JPH10213697A - Decontamination method of radioactive waste - Google Patents
Decontamination method of radioactive wasteInfo
- Publication number
- JPH10213697A JPH10213697A JP1502897A JP1502897A JPH10213697A JP H10213697 A JPH10213697 A JP H10213697A JP 1502897 A JP1502897 A JP 1502897A JP 1502897 A JP1502897 A JP 1502897A JP H10213697 A JPH10213697 A JP H10213697A
- Authority
- JP
- Japan
- Prior art keywords
- radioactive
- halide
- chloride
- solvent
- 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
- 238000005202 decontamination Methods 0.000 title claims abstract description 24
- 239000002901 radioactive waste Substances 0.000 title claims description 32
- 230000003588 decontaminative effect Effects 0.000 claims abstract description 20
- 239000002904 solvent Substances 0.000 claims abstract description 16
- 230000002285 radioactive effect Effects 0.000 claims abstract description 14
- 150000004820 halides Chemical class 0.000 claims description 40
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims description 21
- 238000000034 method Methods 0.000 claims description 18
- 229910001868 water Inorganic materials 0.000 claims description 18
- 229910052761 rare earth metal Inorganic materials 0.000 claims description 9
- 229910052768 actinide Inorganic materials 0.000 claims description 7
- 150000001255 actinides Chemical class 0.000 claims description 6
- 229910052792 caesium Inorganic materials 0.000 claims description 6
- TVFDJXOCXUVLDH-UHFFFAOYSA-N caesium atom Chemical compound [Cs] TVFDJXOCXUVLDH-UHFFFAOYSA-N 0.000 claims description 6
- 238000001704 evaporation Methods 0.000 claims description 6
- 239000007787 solid Substances 0.000 claims description 6
- 229910052712 strontium Inorganic materials 0.000 claims description 4
- CIOAGBVUUVVLOB-UHFFFAOYSA-N strontium atom Chemical compound [Sr] CIOAGBVUUVVLOB-UHFFFAOYSA-N 0.000 claims description 4
- 239000003463 adsorbent Substances 0.000 claims description 2
- 238000001816 cooling Methods 0.000 claims description 2
- 239000000941 radioactive substance Substances 0.000 claims description 2
- 239000002699 waste material Substances 0.000 abstract description 34
- VEXZGXHMUGYJMC-UHFFFAOYSA-M Chloride anion Chemical compound [Cl-] VEXZGXHMUGYJMC-UHFFFAOYSA-M 0.000 abstract description 28
- FAPWRFPIFSIZLT-UHFFFAOYSA-M Sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 abstract description 28
- 150000003839 salts Chemical class 0.000 abstract description 28
- 239000011780 sodium chloride Substances 0.000 abstract description 14
- 238000002844 melting Methods 0.000 abstract description 4
- 230000008018 melting Effects 0.000 abstract description 4
- 238000004064 recycling Methods 0.000 abstract 1
- WCUXLLCKKVVCTQ-UHFFFAOYSA-M Potassium chloride Chemical compound [Cl-].[K+] WCUXLLCKKVVCTQ-UHFFFAOYSA-M 0.000 description 54
- KWGKDLIKAYFUFQ-UHFFFAOYSA-M lithium chloride Chemical compound [Li+].[Cl-] KWGKDLIKAYFUFQ-UHFFFAOYSA-M 0.000 description 49
- 239000000243 solution Substances 0.000 description 28
- 235000011164 potassium chloride Nutrition 0.000 description 27
- 239000001103 potassium chloride Substances 0.000 description 27
- 238000006243 chemical reaction Methods 0.000 description 15
- 238000012958 reprocessing Methods 0.000 description 12
- 239000002244 precipitate Substances 0.000 description 9
- 239000002915 spent fuel radioactive waste Substances 0.000 description 9
- 238000011084 recovery Methods 0.000 description 7
- 229910052783 alkali metal Inorganic materials 0.000 description 6
- 150000001340 alkali metals Chemical class 0.000 description 6
- 238000010586 diagram Methods 0.000 description 6
- 229920006395 saturated elastomer Polymers 0.000 description 5
- 229910021536 Zeolite Inorganic materials 0.000 description 4
- 150000001805 chlorine compounds Chemical class 0.000 description 4
- HNPSIPDUKPIQMN-UHFFFAOYSA-N dioxosilane;oxo(oxoalumanyloxy)alumane Chemical compound O=[Si]=O.O=[Al]O[Al]=O HNPSIPDUKPIQMN-UHFFFAOYSA-N 0.000 description 4
- 239000000843 powder Substances 0.000 description 4
- 238000001556 precipitation Methods 0.000 description 4
- 239000010457 zeolite Substances 0.000 description 4
- 229910052784 alkaline earth metal Inorganic materials 0.000 description 3
