JPS63208800A - Method of recovering radioactive element extraction agent - Google Patents
Method of recovering radioactive element extraction agentInfo
- Publication number
- JPS63208800A JPS63208800A JP62041193A JP4119387A JPS63208800A JP S63208800 A JPS63208800 A JP S63208800A JP 62041193 A JP62041193 A JP 62041193A JP 4119387 A JP4119387 A JP 4119387A JP S63208800 A JPS63208800 A JP S63208800A
- Authority
- JP
- Japan
- Prior art keywords
- extractant
- waste liquid
- column
- radioactive waste
- transuranic elements
- 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.)
- Granted
Links
- 238000000034 method Methods 0.000 title claims description 20
- 230000002285 radioactive effect Effects 0.000 title claims description 3
- 238000000605 extraction Methods 0.000 title description 8
- 239000007788 liquid Substances 0.000 claims description 42
- 239000002901 radioactive waste Substances 0.000 claims description 19
- 239000002699 waste material Substances 0.000 claims description 17
- OKKJLVBELUTLKV-UHFFFAOYSA-N Methanol Chemical compound OC OKKJLVBELUTLKV-UHFFFAOYSA-N 0.000 claims description 15
- 238000011282 treatment Methods 0.000 claims description 13
- CSCPPACGZOOCGX-UHFFFAOYSA-N Acetone Chemical compound CC(C)=O CSCPPACGZOOCGX-UHFFFAOYSA-N 0.000 claims description 12
- 239000007787 solid Substances 0.000 claims description 11
- 239000003463 adsorbent Substances 0.000 claims description 9
- SGZRFMMIONYDQU-UHFFFAOYSA-N n,n-bis(2-methylpropyl)-2-[octyl(phenyl)phosphoryl]acetamide Chemical compound CCCCCCCCP(=O)(CC(=O)N(CC(C)C)CC(C)C)C1=CC=CC=C1 SGZRFMMIONYDQU-UHFFFAOYSA-N 0.000 claims description 8
- 239000003480 eluent Substances 0.000 claims description 6
- 239000003795 chemical substances by application Substances 0.000 claims description 4
- 239000002253 acid Substances 0.000 claims description 3
- 238000010828 elution Methods 0.000 claims description 2
- -1 carbamoyl methylene pho sph-onate Chemical compound 0.000 claims 1
- 125000000325 methylidene group Chemical group [H]C([H])=* 0.000 claims 1
- 239000012071 phase Substances 0.000 description 10
- 238000003860 storage Methods 0.000 description 10
- NWUYHJFMYQTDRP-UHFFFAOYSA-N 1,2-bis(ethenyl)benzene;1-ethenyl-2-ethylbenzene;styrene Chemical compound C=CC1=CC=CC=C1.CCC1=CC=CC=C1C=C.C=CC1=CC=CC=C1C=C NWUYHJFMYQTDRP-UHFFFAOYSA-N 0.000 description 6
- 239000003456 ion exchange resin Substances 0.000 description 6
- 229920003303 ion-exchange polymer Polymers 0.000 description 6
- 239000011347 resin Substances 0.000 description 6
- 229920005989 resin Polymers 0.000 description 6
- 238000005406 washing Methods 0.000 description 6
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 description 4
- 229910017604 nitric acid Inorganic materials 0.000 description 4
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 4
- MUBZPKHOEPUJKR-UHFFFAOYSA-N Oxalic acid Chemical compound OC(=O)C(O)=O MUBZPKHOEPUJKR-UHFFFAOYSA-N 0.000 description 3
- 239000003513 alkali Substances 0.000 description 3
- 239000008346 aqueous phase Substances 0.000 description 3
- 239000002927 high level radioactive waste Substances 0.000 description 3
- 239000012535 impurity Substances 0.000 description 3
