JPH0566289A - Method for separating technetium 99 in high level radioactive waste solution - Google Patents
Method for separating technetium 99 in high level radioactive waste solutionInfo
- Publication number
- JPH0566289A JPH0566289A JP22668891A JP22668891A JPH0566289A JP H0566289 A JPH0566289 A JP H0566289A JP 22668891 A JP22668891 A JP 22668891A JP 22668891 A JP22668891 A JP 22668891A JP H0566289 A JPH0566289 A JP H0566289A
- Authority
- JP
- Japan
- Prior art keywords
- technetium
- elements
- level radioactive
- waste solution
- anion exchange
- 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.)
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Classifications
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- 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
-
- 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
- Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
- Y02W30/00—Technologies for solid waste management
- Y02W30/50—Reuse, recycling or recovery technologies
Landscapes
- Inorganic Compounds Of Heavy Metals (AREA)
- Removal Of Specific Substances (AREA)
- Manufacture And Refinement Of Metals (AREA)
Abstract
Description
【０００１】[0001]
【産業上の利用分野】本発明は核燃料再処理施設から発
生する高レベル放射性廃液中のテクネチウム９９核種の
分離方法に関する。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for separating technetium-99 nuclides from a high-level radioactive liquid waste generated from a nuclear fuel reprocessing facility.
【０００２】[0002]
【従来の技術】核燃料再処理施設で使用済み核燃料から
ウランとプルトニウムを回収した後の高レベル放射性廃
液中には、セシウム（Cs）等のアルカリ金属元素、スト
ロンチウム（Sr）、バリウム（Ba）等のアルカリ土類金
属元素、セリウム（Ce）、ユーロピウム（Eu）、プラセ
オジム（Pr）等の希土類元素、α放射体で放射能強度は
弱いが長半減期核種であるウラン(U) 、プルトニウム
（Pu）、アメリシウム（Am）、キュリウム（Cm）等のア
クチニド元素、ルテニウム（Ru）、ロジウム（Rh）、パ
ラジウム（Pd）等の有用貴金属元素である白金族元素、
テクネチウム（Tc）およびモリブデン（Mo）その他が溶
解しているが、従来の高レベル放射性廃液の処理はこれ
らのを元素群を一括してガラス固化体として処分してい
た。2. Description of the Related Art Alkali metal elements such as cesium (Cs), strontium (Sr), barium (Ba), etc. are contained in high-level radioactive liquid waste after uranium and plutonium are recovered from spent nuclear fuel at a nuclear fuel reprocessing facility. Alkaline earth metal elements, rare earth elements such as cerium (Ce), europium (Eu), praseodymium (Pr), and uranium (U) and plutonium (Pu) ), Americium (Am), curium (Cm) and other actinide elements, ruthenium (Ru), rhodium (Rh), palladium (Pd) and other useful precious metal elements, platinum group elements,
Although technetium (Tc), molybdenum (Mo), etc. are dissolved, in the conventional treatment of high-level radioactive liquid waste, these elements were collectively disposed as vitrified solids.
【０００３】[0003]
【発明が解決しようとする課題】しかしながら、このよ
うな高レベル放射性廃液中には強いβ放射体で超長半減
期核分裂生成物であるテクネチウム９９が含まれてお
り、貯蔵中にキャニスタが経年変化、腐食または地殻変
動等で破損した場合、テクネチウム９９の浸出と地層移
動により環境・人体に対して長期間にわたる影響を与え
る恐れがある。また、放射能レベルや発熱量または半減
期の異なる元素群の一括処理は廃棄物固化体の増加、貯
蔵年数の長期化につながり、最終的な生活圏からの隔離
である地層処分技術そのものの実現を困難にする課題が
あった。このようなことから、放射能レベルや発熱量ま
たは半減期の異なる元素群をそれぞれ化学的・物理的特
性にあわせて分離する技術開発が必要となっている。However, such high-level radioactive liquid waste contains technetium-99, which is a strong β-emitter and a very long half-life fission product, and the canister changes over time during storage. When damaged by corrosion, crustal movement, etc., leaching of technetium 99 and movement of the stratum may have a long-term effect on the environment and human body. In addition, collective treatment of elements with different levels of radioactivity, calorific value, or half-life leads to an increase in solid waste and a longer storage period, and the realization of the geological disposal technology itself, which is the final isolation from living areas. There was a problem that made it difficult. For this reason, it is necessary to develop a technology for separating element groups having different radioactivity levels, calorific values, or half-lives according to their chemical and physical characteristics.