- -1 KC (1 Chemical compound 0.000 description 2
- WHXSMMKQMYFTQS-UHFFFAOYSA-N Lithium Chemical compound [Li] WHXSMMKQMYFTQS-UHFFFAOYSA-N 0.000 description 2
- 229910018957 MClx Inorganic materials 0.000 description 2
- 239000007864 aqueous solution Substances 0.000 description 2
- 238000005341 cation exchange Methods 0.000 description 2
- 230000000694 effects Effects 0.000 description 2
- 230000008020 evaporation Effects 0.000 description 2
- 238000005342 ion exchange Methods 0.000 description 2
- 238000010298 pulverizing process Methods 0.000 description 2
- 239000002516 radical scavenger Substances 0.000 description 2
- 239000012857 radioactive material Substances 0.000 description 2
- 239000013049 sediment Substances 0.000 description 2
- 238000004062 sedimentation Methods 0.000 description 2
- 239000010414 supernatant solution Substances 0.000 description 2
- 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 1
- 235000006481 Colocasia esculenta Nutrition 0.000 description 1
- 240000004270 Colocasia esculenta var. antiquorum Species 0.000 description 1
- 229910020549 KCl—NaCl Inorganic materials 0.000 description 1
- ZLMJMSJWJFRBEC-UHFFFAOYSA-N Potassium Chemical compound [K] ZLMJMSJWJFRBEC-UHFFFAOYSA-N 0.000 description 1
- 239000003513 alkali Substances 0.000 description 1
- 150000001342 alkaline earth metals Chemical class 0.000 description 1
- 150000004645 aluminates Chemical class 0.000 description 1
- 239000003729 cation exchange resin Substances 0.000 description 1
- 150000001875 compounds Chemical class 0.000 description 1
- 238000005260 corrosion Methods 0.000 description 1
- 230000007797 corrosion Effects 0.000 description 1
- 230000004992 fission Effects 0.000 description 1
- 239000000446 fuel Substances 0.000 description 1
- 229910052736 halogen Inorganic materials 0.000 description 1
- 150000002367 halogens Chemical class 0.000 description 1
- 238000010438 heat treatment Methods 0.000 description 1
- 229910052744 lithium Inorganic materials 0.000 description 1
- 239000000463 material Substances 0.000 description 1
- 229910052751 metal Inorganic materials 0.000 description 1
- 239000002184 metal Substances 0.000 description 1
- 238000002156 mixing Methods 0.000 description 1
- 239000000203 mixture Substances 0.000 description 1
- 229910052700 potassium Inorganic materials 0.000 description 1
- 239000011591 potassium Substances 0.000 description 1
- 239000000047 product Substances 0.000 description 1
- 238000000746 purification Methods 0.000 description 1
- 150000002910 rare earth metals Chemical class 0.000 description 1
- 238000000926 separation method Methods 0.000 description 1
- 239000011734 sodium Substances 0.000 description 1
- 238000001179 sorption measurement Methods 0.000 description 1
Landscapes
- Processing Of Solid Wastes (AREA)
Abstract
(57)【要約】
【課題】除染した塩廃棄物中に塩化ナトリウムが混入す
るため、リサイクル利用が難しく、また、塩化物の融点
以上の高温で処理を行う必要がある。
【解決手段】塩廃棄物を溶媒に溶解した後に室温付近の
除染操作により非放射性元素または放射性元素の回収を
行う。
(57) [Summary] [Problem] Since sodium chloride is mixed in decontaminated salt waste, recycling is difficult, and it is necessary to perform treatment at a high temperature not lower than the melting point of chloride. SOLUTION: After dissolving salt waste in a solvent, a non-radioactive element or a radioactive element is recovered by a decontamination operation near room temperature.
Description
【０００１】[0001]
【発明の属する技術分野】本発明は放射性廃棄物除染処
理方法に関する。The present invention relates to a radioactive waste decontamination method.
【０００２】[0002]
【従来の技術】放射性元素を含むハロゲン化物の除染方
法は、例えば、M. A. Lewis et al.，J. Am. Ceram. So
c.,７６（１１），２８２６−２８３２(１９９３）に記
載されている。この方法では、金属燃料の乾式再処理に
おいて発生する塩廃棄物（LiCl，ＫＣｌ，ＮａＣｌ，核
分裂生成物の塩化物，アクチニド元素の塩化物）を溶融
させ、これをゼオライトのカラムに通し、ゼオライトの
イオン交換を利用して塩廃棄物中の放射性核種をゼオラ
イトに吸着，除染する。回収した塩（ＬｉＣｌ，ＫＣ
ｌ，ＮａＣｌ）は再び乾式再処理の主工程へリサイクル
することができる。放射性元素を含むハロゲン化物の別
の除染方法は、小山正史，瀬戸千秋；特開平8−110397
号公報に記載されている。この方法では、ハロゲン化物
の放射性廃棄物を融点以上に保ち、この状態で、ＬｉＡ
ｌＯ2 またはＫＡｌＯ2 またはKSO4またはＬｉ12ＺｒＯ
3をスカベンジャー体として溶融塩中に添加することに
よってハロゲン化物廃棄物中に含まれるアルカリ土類元
素，希土類元素，アクチニド元素のハロゲン化物を選択
的にアルミン酸またはジルコン酸または酸化物として沈
殿分離させる。2. Description of the Related Art Methods for decontamination of halides containing radioactive elements are described, for example, in MA Lewis et al., J. Am. Ceram.