- 238000011084 recovery Methods 0.000 description 3
- 238000001179 sorption measurement Methods 0.000 description 3
- 239000000126 substance Substances 0.000 description 3
- 229910052768 actinide Inorganic materials 0.000 description 2
- 150000001255 actinides Chemical class 0.000 description 2
- 239000003957 anion exchange resin Substances 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 230000004992 fission Effects 0.000 description 2
- 238000012958 reprocessing Methods 0.000 description 2
- 238000000926 separation method Methods 0.000 description 2
- VWDWKYIASSYTQR-UHFFFAOYSA-N sodium nitrate Chemical compound [Na+].[O-][N+]([O-])=O VWDWKYIASSYTQR-UHFFFAOYSA-N 0.000 description 2
- 239000002915 spent fuel radioactive waste Substances 0.000 description 2
- STCOOQWBFONSKY-UHFFFAOYSA-N tributyl phosphate Chemical compound CCCCOP(=O)(OCCCC)OCCCC STCOOQWBFONSKY-UHFFFAOYSA-N 0.000 description 2
- IOPIMTCNBFTMDS-UHFFFAOYSA-N C=[P] Chemical compound C=[P] IOPIMTCNBFTMDS-UHFFFAOYSA-N 0.000 description 1
- OAICVXFJPJFONN-UHFFFAOYSA-N Phosphorus Chemical compound [P] OAICVXFJPJFONN-UHFFFAOYSA-N 0.000 description 1
- 101710104624 Proline/betaine transporter Proteins 0.000 description 1
- FAPWRFPIFSIZLT-UHFFFAOYSA-M Sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 description 1
- 238000003916 acid precipitation Methods 0.000 description 1
- 238000005349 anion exchange Methods 0.000 description 1
- 238000011001 backwashing Methods 0.000 description 1
- 125000003917 carbamoyl group Chemical group [H]N([H])C(*)=O 0.000 description 1
- 238000004140 cleaning Methods 0.000 description 1
- 150000001875 compounds Chemical class 0.000 description 1
- 230000006866 deterioration Effects 0.000 description 1
- 239000012153 distilled water Substances 0.000 description 1
- 238000001035 drying Methods 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 238000005470 impregnation Methods 0.000 description 1
- 239000007791 liquid phase Substances 0.000 description 1
- 231100001252 long-term toxicity Toxicity 0.000 description 1
- 150000002903 organophosphorus compounds Chemical class 0.000 description 1
- 235000006408 oxalic acid Nutrition 0.000 description 1
- 230000033116 oxidation-reduction process Effects 0.000 description 1
- 235000010344 sodium nitrate Nutrition 0.000 description 1
- 239000004317 sodium nitrate Substances 0.000 description 1
- 238000007711 solidification Methods 0.000 description 1
- 230000008023 solidification Effects 0.000 description 1
- 238000000638 solvent extraction Methods 0.000 description 1
- 238000004017 vitrification Methods 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
-
- 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)
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」
この発明は、超ウラン元素（ＴＲＵ）を含む放射性廃液
を抽出剤を用いて容易かつ経済的に処理する施設におけ
る。溶出抽出剤の回収方法に関するものである。DETAILED DESCRIPTION OF THE INVENTION "Industrial Application Field" The present invention is directed to a facility that easily and economically processes radioactive waste liquid containing transuranium elements (TRU) using an extractant. The present invention relates to a method for recovering an eluted extractant.
