JPH01215727A - Adsorptive separation of technetium - Google Patents
Adsorptive separation of technetiumInfo
- Publication number
- JPH01215727A JPH01215727A JP3762588A JP3762588A JPH01215727A JP H01215727 A JPH01215727 A JP H01215727A JP 3762588 A JP3762588 A JP 3762588A JP 3762588 A JP3762588 A JP 3762588A JP H01215727 A JPH01215727 A JP H01215727A
- Authority
- JP
- Japan
- Prior art keywords
- technetium
- activated carbon
- radioactive waste
- level radioactive
- nitric acid
- 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.)
- Pending
Links
- 229910052713 technetium Inorganic materials 0.000 title claims description 44
- GKLVYJBZJHMRIY-UHFFFAOYSA-N technetium atom Chemical compound [Tc] GKLVYJBZJHMRIY-UHFFFAOYSA-N 0.000 title claims description 44
- 238000000926 separation method Methods 0.000 title claims description 8
- 230000000274 adsorptive effect Effects 0.000 title description 2
- OKTJSMMVPCPJKN-UHFFFAOYSA-N Carbon Chemical compound [C] OKTJSMMVPCPJKN-UHFFFAOYSA-N 0.000 claims description 36
- 239000002927 high level radioactive waste Substances 0.000 claims description 15
- 239000007788 liquid Substances 0.000 claims description 14
- 238000000034 method Methods 0.000 claims description 13
- 238000001179 sorption measurement Methods 0.000 claims description 11
- 239000002915 spent fuel radioactive waste Substances 0.000 claims description 9
- 238000012958 reprocessing Methods 0.000 claims description 3
- 239000003463 adsorbent Substances 0.000 claims description 2
- 239000000243 solution Substances 0.000 description 18
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 description 14
- 229910017604 nitric acid Inorganic materials 0.000 description 14
- NHNBFGGVMKEFGY-UHFFFAOYSA-N Nitrate Chemical compound [O-][N+]([O-])=O NHNBFGGVMKEFGY-UHFFFAOYSA-N 0.000 description 6
- 239000003245 coal Substances 0.000 description 6
- BASFCYQUMIYNBI-UHFFFAOYSA-N platinum Chemical group [Pt] BASFCYQUMIYNBI-UHFFFAOYSA-N 0.000 description 5
- 238000005192 partition Methods 0.000 description 4
- 238000002474 experimental method Methods 0.000 description 3
- 239000012530 fluid Substances 0.000 description 3
- 238000001556 precipitation Methods 0.000 description 3
- 229910002651 NO3 Inorganic materials 0.000 description 2
- 239000003929 acidic solution Substances 0.000 description 2
- 230000015572 biosynthetic process Effects 0.000 description 2
- 229910052799 carbon Inorganic materials 0.000 description 2
- 239000003153 chemical reaction reagent Substances 0.000 description 2
- JHIVVAPYMSGYDF-UHFFFAOYSA-N cyclohexanone Chemical compound O=C1CCCCC1 JHIVVAPYMSGYDF-UHFFFAOYSA-N 0.000 description 2
- 238000005516 engineering process Methods 0.000 description 2
- 238000000605 extraction Methods 0.000 description 2
- 239000000463 material Substances 0.000 description 2
- 229910021645 metal ion Inorganic materials 0.000 description 2
- 230000005855 radiation Effects 0.000 description 2
- 238000011084 recovery Methods 0.000 description 2
- VWDWKYIASSYTQR-UHFFFAOYSA-N sodium nitrate Chemical compound [Na+].[O-][N+]([O-])=O VWDWKYIASSYTQR-UHFFFAOYSA-N 0.000 description 2
- 238000000638 solvent extraction Methods 0.000 description 2
- DHEQXMRUPNDRPG-UHFFFAOYSA-N strontium nitrate Chemical compound [Sr+2].[O-][N+]([O-])=O.[O-][N+]([O-])=O DHEQXMRUPNDRPG-UHFFFAOYSA-N 0.000 description 2
- 239000002699 waste material Substances 0.000 description 2
- PAWQVTBBRAZDMG-UHFFFAOYSA-N 2-(3-bromo-2-fluorophenyl)acetic acid Chemical compound OC(=O)CC1=CC=CC(Br)=C1F PAWQVTBBRAZDMG-UHFFFAOYSA-N 0.000 description 1