c., 76 (11), 2826-2832 (1993). In this method, salt waste (LiCl, KCl, NaCl, fission product chloride, actinide element chloride) generated in the dry reprocessing of metal fuel is melted and passed through a zeolite column to form a zeolite. Adsorption and decontamination of radionuclides in salt waste to zeolite using ion exchange. Recovered salt (LiCl, KC
(1, NaCl) can be recycled again to the main process of dry reprocessing. Another method for decontamination of halides containing radioactive elements is described in Masashi Koyama, Chiaki Seto;
No., published in Japanese Unexamined Patent Publication No. In this method, the halide radioactive waste is kept above the melting point and, in this state, LiA
10 2 or KAlO 2 or KSO 4 or Li 12 ZrO
Addition of 3 as a scavenger to the molten salt selectively precipitates and separates alkaline earth, rare earth and actinide halides contained in halide waste as aluminate or zirconate or oxide. .
【０００３】[0003]
【発明が解決しようとする課題】従来技術によれば、ハ
ロゲン化物の放射性廃棄物中からアルカリ土類金属元
素，希土類元素，アクチニド元素を除去し、回収したＬ
ｉＣｌ−ＫＣｌ−ＮａＣｌを乾式再処理の主工程などに
再利用することが可能である。しかし、アルカリ金属元
素であるＮａの除染係数が低いため、再利用するハロゲ
ン化物中にＮａＣｌが混入しハロゲン化物の組成及び融
点が上昇する。また、ゼオライトによるイオン交換処理
もスカベンジャー体の添加による沈殿精製もハロゲン化
物の融点以上の高温でそれぞれ操作を行う必要があるた
め、処理施設が複雑な構造になったり、ハロゲン化物に
耐食性のある構造材料を使う必要がある。According to the prior art, alkaline earth metal elements, rare earth elements, and actinide elements are removed from radioactive waste of halides and recovered L is recovered.
iCl-KCl-NaCl can be reused in the main step of dry reprocessing and the like. However, since the decontamination coefficient of Na, which is an alkali metal element, is low, NaCl is mixed in the halide to be recycled, and the composition and melting point of the halide increase. In addition, both the ion exchange treatment with zeolite and the purification by precipitation with the addition of a scavenger need to be performed at a high temperature higher than the melting point of the halide, so that the treatment facility becomes complicated and the structure has corrosion resistance to halide. You need to use materials.
【０００４】本発明の目的は、ハロゲン化物の放射性廃
棄物中のアルカリ金属元素の除染が可能で、かつ室温付
近での操作が行えるハロゲン化物の放射性廃棄物の処理
方法を提供することにある。An object of the present invention is to provide a method for treating a halide radioactive waste which can decontaminate an alkali metal element in the halide radioactive waste and which can be operated near room temperature. .
【０００５】[0005]
【課題を解決するための手段】本発明では、使用済み燃
料の乾式再処理などで発生するアルカリ金属元素，アル
カリ土類金属元素，希土類元素の塩化物などのハロゲン
化物から成る放射性物質を含むハロゲン化物廃棄物を一
端水などの溶媒に溶解させた後に、室温付近での分離操
作で溶液中から放射性物質および／または非放射性物質
を回収することにより、ＬｉＣｌ，ＫＣｌが再利用可能
となるようにしている。According to the present invention, a halogen containing a radioactive substance composed of a halide such as an alkali metal element, an alkaline earth metal element or a rare earth element chloride generated by dry reprocessing of spent fuel or the like is provided. After dissolving the compound waste in a solvent such as water, the radioactive material and / or non-radioactive material is recovered from the solution by a separation operation near room temperature so that LiCl and KCl can be reused. ing.
【０００６】本発明の放射性廃棄物の処理方法は、水な
どの溶媒にハロゲン化物を溶解させた後、放射性廃棄物
中の大部分を占めるＬｉＣｌ，ＫＣｌが飽和し、かつそ
の他の放射性元素等が飽和溶解度に達しない状態まで溶
液を真空蒸発させるかまたは冷却することにより、Ｌｉ
ＣｌおよびＫＣｌを析出させ沈殿を分離回収することで
ハロゲン化廃棄物を除染し、回収したＬｉＣｌ，ＫＣｌ
を再利用できる。ハロゲン化物廃棄物を水に溶解させた
場合、廃棄物中の希土類元素およびアクチニド元素は酸
化物沈殿として予め分離回収することが可能である。In the method for treating radioactive waste according to the present invention, after dissolving a halide in a solvent such as water, LiCl and KCl, which occupy most of the radioactive waste, are saturated, and other radioactive elements and the like are removed. By evaporating the solution in vacuo or cooling to a point where saturation solubility is not reached, Li
Cl and KCl are precipitated, and the precipitate is separated and recovered to decontaminate the halogenated waste, and the recovered LiCl, KCl
Can be reused. When the halide waste is dissolved in water, the rare earth element and the actinide element in the waste can be separated and recovered in advance as an oxide precipitate.
【０００７】また、本発明では、ハロゲン化物廃棄物を
水蒸気分圧を調整した雰囲気下に置いた場合、ハロゲン
化物の潮解性により飽和状態に近い水溶液を得ることが
可能である。この時、特に限定されるものではないが、
雰囲気中の水蒸気分圧は飽和蒸気圧とすることが望まし
い。Further, in the present invention, when the halide waste is placed in an atmosphere in which the partial pressure of steam is adjusted, it is possible to obtain an aqueous solution which is almost saturated due to the deliquescence of the halide. At this time, although not particularly limited,
It is desirable that the partial pressure of water vapor in the atmosphere be a saturated vapor pressure.