「従来の技術」
周知のように、使用済み核燃料に対しては、再処理によ
りＵ−Ｐｕを回収し、残りの高レベル放射性廃液をガラ
ス固化する方法が我が国において採択されている。廃棄
物の処理処分という点から、高レベル放射性廃液中に含
まれる長寿命α放射体である超ウラン元素（ＴＲＵ）を
分離し、これを効果的に貯蔵管理または消滅処理するの
からつとら妥当な考え方であり、従来、再処理は、使用
済み燃料を硝酸に溶解して種々の酸化還元処理を行なぃ
、ＴＢＰ（リン酸トリブチル）を用いた溶媒抽出でＵと
Ｐｕとを分離する、いわゆるＰｕｒｅｘ法が用いられて
いる。"Prior Art" As is well known, Japan has adopted a method for recovering U-Pu from spent nuclear fuel through reprocessing and vitrifying the remaining high-level radioactive waste liquid. From the point of view of waste treatment and disposal, it is appropriate to separate transuranium elements (TRU), which are long-lived alpha emitters, contained in high-level radioactive waste liquid and effectively store and manage or eliminate them. Conventionally, reprocessing involves dissolving spent fuel in nitric acid and performing various oxidation-reduction treatments, and separating U and Pu through solvent extraction using TBP (tributyl phosphate). The so-called Purex method is used.
ところで、前記従来の放射性廃液の処理方法において、
Ｕ、Ｐｕ除去後の高レベル放射性廃液は、これをガラス
固化して貯蔵管理しているが、長期的な毒性を持つＴ
ＲＵが含まれているため、放射性廃棄物（ガラス固化物
）の貯蔵管理に膨大なコストを要している。By the way, in the conventional radioactive waste liquid treatment method,
The high-level radioactive waste liquid after removing U and Pu is stored and managed by vitrifying it, but T, which has long-term toxicity,
Because it contains RU, storage and management of radioactive waste (vitrified waste) requires enormous costs.
そこで、本願発明者は、ＴＲＵを含む放射性廃液からＴ
ＲＵを容易かつ経済的に分離除去し、廃液固化体を低減
することができる以下のような構成の放射性廃液の処理
方法を発明した。Therefore, the inventor of the present application has developed a method to extract TRU from radioactive waste liquid containing TRU.
We have invented a method for treating radioactive waste liquid having the following configuration, which can easily and economically separate and remove RU and reduce the amount of solidified waste liquid.
この処理方法は、まず、高酸濃度にした放射性廃液から
陰イオン交換等によりＵ、−Ｐｕを分離除去し、Ｕ、Ｐ
ｕ除去後のＴＲＵを含む放射性廃液をＣＭ Ｐ （ｃａ
ｒｂａｍｏｙｌ ｍｅｔｈｙｌｅｎｅ ｐｈｏｓｐｈｏ
ｎａｔｅ）またはＣＭ Ｐ Ｏ（ｃａｒｂａａｏｙｌ
５ｅｔｈｙｌｅｎｅ ｐｈｏｓｐｈｉｎｅｏｘｉｄｅ）
を含浸させた固体支持体（カラム）中を通過させること
によりこの固体支持体に液中のＴＲＵを吸着させて前記
廃液からＴｒ（Ｕを除去するとともに、前記固体支持体
に吸着したＴＲＵを溶離して回収することを特徴とする
方法である。This treatment method first separates and removes U and -Pu from radioactive waste liquid with a high acid concentration by anion exchange, etc.
The radioactive waste liquid containing TRU after u removal is CMP (ca
rbamoyl methylene phosphor
nate) or CM PO (carbaaoyl
5ethylene phosphone oxide)
By passing it through a solid support (column) impregnated with Tr (U), the solid support adsorbs TRU in the liquid to remove Tr (U) from the waste liquid and elute the TRU adsorbed on the solid support. This method is characterized in that it is collected by
前記構成におけるＣ Ｍ Ｐ （ｃａｒｂａａ＋ｏｙｌ
ａｅｔｈｙｌｅｎｅｐｈｏｓｐｈｏｎａｔｅ）または
ＣＭ Ｐ Ｏ（ｃａｒｂａｍｏｙｌ ５ｅｔｈｙ−１ｅ
ｎｅ ｐｈｏｓｐｈｉｒ＋ｅ ｏｘｉｄｅ）は、３価、
４価、６価のアクチノイド元素の抽出、特に３価のアク
チノイド元素の抽出が可能な化合物（二座配位系有機リ
ン化合物）として最近注目され始めた抽出剤である。こ
の方法においては、このＣＭＰＳＣＭＰＯをイオン交換
樹脂等の固体支持体に含浸させろことによって、比較的
高価なＣＭＰ％ＣＭＰＯの流損失を防止するとともに、
ＣＭＰ、ＣＭＰＯに抽出させたＴＲＵの固定を容易にし
ている。C M P (carbaa+oil
aethylenephosphonate) or CM PO (carbamoyl 5ethy-1e
ne phosphor+e oxide) is trivalent,
It is an extractant that has recently begun to attract attention as a compound (bidentate organic phosphorus compound) capable of extracting tetravalent and hexavalent actinide elements, especially trivalent actinide elements. In this method, by impregnating this CMPSCMPO into a solid support such as an ion exchange resin, flow loss of relatively expensive CMP%CMPO is prevented, and
This facilitates the fixation of TRU extracted by CMP and CMPO.
