JPH04301799A - Treating method for radioactive waste fluid - Google Patents
Treating method for radioactive waste fluidInfo
- Publication number
- JPH04301799A JPH04301799A JP9169891A JP9169891A JPH04301799A JP H04301799 A JPH04301799 A JP H04301799A JP 9169891 A JP9169891 A JP 9169891A JP 9169891 A JP9169891 A JP 9169891A JP H04301799 A JPH04301799 A JP H04301799A
- Authority
- JP
- Japan
- Prior art keywords
- water glass
- heavy metals
- added
- waste fluid
- radioactive waste
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Pending
Links
- 239000002901 radioactive waste Substances 0.000 title claims abstract description 22
- 238000000034 method Methods 0.000 title claims description 34
- 239000012530 fluid Substances 0.000 title abstract 10
- 229910001385 heavy metal Inorganic materials 0.000 claims abstract description 36
- 239000002244 precipitate Substances 0.000 claims abstract description 35
- 235000019353 potassium silicate Nutrition 0.000 claims abstract description 34
- NTHWMYGWWRZVTN-UHFFFAOYSA-N sodium silicate Chemical compound [Na+].[Na+].[O-][Si]([O-])=O NTHWMYGWWRZVTN-UHFFFAOYSA-N 0.000 claims abstract description 33
- 229910052742 iron Inorganic materials 0.000 claims abstract description 17
- 229910052804 chromium Inorganic materials 0.000 claims abstract description 15
- 229910052759 nickel Inorganic materials 0.000 claims abstract description 15
- NBIIXXVUZAFLBC-UHFFFAOYSA-N Phosphoric acid Chemical compound OP(O)(O)=O NBIIXXVUZAFLBC-UHFFFAOYSA-N 0.000 claims abstract description 13
- QAOWNCQODCNURD-UHFFFAOYSA-N Sulfuric acid Chemical compound OS(O)(=O)=O QAOWNCQODCNURD-UHFFFAOYSA-N 0.000 claims abstract description 13
- 159000000003 magnesium salts Chemical class 0.000 claims abstract description 9
- 229910000147 aluminium phosphate Inorganic materials 0.000 claims abstract description 4
- 239000007788 liquid Substances 0.000 claims description 31
- 229910052770 Uranium Inorganic materials 0.000 claims description 13
- 239000012670 alkaline solution Substances 0.000 claims description 5
- 239000002699 waste material Substances 0.000 abstract description 27
- HEMHJVSKTPXQMS-UHFFFAOYSA-M Sodium hydroxide Chemical compound [OH-].[Na+] HEMHJVSKTPXQMS-UHFFFAOYSA-M 0.000 abstract description 18
- 239000002904 solvent Substances 0.000 abstract description 12
- 239000011777 magnesium Substances 0.000 abstract description 9
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 abstract description 7
- 239000002253 acid Substances 0.000 abstract description 7
- 229910017604 nitric acid Inorganic materials 0.000 abstract description 7