【０００４】本発明は、かかる点に対処してなされたも
ので、核燃料再処理施設から発生する高レベル放射性廃
液中の核分裂生成物をアクチニド元素および希土類元素
のグループとアルカリ金属元素、アルカリ土類金属元
素、モリブデン、テクネチウム９９および白金族元素の
グループに群分離し、さらにアルカリ金属元素、アルカ
リ土類金属元素、モリブデン、テクネチウム９９および
白金族元素のグループ中のテクネチウム９９を分離回収
する方法を提供することを目的とするものである。The present invention has been made in consideration of such a point, and the fission products in the high-level radioactive liquid waste generated from the nuclear fuel reprocessing facility are treated with a group of actinide elements and rare earth elements, alkali metal elements and alkaline earth elements. A method for separating into groups of metal elements, molybdenum, technetium 99 and platinum group elements, and further separating and recovering alkali metal elements, alkaline earth metal elements, molybdenum, technetium 99 and technetium 99 in groups of platinum group elements is provided. The purpose is to do.
【０００５】[0005]
【課題を解決するための手段】すなわち、本発明の高レ
ベル放射性廃液中のテクネチウム９９の分離方法は、核
燃料再処理施設から発生するアクチニド元素、希土類元
素、アルカリ金属元素、アルカリ土類金属元素、テクネ
チウム９９および白金族元素等が溶解している高レベル
放射性廃液中より湿式分離法で前記アクチニド元素およ
び希土類元素を分離除去した後、前記アルカリ金属元
素、アルカリ土類金属元素、テクネチウム９９および白
金族元素等を含む廃液を陰イオン交換樹脂に接触させて
前記廃液中に唯一陰イオン化学種として存在するテクネ
チウム９９核種を選択的に吸着させることを特徴とす
る。That is, the method for separating technetium 99 in a high-level radioactive waste liquid according to the present invention is performed by an actinide element, a rare earth element, an alkali metal element, an alkaline earth metal element generated from a nuclear fuel reprocessing facility, After the actinide element and the rare earth element are separated and removed by a wet separation method from a high-level radioactive waste liquid in which technetium 99 and a platinum group element and the like are dissolved, the alkali metal element, the alkaline earth metal element, the technetium 99 and the platinum group element are removed. It is characterized in that a waste liquid containing an element or the like is brought into contact with an anion exchange resin to selectively adsorb technetium 99 nuclide which is the only anionic chemical species present in the waste liquid.
【０００６】[0006]
【作用】核燃料再処理施設で発生する高レベル放射性廃
液に含まれるアクチニド元素および希土類元素は湿式分
離法で塩として沈澱し、アルカリ金属元素、アルカリ土
類金属元素、テクネチウム９９および白金族元素等はロ
液もしくは水層に残留する。このロ液もしくは水層部分
を陰イオン交換工程に導入することによって、陰イオン
化学種として存在するテクネチウム９９のみを分離回収
する。[Function] Actinide elements and rare earth elements contained in the high-level radioactive waste liquid generated at the nuclear fuel reprocessing facility are precipitated as salts by the wet separation method, and alkali metal elements, alkaline earth metal elements, technetium 99, platinum group elements, etc. It remains in the liquid or aqueous layer. By introducing this liquid or water layer portion into the anion exchange step, only technetium 99 existing as an anion chemical species is separated and recovered.
【０００７】[0007]
【実施例】以下、図１に示す本発明の一実施例について
説明する。DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention shown in FIG. 1 will be described below.
【０００８】図１は、高レベル放射性廃液からテクネチ
ウムを選択的に分離回収し、消滅処理に持ち込むまでの
処理工程を示すもので、高レベル放射性廃液１にシュウ
酸を添加して、廃液１中のアクチニド元素および希土類
元素を沈殿させ、溶液相２と沈殿相３に分離する湿式分
離工程４と、アルカリ金属元素、アルカリ土類金属元
素、白金族元素、テクネチウム９９およびモリブデンそ
の他を含む溶液相を陰イオン交換樹脂に接触させてテク
ネチウム９９のみ吸着させ分離するイオン交換工程５
と、分離したテクネチウム９９を軽水炉（ＬＷＲ）等に
て核種変換する消滅処理工程６とから成っている。次に
本実施例の各処理工程について詳細に説明する。[0008] Fig. 1 shows a treatment process of selectively separating and recovering technetium from a high-level radioactive waste liquid and bringing it to an extinction process. In the waste liquid 1, oxalic acid was added to the high-level radioactive waste liquid 1. Of the actinide element and the rare earth element, and a wet separation step 4 of separating into a solution phase 2 and a precipitation phase 3 and a solution phase containing an alkali metal element, an alkaline earth metal element, a platinum group element, technetium 99, molybdenum and the like. Ion exchange step 5 of contacting with anion exchange resin and adsorbing and separating only technetium 99
And an extinction treatment step 6 for converting the separated technetium 99 into a nuclide in a light water reactor (LWR) or the like. Next, each processing step of this embodiment will be described in detail.