また、前記ＣＭＰまたはＣＭＰＯの含浸固体支持体には
、ＣＭＰまたはＣＭＰＯの含浸量が多く、ＣＭＰまたは
ＣＭＰＯが溶出しにくい樹脂を使用する。例えば、Ａｍ
ｂｅｒｌｉｔｅ ＸＡＤ−４（商品名；非極性のポリス
チレン−ＤＶＢ樹脂、巨大網状構造）が好適である。こ
のような固体支持体にＣＭＰまたはＣＭＰＯ（抽出剤）
を含浸させて構成する吸着剤は、まず前記樹脂をアセト
ンで洗浄して不純物を除去し・これを減圧乾燥したもの
に抽出剤を含浸させて調製する。Further, for the solid support impregnated with CMP or CMPO, a resin is used that has a large amount of CMP or CMPO impregnated and that CMP or CMPO is difficult to elute. For example, Am
berlite XAD-4 (trade name; non-polar polystyrene-DVB resin, giant network structure) is preferred. CMP or CMPO (extractant) on such a solid support
An adsorbent impregnated with is prepared by first washing the resin with acetone to remove impurities, drying it under reduced pressure, and impregnating the resin with an extractant.
このようにして調製した吸着剤を塔内に充填してカラム
を形成し、このカラム中にＵ、Ｐｕ除去後のＴＲＵを含
む放射性廃液を通過させれば・液中のＴＲＵをカラムに
吸着させることができ、これによって容易にＴ［ＩＵの
除去を行なうことができろ。カラムに吸着させたＴＲＵ
は希酸溶液の洗浄により容易に溶離することができ、Ｔ
ｎＵを他の核分裂生成物から分離して回収することがで
きる。If the adsorbent prepared in this way is packed into a column to form a column, and the radioactive waste liquid containing TRU after U and Pu removal is passed through this column, the TRU in the liquid will be adsorbed to the column. This allows for easy removal of T[IU. TRU adsorbed on column
can be easily eluted by washing with dilute acid solution, and T
nU can be separated and recovered from other fission products.
７［発明が解決しようとする問題点］
ところで、前記のようにして支持体に含浸された抽出剤
は前記支持体のカラムに通水することにより、その溶解
度（〜５０Ｇｐｐｍ）に応じて溶出されてい（。調製し
た吸着カラムに蒸留水を通すと、通水初期に多量の抽出
剤が流出し、その後は一定濃度（４２０ＰＰＩＩ）での
溶出となる。一定漢度での溶出は抽出剤の水に対する溶
解度によるものであるが、初期の多量の流出は抽出剤の
含浸工程で余剰抽出剤として除去しきれないものが流出
したことによるものである。7 [Problems to be Solved by the Invention] By the way, the extractant impregnated into the support as described above is eluted according to its solubility (~50 Gppm) by passing water through the column of the support. (When distilled water is passed through the prepared adsorption column, a large amount of extractant flows out at the beginning of the water flow, and thereafter it elutes at a constant concentration (420 PPII). The initial large amount of outflow is due to the outflow of excess extractant that could not be completely removed during the extractant impregnation process.