【０００８】また、本発明では、ハロゲン化物の廃棄物
を溶解させた溶液をセシウムおよび／またはストロンチ
ウムを選択的に吸着する吸着剤を充填したカラムを通し
てセシウムおよび／またはストロンチウムを吸着させる
ことによりハロゲン化物廃棄物の溶液を除染することに
より、ハロゲン化物廃棄物中の非放射性元素を再利用可
能にすることができる。Further, in the present invention, a halide solution is dissolved by adsorbing cesium and / or strontium through a column packed with an adsorbent for selectively adsorbing cesium and / or strontium. By decontaminating the waste solution, the non-radioactive elements in the halide waste can be reused.
【０００９】[0009]
（実施例１）本発明の一実施例であるハロゲン化物放射
性廃棄物の除染処理方法が実施される除染装置を図１を
用いて説明する。このハロゲン化物放射性廃棄物の除染
処理装置は気密反応槽１，塩廃棄物粉砕・供給装置２，
水供給装置３，真空ポンプ４，溶液貯槽５，元素濃度モ
ニタ６から成る。(Embodiment 1) A decontamination apparatus in which a method for decontaminating halide radioactive waste according to one embodiment of the present invention will be described with reference to FIG. This apparatus for decontamination of halide radioactive waste comprises an airtight reactor 1, a salt waste crushing / supplying apparatus 2,
It comprises a water supply device 3, a vacuum pump 4, a solution storage tank 5, and an element concentration monitor 6.
【００１０】溶融塩化物を用いた使用済み原子燃料の再
処理で発生する塩廃棄物を塩廃棄物粉砕・供給装置２に
より粉末として気密反応槽１へ供給する。続いて水供給
装置３より気密反応槽１へ水を供給し塩廃棄物を溶解す
る。さらに真空ポンプ４を動作させることにより気密反
応槽１内を真空に保ち、溶液７を蒸発させる。溶液中の
塩濃度が飽和溶解度に達すると塩化物が析出し始める。
図２に各塩化物の水への溶解度を示す。この時、溶融塩
化物を用いた使用済み原子燃料の再処理で発生する塩廃
棄物中では塩化リチウムおよび塩化カリウムが他の塩化
物に比べて濃度が高く、かつ、塩化カリウムの溶解度は
塩化リチウムの溶解度より低いため、まず塩化カリウム
が析出し、続いて塩化リチウムが析出し始める。元素濃
度モニタ６で溶液中の元素濃度をモニタし、塩化リチウ
ムおよび塩化カリウム以外のいずれかの塩化物が飽和溶
解度に達した時点で真空ポンプ４を停止させ、溶液を貯
槽５に回収すれば塩化リチウムおよび塩化カリウム以外
の塩化物を含まない固体が得られる。図３に２５℃で塩
化ナトリウムを析出させないように真空ポンプ４を操作
した場合の各塩化物の収率を示す。図の横軸は塩化物廃
棄物中に含まれる塩化物の塩化ナトリウムに対する重量
比である。また、縦軸は２５℃で溶媒を蒸発させていく
場合、塩化ナトリウムを混入させることなく回収できる
塩化物の最大値である。例えばＭＣｌｘ／ＮａＣｌ＝１
０，Ｍ＝Ｋ（カリウム）の場合、初め塩化カリウムが塩
化ナトリウムの１０倍溶液中に溶解している。この溶媒
を徐々に蒸発させると飽和溶解度に達した塩化カリウム
が析出し、塩化ナトリウムが飽和溶解度に達して塩化ナ
トリウムが析出し始めるまでに、溶液中の塩化カリウム
の９０％を回収することができる。同様にＭＣｌｘ／Ｎ
ａＣｌ＝１０，Ｍ＝Ｌｉ（リチウム）の場合では７６％
の塩化リチウムを回収することができる。[0010] Salt waste generated by the reprocessing of spent nuclear fuel using molten chloride is supplied to the hermetic reaction tank 1 as powder by a salt waste crushing / supplying device 2. Subsequently, water is supplied from the water supply device 3 to the hermetic reaction tank 1 to dissolve salt waste. Further, by operating the vacuum pump 4, the inside of the hermetic reaction tank 1 is maintained at a vacuum, and the solution 7 is evaporated. Chloride begins to precipitate when the salt concentration in the solution reaches saturation solubility.