- QGZKDVFQNNGYKY-UHFFFAOYSA-N Ammonia Chemical compound N QGZKDVFQNNGYKY-UHFFFAOYSA-N 0.000 abstract description 6
- 150000007513 acids Chemical class 0.000 abstract description 5
- 239000003513 alkali Substances 0.000 abstract description 4
- 235000019795 sodium metasilicate Nutrition 0.000 abstract description 4
- 229910052911 sodium silicate Inorganic materials 0.000 abstract description 4
- 229910021529 ammonia Inorganic materials 0.000 abstract description 3
- 230000003472 neutralizing effect Effects 0.000 abstract 1
- XEEYBQQBJWHFJM-UHFFFAOYSA-N iron Substances [Fe] XEEYBQQBJWHFJM-UHFFFAOYSA-N 0.000 description 16
- 239000011651 chromium Substances 0.000 description 15
- 239000000243 solution Substances 0.000 description 15
- PXHVJJICTQNCMI-UHFFFAOYSA-N nickel Substances [Ni] PXHVJJICTQNCMI-UHFFFAOYSA-N 0.000 description 14
- 238000000605 extraction Methods 0.000 description 13
- 239000000706 filtrate Substances 0.000 description 12
- 239000003792 electrolyte Substances 0.000 description 8
- 238000001914 filtration Methods 0.000 description 7
- 229910052750 molybdenum Inorganic materials 0.000 description 7
- 235000011007 phosphoric acid Nutrition 0.000 description 7
- 239000004568 cement Substances 0.000 description 6
- 230000000694 effects Effects 0.000 description 5
- 235000011149 sulphuric acid Nutrition 0.000 description 5
- STCOOQWBFONSKY-UHFFFAOYSA-N tributyl phosphate Chemical compound CCCCOP(=O)(OCCCC)OCCCC STCOOQWBFONSKY-UHFFFAOYSA-N 0.000 description 5
- 229910000975 Carbon steel Inorganic materials 0.000 description 4
- 239000008346 aqueous phase Substances 0.000 description 4
- 239000010962 carbon steel Substances 0.000 description 4
- 239000008151 electrolyte solution Substances 0.000 description 4
- 238000005498 polishing Methods 0.000 description 4
- 238000011084 recovery Methods 0.000 description 4
- 230000033228 biological regulation Effects 0.000 description 3
- 238000005202 decontamination Methods 0.000 description 3
- 229910052751 metal Inorganic materials 0.000 description 3
- 239000002184 metal Substances 0.000 description 3
- 150000002739 metals Chemical class 0.000 description 3
- XTAZYLNFDRKIHJ-UHFFFAOYSA-N n,n-dioctyloctan-1-amine Chemical compound CCCCCCCCN(CCCCCCCC)CCCCCCCC XTAZYLNFDRKIHJ-UHFFFAOYSA-N 0.000 description 3
- 238000006386 neutralization reaction Methods 0.000 description 3
- 229910052757 nitrogen Inorganic materials 0.000 description 3
- 230000001105 regulatory effect Effects 0.000 description 3
- 150000003839 salts Chemical class 0.000 description 3
- 239000010935 stainless steel Substances 0.000 description 3