このような初期の流出抽出剤およびその後の溶出抽出剤
は、抽出剤そのものが大変高価なものなので、そのまま
流出させたのでは運転コストの増大を招き、好ましくな
い。Since the extractant itself is very expensive, such an initial effluent extractant and a subsequent eluted extractant are undesirable because it would increase operating costs if they were allowed to flow out as they are.
「問題点を解決するための手段」
ＣＭＰまたはＣＭＰＯ（抽出剤）の含浸支持体として用
いる樹脂（例えば、Ａａｂｅｒｌｉｔｅ ＸＡＤ−４）
は、元来、有機物による汚染水から高分子の有機物を吸
着、回収するためのものであり、吸着した高分子有機物
はアセトンあるいはメタノール溶液等を用いて樹脂相か
ら液相に移動させることができるものである。"Means for Solving the Problem" Resin used as CMP or CMPO (extractant) impregnated support (e.g. Aaberlite XAD-4)
is originally intended for adsorbing and recovering high-molecular organic substances from water contaminated with organic substances, and the adsorbed high-molecular organic substances can be transferred from the resin phase to the liquid phase using acetone or methanol solution, etc. It is something.
従って、この原理を利用し、抽出剤含浸カラムから水相
への溶解度に応じて溶出した抽出剤については、前記抽
出剤含浸カラムの下流に吸着剤カラムを設けて溶出抽出
剤を吸着させ、この溶出抽出剤を吸着した吸着剤カラム
をこのカラムから超ウラン元素の溶離した後にアセトン
やメタノール溶液等の溶出剤で溶出剤相に移し、この抽
出剤含有相を必要に応じてアルカリ洗浄等を行ない、抽
出剤含有相と水相とに分離し、劣化不純物を除去し、そ
の後、抽出剤含有相を減圧下に置くことによって抽出剤
を揮発回収する。Therefore, using this principle, for the extractant eluted from the extractant-impregnated column according to its solubility in the aqueous phase, an adsorbent column is provided downstream of the extractant-impregnated column to adsorb the eluted extractant. After the transuranium elements are eluted from the adsorbent column that has adsorbed the eluting extractant, it is transferred to the eluent phase using an eluent such as acetone or methanol solution, and this extractant-containing phase is washed with alkali, etc. as necessary. The extraction agent is separated into an extractant-containing phase and an aqueous phase, deterioration impurities are removed, and the extractant is then volatilized and recovered by placing the extractant-containing phase under reduced pressure.
なお、初期の多量の流出抽出剤については、固体支持体
に含浸させるために固体支持体カラム中に満たした抽出
液をカラムから抜いた後、カラムをカラムボリュームの
ｌθ倍程度で洗浄し、この洗浄液を前記吸着剤カラムに
通して、抽出剤を吸着し、その後、この吸着カラムをア
セトンやメタノール溶液等の溶出剤で洗浄し、続いて、
抽出剤含有相を減圧下に置くことによって抽出剤を揮発
回収する。In addition, regarding the initial large amount of extractant flowing out, after removing the extract solution filled in the solid support column to impregnate the solid support from the column, wash the column with approximately lθ times the column volume. The washing solution is passed through the adsorbent column to adsorb the extractant, and the adsorption column is then washed with an eluent such as acetone or methanol solution, followed by
The extractant is volatilized and recovered by placing the extractant-containing phase under reduced pressure.
１作用」
このように、本発明の抽出剤回収方法によれば、高価な
抽出剤を放射性廃液処理系の外に流出させて損失するこ
とがないので、放射性元素抽出能力の高い抽出剤の特質
を低コストに生かすことができ、放射性廃液の経済的処
理に大きく寄与することができる。As described above, according to the extractant recovery method of the present invention, the expensive extractant will not be lost by flowing out of the radioactive waste liquid treatment system. can be utilized at low cost and can greatly contribute to the economical treatment of radioactive waste liquid.