【０００９】本実施例の湿式分離工程４においては、シ
ュウ酸を試薬とする沈澱法を適用する。高レベル放射性
廃液１中のアクチニド元素はＵ4+、Ｐｕ3+、Ｎｐ4+、Ａ
ｍ3+、Ｃｍ3+の状態で溶解しており、希土類元素ととも
にシュウ酸塩として沈澱する。なお、前工程として高レ
ベル放射性廃液１中に含まれている硝酸をギ酸の還元作
用で分解する脱硝工程が従来必要とされていたが、沈殿
試薬として用いるシュウ酸には還元作用があるため、本
実施例ではこのシュウ酸を大過剰量添加することによ
り、シュウ酸によって硝酸を分解せしめ、ギ酸による脱
硝工程を削減することが可能となる。高レベル放射性廃
液１においては、Ｕ4+およびＮｐ4+はシュウ酸により Ｕ4+ →Ｕ3+ Ｎｐ4+→Ｎｐ3+ まで還元され、前記アクチニド元素はそれぞれ、以下の
化学反応式 ２Ｕ3+ ＋３（ＣＯＯＨ）2 →Ｕ2 （Ｃ2 Ｏ4 ）3 ↓＋６Ｈ+ ２Ｐｕ3+＋３（ＣＯＯＨ）2 →Ｐｕ2 （Ｃ2 Ｏ4 ）3 ↓＋６Ｈ+ ２Ｎｐ3+＋３（ＣＯＯＨ）2 →Ｎｐ2 （Ｃ2 Ｏ4 ）3 ↓＋６Ｈ+ ２Ａｍ3+＋３（ＣＯＯＨ）2 →Ａｍ2 （Ｃ2 Ｏ4 ）3 ↓＋６Ｈ+ ２Ｃｍ3+＋３（ＣＯＯＨ）2 →Ｃｍ2 （Ｃ2 Ｏ4 ）3 ↓＋６Ｈ+ にしたがって、シュウ酸ウラナス、シュウ酸プルトニウ
ム、シュウ酸ネプツニウム、シュウ酸アメリシウム、シ
ュウ酸キュリウムとして沈澱する。希土類元素はシュウ
酸の添加により、以下に反応式を示すように、 ２Ｃｅ3+＋３（ＣＯＯＨ）2 →Ｃｅ2 （Ｃ2 Ｏ4 ）3 ↓＋６Ｈ+ ２Ｅｕ3+＋３（ＣＯＯＨ）2 →Ｅｕ2 （Ｃ2 Ｏ4 ）3 ↓＋６Ｈ+ ２Ｐｒ3+＋３（ＣＯＯＨ）2 →Ｐｒ2 （Ｃ2 Ｏ4 ）3 ↓＋６Ｈ+ シュウ酸セリウム、シュウ酸ユーロピウム、シュウ酸プ
ラセオジムとなり沈澱する。このようにして生じた沈澱
物はロ別され、長半減期核種であるアクチニド元素を含
むため、例えば、ガラス個化体に個化処理されて貯蔵さ
れる。In the wet separation step 4 of this embodiment, a precipitation method using oxalic acid as a reagent is applied. The actinide elements in the high-level radioactive liquid waste 1 are U 4+ , Pu 3+ , Np 4+ , A
It is dissolved in the state of m 3+ and Cm 3+ and precipitates as an oxalate together with the rare earth element. A denitration step of decomposing nitric acid contained in the high-level radioactive waste liquid 1 by a reducing action of formic acid has been conventionally required as a previous step, but oxalic acid used as a precipitation reagent has a reducing action. In this embodiment, by adding a large excess amount of oxalic acid, nitric acid is decomposed by oxalic acid, and it becomes possible to reduce the denitration step using formic acid. In the high level radioactive liquid waste 1, U 4+ and Np 4+ are reduced to U 4+ → U 3+ Np 4+ → Np 3+ by oxalic acid, and the actinide elements are respectively represented by the following chemical reaction formula 2U 3 + +3 (COOH) 2 → U 2 (C 2 O 4 ) 3 ↓ + 6H + 2Pu 3+ +3 (COOH) 2 → Pu 2 (C 2 O 4 ) 3 ↓ + 6H + 2Np 3+ +3 (COOH) 2 → Np 2 (C 2 O 4) 3 ↓ + 6H + 2Am 3+ +3 (COOH) 2 → Am 2 (C 2 O 4) 3 ↓ + 6H + 2Cm 3+ +3 (COOH) 2 → Cm 2 (C 2 O 4) 3 According to ↓ + 6H + , it precipitates as Uranus oxalate, Plutonium oxalate, Neptunium oxalate, Americium oxalate, Curium oxalate. By adding oxalic acid, the rare earth element is 2Ce 3+ +3 (COOH) 2 → Ce 2 (C 2 O 4 ) 3 ↓ + 6H + 2Eu 3+ +3 (COOH) 2 → Eu 2 as shown in the reaction formula below. (C 2 O 4 ) 3 ↓ + 6H + 2Pr 3+ +3 (COOH) 2 → Pr 2 (C 2 O 4 ) 3 ↓ + 6H + Cerium oxalate, europium oxalate, praseodymium oxalate and precipitate. The precipitate thus produced is separated by filtration and contains the actinide element, which is a long half-life nuclide, so that it is singulated into a vitrified product and stored.