- VEXZGXHMUGYJMC-UHFFFAOYSA-M Chloride anion Chemical compound [Cl-] VEXZGXHMUGYJMC-UHFFFAOYSA-M 0.000 description 1
- 229910000831 Steel Inorganic materials 0.000 description 1
- HCHKCACWOHOZIP-UHFFFAOYSA-N Zinc Chemical compound [Zn] HCHKCACWOHOZIP-UHFFFAOYSA-N 0.000 description 1
- 239000002253 acid Substances 0.000 description 1
- 239000003054 catalyst Substances 0.000 description 1
- 150000001768 cations Chemical class 0.000 description 1
- 239000003610 charcoal Substances 0.000 description 1
- 239000003638 chemical reducing agent Substances 0.000 description 1
- 238000000975 co-precipitation Methods 0.000 description 1
- 238000007796 conventional method Methods 0.000 description 1
- 230000007797 corrosion Effects 0.000 description 1
- 238000005260 corrosion Methods 0.000 description 1
- 238000003795 desorption Methods 0.000 description 1
- 230000006866 deterioration Effects 0.000 description 1
- 238000010790 dilution Methods 0.000 description 1
- 239000012895 dilution Substances 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 230000007613 environmental effect Effects 0.000 description 1
- 238000005755 formation reaction Methods 0.000 description 1
- 238000010438 heat treatment Methods 0.000 description 1
- 150000002500 ions Chemical class 0.000 description 1
- 238000005259 measurement Methods 0.000 description 1
- ZUZLIXGTXQBUDC-UHFFFAOYSA-N methyltrioctylammonium Chemical compound CCCCCCCC[N+](C)(CCCCCCCC)CCCCCCCC ZUZLIXGTXQBUDC-UHFFFAOYSA-N 0.000 description 1
- XTAZYLNFDRKIHJ-UHFFFAOYSA-N n,n-dioctyloctan-1-amine Chemical compound CCCCCCCCN(CCCCCCCC)CCCCCCCC XTAZYLNFDRKIHJ-UHFFFAOYSA-N 0.000 description 1
- -1 nitrate ions Chemical class 0.000 description 1
- 239000003758 nuclear fuel Substances 0.000 description 1
- 150000002894 organic compounds Chemical class 0.000 description 1
- XYFCBTPGUUZFHI-UHFFFAOYSA-O phosphonium Chemical compound [PH4+] XYFCBTPGUUZFHI-UHFFFAOYSA-O 0.000 description 1
- 239000000843 powder Substances 0.000 description 1
- 239000000700 radioactive tracer Substances 0.000 description 1
- 238000011160 research Methods 0.000 description 1
- 150000003839 salts Chemical class 0.000 description 1
- 239000004065 semiconductor Substances 0.000 description 1
- 235000010344 sodium nitrate Nutrition 0.000 description 1
- 239000004317 sodium nitrate Substances 0.000 description 1
- 239000010959 steel Substances 0.000 description 1
- 239000006228 supernatant Substances 0.000 description 1
- 238000003786 synthesis reaction Methods 0.000 description 1
- NRZGVGVFPHPXEO-UHFFFAOYSA-M tetraphenylarsanium;chloride Chemical compound [Cl-].C1=CC=CC=C1[As+](C=1C=CC=CC=1)(C=1C=CC=CC=1)C1=CC=CC=C1 NRZGVGVFPHPXEO-UHFFFAOYSA-M 0.000 description 1
- WAGFXJQAIZNSEQ-UHFFFAOYSA-M tetraphenylphosphonium chloride Chemical compound [Cl-].C1=CC=CC=C1[P+](C=1C=CC=CC=1)(C=1C=CC=CC=1)C1=CC=CC=C1 WAGFXJQAIZNSEQ-UHFFFAOYSA-M 0.000 description 1
- STCOOQWBFONSKY-UHFFFAOYSA-N tributyl phosphate Chemical compound CCCCOP(=O)(OCCCC)OCCCC STCOOQWBFONSKY-UHFFFAOYSA-N 0.000 description 1
- 238000005406 washing Methods 0.000 description 1
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 1
- 229910052725 zinc Inorganic materials 0.000 description 1
- 239000011701 zinc Substances 0.000 description 1
Landscapes
- Treatment Of Liquids With Adsorbents In General (AREA)
- Inorganic Compounds Of Heavy Metals (AREA)
- Solid-Sorbent Or Filter-Aiding Compositions (AREA)
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】 （産業上の利用分野） 本発明はテクネチウムの吸着分離法に関する。[Detailed description of the invention] (Industrial application field) The present invention relates to a method for adsorptive separation of technetium.