以下、この発明を実施例によりさらに詳しく説明する。Hereinafter, this invention will be explained in more detail with reference to Examples.
「実施例」
第１図に本発明方法を実施するに好適な放射性廃液の処
理装置の概略構成図を示すものである。"Example" FIG. 1 shows a schematic diagram of a radioactive waste liquid treatment apparatus suitable for carrying out the method of the present invention.
周知のように、放射性廃液は、多量のアメリンラム；Ａ
１１を含んでいる。この放射性廃液は、通常、前記Ａｍ
の他にＵ（ＶＩ）、Ｐｕ（■）を含んでおり、これらの
濃度が高い場合にはＡａ＋の吸着容量に影響を及ぼすこ
とが考えられる。そこで、まず、図に示すように、一旦
、廃液供給槽１こ貯えた廃液をポンプ２により濃硝酸溶
液とともに陰イオン交換樹脂塔３に流して、Ｕ、Ｐｕ元
素を除去する。除去したＵ、Ｐｕ成分は希硝酸溶液によ
る逆洗により塔３内のイオン交換樹脂から溶離し、Ｕ−
Ｐｕ貯留槽４に貯えて適宜リサイクルする。As is well known, radioactive waste liquid contains a large amount of amerine rum;
Contains 11. This radioactive waste liquid is usually
In addition, it contains U (VI) and Pu (■), and if their concentrations are high, it is thought that they will affect the adsorption capacity of Aa+. Therefore, first, as shown in the figure, the waste liquid that has been stored in one waste liquid supply tank is sent to the anion exchange resin column 3 together with a concentrated nitric acid solution using the pump 2 to remove U and Pu elements. The removed U and Pu components are eluted from the ion exchange resin in column 3 by backwashing with dilute nitric acid solution, and the U-
It is stored in the Pu storage tank 4 and recycled as appropriate.
Ｕ、Ｐｕ除去後の流出液（Ａａ＋廃液）は、廃液貯留１
５に一時貯留し、この貯留槽５からポンプ６により抽出
部７に供給する。抽出部７は、ｌバッチあたりの廃液処
理量を増すために７ａ（７ａ）、７ｂの２段とし、前段
の塔７ａ、７ａにはＣＭＰ含浸イオン交換樹脂またはＣ
ＭＰＯ含浸イオン交換樹脂を充填し、後段の塔７ｂに溶
出した抽出剤を回収・保持するためのバックアップカラ
ム（吸着剤カラム）を設ける。このバックアップカラム
により、後述するように、前段の塔７ａ、７ａから溶出
した抽出剤を吸着、回収し、抽出剤の溶出によるコスト
の損失を大きく低減化することかできる。The effluent (Aa + waste liquid) after U and Pu removal is transferred to waste liquid storage 1.
The liquid is temporarily stored in a tank 5, and is supplied from this storage tank 5 to an extraction section 7 by a pump 6. The extraction section 7 has two stages, 7a (7a) and 7b, in order to increase the amount of waste liquid processed per 1 batch, and the former columns 7a, 7a are equipped with CMP-impregnated ion exchange resin or CMP-impregnated ion exchange resin.
A backup column (adsorbent column) filled with MPO-impregnated ion exchange resin and used to collect and retain the extractant eluted into the latter column 7b is provided. With this backup column, as will be described later, it is possible to adsorb and recover the extractant eluted from the preceding columns 7a and 7a, thereby greatly reducing the cost loss due to the elution of the extractant.
前記抽出部７からのＡｍが分離除去された流出液は、一
旦ＦＰ廃液貯留槽８へ送り、例えば、硝酸ナトリウムに
対する処理を施して、暫定固化処理を行ない、核分裂生
成物を同化することができる。The effluent from the extraction section 7, from which Am has been separated and removed, is temporarily sent to the FP waste liquid storage tank 8, where it is treated with, for example, sodium nitrate, to perform temporary solidification treatment, and the fission products can be assimilated. .