- 229910001220 stainless steel Inorganic materials 0.000 description 3
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 3
- 238000005345 coagulation Methods 0.000 description 2
- 230000015271 coagulation Effects 0.000 description 2
- 230000003588 decontaminative effect Effects 0.000 description 2
- 230000007423 decrease Effects 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 239000003085 diluting agent Substances 0.000 description 2
- 238000004090 dissolution Methods 0.000 description 2
- SNRUBQQJIBEYMU-UHFFFAOYSA-N dodecane Chemical compound CCCCCCCCCCCC SNRUBQQJIBEYMU-UHFFFAOYSA-N 0.000 description 2
- XLYOFNOQVPJJNP-UHFFFAOYSA-M hydroxide Chemical compound [OH-] XLYOFNOQVPJJNP-UHFFFAOYSA-M 0.000 description 2
- 238000002156 mixing Methods 0.000 description 2
- 239000000203 mixture Substances 0.000 description 2
- 229940094933 n-dodecane Drugs 0.000 description 2
- 239000003758 nuclear fuel Substances 0.000 description 2
- 238000001556 precipitation Methods 0.000 description 2
- 230000002285 radioactive effect Effects 0.000 description 2
- 238000003756 stirring Methods 0.000 description 2
- VYZAMTAEIAYCRO-UHFFFAOYSA-N Chromium Chemical compound [Cr] VYZAMTAEIAYCRO-UHFFFAOYSA-N 0.000 description 1
- ZOKXTWBITQBERF-UHFFFAOYSA-N Molybdenum Chemical compound [Mo] ZOKXTWBITQBERF-UHFFFAOYSA-N 0.000 description 1
- 229910021333 Na2Si Inorganic materials 0.000 description 1
- 150000003863 ammonium salts Chemical class 0.000 description 1
- 239000012752 auxiliary agent Substances 0.000 description 1
- 150000001721 carbon Chemical class 0.000 description 1
- 238000001514 detection method Methods 0.000 description 1
- 229910021645 metal ion Inorganic materials 0.000 description 1
- 239000011733 molybdenum Substances 0.000 description 1
- 238000010979 pH adjustment Methods 0.000 description 1
- 238000000638 solvent extraction Methods 0.000 description 1
- 239000000126 substance Substances 0.000 description 1
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 1
Landscapes
- Removal Of Specific Substances (AREA)
Abstract
Description
【発明の詳細な説明】
【０００１】
【産業上の利用分野】本発明はＦｅ、Ｎｉ、Ｃｒなどの
重金属とＵを含む放射性廃液の処理方法に関する。
【０００２】
【従来の技術】従来、Ｕで汚染された炭素鋼およびステ
ンレス鋼などの金属の除染方法のひとつとして、電界研
磨法が知られており、この電界研磨法に用いられる電解
液として、３０〜７０％濃度のＨ２ＳＯ４やＨ３ＰＯ４
が知られている。前記電界研磨法を実施して金属の除染
を行なうと、処理後の電解液中にはＵを始め、Ｆｅ、Ｎ
ｉ、Ｃｒなどの重金属が含まれるようになる。この電解
液は、一定量以上の元素が溶解すると除染効果が低下す
るので、除染効果が低下した時点で交換するようにして
いる。
【０００３】通常、このような使用済みの電解液は、金
属イオンの濃度が高く、処理が困難なために、放射性廃
棄物としてそのまま保管されている場合が多く、処理面
で問題になっている。
【０００４】従来一般に、溶液中に溶解しているＵを抽
出する方法として溶媒抽出法が採用されており、前記の
ようなＨ２ＳＯ４系やＨ３ＰＯ４系の溶液からＵを抽出
する際の溶媒としては、トリオクチルアミン（ＴＯＡ）
を用いることが多く、ＨＮＯ３系の溶液からＵを抽出す
る際の溶媒としては、トリブチル燐酸（ＴＢＰ）を用い
ることが多い。
【０００５】