更に詳しくは、本発明は、使用済核燃料の再処理によっ
て発生する高レベル放射性廃液、使用済核燃料溶解液等
のテクネチウムを含む廃液から、テクネチウムを分離回
収するため、活性炭をテクネチウムの吸着剤として使用
することを特徴とするテクネチウムの吸着分離法に関す
る。More specifically, the present invention uses activated carbon as an adsorbent for technetium in order to separate and recover technetium from technetium-containing waste fluids such as high-level radioactive waste fluids and spent nuclear fuel dissolved fluids generated by reprocessing spent nuclear fuels. This invention relates to a method for adsorption and separation of technetium.
（従来の技術）
テクネチウムは、天然には全く存在しない元素で、原子
炉や加速装置によって人工的に得られる元素であるが、
使用済核燃料の再処理に伴って発生する高レベル放射性
廃液中又は使用済核燃料溶解液中には、使用済核燃料１
トン当たり約８００ｇが含まれている。(Prior technology) Technetium is an element that does not exist in nature at all, and is obtained artificially through nuclear reactors and accelerators.
Spent nuclear fuel 1 is contained in the high-level radioactive waste liquid or spent nuclear fuel solution generated during the reprocessing of spent nuclear fuel.
Contains about 800g per ton.
高レベル放射性廃液又は使用済核燃料溶解液からのテク
ネチウムの分離回収は、高レベル放射性廃液をガラス固
化体として地層処分する場合の被曝などによる環境影響
を出来るだけ少なくするという観点から、また、回収し
たテクネチウムを鋼材の腐食防止、或いは白金族元素と
同様に有機化合物の合成分野での触媒として有効に利用
できるという観点から、最近特に注目されている。この
ため、高レベル放射性廃液又は使用済核燃料溶解液から
のテクネチウムの分離回収は特に有効である。Separation and recovery of technetium from high-level radioactive waste liquid or spent nuclear fuel solution is carried out from the viewpoint of minimizing the environmental impact due to radiation exposure when high-level radioactive waste liquid is disposed of as a vitrified material in geological formations. Technetium has recently attracted particular attention from the viewpoint that it can be effectively used to prevent corrosion of steel materials or as a catalyst in the synthesis of organic compounds, similar to platinum group elements. For this reason, separation and recovery of technetium from high-level radioactive waste liquid or spent nuclear fuel solution is particularly effective.
従来用いられてきたテクネチウムの分離回収法には、リ
ン酸トリブチル、トリーｎ−オクチルアミン、硝酸トリ
カプリル−メチルアンモニウム、塩化テトラフェニルア
ルソニウム、シクロヘキサノン等による溶媒抽出法、塩
化テトラフェニルフォスフオニウムによる沈澱法などが
ある。Conventionally used methods for separating and recovering technetium include solvent extraction using tributyl phosphate, tri-n-octylamine, tricapryl-methylammonium nitrate, tetraphenylarsonium chloride, cyclohexanone, etc., and precipitation using tetraphenylphosphonium chloride. There are laws, etc.
これらの方法の欠点は、溶媒抽出法では抽出試薬の放射
線による劣化が比較的はやく、劣化に伴い発生する使用
不能になった抽出試薬の処理処分に新たな技術を必要と
すること、塩化テトラフェニルフォスフオニウムによる
沈澱法では酸性溶液からのテクネチウムの分離回収が不
可能であることである。The disadvantages of these methods are that in the solvent extraction method, the extraction reagent deteriorates relatively quickly due to radiation, and new technology is required to dispose of the extraction reagent that is no longer usable due to deterioration, and tetraphenyl chloride The problem is that it is impossible to separate and recover technetium from an acidic solution using the precipitation method using phosphonium.
このような欠点のないテクネチウムの回収法として、テ
クネチウムを含む溶液に還元剤と白金族元素を共存させ
て加熱することによって沈澱又は共沈の形でテクネチウ
ムを分離回収する方法が発明（特開昭６２−９３３２０
号公報）されているが、この方法には白金族元素の濃度
が０．００５Ｍ以下の溶液では適用できないという欠点
がある。As a method for recovering technetium without such drawbacks, a method was invented in which technetium is separated and recovered in the form of precipitation or coprecipitation by heating a solution containing technetium in the coexistence of a reducing agent and a platinum group element (Japanese Unexamined Patent Publication No. 62-93320
However, this method has the drawback that it cannot be applied to solutions with a platinum group element concentration of 0.005M or less.