【００１０】上記沈澱物を除去したあとのロ液には、ア
ルカリ金属元素・アルカリ土類金属元素がＣｓ+ 、Ｓｒ
2+、Ｂａ2+の状態で溶解している。白金族元素およびそ
の他の元素ではＴｃがＴｃＯ4-であり、Ｒｕ、Ｒｈ、Ｐ
ｄは酸化物、水酸化物の形態で溶解している。陰イオン
交換工程５では上記ロ液を陰イオン交換カラムに導入し
て、ＴｃＯ4-なる形態のテクネチウム９９を吸着させ
る。陰イオン交換樹脂としては交換速度が速く吸着能の
大きいポーラス型の樹脂を採用する。ＴｃＯ4-の吸着は
希硝酸（低ｐＨ域）で行い、また溶出は濃硝酸（高ｐＨ
域）で行う。After the precipitate is removed, the solution containing alkali metal elements and alkaline earth metal elements is Cs +. , Sr
It is dissolved in the state of 2+ and Ba 2+ . In the platinum group element and other elements, Tc is TcO 4- , and Ru, Rh, P
d is dissolved in the form of oxide or hydroxide. In the anion exchange step 5, the above filtrate is introduced into the anion exchange column to adsorb technetium 99 in the form of TcO 4- . As the anion exchange resin, a porous type resin having a high exchange rate and a large adsorption capacity is adopted. TcO 4- is adsorbed with dilute nitric acid (low pH range), and eluted with concentrated nitric acid (high pH range).
Area).
【００１１】このようにして分離したテクネチウムは、
次の消滅処理工程６において、例えば燃料中に混入され
て軽水炉（ＬＷＲ）等によって核種変換される。すなわ
ち、中性子の照射により超長半減期核分裂生成物である
テクネチウム９９は安定同位体であるルテニウム９９
（Ｒｕ−９９）に変換される。The technetium thus separated is
In the next extinction processing step 6, for example, it is mixed into fuel and converted into a nuclide by a light water reactor (LWR) or the like. That is, technetium-99, which is an ultralong half-life fission product by neutron irradiation, is ruthenium-99, which is a stable isotope.
(Ru-99).
【００１２】このような高レベル放射性廃液の処理方法
を用いれば、固化処分体の減少および隔離期間の短縮、
さらに地層処分技術をより具体的な技術として実現する
ことができる。表１に、上記実施例にしたがって高レベ
ル放射性廃液を処理し、アクチニド元素および希土類元
素のみガラス固化体とした場合と、従来例の高レベル放
射性廃液をそのままガラス固化体とした場合について、
貯蔵期間に対する効果を比較して示す。従来例では高レ
ベル放射性廃液をそのままガラス固化体とし、30年から
50年の冷却期間をおいた後、地下数百メートルの地底に
処分した。By using such a method for treating high-level radioactive liquid waste, it is possible to reduce the number of solid wastes and the isolation period.
Furthermore, the geological disposal technology can be realized as a more specific technology. Table 1 shows a case where the high-level radioactive waste liquid was treated according to the above-mentioned example and only the actinide element and the rare earth element were made into the vitrified body, and the high-level radioactive waste liquid of the conventional example was directly made into the vitrified body.
