as an assimilation agent.
この方法では、セメント同化法の場合のように、水分も
共に固化されることはなく、セメント同化法に比して同
化体の発生量を１１５程度にまで減容することができ、
機械的性質等に優れているという利点があるが、使用さ
れるプラスチックの単量体の多くが常温で揮発性である
ため、同化処理過程で火災発生の危険があり、また同化
剤が有機材料であるために得られる固化体が耐火性に乏
しくなるという欠点がある。In this method, unlike in the case of the cement assimilation method, water is not solidified as well, and the volume of assimilates generated can be reduced to about 115 compared to the cement assimilation method.
Although it has the advantage of having excellent mechanical properties, many of the monomers used in the plastics used are volatile at room temperature, so there is a risk of fire during the assimilation process, and the assimilation agent does not mix well with organic materials. This has the disadvantage that the resulting solidified product has poor fire resistance.
（ニ）ペレット化法
この方法は、放射性廃棄物を濃縮乾燥して硫酸ナトリウ
ムを含む放射性廃棄物とした後、これをペレット化して
中間貯蔵する方法である。(d) Pelletization method This method is a method in which radioactive waste is concentrated and dried to produce radioactive waste containing sodium sulfate, which is then pelletized and intermediately stored.
この方法によれば、同化体の発生量はプラスチック同化
法における発生量よりも減容される利点があるが、ペレ
ット化までに乾燥や成形等の数工程を必要とし、また大
型の設備を必要とするという欠点がある。This method has the advantage of reducing the volume of assimilates compared to the plastic assimilation method, but requires several steps such as drying and molding before pelletizing, and requires large equipment. There is a drawback that.
（ホ）ガラス同化法
この方法は、放射性廃棄物を濃縮乾燥して得られた粉体
をカラス化剤とともに加熱溶融さした後、冷却固化させ
る方法でおる。(e) Glass assimilation method In this method, a powder obtained by concentrating and drying radioactive waste is heated and melted together with a glassing agent, and then cooled and solidified.
この方法では、プラスチック同化法の１〜２倍の減容率
が得られる反面、加熱溶融温度が１２００℃以上と極め
て高温であるため、特殊な溶融装置を必要としその操作
や保守が難しいという欠点がおる。Although this method can achieve a volume reduction rate of 1 to 2 times the plastic assimilation method, the drawback is that it requires special melting equipment, which is difficult to operate and maintain because the heating melting temperature is extremely high, over 1200°C. There is.
［発明の目的］
本発明者等は、このような従来の欠点を解消すべく鋭意
研究をすすめた結果、原子力発電所等の放射性物質取り
扱い施設で発生する無機塩類を含む放射性廃棄物を、無
機塩類の融点以上の温度にまで加熱し溶融させて固融の
密度（真密度）になるまで収縮さけ、ここで生成した高
温の溶融物を冷却して固化させることにより、減容率が
高く飛散性のない中間貯蔵体が得られることを見出した
。[Purpose of the Invention] As a result of intensive research aimed at resolving these conventional drawbacks, the present inventors have discovered that radioactive waste containing inorganic salts generated at facilities handling radioactive materials such as nuclear power plants can be treated with inorganic salts. By heating and melting salts to a temperature higher than their melting point and shrinking them until they reach the density of a solid (true density), the high-temperature molten material produced here is cooled and solidified, resulting in a high volume reduction rate and scattering. It has been found that a neutral intermediate storage body can be obtained.
すなわち、たとえばＢＷＲ原子力発電所で発生する液体
放射性廃棄物の中には、硫酸ナトリウムが固形分比で５
０〜１００重量％の割合で含まれており、この硫酸ナト
リウムは８８４℃以上の温度で溶融する。この硫酸ナト
リウム溶融物は、粘性が極めて低くかつ固有の密度（真
密度）まで収縮しており、その後の冷却により粉体化す
ることなくＮａ＋と５Ｏ４−−がイオン結合した結晶同
化体を生成する。For example, liquid radioactive waste generated at a BWR nuclear power plant contains sodium sulfate with a solid content of 5.
It is contained in a proportion of 0 to 100% by weight, and this sodium sulfate melts at a temperature of 884°C or higher. This sodium sulfate melt has an extremely low viscosity and has shrunk to a specific density (true density), and upon subsequent cooling, it produces a crystalline assimilate in which Na+ and 5O4-- are ionically bonded without turning into powder. .