（発明が解決しようとする問題点）
本発明の目的は、このような欠点のないテクネチウムの
分離回収法を提供することである。(Problems to be Solved by the Invention) An object of the present invention is to provide a method for separating and recovering technetium that does not have such drawbacks.
（問題点をするための手段）
本願発明者等は、鋭意研究の結果、テクネチウムを含む
溶液に活性炭を添加するか、或いは、テクネチウムを含
む溶液を活性炭を充填したカラムに通すことによって、
テクネチウムを比較的高濃度の酸性溶液からでも、活性
炭により吸着分離できることを発見し、この方法におい
ては、活性炭の放射線による劣化は少なく、使用不能に
なった活性炭は焼却などの従来技術により処理すること
で廃棄物発生量を極端に少なくできること、及び一般に
２〜３Ｍ硝酸溶液として貯蔵される高レベル放射性廃液
の酸濃度をほとんど低下させることなく、テクネチウム
を分離回収できることを発見し、本発明の目的を達成し
た。(Means for solving the problem) As a result of intensive research, the inventors of the present application have found that by adding activated carbon to a solution containing technetium, or passing a solution containing technetium through a column filled with activated carbon,
They discovered that technetium can be adsorbed and separated using activated carbon even from relatively highly concentrated acidic solutions. In this method, the activated carbon deteriorates little due to radiation, and activated carbon that is no longer usable can be disposed of using conventional techniques such as incineration. We have discovered that the amount of waste generated can be extremely reduced, and that technetium can be separated and recovered without substantially reducing the acid concentration of high-level radioactive waste liquid, which is generally stored as a 2-3M nitric acid solution. Achieved.
従って、本発明は、高レベル放射性廃液、使用済核燃料
溶解液等のテクネチウムを含む溶液に活性炭を添加する
、或いはテクネチウムを含む溶液を活性炭を充填したカ
ラムに通すことによって、テクネチウムを活性炭により
吸着分離する方法から成る発明である。Therefore, the present invention is capable of adsorbing and separating technetium by adding activated carbon to a solution containing technetium, such as a high-level radioactive waste liquid or a spent nuclear fuel solution, or by passing a solution containing technetium through a column packed with activated carbon. This invention consists of a method for
（実施例）
本発明を実施例についてさらに具体的に説明する。しか
し、本発明は実施例によって限定されるものではない。(Example) The present invention will be described in more detail with reference to Examples. However, the present invention is not limited to the examples.
ス崖±
活性炭０．２５ ｇを入れた２０ｍ１のバイアルビンに
過テクネチウム酸イオンを含む硝酸或いは種々の金属イ
オンを含む硝酸溶液１０ｎ＋１を添加した。10n+1 of a nitric acid solution containing pertechnetate ions or a nitric acid solution containing various metal ions was added to a 20 ml vial containing 0.25 g of activated carbon.
この試料を２５℃に設定したエアバス中に入れ、回転ロ
ーラー上で１時間攪拌した。This sample was placed in an air bath set at 25°C and stirred on a rotating roller for 1 hour.
試料を取り出して遠心分離機にかけた後、上澄液の１ｍ
ｌをサンプリングし、トレーサーとして加えた９Ｓ−Ｔ
ｃの放射能強度をＧｅ半導体検出器で測定することによ
り、テクネチウムの活性炭への分配係数（Ｋｄ）及び平
衡吸着量を求めた。After removing the sample and centrifuging, 1 m of supernatant
9S-T was sampled and added as a tracer.
By measuring the radioactivity intensity of c with a Ge semiconductor detector, the distribution coefficient (Kd) of technetium to activated carbon and the equilibrium adsorption amount were determined.
ここでＫｄは次の式で表される。Here, Kd is expressed by the following formula.