【発明が解決しようとする課題】しかしながらこの種の
廃液をＴＯＡやＴＢＰで処理してＵを回収した後の抽出
廃液の中にも小量のＵや重金属が残留している。従来、
この小量のＵや重金属を除去するには、これらを溶液か
ら沈殿物とし、これをセメント固化させた後に除去して
いる。ところが、この沈澱物除去後の廃液の中にも微量
のＵと重金属が含まれているので、この微量のＵと重金
属を除去しなければ廃液を廃棄できない問題がある。
【０００６】従来、前記微量のＵと重金属の除去を行な
う方法として、廃液中の重金属を凝集沈殿で除去する方
法が知られ、凝集沈殿法の一つに水ガラス法がある。即
ち、水ガラスの主成分であるＮａ２ＳｉＯ３は、強酸と
反応して沈殿を生じたり、あるいは、ゲル状態となって
重金属を凝集沈殿することが知られているので、この性
質を利用して沈殿除去を行なうことができるのである。
なお、このような水ガラス殿物は、強酸以外の物質、例
えば、アンモニウム塩の存在によっても生成させ得るこ
とが知られている。
【０００７】しかしながら、前記の沈殿物除去後の廃液
は、水ガラスを添加しただけでは沈殿を生じないことが
明らかになっている。従って何等かの補助剤を添加して
水ガラス殿物を生成させる必要があるが、廃棄時の最終
的なｐＨ調整を考慮すると、強酸を添加することは得策
ではなく、また、アンモニアを添加することは排出規制
があるために問題が多い。
【０００８】本発明は前記課題を解決するためになされ
たもので、微量のＵと重金属を含む硫酸系あるいはリン
酸系の廃液に強酸やアンモニアを加えることなく水ガラ
ス殿物を生じさせてＵと重金属の除去ができる放射性廃
液の処理方法を提供することを目的とする。
【０００９】
【課題を解決するための手段】請求項１記載の発明は前
記課題を解決するために、Ｆｅ、Ｎｉ、Ｃｒなどの重金
属とＵを含む硫酸系あるいはリン酸系の放射性廃液に、
水溶性のマグネシウム塩を溶解した後に水ガラスを添加
し、更にアルカリ溶液を添加してｐＨを８〜９に調整し
、水ガラス殿物を生成させてＦｅ、Ｎｉ、Ｃｒなどの重
金属とＵを除去するものである。
【００１０】
【作用】微量の重金属とＵを含む放射性廃液に、Ｍｇ（
ＮＯ３）２・６Ｈ２Ｏなどの水溶性のマグネシウム塩を
溶解した場合、Ｍｇは、弱アルカリ領域でＮａ２ＳｉＯ
３あるいは重金属と複塩を形成し、水ガラス殿物を生じ
る要因となる。そして、この水ガラス殿物にはＦｅ、Ｎ
ｉ、Ｃｒなどの重金属とＵが凝集される。よってこの水
ガラス殿物を除去することで廃液処理ができる。
【００１１】以下、図面を参照して本発明を更に詳細に
説明する。図１はＵと重金属を含むＨ２ＳＯ４系あるい
はＨ３ＰＯ４系の放射性の使用済電解廃液を処理する工
程に本発明方法を適用した一実施例を示すフロー図であ
る。
ここで処理される放射性使用済電解廃液とは、Ｆｅ（鉄
）、Ｎｉ（ニッケル）、Ｃｒ（クロム）、Ｍｏ（モリブ
デン）などの重金属とＵ（ウラン）を含むもので、原子
力施設においてＵに汚染された炭素鋼あるいはステンレ
ス鋼を処理する場合に、これらを電界研磨法によって処
理した後に残されるものである。この使用済電解廃液に
は、Ｈ２ＳＯ４やＨ３ＰＯ４が高濃度で含まれ、Ｕや重
金属も多量含まれている。
【００１２】前記使用済電解廃液には多量のＵと重金属
が含まれている。この使用済電解廃液を処理して当該廃
液からＵと重金属を除去することで微量のＵと重金属を
含む放射性廃液が、生じ、この放射性廃液からＵと重金
属を除去する方法が本発明方法である。そこで本発明方
法を説明する前に使用済電解廃液を処理してＵと重金属
を除去する方法について図１を基に以下に説明する。
【００１３】まず、使用済電解廃液にＨＮＯ３などの強
酸を混合工程１において混合し、次に抽出工程２におい
て溶媒を添加して抽出処理を行ない、抽出廃液を中和工
程３においてＮａＯＨなどのアルカリ溶液で中和した後
に瀘過工程４で瀘過する。抽出工程２で得られたＵを含
む溶媒は、逆抽出工程５で水を加えて逆抽出を行なって
大部分のＵを回収するとともにここで生じた溶媒は再使
用する。瀘過後の沈殿物はセメントによる固化工程１０
によりセメント固化してセメント固化体を得、これは放
射性廃棄物として保管する。
【００１４】前記瀘過工程４で生じる瀘液が本発明方法
で処理するべき放射性廃液である。
【００１５】この放射性廃液（瀘液）を本発明方法によ
って処理するには、溶解工程６において放射性廃液にＭ
ｇ（ＮＯ３）２・６Ｈ２Ｏなどの水溶性のマグネシウム
塩を添加して溶解し、更にＮａ２ＳｉＯ３を主成分とす
る水ガラスを添加する。ここで添加溶解するマグネシウ
ム塩としてＭｇ（ＮＯ３）２・６Ｈ２Ｏの他に、ＭｇＣ
ｌ２・６Ｈ２Ｏなどを用いても良い。前記のマグネシウ
ム塩においてＭｇは、弱アルカリ性領域で、Ｎａ２Ｓｉ
Ｏ３あるいは重金属と複塩を形成し、水ガラス殿物を生
じる要因となる。
【００１６】次に調整工程７においてＮａＯＨなどのア
ルカリ溶液を添加してｐＨを８〜９に調節し、水ガラス
殿物を生じさせる。ここで添加するアルカリ溶液は、Ｎ
ａＯＨの他にＫＯＨなどでも良い。この後に瀘過工程８
において瀘過を行ない、処理済廃液と水ガラス殿物に分
離し、処理済廃液は排出し、水ガラス殿物は保管する。
この水ガラス殿物には放射性廃液中の微量のＵと重金属
が凝集されている。
【００１７】
【実施例】「実施例１」Ｈ２ＳＯ４濃度約３０％、Ｕと
ＦｅとＮｉおよびＭｏをそれぞれ約５ｇ／リットル含有
する使用済電解液を処理した例を示す。まず、以下の手
順により使用済電解液中のＵを回収した。
【００１８】■前記使用済電解液１．８リットルに６２
％ＨＮＯ３を添加する。（ＨＮＯ３濃度１．４Ｎ）■Ｈ
ＮＯ３を添加した使用済電解液（Ｕ含有量４．８６ｇ／
リットル）１容量に対し、ＴＢＰと希釈剤のｎドデカン
を容量比３：７で混合した溶媒１容量を加え、分液ロー
トを用いて室温で５分間震盪し、水相中のＵを抽出する
。
■Ｕを含有する溶媒１容量に対し、純水１容量を加え、