本発明はこのような知見に基いてなされたもので、上述
した従来の放射性廃棄物の同化法の欠点を解消した放射
性廃棄物中間貯蔵体の製造方法を提供することを目的と
する。The present invention was made based on such knowledge, and an object of the present invention is to provide a method for manufacturing a radioactive waste intermediate storage body that eliminates the drawbacks of the conventional radioactive waste assimilation method described above.
［発明の概要］
すなわち本発明の放射性廃棄物中間貯蔵体の製造方法は
、原子力施設で発生した無機塩類を含む放射性廃棄物を
、前記無機塩類の融点以上にまで加熱して溶融させた後
、この溶融物を耐熱耐食性のキャニスタ−内で冷却して
固化させることを特徴とする第１の発明と、無機塩類を
含む放射性廃棄物を金属キャニスタ−に入れ、このキレ
ニスターを前記無機塩類の融点以上の温度まで高周波誘
導加熱して前記放射性廃棄物を溶融させた後、溶融物を
前記キャニスタ−中で冷却して固化させることを特徴と
する第２の発明に関するものである。[Summary of the Invention] That is, the method for manufacturing a radioactive waste intermediate storage body of the present invention involves heating and melting radioactive waste containing inorganic salts generated at a nuclear facility to a temperature equal to or higher than the melting point of the inorganic salts; The first invention is characterized in that the molten material is cooled and solidified in a heat-resistant and corrosion-resistant canister, and the radioactive waste containing inorganic salts is placed in a metal canister, and the melt is heated to a temperature higher than the melting point of the inorganic salts. The second invention is characterized in that, after the radioactive waste is melted by high-frequency induction heating to a temperature of , the melt is cooled and solidified in the canister.
第１の発明においては、まず原子力施設で発生した無機
塩類たとえば硫酸ナトリウムを含む液体放射性廃棄物を
常法によって乾燥して粉体化する。In the first invention, first, liquid radioactive waste containing inorganic salts such as sodium sulfate generated at a nuclear facility is dried and powdered by a conventional method.
このとき液体放射性廃棄物中にイオン交換樹脂フィルタ
等の有機質物質がスラリー状で分散されていてもよい。At this time, an organic substance such as an ion exchange resin filter may be dispersed in the liquid radioactive waste in the form of a slurry.
通常、５Ａ酸ナトリウムを乾燥させると、粒径が約１０
０μｍで見かけ密度が約１．０の粉体となる。Normally, when sodium 5A acid is dried, the particle size is about 10
At 0 μm, the powder has an apparent density of about 1.0.
次にこの乾燥粉体を融点以上の温度で加熱して溶融させ
る。硫酸ナトリウムの粉体は、８８４℃（融点）以上の
温度に加熱されると低粘度で真密度を有する溶融物とな
る。Next, this dry powder is heated to a temperature above its melting point to melt it. When a powder of sodium sulfate is heated to a temperature of 884° C. (melting point) or higher, it becomes a melt having a low viscosity and true density.
次いで、いかなる同化材も添加することなく、そのまま
冷却することにより粉体化せずに塊状の同化体となる。Next, by cooling the mixture as it is without adding any assimilating material, it becomes a lump-like assimilated material without being pulverized.
なお、液体放射性廃棄物には、このような無機塩類の他
に鉄を主成分とする金属腐食生成物がわずかに含まれて
いるが、これは無機塩類の溶融温度までに安定な酸化物
となり、溶融物中に均一に分散されるので、これが同化
処理に影響を及ぼすことはない。In addition to these inorganic salts, liquid radioactive waste contains a small amount of metal corrosion products whose main component is iron, but these become stable oxides at the melting temperature of the inorganic salts. , this does not affect the assimilation process since it is uniformly dispersed in the melt.