活性炭中の９％＋″Ｔｃ濃度（μＣｉ／ｇ ）Ｋｄ
−
溶液中の ９！ｌｌｌＴｃ濃度（１１Ｃｉ／＋１）なお
、ここで使用した活性炭は、市販の亜鉛賦活粉末活性炭
（以下未処理炭と呼ぶ）をそのまま、或いは４Ｍ硝酸で
洗浄後１０５℃で乾燥したもの（処理炭と呼ぶ）を用い
た。Ｋｄの測定ではトレーサとして””Ｔｃ （ＴｃＯ
，−形）を用いたが、高レベル放射性廃液で考えられる
テクネチウム濃度（０，０１５Ｍ）での平衡吸着量の測
定ではさらに４４Ｔｃ （ＴｃＯ４−形）を添加して行
った。実験により得られた硝酸濃度とテクネチウムのＫ
ｄとの関係を第１図に示した０図から明らかなように未
処理炭では２Ｍ以下、処理炭では０．５Ｍ以下の硝酸溶
液からテクネチウムをＫｄが１００以上で吸着分離でき
ることがわかる。9%+''Tc concentration in activated carbon (μCi/g) Kd
- 9 in solution! lllTc concentration (11Ci/+1) The activated carbon used here was commercially available zinc-activated powder activated carbon (hereinafter referred to as untreated carbon), or was dried at 105°C after washing with 4M nitric acid (referred to as treated carbon). ) was used. In the measurement of Kd, “”Tc (TcO
, - form) was used, but 44Tc (TcO4 - form) was further added to measure the equilibrium adsorption amount at a technetium concentration (0,015 M), which is considered to be a high-level radioactive waste liquid. Nitric acid concentration and technetium K obtained by experiment
As is clear from Figure 1, which shows the relationship with d, technetium can be adsorbed and separated from a nitric acid solution of 2M or less for untreated coal and 0.5M or less for treated coal when Kd is 100 or more.
テクネチウムの吸着に及ぼす共存塩の影響について調べ
た実験から得られた硝酸イオン濃度とテクネチウムのＫ
ｄとの関係を第２図に示した。第２図には処理炭の場合
を示しているが、テクネチウムのＫｄは陽イオンによら
ず硝酸イオン濃度の一１乗に比例することがわかる。高
レベル放射性廃液中に存在する硝酸イオン濃度は硝酸に
よるものを除けばＩＭ程度であるから、２Ｍ硝酸濃度の
高レベル放射性廃液を対象とした場合には、硝酸イオン
濃度は約３Ｍとなり、Ｋｄが１０以上で吸着分離できる
。なお、テクネチウムの吸着についてさらに大きなｋｄ
を得るためには、高レベル放射性廃液の硝酸濃度を脱硝
や水による希釈によって低下させることがより効果的で
ある。Nitrate ion concentration and K of technetium obtained from an experiment investigating the influence of coexisting salts on the adsorption of technetium
The relationship with d is shown in Figure 2. FIG. 2 shows the case of treated charcoal, and it can be seen that the Kd of technetium is proportional to the 11th power of the nitrate ion concentration, regardless of the cation. The concentration of nitrate ions present in high-level radioactive waste liquid is about IM except for that caused by nitric acid, so if a high-level radioactive waste liquid with a 2M nitric acid concentration is used, the nitrate ion concentration will be approximately 3M, and Kd will be At 10 or more, adsorption separation is possible. Furthermore, a larger kd for technetium adsorption is required.
In order to obtain this, it is more effective to reduce the nitric acid concentration of high-level radioactive waste liquid by denitrification or dilution with water.
高レベル放射性廃液で考えられるテクネチウム濃度（０
，０１５Ｍ）での平衡吸着量に関する実験では、０．５
Ｍ硝酸溶液中で未処理炭についての平衡吸着量は０．６
０ ａ＋ｅｑ／ｇ 、処理炭について０．３３ｍｅｑ／
ｇとなった・
高レベル放射性廃液に含むと予想される元素のうちで活
性炭に吸着する元素はテクネチウム以外に白金族元素が
あげられるが、白金族元素はテクネチウムに比べて吸着
され難く、テクネチウムの吸着に及ぼす影響はほとんど
ない。このため活性炭は高レベル放射性廃液からのテク
ネチウムの選択的な吸着分離に特に有効である。Possible technetium concentration in high-level radioactive waste liquid (0
, 015M), the equilibrium adsorption amount was 0.5
The equilibrium adsorption amount for untreated coal in M nitric acid solution is 0.6
0 a+eq/g, 0.33 meq/g for treated coal
Among the elements expected to be contained in high-level radioactive waste liquid, the elements other than technetium that can be adsorbed on activated carbon include platinum group elements, but platinum group elements are less adsorbed than technetium, and It has little effect on adsorption. Activated carbon is therefore particularly effective for selective adsorption and separation of technetium from high-level radioactive waste liquids.
なお、活性炭に吸着されたテクネチウムは４Ｍ以上の硝
酸を使用すること、ないしは脱着時の溶液の温度を１０
０℃程度にまで上昇することによって脱着することがで
きる。For technetium adsorbed on activated carbon, use nitric acid with a concentration of 4M or higher, or lower the temperature of the solution during desorption to 10%.
It can be desorbed by increasing the temperature to about 0°C.