分液ロートを用いて室温で５分間震盪し、溶媒中のＵを
逆抽出する。
■抽出操作を４段、逆抽出操作をそれぞれの抽出段ごと
に水相にＵが検出されなくなるまで（最高６段）繰り返
す。
【００１９】以上の手段で回収した合計９．５２ｇ（回
収率約９８％）のＵ中のＦｅ、Ｎｉ、Ｍｏの含有量を表
１に示すが、いずれの値も核燃料分野における含有上限
値以下である。
【００２０】
【表１】
【００２１】次いでＵ回収後の抽出廃液に、２５％Ｎａ
ＯＨ溶液を加えてｐＨを６〜７．０に調整し、含有する
重金属（主としてＦｅ、Ｎｉ、ＣｒおよびＭｏ）を水酸
化物の沈殿として瀘別・分離し、沈殿物はセメント固化
した。
【００２２】この際に生じた瀘液を放射性廃液として見
立て、これに本発明方法を適用し、処理する。■瀘液１
リットルに対して４ｇのＭｇ（ＮＯ３）２・６Ｈ２Ｏを
添加して溶解する。■瀘液１リットルあたり、２ｇの水
ガラスを添加し、更にＮａＯＨを添加することにより瀘
液のｐＨを８〜９に調整した。■１時間攪拌後、２４時
間放置した。■２４時間の放置後、水ガラス殿物を定量
、瀘紙により瀘過した。
【００２３】以上の工程で処理された瀘液中のＵおよび
Ｃｒの処理前後の含有量を表２に示す。
【００２４】
【表２】
【００２５】Ｕの規制値は濃縮度によって若干異なるが
、約１ｐｐｍ、総量Ｃｒも同じく１ｐｐｍであることか
ら、表２に示す結果は規制値を十分に下回っていること
が判明した。このことから、本発明方法の効果を確認で
きた。
【００２６】「実施例２」Ｈ３ＰＯ４濃度約３０％、Ｕ
とＦｅとＮｉおよびＭｏをそれぞれ約５ｇ／リットル含
有する使用済電解液を処理した例を示す。
【００２７】まず、以下の手順により使用済電解液中の
Ｕを回収した。■使用済電解液１リットルに６２％ＨＮ
Ｏ３を添加する。（ＨＮＯ３濃度６．９Ｎ）■ＨＮＯ３
を添加した使用済電解液（Ｕ含有量２．４２ｇ／リット
ル）１容量に対し、ＴＢＰと希釈剤のｎドデカンを容量
比３：７で混合した溶媒１容量を加え、分液ロートを用
いて室温で５分間震盪し、水相中のＵを抽出する。■Ｕ
を含有する溶媒１容量に対し、純水１容量を加え、分液
ロートを用いて室温で５分間震盪し、溶媒中のＵを逆抽
出する。■抽出操作を４段、逆抽出操作をそれぞれの抽
出段ごとに水相にＵが検出されなくなるまで（最高５段
）繰り返す。
【００２８】以上の手段で回収した合計５．６８ｇ（回
収率約９７％）のＵ中のＦｅ、Ｎｉ、Ｍｏの含有量を表
３に示すが、いずれの値も核燃料分野における含有上限
値以下である。
【００２９】
【表３】
【００３０】次いで、Ｕ回収後の抽出廃液に、２５％Ｎ
ａＯＨ溶液を加えてｐＨを６．５〜７．０に調整し、含
有する重金属（主としてＦｅ、Ｎｉ、ＣｒおよびＭｏ）
を水酸化物の沈殿として瀘別・分離し、沈殿物はセメン
ト固化した。
【００３１】この際に生じた瀘液を放射性廃液として見
立て、これに本発明方法を適用し、処理する。■瀘液１
リットルに対して４ｇのＭｇ（ＮＯ３）２・６Ｈ２Ｏを
添加して溶解する。■瀘液１リットルあたり、２ｇの水
ガラスを添加し、更にＮａＯＨを添加することにより瀘
液のｐＨを８〜９に調整した。■１時間攪拌後、２４時
間放置した。■２４時間の放置後、水ガラス殿物を定量
、瀘紙により瀘過した。
【００３２】以上の工程で処理された瀘液中のＵおよび
Ｃｒの処理前後の含有量を表４に示す。
【表４】
【００３３】Ｕの規制値は濃縮度によって
若干異なるが、約１ｐｐｍ、総量Ｃｒも同じく１ｐｐｍ
であることから、表４に示す結果は規制値を十分に下回
っていることが判明した。このことから、本発明方法の
効果を確認できた。
【００３４】
【発明の効果】以上説明したように本発明によれば、Ｕ
に汚染された炭素鋼やステンレス鋼などの金属を電界研
磨して生じる廃液を廃液処理して大部分のＵと重金属を
除去した後において、微量の重金属とＵを含む放射性廃
液に、Ｍｇ（ＮＯ３）２・６Ｈ２Ｏなどの水溶性のマグ
ネシウム塩を溶解し、その後にアルカリ溶液を添加して
ｐＨ調整し、水ガラス殿物を生成させることができる。
この場合、Ｍｇは弱アルカリ性でＮａ２ＳｉＯ３あるい
は重金属と複塩を形成し、水ガラス殿物を生じる要因と
なる。従ってこの水ガラス殿物にはＦｅ、Ｎｉ、Ｃｒな
どの重金属とＵが凝集される。よってこの水ガラス殿物
を放射性廃液から除去することでＦｅ、Ｎｉ、Ｃｒなど
の重金属とＵを除去することができ、放射性廃液の処理
ができる効果がある。また、本発明方法によれば、廃液
中のＵを検出限界以下に完全に除去できるとともに、重
金属を規制値以下に減少させることができる。Description: [0001] The present invention relates to a method for treating radioactive waste liquid containing heavy metals such as Fe, Ni, and Cr and U. [0002] Conventionally, an electrolytic polishing method has been known as one of the decontamination methods for metals such as carbon steel and stainless steel contaminated with U, and the electrolytic solution used in this electrolytic polishing method is , 30-70% concentration of H2SO4 and H3PO4
It has been known. When the electrolytic polishing method described above is carried out to decontaminate metals, U, Fe, N, etc. are present in the electrolyte after treatment.