一方、抽出塔７ａ、７ａ内のカラムに吸着されているＡ
１等のＴＲＵは希硝酸溶液による順法により溶離し、こ
のＡｍを含む廃液は、一時廃液貯留槽９に貯え、ガラス
固化処理を行ない処分するか、あるいは必要な場合には
、シュウ酸沈澱法等によりＡｌ１１を回収する。On the other hand, A adsorbed on the columns in the extraction towers 7a and 7a
The first TRU is eluted using a dilute nitric acid solution, and the waste liquid containing Am is temporarily stored in the waste liquid storage tank 9 and disposed of by vitrification treatment, or if necessary, oxalic acid precipitation method. Al11 is recovered by et al.
前記のようにして’ｒ ＲＵの回収が終了した時点で、
図に示すように、前記塔７ａの下流のバックアップカラ
ム７ｂをアセトンやメタノール溶液等の溶出剤１０で逆
洗し、洗浄液をアルカリ洗浄槽１１にてアルカリ洗浄後
、続いて液−液分離相１２で抽出剤含有相と水相とに分
離してＣＭＰ劣化不純物を除去し、この洗浄後の抽出剤
含有相を減圧槽１３に導き、減圧下で揮発させることに
よって抽出剤を回収する。回収したＣＭＰは塔７ａ、７
ａ内のイオン交換樹脂への再含浸などに使用する。When the recovery of 'r RU is completed as described above,
As shown in the figure, the backup column 7b downstream of the column 7a is backwashed with an eluent 10 such as acetone or methanol solution, and the washing liquid is washed with an alkali in an alkali washing tank 11, followed by a liquid-liquid separation phase 12. The extracted agent-containing phase is separated into an extractant-containing phase and an aqueous phase to remove CMP-degraded impurities, and the extracted agent-containing phase after washing is led to a vacuum tank 13 and evaporated under reduced pressure to recover the extractant. The recovered CMP is sent to towers 7a and 7.
Used for re-impregnating the ion exchange resin in a.
「発明の効果」
以上説明したように、この発明に係る抽出剤回収方法に
よれば、高価な抽出剤を放射性廃液処理系の外に流出さ
せて損失することがないので、放射性廃液の処理施設に
おいて放射性元素抽出能力の高い抽出剤の特質を低コス
トに生かすことができ、放射性廃液の経済的処理に大き
く寄与することができる。"Effects of the Invention" As explained above, according to the extractant recovery method according to the present invention, the expensive extractant will not be lost by flowing out of the radioactive waste liquid treatment system, so The characteristics of an extractant with a high ability to extract radioactive elements can be utilized at low cost, making it possible to greatly contribute to the economical treatment of radioactive waste liquid.
第１図はこの発明方法を実施するのに好適な放射性廃液
の処理装置の一例を示すもので、同装置の概略構成図で
ある。
ｌ・・・・・・廃液供給槽、２．６・・・・・・ポンプ
、３・・・・・・陰イオン交換樹脂塔、
４・・・・・・Ｕ−Ｐｕ貯留槽、５・・目・・廃液貯留
槽、７・・・・・・抽出部、７ａ・・・・・・抽出塔、
７ｂ・・・・・・バックアップカラム（吸着剤カラム）
、８・・・・・・ＦＰ廃液貯留槽、９・・・・・・廃液
貯留槽、ｉｏ・・・・・・ＣＭＰ溶出剤、１１・旧・・
アルカリ洗浄槽、１２・・・・・・液−液分離槽、１３
・・・・・・減圧槽。FIG. 1 shows an example of a radioactive waste liquid treatment apparatus suitable for carrying out the method of the present invention, and is a schematic diagram of the apparatus. 1... Waste liquid supply tank, 2.6... Pump, 3... Anion exchange resin tower, 4... U-Pu storage tank, 5.・Eye: Waste liquid storage tank, 7: Extraction section, 7a: Extraction tower,
7b・・・Backup column (adsorbent column)
, 8...FP waste liquid storage tank, 9... Waste liquid storage tank, io...CMP eluent, 11. Old...