第１図は硝酸濃度とテクネチウムの分配係数との関係を
示すグラフで、横軸は硝酸濃度（Ｍ＞、縦軸はテクネチ
ウムの分配係数（ｍｌ／ｇ）である。
図中の白丸の点は未処理炭について、黒丸の点は処理炭
についての実験値を示す。
第２図は硝酸イオン濃度とテクネチウムの分配係数との
関係を示すグラフで、
横軸は硝酸イオン濃度（Ｍ）、
縦軸はテクネチウムの分配係数（ｍｌ／ｇ）である。
なお、図中のデータは種々の金属イオンを含む０．５Ｍ
硝酸溶液での実験から得られたもので、白丸の点は硝酸
ナトリウムを、黒丸の点は硝酸ストロンチウムを、三角
の点は硝酸アンモニウムをそれぞれ含む場合を示す。Figure 1 is a graph showing the relationship between nitric acid concentration and technetium partition coefficient, where the horizontal axis is nitric acid concentration (M>) and the vertical axis is technetium partition coefficient (ml/g). For untreated coal, black circles indicate experimental values for treated coal. Figure 2 is a graph showing the relationship between nitrate ion concentration and technetium partition coefficient, where the horizontal axis is nitrate ion concentration (M) and the vertical axis is the nitrate ion concentration (M). is the partition coefficient (ml/g) of technetium.The data in the figure is 0.5M containing various metal ions.
The results were obtained from an experiment using a nitric acid solution; white circles indicate cases containing sodium nitrate, black circles indicate cases containing strontium nitrate, and triangular points indicate cases containing ammonium nitrate.
Claims (1)
廃液、使用済核燃料溶解液等のテクネチウムを含む溶液
から、テクネチウムを分離回収するため、活性炭をテク
ネチウムの吸着剤として使用することを特徴とするテク
ネチウムの吸着分離法。The method is characterized in that activated carbon is used as an adsorbent for technetium in order to separate and recover technetium from solutions containing technetium such as high-level radioactive waste liquid and spent nuclear fuel solution generated during reprocessing of spent nuclear fuel. Technetium adsorption separation method.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP3762588A JPH01215727A (en) | 1988-02-22 | 1988-02-22 | Adsorptive separation of technetium |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP3762588A JPH01215727A (en) | 1988-02-22 | 1988-02-22 | Adsorptive separation of technetium |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPH01215727A true JPH01215727A (en) | 1989-08-29 |
Family
ID=12502815
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP3762588A Pending JPH01215727A (en) | 1988-02-22 | 1988-02-22 | Adsorptive separation of technetium |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH01215727A (en) |
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2010133938A (en) * | 2008-10-28 | 2010-06-17 | Japan Atomic Energy Agency | Method of separating trivalent lanthanide and trivalent actinide |
| WO2010146722A1 (en) * | 2009-06-19 | 2010-12-23 | 株式会社化研 | Method and system for concentrating radioactive technetium as raw material for radioactive drug and labeling compound therefor and collecting the same by elution |
| JP2014512258A (en) * | 2011-02-25 | 2014-05-22 | ウィリアム・マーシュ・ライス・ユニバーシティ | Sorption and separation of various substances by graphene oxide |
-
1988
- 1988-02-22 JP JP3762588A patent/JPH01215727A/en active Pending
Cited By (6)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2010133938A (en) * | 2008-10-28 | 2010-06-17 | Japan Atomic Energy Agency | Method of separating trivalent lanthanide and trivalent actinide |
| WO2010146722A1 (en) * | 2009-06-19 | 2010-12-23 | 株式会社化研 | Method and system for concentrating radioactive technetium as raw material for radioactive drug and labeling compound therefor and collecting the same by elution |
| JP2011002370A (en) * | 2009-06-19 | 2011-01-06 | Kaken:Kk | Method and system for concentration and elution recovery of radioactive technetium as material for radiopharmaceutical medicine and labeled compound of the same |
| US9236153B2 (en) | 2009-06-19 | 2016-01-12 | Kaken Co., Ltd. | Method of recovering enriched radioactive technetium and system therefor |
| EP2444106A4 (en) * | 2009-06-19 | 2016-04-20 | Kaken Co Ltd | Method and system for concentrating radioactive technetium as raw material for radioactive drug and labeling compound therefor and collecting the same by elution |
| JP2014512258A (en) * | 2011-02-25 | 2014-05-22 | ウィリアム・マーシュ・ライス・ユニバーシティ | Sorption and separation of various substances by graphene oxide |
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