Heavy metals such as i and Cr are included. The decontamination effect of this electrolytic solution decreases when more than a certain amount of elements are dissolved, so it is replaced when the decontamination effect decreases. [0003] Normally, such used electrolytes have a high concentration of metal ions and are difficult to dispose of, so they are often stored as radioactive waste, which poses a problem in terms of disposal. . [0004] Conventionally, a solvent extraction method has generally been adopted as a method for extracting U dissolved in a solution, and as a solvent when extracting U from the above-mentioned H2SO4-based or H3PO4-based solutions, Trioctylamine (TOA)
is often used, and tributyl phosphate (TBP) is often used as a solvent when extracting U from an HNO3-based solution. [0005] However, after treating this type of waste liquid with TOA or TBP and recovering U, a small amount of U and heavy metals remain in the extracted waste liquid. Conventionally,
In order to remove this small amount of U and heavy metals, they are made into a precipitate from the solution, which is solidified into cement and then removed. However, since the waste liquid after removing the precipitate also contains trace amounts of U and heavy metals, there is a problem that the waste liquid cannot be disposed of unless these trace amounts of U and heavy metals are removed. Conventionally, as a method for removing trace amounts of U and heavy metals, there has been known a method of removing heavy metals from waste liquid by coagulation and precipitation, and one of the coagulation and precipitation methods is the water glass method. In other words, Na2SiO3, which is the main component of water glass, is known to react with strong acids to form a precipitate, or become a gel to coagulate and precipitate heavy metals, so this property can be used to remove the precipitate. It is possible to do this. It is known that such water glass precipitates can also be produced by the presence of substances other than strong acids, such as ammonium salts. [0007] However, it has been revealed that the waste liquid after removing the precipitate does not produce precipitate simply by adding water glass. Therefore, it is necessary to add some kind of auxiliary agent to generate water glass precipitate, but considering the final pH adjustment at the time of disposal, it is not a good idea to add a strong acid, and it is not advisable to add ammonia. This is problematic due to emission regulations. The present invention was made in order to solve the above-mentioned problems, and it is possible to produce water glass precipitates without adding strong acids or ammonia to sulfuric acid-based or phosphoric acid-based waste liquids containing trace amounts of U and heavy metals. The purpose of the present invention is to provide a method for treating radioactive waste liquid that can remove radioactive waste and heavy metals. [Means for Solving the Problems] In order to solve the above-mentioned problems, the invention as set forth in claim 1 provides a solution to a sulfuric acid-based or phosphoric acid-based radioactive waste liquid containing heavy metals such as Fe, Ni, and Cr and U.
After dissolving the water-soluble magnesium salt, water glass is added, and an alkaline solution is further added to adjust the pH to 8 to 9 to form a water glass precipitate, which removes heavy metals such as Fe, Ni, and Cr and U. It is to be removed. [Operation] Mg (
When a water-soluble magnesium salt such as NO3)2.6H2O is dissolved, Mg becomes Na2SiO in the weak alkaline region.
3 or forms a double salt with heavy metals, which causes water glass precipitate. In this water glass precipitate, Fe, N
Heavy metals such as i, Cr and U are aggregated. Therefore, waste liquid can be treated by removing this water glass precipitate. The present invention will be explained in more detail below with reference to the drawings. FIG. 1 is a flow diagram showing an embodiment in which the method of the present invention is applied to a process of treating a radioactive spent electrolytic waste solution of H2SO4 or H3PO4 containing U and heavy metals. The radioactive spent electrolytic waste liquid treated here contains heavy metals such as Fe (iron), Ni (nickel), Cr (chromium), Mo (molybdenum), and U (uranium), and is When processing contaminated carbon steel or stainless steel, it is what remains after processing these by electropolishing. This used electrolytic waste solution contains H2SO4 and H3PO4 at high concentrations, and also contains large amounts of U and heavy metals. [0012] The spent electrolytic waste liquid contains a large amount of U and heavy metals. By treating this used electrolytic waste solution and removing U and heavy metals from the waste solution, a radioactive waste solution containing trace amounts of U and heavy metals is generated, and the method of the present invention is to remove U and heavy metals from this radioactive waste solution. . Therefore, before explaining the method of the present invention, a method for treating a used electrolytic waste solution to remove U and heavy metals will be described below with reference to FIG. 1. [0013] First, a strong acid such as HNO3 is mixed with the used electrolytic waste liquid in the mixing step 1, then a solvent is added in the extraction step 2 for extraction treatment, and the extracted waste liquid is mixed with an alkali such as NaOH in the neutralization step 3. After neutralization with the solution, it is filtered in a filtration step 4. The U-containing solvent obtained in extraction step 2 is back-extracted with water added in back-extraction step 5 to recover most of the U, and the solvent produced here is reused. The precipitate after filtration is solidified with cement step 10
The solidified cement is obtained by solidifying the cement, which is stored as radioactive waste. The filtrate produced in the filtration step 4 is the radioactive waste liquid to be treated in the method of the present invention. In order to treat this radioactive waste liquid (filtrate) according to the method of the present invention, M is added to the radioactive waste liquid in the dissolution step 6.