こうして生成された同化体は、表面積が粉体よりはるか
に小さく水にかなり溶けにくくなっているが、長期間湿
気にさらされたり水中に浸漬したりすると溶出して好ま
しくない。したがって同化体は最終的に処分されるまで
の間、耐熱性があり、かつ中性塩である同化体と反応し
ないセラミックや金属からなるキャニスタ−内に封止し
て保管することが望ましい。キャニスタ−内への充填は
溶融物を溶融状態・を保持しながら行ってもよいが、溶
融炉をそのままキャニスタ−とし、溶融物が炉内に充填
された時に加熱を停止するとともにそのまま冷却する方
法をとることが望ましい。The assimilate thus produced has a much smaller surface area than powder and is considerably less soluble in water, but it is undesirable because it will elute if exposed to moisture or immersed in water for a long period of time. Therefore, it is desirable that the assimilate is sealed and stored in a canister made of ceramic or metal, which is heat resistant and does not react with the assimilate, which is a neutral salt, until it is finally disposed of. The canister may be filled with the molten material while maintaining it in a molten state, but there is a method in which the melting furnace is used as the canister, and when the molten material is filled into the furnace, heating is stopped and the melt is cooled as it is. It is desirable to take
第２の本発明においては、このような加熱溶融および冷
却手段をとる場合の熱源として、高周波誘導加熱装置を
備えたステンレス鋼やＮｉ −Ｃｒ系合金のような金属
るつぼが溶融炉として用いられる。In the second aspect of the invention, a metal crucible such as stainless steel or Ni-Cr alloy equipped with a high-frequency induction heating device is used as a melting furnace as a heat source when such heating melting and cooling means are used.
すなわち、金属るつぼを高周波誘導加熱により直接加熱
し、その輻射熱で中にいれた硫酸ナトリウム等の乾燥粉
体を溶融し、るつぼ内が溶融物で満たされた時点で加熱
を停止してそのまま冷却する。このような金属製るつぼ
を用いてこのるつぼをキャニスタ−として兼用さける場
合には、るつぼの酸化を防止するために加熱溶融および
冷却を窒素ガスやアルゴンガスのような不活性ガス雰囲
気中で行うことが望ましい。また、不活性ガス雰囲気中
のステンレス鋼は１２００℃以下の温度では酸化がみら
れず、（ラボスケールの実験）、ざらに硫酸ナトリウム
の溶融物は１２００’Ｃ以上の温度で少しずつ蒸発する
ことから、加熱溶融温度は、８８４〜１２００″Ｃの範
囲とすることが望ましい。In other words, a metal crucible is directly heated by high-frequency induction heating, and the radiant heat melts the dry powder such as sodium sulfate, and when the crucible is filled with the molten material, the heating is stopped and the crucible is allowed to cool. . When using such a metal crucible and avoiding using it as a canister, heat melting and cooling should be performed in an inert gas atmosphere such as nitrogen gas or argon gas to prevent oxidation of the crucible. is desirable. In addition, stainless steel in an inert gas atmosphere does not oxidize at temperatures below 1200'C (laboratory scale experiment), and molten sodium sulfate evaporates little by little at temperatures above 1200'C. Therefore, it is desirable that the heating melting temperature is in the range of 884 to 1200''C.
またさらにこのような金属るつぼを高周波誘導加熱の方
法により加熱する場合には、高周波電源の周波数は、内
径が２０〜６０Ｃｍの普通の大ぎざのるつぼで、１０〜
４００ ｋＨｚの範囲が適当である。電源周波数が１０
ｋＨｚ未満のときには、溶融塩の表面が上部方向にふく
らんで、るつぼ内に粉体を充分に供給できなくなり、４
００ｋＨｚを越えると発熱が不十分になるので好ましく
ない。Furthermore, when heating such a metal crucible by a high frequency induction heating method, the frequency of the high frequency power source is 10 to
A range of 400 kHz is suitable. Power frequency is 10
When the frequency is less than kHz, the surface of the molten salt bulges upward, making it impossible to supply powder into the crucible, resulting in
If it exceeds 00 kHz, heat generation becomes insufficient, which is not preferable.
本発明においては、このようにして無機塩類を含む放射
性廃棄物が固化処理され放射能飛散性のない中間貯蔵体
が生成されるが、この方法はＢＷＲ原子力発電所で発生
する硫酸ナトリウムを含む放射性廃棄物ばかりでなく、
ＰＷＲ原子力発電所から発生するホウ酸ナトリウムを含
む放射性廃棄物の処理方法としても有用である。In the present invention, radioactive waste containing inorganic salts is solidified in this way to produce an intermediate storage medium that does not scatter radioactivity. Not only waste
It is also useful as a method for treating radioactive waste containing sodium borate generated from PWR nuclear power plants.
なお以上は、これらの無機塩類を含む液体放射性廃棄物
を一旦濃縮乾燥して得られた粉体を、溶融炉に供給し加
熱溶融させた例であるが、本発明は必ずしもこれに限定
されるものではなく、あらかじめ溶融炉内で少量の硫酸
ナトリウム等を溶融させておき、その溶融面に液体放射
性廃棄物をそのまま供給して水分を蒸発させつつ放射性
廃棄物中の無機塩類を溶融する方法をとることもできる
。The above is an example in which a powder obtained by once concentrating and drying liquid radioactive waste containing these inorganic salts is supplied to a melting furnace and heated and melted, but the present invention is not necessarily limited to this. Instead of melting a small amount of sodium sulfate, etc. in a melting furnace in advance, liquid radioactive waste is directly supplied to the melted surface, and the water is evaporated while melting the inorganic salts in the radioactive waste. You can also take it.