FIG. 2 shows the solubility of each chloride in water. At this time, in salt waste generated by reprocessing spent nuclear fuel using molten chloride, lithium chloride and potassium chloride are higher in concentration than other chlorides, and the solubility of potassium chloride is lithium chloride. , Potassium chloride precipitates first, and then lithium chloride starts to precipitate. The element concentration in the solution is monitored by the element concentration monitor 6, and when any chloride other than lithium chloride and potassium chloride reaches the saturation solubility, the vacuum pump 4 is stopped. A chloride-free solid other than lithium and potassium chloride is obtained. FIG. 3 shows the yield of each chloride when the vacuum pump 4 was operated at 25 ° C. so as not to precipitate sodium chloride. The horizontal axis of the figure is the weight ratio of chloride contained in chloride waste to sodium chloride. The vertical axis represents the maximum value of chloride that can be recovered without mixing sodium chloride when the solvent is evaporated at 25 ° C. For example, MClx / NaCl = 1
When 0, M = K (potassium), potassium chloride is initially dissolved in a 10-fold solution of sodium chloride. When the solvent is gradually evaporated, potassium chloride reaching the saturation solubility is precipitated, and 90% of the potassium chloride in the solution can be recovered before the sodium chloride reaches the saturation solubility and the sodium chloride starts to precipitate. . Similarly, MClx / N
76% when aCl = 10 and M = Li (lithium)
Of lithium chloride can be recovered.
【００１１】回収した塩化リチウム及び塩化カリウムは
塩化ナトリウムを分離しているため、再び溶融塩化物を
用いた使用済み原子燃料の再処理に利用することが可能
である。同時に、塩廃棄物中から非放射性元素を分離す
るため、放射性廃棄物の量を大幅に低減させ得る。Since the recovered lithium chloride and potassium chloride separate sodium chloride, it can be reused for reprocessing spent nuclear fuel using molten chloride. At the same time, the amount of radioactive waste can be significantly reduced because non-radioactive elements are separated from the salt waste.
【００１２】（実施例２）本発明の別の実施例であるハ
ロゲン化物放射性廃棄物の除染処理方法が実施される除
染装置の構成を図４を用いて説明する。このハロゲン化
物放射性廃棄物の除染処理装置は気密反応槽１，塩廃棄
物粉砕・供給装置２，水供給装置３，真空ポンプ４，溶
液貯槽５，元素濃度モニタ６，酸化物沈殿回収槽１１お
よび溶液移送ポンプ１２から成る。(Embodiment 2) The configuration of a decontamination apparatus in which a method for decontamination of halide radioactive waste according to another embodiment of the present invention will be described with reference to FIG. The halide radioactive waste decontamination treatment apparatus includes an airtight reaction tank 1, a salt waste crushing / supplying apparatus 2, a water supply apparatus 3, a vacuum pump 4, a solution storage tank 5, an element concentration monitor 6, and an oxide sedimentation recovery tank 11. And a solution transfer pump 12.
【００１３】溶融塩化物を用いた使用済み原子燃料の再
処理で発生する塩廃棄物を塩廃棄物粉砕・供給装置２に
より粉末として酸化物沈殿回収槽１１へ供給する。続い
て水供給装置３より酸化物沈殿回収槽１１へ水を供給し
塩廃棄物を溶解する。水に溶解した塩化物中の希土類元
素およびアクチニド元素は中性付近の溶液中では酸化物
を形成するため、槽１１内に不溶性の酸化物沈殿１３と
して回収される。続いて上澄みの溶液を溶液移送ポンプ
１２にて気密反応槽１に供給した後、真空ポンプ４で気
密反応槽１内を真空に保ち、溶液７を蒸発させる。この
時、溶融塩化物を用いた使用済み原子燃料の再処理で発
生する塩廃棄物中では塩化リチウムおよび塩化カリウム
が他の塩化物に比べて濃度が高く、かつ、塩化カリウム
の溶解度は塩化リチウムの溶解度より低いため、まず塩
化カリウムが析出し、続いて塩化リチウムが析出し始め
る。元素濃度モニタ６で溶液中の元素濃度をモニタし、
塩化リチウムおよび塩化カリウム以外のいずれかの塩化
物が飽和溶解度に達した時点で真空ポンプ４を停止さ
せ、溶液を貯槽５に回収すれば塩化リチウムおよび塩化
カリウムのみからなる固体が得られる。The salt waste generated by the reprocessing of the spent nuclear fuel using the molten chloride is supplied to the oxide sediment recovery tank 11 as powder by the salt waste pulverizing / supplying device 2. Subsequently, water is supplied from the water supply device 3 to the oxide precipitation recovery tank 11 to dissolve salt waste. The rare earth element and the actinide element in the chloride dissolved in water form an oxide in a solution near neutrality, and are thus recovered as an insoluble oxide precipitate 13 in the tank 11. Subsequently, the supernatant solution is supplied to the hermetic reaction tank 1 by the solution transfer pump 12, and then the inside of the hermetic reaction tank 1 is kept at a vacuum by the vacuum pump 4 to evaporate the solution 7. At this time, in salt waste generated by reprocessing spent nuclear fuel using molten chloride, lithium chloride and potassium chloride are higher in concentration than other chlorides, and the solubility of potassium chloride is lithium chloride. , Potassium chloride precipitates first, and then lithium chloride starts to precipitate. The element concentration in the solution is monitored by the element concentration monitor 6,
When any chloride other than lithium chloride and potassium chloride reaches the saturation solubility, the vacuum pump 4 is stopped, and the solution is collected in the storage tank 5 to obtain a solid consisting only of lithium chloride and potassium chloride.