The effect on storage time is shown in comparison. In the conventional example, the high-level radioactive waste liquid was directly made into a vitrified body,
After a cooling period of 50 years, it was disposed in the underground several hundred meters underground.
【００１３】[0013]
【表１】 [Table 1]
【００１４】表１から明らかなように、本実施例の処理
方法によれば、従来の処理方法に比べて地層処分年数を
1/1000まで減少させることができる。これは群分離によ
って、超長半減期核分裂生成物であるテクネチウム９９
を除いたアクチニド元素・希土類元素のみを固化体とし
て貯蔵するためである。As is clear from Table 1, according to the treatment method of this embodiment, the number of years of geological disposal is longer than that of the conventional treatment method.
Can be reduced to 1/1000. This is technetium-99, which is a very long half-life fission product due to group separation.
This is because only the actinide element / rare earth element except for is stored as a solidified body.
【００１５】また表２に、同様にして、ガラス固化体の
貯蔵施設の人工バリアのうちキャニスタとオーバーパッ
クを除いたコンクリート外壁の厚みを算出した結果を比
較して示す。Table 2 also shows a comparison of the results of the calculation of the thickness of the concrete outer wall excluding the canister and the overpack in the artificial barrier of the storage facility for the vitrified body in the same manner.
【００１６】[0016]
【表２】 [Table 2]
【００１７】表２から本発明の実施例の方が従来例に比
べてバリア厚さが1/100 になることが分かる。これは高
レベル放射性廃液から放射能の強いテクネチウム９９を
分離回収することにより、ガラス固化体が比較的放射能
の弱いアクチニド元素・希土類元素のみからなるためで
ある。From Table 2, it can be seen that the barrier thickness of the embodiment of the present invention is 1/100 of that of the conventional example. This is because by virtue of separating and recovering technetium 99 having high radioactivity from the high-level radioactive waste liquid, the vitrified body is composed of only actinide element / rare earth element having relatively low radioactivity.
【００１８】以上説明したように、この実施例では、高
レベル放射性廃液を湿式分離法で群分離する際に大過剰
量のシュウ酸を用いることにより、脱硝工程と湿式分離
工程を同時に行うことができ、したがって有機溶媒を用
いた溶媒抽出法や従来のシュウ酸塩沈殿法では必要とし
ていたギ酸による脱硝工程を削減することができるた
め、放射性廃溶媒を低減することができる。また、高レ
ベル放射性廃液からテクネチウム９９を選択的に分離回
収することで、処分固化体の量および貯蔵年数を大幅に
減少させることができ、かつアルカリ金属元素群および
有用貴金属元素である白金族元素群の再利用をより容易
にする。As described above, in this embodiment, the denitration step and the wet separation step can be performed simultaneously by using a large excess amount of oxalic acid when the high-level radioactive waste liquid is group-separated by the wet separation method. Therefore, it is possible to reduce the denitration step using formic acid, which is required in the solvent extraction method using an organic solvent and the conventional oxalate precipitation method, and thus it is possible to reduce the radioactive waste solvent. In addition, by selectively separating and recovering technetium-99 from the high-level radioactive waste liquid, the amount of solidified waste and the number of years of storage can be significantly reduced, and the alkali metal element group and the platinum group element, which is a useful precious metal element, can be significantly reduced. Makes group re-use easier.
【００１９】なお、本実施例の湿式分離工程４において
は、アクチニド元素・希土類元素の沈澱試薬としてシュ
ウ酸を用いたが、その代わりにフッ化水素、ヨウ素酸カ
リウム、アンモニアを用いることも可能である。また消
滅処理工程６においてはテクネチウム９９の中性子吸収
断面積から軽水炉の代わりに高速増殖炉の使用も考えら
れる。さらには加速器、専焼炉なども用いることが可能
である。In the wet separation step 4 of this embodiment, oxalic acid was used as the actinide / rare earth element precipitation reagent, but hydrogen fluoride, potassium iodate, or ammonia can be used instead. is there. Further, in the extinction treatment step 6, it is possible to use a fast breeder reactor instead of the light water reactor due to the neutron absorption cross section of technetium 99. Further, it is possible to use an accelerator, a special baking furnace, or the like.