A water-soluble magnesium salt such as g(NO3)2.6H2O is added and dissolved, and water glass containing Na2SiO3 as a main component is further added. Here, as the magnesium salt added and dissolved, in addition to Mg(NO3)2.6H2O, MgC
l2.6H2O or the like may also be used. In the above magnesium salt, Mg is in a weakly alkaline region, and Na2Si
It forms a double salt with O3 or heavy metals, which causes water glass precipitate. Next, in the adjustment step 7, an alkaline solution such as NaOH is added to adjust the pH to 8 to 9 to form a water glass precipitate. The alkaline solution added here is N
In addition to aOH, KOH or the like may also be used. After this, filtration step 8
The treated waste liquid and water glass precipitate are separated by filtration, the treated waste liquid is discharged, and the water glass precipitate is stored. A trace amount of U and heavy metals in the radioactive waste liquid are aggregated in this water glass precipitate. [Example 1] An example in which a used electrolytic solution containing approximately 30% H2SO4 concentration and approximately 5 g/liter of each of U, Fe, Ni, and Mo is treated will be described. First, U in the used electrolyte was recovered by the following procedure. ■ 62 liters of the used electrolyte
Add %HNO3. (HNO3 concentration 1.4N) ■H
Spent electrolyte with NO3 added (U content 4.86g/
Add 1 volume of a solvent consisting of a mixture of TBP and diluent n-dodecane at a volume ratio of 3:7 to 1 volume (liter), and shake for 5 minutes at room temperature using a separating funnel to extract U in the aqueous phase. . ■Add 1 volume of pure water to 1 volume of solvent containing U,
Shake for 5 minutes at room temperature using a separating funnel to back-extract U in the solvent. ■ Repeat the extraction operation in 4 stages and the back extraction operation for each extraction stage until no U is detected in the aqueous phase (up to 6 stages). The contents of Fe, Ni, and Mo in the total of 9.52 g (recovery rate of about 98%) of U recovered by the above method are shown in Table 1, and all values are below the upper limit for content in the nuclear fuel field. It is. [Table 1] Next, 25% Na was added to the extraction waste liquid after U recovery.
The pH was adjusted to 6 to 7.0 by adding an OH solution, and the heavy metals (mainly Fe, Ni, Cr, and Mo) contained were filtered and separated as hydroxide precipitates, and the precipitates were solidified into cement. [0022] The filtrate produced at this time is regarded as a radioactive waste liquid, and the method of the present invention is applied to it to treat it. ■Filtrate 1
Add and dissolve 4 g of Mg(NO3)2.6H2O per liter. (2) 2 g of water glass was added per liter of the filtrate, and the pH of the filtrate was adjusted to 8-9 by further adding NaOH. (2) After stirring for 1 hour, it was left to stand for 24 hours. (2) After standing for 24 hours, the water glass precipitate was quantitatively determined and filtered using filter paper. Table 2 shows the contents of U and Cr in the filtrate treated in the above steps before and after the treatment. [Table 2] [0025] The regulatory value for U varies slightly depending on the concentration, but it is approximately 1 ppm, and the total amount of Cr is also 1 ppm, so the results shown in Table 2 are well below the regulatory value. There was found. From this, the effect of the method of the present invention was confirmed. "Example 2" H3PO4 concentration approximately 30%, U
An example is shown in which a used electrolytic solution containing approximately 5 g/liter of each of Fe, Ni, and Mo was treated. First, U in the used electrolyte was recovered by the following procedure. ■62% HN in 1 liter of used electrolyte
Add O3. (HNO3 concentration 6.9N) ■HNO3
To 1 volume of the used electrolyte (U content: 2.42 g/liter), add 1 volume of a solvent containing TBP and n-dodecane as a diluent at a volume ratio of 3:7, and use a separating funnel to Shake at room temperature for 5 minutes to extract U in the aqueous phase. ■U
1 volume of pure water is added to 1 volume of the solvent containing , and the mixture is shaken for 5 minutes at room temperature using a separating funnel to back-extract U in the solvent. ■ Repeat the extraction operation in 4 stages and the back extraction operation for each extraction stage until no U is detected in the aqueous phase (up to 5 stages). The contents of Fe, Ni, and Mo in the total of 5.68 g (recovery rate of about 97%) of U recovered by the above method are shown in Table 3, and all values are below the upper limit for content in the nuclear fuel field. It is. [Table 3] Next, 25% N was added to the extraction waste liquid after U recovery.