【００１４】本実施例では、回収する塩化リチウム及び
塩化カリウムからアルカリ金属元素，アルカリ土類元素
を除くことに加え、さらに希土類元素を取り除くことが
可能である。In this embodiment, in addition to removing the alkali metal element and the alkaline earth element from the recovered lithium chloride and potassium chloride, it is possible to further remove the rare earth element.
【００１５】（実施例３）本発明の別の実施例であるハ
ロゲン化物放射性廃棄物の除染処理方法が実施される除
染装置の構成を図５を用いて説明する。このハロゲン化
物放射性廃棄物の除染処理装置は気密反応槽１，塩廃棄
物粉砕・供給装置２，水供給装置３，真空ポンプ４，溶
液貯槽５，元素濃度モニタ６，酸化物沈殿回収槽１１，
溶液移送ポンプ１２，陽イオン交換カラム１４，蒸発槽
１５およびヒータ１６から成る。(Embodiment 3) The structure of a decontamination apparatus in which a method for decontamination of halide radioactive waste according to another embodiment of the present invention will be described with reference to FIG. The halide radioactive waste decontamination treatment apparatus includes an airtight reaction tank 1, a salt waste crushing / supplying apparatus 2, a water supply apparatus 3, a vacuum pump 4, a solution storage tank 5, an element concentration monitor 6, and an oxide sedimentation recovery tank 11. ,
It comprises a solution transfer pump 12, a cation exchange column 14, an evaporation tank 15, and a heater 16.
【００１６】溶融塩化物を用いた使用済み原子燃料の再
処理で発生する塩廃棄物を塩廃棄物粉砕・供給装置２に
より粉末として酸化物沈殿回収槽１１へ供給する。続い
て水供給装置３より酸化物沈殿回収槽１１へ水を供給し
塩廃棄物を溶解する。水に溶解した塩化物中の希土類元
素およびアクチニド元素は中性付近の溶液中では酸化物
を形成するため、槽１１内に不溶性の酸化物沈殿１３と
して回収される。続いて上澄みの溶液を溶液移送ポンプ
１２で、セシウムおよびアルカリ土類金属元素を選択的
に吸着する陽イオン交換樹脂を充填した陽イオン交換カ
ラム１４に供給する。カラム１４を通過させセシウムお
よびアルカリ土類元素を除去した流出液を蒸発槽１５に
供給する。ヒータ１６により蒸発槽１５を加熱し、溶液
の水分を蒸発させることにより塩化リチウムおよび塩化
カリウムからなる固体が得られる。The salt waste generated by the reprocessing of the spent nuclear fuel using the molten chloride is supplied to the oxide sediment recovery tank 11 as powder by the salt waste pulverizing / supplying device 2. Subsequently, water is supplied from the water supply device 3 to the oxide precipitation recovery tank 11 to dissolve salt waste. The rare earth element and the actinide element in the chloride dissolved in water form an oxide in a solution near neutrality, and are thus recovered as an insoluble oxide precipitate 13 in the tank 11. Subsequently, the supernatant solution is supplied by a solution transfer pump 12 to a cation exchange column 14 filled with a cation exchange resin that selectively adsorbs cesium and alkaline earth metal elements. The effluent from which the cesium and alkaline earth elements have been removed by passing through the column 14 is supplied to the evaporation tank 15. A solid consisting of lithium chloride and potassium chloride is obtained by heating the evaporating tank 15 with the heater 16 and evaporating the water content of the solution.
【００１７】本実施例によれば、回収する塩化リチウム
および塩化カリウム中からアルカリ金属元素，アルカリ
度類元素及び／または希土類元素を除去する実施例１，
２の効果に加え、アルカリ金属元素及びアルカリ土類元
素をより効果的に除去することが可能である。According to this embodiment, an alkali metal element, an alkali element and / or a rare earth element are removed from the recovered lithium chloride and potassium chloride.
In addition to the effect of 2, the alkali metal element and the alkaline earth element can be more effectively removed.
【００１８】（実施例４）本発明の別の実施例であるハ
ロゲン化物放射性廃棄物の除染処理方法が実施される除
染装置の構成を図５を用いて説明する。このハロゲン化
物放射性廃棄物の除染処理装置は気密反応槽１，塩廃棄
物粉砕・供給装置２，溶液貯槽５，元素濃度モニタ６、
および水蒸気圧調整装置２１から成る。(Embodiment 4) The structure of a decontamination apparatus in which a method for decontamination of halide radioactive waste according to another embodiment of the present invention will be described with reference to FIG. This halide radioactive waste decontamination treatment apparatus comprises an airtight reaction tank 1, a salt waste crushing / supplying apparatus 2, a solution storage tank 5, an element concentration monitor 6,
And a steam pressure adjusting device 21.