【００２０】[0020]
【発明の効果】以上の説明からも明らかなように、本発
明によれば、高レベル放射性廃液を群分離し、さらに強
いβ放射体である超長半減期核分裂生成物であるテクネ
チウム９９を分離・消滅処理することにより、固化体の
減容ならびに貯蔵期間の短縮が可能となる。また、アク
チニド元素・希土類元素の群を分離除去した後の廃液か
ら、高放射能のテクネチウム９９を選択的に抽出除去す
ることで、アルカリ金属、アルカリ土類金属元素および
白金族元素等の有用元素の分離回収が容易になる。加え
て、消滅処理によりテクネチウム９９の有効利用も可能
となる。As is apparent from the above description, according to the present invention, high-level radioactive liquid waste is separated into groups, and technetium 99, which is a very long half-life fission product, which is a strong β-emitter, is separated. -By eliminating the waste, it is possible to reduce the volume of the solidified product and shorten the storage period. In addition, by selectively extracting and removing technetium 99 with high radioactivity from the waste liquid after separating and removing the group of actinide element and rare earth element, useful elements such as alkali metal, alkaline earth metal element and platinum group element It becomes easy to separate and collect. In addition, the annihilation process enables effective use of technetium 99.
【図１】本発明の高レベル放射性廃液中のテクネチウム
９９の分離方法の一実施例を示すフロー図である。FIG. 1 is a flow chart showing an example of a method for separating technetium 99 in a high-level radioactive liquid waste according to the present invention.
Claims (1)
ド元素、希土類元素、アルカリ金属元素、アルカリ土類
金属元素、テクネチウム９９および白金族元素等が溶解
している高レベル放射性廃液中より湿式分離法で前記ア
クチニド元素および希土類元素を分離除去した後、前記
アルカリ金属元素、アルカリ土類金属元素、テクネチウ
ム９９および白金族元素等を含む廃液を陰イオン交換樹
脂に接触させて前記廃液中に唯一陰イオン化学種として
存在するテクネチウム９９核種を選択的に吸着させるこ
とを特徴とする高レベル放射性廃液中のテクネチウム９
９の分離方法。1. A wet separation method from a high-level radioactive waste liquid in which actinide elements, rare earth elements, alkali metal elements, alkaline earth metal elements, technetium 99, platinum group elements, etc. generated from nuclear fuel reprocessing facilities are dissolved. After the actinide element and the rare earth element are separated and removed, a waste solution containing the alkali metal element, the alkaline earth metal element, technetium 99, a platinum group element, etc. is contacted with an anion exchange resin, and the only anion chemical is present in the waste solution. Technetium-99 in high-level radioactive liquid waste, characterized by selectively adsorbing technetium-99 as a seed
9. Separation method of 9.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP22668891A JP3034353B2 (en) | 1991-09-06 | 1991-09-06 | Method for separating technetium-99 from high-level radioactive liquid waste |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP22668891A JP3034353B2 (en) | 1991-09-06 | 1991-09-06 | Method for separating technetium-99 from high-level radioactive liquid waste |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPH0566289A true JPH0566289A (en) | 1993-03-19 |
| JP3034353B2 JP3034353B2 (en) | 2000-04-17 |
Family
ID=16849099
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP22668891A Expired - Lifetime JP3034353B2 (en) | 1991-09-06 | 1991-09-06 | Method for separating technetium-99 from high-level radioactive liquid waste |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JP3034353B2 (en) |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2007254170A (en) * | 2006-03-20 | 2007-10-04 | Inst Nuclear Energy Research Rocaec | APPARATUS AND METHOD FOR CONCENTRATING TECHNETIUM-99m PERTECHNETATE |
| JP2016122006A (en) * | 2010-03-09 | 2016-07-07 | クリオン インコーポレイテッド | Isotope-specific separation and vitrification using ion-specific media |
-
1991
- 1991-09-06 JP JP22668891A patent/JP3034353B2/en not_active Expired - Lifetime
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2007254170A (en) * | 2006-03-20 | 2007-10-04 | Inst Nuclear Energy Research Rocaec | APPARATUS AND METHOD FOR CONCENTRATING TECHNETIUM-99m PERTECHNETATE |
| JP4578425B2 (en) * | 2006-03-20 | 2010-11-10 | 行政院原子能委員會核能研究所 | Concentration apparatus and method for technetium-99m pertechnetate solution |
| JP2016122006A (en) * | 2010-03-09 | 2016-07-07 | クリオン インコーポレイテッド | Isotope-specific separation and vitrification using ion-specific media |
Also Published As
| Publication number | Publication date |
|---|---|
| JP3034353B2 (en) | 2000-04-17 |
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