この方法では、液体放射性廃棄物中に含まれる水の蒸発
により大きなエネルギーが失われるが、反面乾燥工程を
省くことができるという利点がある。In this method, a large amount of energy is lost due to the evaporation of the water contained in the liquid radioactive waste, but on the other hand, it has the advantage that a drying step can be omitted.
［発明の実施例］ 以下本発明の実施例について説明する。[Embodiments of the invention] Examples of the present invention will be described below.
実施例１〜５
ＢＷＲ原子力発電所で発生した液体放射性廃棄物を通常
の竪型薄膜乾燥機にかけて乾燥し、次表に示す組成およ
び比重を有する乾燥粉体を得た。Examples 1 to 5 Liquid radioactive waste generated at the BWR nuclear power plant was dried in a conventional vertical thin film dryer to obtain dry powder having the composition and specific gravity shown in the following table.
次いで得られた乾燥粉体をそれぞれ電気炉中に投入し、
８９０〜１２００℃の温度で加熱して溶融した後、これ
らの溶融物をただちに鉄製の鋳型に注入してそのまま冷
却した。Next, each of the obtained dry powders was put into an electric furnace,
After heating and melting at a temperature of 890 to 1200° C., the melt was immediately poured into an iron mold and allowed to cool.
冷却後いずれも全体が一体にかたまり塊状に固化した同
化体が得られた。こうして得られた同化体の比重および
圧縮強度を測定した。測定結果を次表に示す。After cooling, assimilates were obtained in which the whole was solidified into a lump. The specific gravity and compressive strength of the assimilate thus obtained were measured. The measurement results are shown in the table below.
（以下余白）
この測定結果から、粉体を加熱溶融した後冷却固化させ
ることにより、比重が約３倍すなわち容積が約１／３に
なり、硫酸ナトリウムを含む放射性廃棄物の著しい減容
比が図られることがわかる。(Left below) From this measurement result, by heating and melting the powder and then cooling and solidifying it, the specific gravity becomes about 3 times as large, or the volume becomes about 1/3, resulting in a significant volume reduction ratio for radioactive waste containing sodium sulfate. I can see that this is being planned.
また減容された固化体は、鋳型にいれたままの状態で中
間貯蔵体として安全に保管することができる。Further, the volume-reduced solidified body can be safely stored as an intermediate storage body while remaining in the mold.
［発明の効果］
以上説明から明らかなように、本発明の放射性廃棄物中
間貯蔵体の製造方法によれば、放射性廃棄物を高い減容
率で処理することができる。また生成した同化体は化学
的に安定で飛散性がないので、耐熱耐食性のキャニスタ
−に収容した状態で中間貯蔵体として安全に保管するこ
とができる。[Effects of the Invention] As is clear from the above description, according to the method for manufacturing a radioactive waste intermediate storage body of the present invention, radioactive waste can be processed at a high volume reduction rate. Furthermore, since the produced assimilate is chemically stable and non-scattering, it can be safely stored as an intermediate storage medium in a heat-resistant and corrosion-resistant canister.
ざらに処理温度が比較的低くて大型あるいは特殊な設備
を必要としないという利点がある。Another advantage is that the processing temperature is relatively low and large or special equipment is not required.
Claims (12)
物を、前記無機塩類の融点以上にまで加熱して溶融させ
た後、この溶融物を耐熱耐食性のキャニスター内で冷却
して固化させることを特徴とする放射性廃棄物中間貯蔵
体の製造方法。(1) Radioactive waste containing inorganic salts generated at nuclear facilities is heated to a temperature higher than the melting point of the inorganic salts to melt it, and then the molten material is cooled and solidified in a heat-resistant and corrosion-resistant canister. A method for producing a characterized intermediate storage body for radioactive waste.
囲第１項記載の放射性廃棄物中間貯蔵体の製造方法。(2) The method for producing a radioactive waste intermediate storage body according to claim 1, wherein the inorganic salt is sodium sulfate.