【００１９】溶融塩化物を用いた使用済み原子燃料の再
処理で発生する塩廃棄物を塩廃棄物粉砕・供給装置２に
より粉末として気密反応槽１へ供給する。続いて水蒸気
圧調整装置２１を用いて気密反応槽１内の水蒸気圧を上
昇させる。好ましくはこの時、水蒸気圧は気密反応槽１
内の温度における飽和蒸気圧とする。反応槽１内の塩化
物は潮解性を有するため気相中の水蒸気を吸収し塩化物
水溶液となる。供給した塩廃棄物の固体が全て溶解した
時点で水蒸気圧調整装置２１で気密反応槽１内の水蒸気
圧を減少させる。この時、好ましくは気相中の水蒸気圧
がゼロになるようにする。この時、溶融塩化物を用いた
使用済み原子燃料の再処理で発生する塩廃棄物中では塩
化リチウムおよび塩化カリウムが他の塩化物に比べて濃
度が高く、かつ、塩化カリウムの溶解度は塩化リチウム
の溶解度より低いため、まず塩化カリウムが析出し、続
いて塩化リチウムが析出し始める。元素濃度モニタ６で
溶液中の元素濃度をモニタし、塩化リチウムおよび塩化
カリウム以外のいずれかの塩化物が飽和溶解度に達した
時点で気密反応槽１内の溶液を貯槽５に回収すれば塩化
リチウムおよび塩化カリウム以外の塩化物を含まない固
体が得られる。Salt waste generated by reprocessing spent nuclear fuel using molten chloride is supplied to the hermetic reaction tank 1 as powder by a salt waste crushing / supplying device 2. Subsequently, the steam pressure in the hermetic reaction tank 1 is increased using the steam pressure adjusting device 21. Preferably, at this time, the steam pressure is set in the hermetic reaction tank 1.
It is the saturated vapor pressure at the temperature inside. Since the chloride in the reaction tank 1 has deliquescence, it absorbs water vapor in the gas phase and becomes a chloride aqueous solution. When all the solids of the supplied salt waste are dissolved, the steam pressure in the hermetic reaction tank 1 is reduced by the steam pressure adjusting device 21. At this time, the water vapor pressure in the gas phase is preferably set to zero. At this time, in salt waste generated by reprocessing spent nuclear fuel using molten chloride, lithium chloride and potassium chloride are higher in concentration than other chlorides, and the solubility of potassium chloride is lithium chloride. , Potassium chloride precipitates first, and then lithium chloride starts to precipitate. The element concentration in the solution is monitored by the element concentration monitor 6, and when any chloride other than lithium chloride and potassium chloride reaches the saturation solubility, the solution in the hermetic reaction tank 1 is collected in the storage tank 5 to obtain lithium chloride. And a solid free of chlorides other than potassium chloride.
【００２０】本実施例によれば、実施例１，２，３の効
果に加え、飽和溶解度に近い高濃度のハロゲン化物溶液
を得ることができる。According to this embodiment, in addition to the effects of the first, second and third embodiments, a high-concentration halide solution close to the saturation solubility can be obtained.
【００２１】[0021]
【発明の効果】本発明によれば、ハロゲン化物の放射性
廃棄物を溶媒に溶解することにより、室温付近の操作に
置いてハロゲン化物の除染を行うことができ、回収した
非放射性のハロゲン化物を再利用することができる。According to the present invention, by dissolving a radioactive waste of a halide in a solvent, the decontamination of the halide can be performed at an operation near room temperature, and the recovered non-radioactive halide can be obtained. Can be reused.
【図１】本発明の一実施例であるハロゲン化物放射性廃
棄物の除染処理装置の系統図。FIG. 1 is a system diagram of an apparatus for decontaminating halide radioactive waste according to an embodiment of the present invention.
【図２】塩化物の飽和溶解度特性図。FIG. 2 is a diagram showing a saturated solubility characteristic of chloride.
【図３】塩化ナトリウムが飽和溶解度に達した時点での
塩化ナトリウム以外の塩化物沈殿の回収率の特性図。FIG. 3 is a characteristic diagram of a recovery rate of chloride precipitates other than sodium chloride when sodium chloride reaches a saturation solubility.
【図４】本発明の好適な一実施例であるハロゲン化物放
射性廃棄物の除染処理装置の系統図。FIG. 4 is a system diagram of an apparatus for decontaminating halide radioactive waste according to a preferred embodiment of the present invention.
【図５】本発明の好適な一実施例であるハロゲン化物放
射性廃棄物の除染処理装置の系統図。FIG. 5 is a system diagram of a halide radioactive waste decontamination treatment apparatus according to a preferred embodiment of the present invention.
【図６】本発明の好適な一実施例であるハロゲン化物放
射性廃棄物の除染処理装置の系統図。FIG. 6 is a system diagram of an apparatus for decontaminating halide radioactive waste according to a preferred embodiment of the present invention.
１…気密反応槽、２…塩廃棄物粉砕・供給装置、３…水
供給装置、４…真空ポンプ、５…溶液貯槽、６…元素濃
度モニタ、７…溶液、８…塩化物沈殿。DESCRIPTION OF SYMBOLS 1 ... Hermetic reaction tank, 2 ... Salt waste crushing and supply apparatus, 3 ... Water supply apparatus, 4 ... Vacuum pump, 5 ... Solution storage tank, 6 ... Element concentration monitor, 7 ... Solution, 8 ... Chloride precipitation.