Alkaline cleaning tank, 12...Liquid-liquid separation tank, 13
・・・・・・Reducing pressure tank.
Claims (1)
Ｕ、Ｐｕ除去後の超ウラン元素を含む放射性廃液をＣＭ
Ｐ（ｃａｒｂａｍｏｙｌ ｍｅｔｈｙｌｅｎｅ ｐｈｏ
ｓｐｈ−ｏｎａｔｅ）またはＣＭＰＯ（ｃａｒｂａｍｏ
ｙｌ ｍｅｔｈｙｌｅｎｅ ｐｈ−ｏｓｐｈｉｎｅ ｏ
ｘｉｄｅ）がらなる抽出剤を含浸させた固体支持体カラ
ム中を通過させることによりこの固体支持体カラムに超
ウラン元素を吸着させて前記廃液から超ウラン元素を除
去するとともに、前記固体支持体に吸着した超ウラン元
素を溶離して回収する放射性廃液処理施設における溶出
抽出剤の回収方法であって、 前記固体支持体カラムの下流に吸着剤カラムを設けて前
記溶出抽出剤を吸着させ、この溶出抽出剤を吸着した吸
着剤カラムを前記超ウラン元素の溶離後にアセトンやメ
タノール溶液等の溶出剤で洗浄し、抽出剤含有相に移し
、前記抽出剤含有相を減圧下に置くことによって前記抽
出剤を回収、再利用することを特徴とする放射性元素抽
出剤の回収方法。[Claims] Separating and removing U and Pu from radioactive waste liquid with high acid concentration,
CM radioactive waste liquid containing transuranic elements after U and Pu removal
P (carbamoyl methylene pho
sph-onate) or CMPO (carbamo
yl methylene ph-osphine o
The transuranic elements are removed from the waste liquid by passing the transuranic elements through a solid support column impregnated with an extractant consisting of xide), and the transuranic elements are adsorbed onto the solid support column. A method for recovering an eluate extractant in a radioactive waste liquid treatment facility in which transuranic elements are eluted and recovered, the method comprising: providing an adsorbent column downstream of the solid support column to adsorb the eluate extractant; After elution of the transuranic elements, the adsorbent column that has adsorbed the agent is washed with an eluent such as acetone or methanol solution, transferred to an extractant-containing phase, and the extractant is removed by placing the extractant-containing phase under reduced pressure. A method for recovering a radioactive element extractant, characterized by recovering and reusing it.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP4119387A JPH0797157B2 (en) | 1987-02-24 | 1987-02-24 | Method for recovering radioactive element extractant |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP4119387A JPH0797157B2 (en) | 1987-02-24 | 1987-02-24 | Method for recovering radioactive element extractant |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS63208800A true JPS63208800A (en) | 1988-08-30 |
| JPH0797157B2 JPH0797157B2 (en) | 1995-10-18 |
Family
ID=12601587
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP4119387A Expired - Lifetime JPH0797157B2 (en) | 1987-02-24 | 1987-02-24 | Method for recovering radioactive element extractant |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH0797157B2 (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN112485273A (en) * | 2020-11-11 | 2021-03-12 | 苏州热工研究院有限公司 | Collecting device and detection method for radioactive iron in water body |
-
1987
- 1987-02-24 JP JP4119387A patent/JPH0797157B2/en not_active Expired - Lifetime
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN112485273A (en) * | 2020-11-11 | 2021-03-12 | 苏州热工研究院有限公司 | Collecting device and detection method for radioactive iron in water body |
Also Published As
| Publication number | Publication date |
|---|---|
| JPH0797157B2 (en) | 1995-10-18 |
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