範囲第１項記載の放射性廃棄物中間貯蔵体の製造方法。(3) The method for producing a radioactive waste intermediate storage body according to claim 1, wherein the inorganic salt is sodium borate.
体としたものである特許請求の範囲第１項ないし第３項
のいずれか１項記載の放射性廃棄物中間貯蔵体の製造方
法。(4) The method for manufacturing a radioactive waste intermediate storage body according to any one of claims 1 to 3, wherein the radioactive waste is obtained by drying liquid radioactive waste and turning it into powder. .
℃である特許請求の範囲第１項ないし第４項のいずれか
１項記載の放射性廃棄物中間貯蔵体の製造方法。(5) The heating melting temperature of radioactive waste is 884-1200
The method for manufacturing a radioactive waste intermediate storage body according to any one of claims 1 to 4, wherein the temperature is .degree.
不活性ガス雰囲気中で行われる特許請求の範囲第１項記
載の放射性廃棄物中間貯蔵体の製造方法。(6) The method for manufacturing a radioactive waste intermediate storage body according to claim 1, wherein the heating melting and cooling of the radioactive waste are both performed in an inert gas atmosphere.
ーに入れ、このキャニスターを前記無機塩類の融点以上
の温度まで高周波誘導加熱して前記放射性廃棄物を溶融
させた後、溶融物を前記キャニスター中で冷却して固化
させることを特徴とする放射性廃棄物中間貯蔵体の製造
方法。(7) Put radioactive waste containing inorganic salts into a metal canister, heat the canister with high-frequency induction to a temperature higher than the melting point of the inorganic salts to melt the radioactive waste, and then pour the molten material into the canister. 1. A method for producing an intermediate storage body for radioactive waste, which comprises cooling and solidifying the intermediate storage body for radioactive waste.
囲第７項記載の放射性廃棄物中間貯蔵体の製造方法。(8) The method for manufacturing a radioactive waste intermediate storage body according to claim 7, wherein the inorganic salt is sodium sulfate.
範囲第７項記載の放射性廃棄物中間貯蔵体の製造方法。(9) The method for producing a radioactive waste intermediate storage body according to claim 7, wherein the inorganic salt is sodium borate.
粉体としたものである特許請求の範囲第７項ないし第９
項のいずれか１項記載の放射性廃棄物中間貯蔵体の製造
方法。(10) Claims 7 to 9, wherein the radioactive waste is obtained by drying liquid radioactive waste and turning it into powder.
2. A method for producing a radioactive waste intermediate storage body according to any one of the above items.
００℃である特許請求の範囲第７項ないし第１０項のい
ずれか１項記載の放射性廃棄物中間貯蔵体の製造方法。(11) The heating melting temperature of radioactive waste is 884-12
The method for manufacturing a radioactive waste intermediate storage body according to any one of claims 7 to 10, wherein the temperature is 00°C.
も不活性ガス雰囲気中で行われる特許請求の範囲第６項
記載の放射性廃棄物中間貯蔵体の製造方法。(12) The method for manufacturing a radioactive waste intermediate storage body according to claim 6, wherein both heating and melting of the radioactive waste and cooling are performed in an inert gas atmosphere.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP16727585A JPS6227699A (en) | 1985-07-29 | 1985-07-29 | Manufacture of radioactive waste intermediate storage body |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP16727585A JPS6227699A (en) | 1985-07-29 | 1985-07-29 | Manufacture of radioactive waste intermediate storage body |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPS6227699A true JPS6227699A (en) | 1987-02-05 |
Family
ID=15846721
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP16727585A Pending JPS6227699A (en) | 1985-07-29 | 1985-07-29 | Manufacture of radioactive waste intermediate storage body |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS6227699A (en) |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US5033533A (en) * | 1989-03-30 | 1991-07-23 | Framatome | Method for producing a container for contaminated metal waste, and a container produced by this method |
| EP0640992A1 (en) * | 1993-08-25 | 1995-03-01 | Doryokuro Kakunenryo Kaihatsu Jigyodan | Method of melting treatment of radioactive miscellaneous solid wastes |
-
1985
- 1985-07-29 JP JP16727585A patent/JPS6227699A/en active Pending
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US5033533A (en) * | 1989-03-30 | 1991-07-23 | Framatome | Method for producing a container for contaminated metal waste, and a container produced by this method |
| EP0640992A1 (en) * | 1993-08-25 | 1995-03-01 | Doryokuro Kakunenryo Kaihatsu Jigyodan | Method of melting treatment of radioactive miscellaneous solid wastes |
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