Adjust the pH to 6.5-7.0 by adding aOH solution and remove the heavy metals contained (mainly Fe, Ni, Cr and Mo).
was filtered and separated as hydroxide precipitate, and the precipitate was solidified with cement. The filtrate produced at this time is regarded as a radioactive waste liquid, and the method of the present invention is applied to it to treat it. ■Filtrate 1
Add and dissolve 4 g of Mg(NO3)2.6H2O per liter. (2) 2 g of water glass was added per liter of the filtrate, and the pH of the filtrate was adjusted to 8-9 by further adding NaOH. (2) After stirring for 1 hour, it was left to stand for 24 hours. (2) After standing for 24 hours, the water glass precipitate was quantitatively determined and filtered using filter paper. Table 4 shows the contents of U and Cr in the filtrate treated in the above steps before and after the treatment. [Table 4] The regulatory value for U varies slightly depending on the concentration, but is approximately 1 ppm, and the total amount of Cr is also 1 ppm.
Therefore, it was found that the results shown in Table 4 were well below the regulation value. From this, the effect of the method of the present invention was confirmed. Effects of the Invention As explained above, according to the present invention, U
After treating the waste liquid produced by electrolytic polishing of metals such as carbon steel and stainless steel contaminated with carbon steel and removing most of the U and heavy metals, Mg (NO3 ) A water glass precipitate can be produced by dissolving a water-soluble magnesium salt such as 2.6H2O and then adding an alkaline solution to adjust the pH. In this case, Mg is weakly alkaline and forms a double salt with Na2SiO3 or heavy metals, causing water glass precipitate. Therefore, heavy metals such as Fe, Ni, Cr, and U are aggregated in this water glass precipitate. Therefore, by removing this water glass precipitate from the radioactive waste liquid, heavy metals such as Fe, Ni, Cr, etc. and U can be removed, and the radioactive waste liquid can be effectively treated. Further, according to the method of the present invention, U in the waste liquid can be completely removed to below the detection limit, and heavy metals can be reduced to below the regulation value.
【図１】図１は本発明方法を説明するためのフロー図で
ある。FIG. 1 is a flow diagram for explaining the method of the present invention.
１ 混合工程 ２ 抽出工程 ３ 中和工程 ４ 瀘過工程 ５ 溶解工程 ６ 調整工程 ７ 瀘過工程 1 Mixing process 2 Extraction process 3 Neutralization process 4. Filtration process 5 Dissolution process 6 Adjustment process 7. Filtration process
Claims (1)
硫酸系あるいはリン酸系の放射性廃液に、水溶性のマグ
ネシウム塩を溶解した後に水ガラスを添加し、更にアル
カリ溶液を添加してｐＨを８〜９に調整し、水ガラス殿
物を生成させてＦｅ、Ｎｉ、Ｃｒなどの重金属とＵを除
去することを特徴とする放射性廃液の処理方法。Claim 1: A water-soluble magnesium salt is dissolved in a sulfuric acid-based or phosphoric acid-based radioactive waste liquid containing heavy metals such as Fe, Ni, and Cr and U, and then water glass is added, and an alkaline solution is further added. A method for treating radioactive waste liquid, which comprises adjusting the pH to 8 to 9, generating a water glass precipitate, and removing heavy metals such as Fe, Ni, Cr, and U.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP9169891A JPH04301799A (en) | 1991-03-29 | 1991-03-29 | Treating method for radioactive waste fluid |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP9169891A JPH04301799A (en) | 1991-03-29 | 1991-03-29 | Treating method for radioactive waste fluid |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPH04301799A true JPH04301799A (en) | 1992-10-26 |
Family
ID=14033739
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP9169891A Pending JPH04301799A (en) | 1991-03-29 | 1991-03-29 | Treating method for radioactive waste fluid |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH04301799A (en) |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| KR100485973B1 (en) * | 2002-11-20 | 2005-05-03 | 주식회사 데콘엔지니어링 | A preliminary acid cleansing device of a high radioactive contamination metal |
| CN114988601A (en) * | 2022-04-22 | 2022-09-02 | 中南大学 | Method for strengthening uranium and arsenic mineralization and improving mineral stability |
-
1991
- 1991-03-29 JP JP9169891A patent/JPH04301799A/en active Pending
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| KR100485973B1 (en) * | 2002-11-20 | 2005-05-03 | 주식회사 데콘엔지니어링 | A preliminary acid cleansing device of a high radioactive contamination metal |
| CN114988601A (en) * | 2022-04-22 | 2022-09-02 | 中南大学 | Method for strengthening uranium and arsenic mineralization and improving mineral stability |
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