Claims (6)
処理方法において、上記放射性廃棄物を溶媒中に溶解し
た後、上記溶媒から放射性物質を分離しハロゲン化放射
性廃棄物を除染することを特徴とする放射性廃棄物の除
染処理方法。1. A method for treating a radioactive waste in the form of a halide, comprising dissolving the radioactive waste in a solvent, separating the radioactive substance from the solvent, and decontaminating the halogenated radioactive waste. Characteristic decontamination method for radioactive waste.
媒である水に溶解させることにより溶液中の希土類元
素，アクチニド元素を酸化物として沈殿させ回収する放
射性廃棄物の除染方法。2. The method for decontaminating radioactive waste according to claim 1, wherein said halide is dissolved in water as a solvent to precipitate and recover rare earth elements and actinide elements in the solution as oxides.
元素を分離する手段として、上記溶媒を蒸発させること
により放射性廃棄物中の非放射性元素を析出沈殿させ、
固体として回収する放射性廃棄物の除染処理方法。3. The method according to claim 1, wherein as a means for separating the non-radioactive element from the solvent, the non-radioactive element in the radioactive waste is precipitated by evaporating the solvent.
A method for decontamination of radioactive waste recovered as a solid.
元素を分離する手段として、上記溶媒を冷却することに
より放射性廃棄物中の非放射性元素を析出沈殿させ、固
体として回収する放射性廃棄物の除染処理方法。4. The method according to claim 1, wherein the non-radioactive element in the radioactive waste is precipitated by cooling the solvent to separate the non-radioactive element from the solvent. Decontamination treatment method.
元素を分離する手段として、上記ハロゲン化物の放射性
元素であるセシウム，ストロンチウムを吸着剤を充填し
たカラムに通すことによってセシウム，ストロンチウム
を回収する放射性廃棄物の除染処理方法。5. The cesium and strontium are recovered by passing cesium and strontium, which are radioactive elements of the halide, through a column filled with an adsorbent, as means for separating non-radioactive elements from the solvent. Decontamination method of radioactive waste.
媒に溶解する手段として、上記ハロゲン化物を水蒸気を
含む雰囲気中に置くことにより上記ハロゲン化物を水溶
液とする放射性廃棄物の除染処理方法。6. A method for decontaminating radioactive waste according to claim 1, wherein the halide is dissolved in a solvent by dissolving the halide in a solvent by placing the halide in an atmosphere containing steam.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP1502897A JPH10213697A (en) | 1997-01-29 | 1997-01-29 | Decontamination method of radioactive waste |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP1502897A JPH10213697A (en) | 1997-01-29 | 1997-01-29 | Decontamination method of radioactive waste |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPH10213697A true JPH10213697A (en) | 1998-08-11 |
Family
ID=11877389
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP1502897A Pending JPH10213697A (en) | 1997-01-29 | 1997-01-29 | Decontamination method of radioactive waste |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH10213697A (en) |
Cited By (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO2014030282A1 (en) | 2012-08-22 | 2014-02-27 | 株式会社モリタ防災テック | Method for decontaminating soil and the like and system for decontaminating soil and the like |
| KR101514570B1 (en) * | 2013-11-25 | 2015-04-23 | 한국원자력연구원 | Separation method of hazardous component and non-hazardous component from highly concentrated metal salts in radioactive waste |
| EP3035341A1 (en) | 2014-12-16 | 2016-06-22 | Hideo Yoshida | Method for decontaminating soil and system for decontaminating soil |
| KR20200022153A (en) | 2018-08-22 | 2020-03-03 | (주)삼현 | A method for decontaminating soil with cesium and system for decontaminating soil with cesium |
| CN111681797A (en) * | 2020-04-30 | 2020-09-18 | 中国辐射防护研究院 | Method for treating radioactive wastewater of retired field of small nuclear facility |
-
1997
- 1997-01-29 JP JP1502897A patent/JPH10213697A/en active Pending
Cited By (9)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO2014030282A1 (en) | 2012-08-22 | 2014-02-27 | 株式会社モリタ防災テック | Method for decontaminating soil and the like and system for decontaminating soil and the like |
| US9770742B2 (en) | 2012-08-22 | 2017-09-26 | Hideo Yoshida | Method for decontaminating soil and the like and system for decontaminating soil and the like |
| KR101514570B1 (en) * | 2013-11-25 | 2015-04-23 | 한국원자력연구원 | Separation method of hazardous component and non-hazardous component from highly concentrated metal salts in radioactive waste |
| EP3035341A1 (en) | 2014-12-16 | 2016-06-22 | Hideo Yoshida | Method for decontaminating soil and system for decontaminating soil |
| KR20160073317A (en) | 2014-12-16 | 2016-06-24 | 히데오 요시다 | Method for decontaminating soil and the like and system for decontaminating soil and the like |
| US9721689B2 (en) | 2014-12-16 | 2017-08-01 | Hideo Yoshida | Method for decontamination of an object |
| US9754692B2 (en) | 2014-12-16 | 2017-09-05 | Hideo Yoshida | System for decontaminating soil and the like |
| KR20200022153A (en) | 2018-08-22 | 2020-03-03 | (주)삼현 | A method for decontaminating soil with cesium and system for decontaminating soil with cesium |
| CN111681797A (en) * | 2020-04-30 | 2020-09-18 | 中国辐射防护研究院 | Method for treating radioactive wastewater of retired field of small nuclear facility |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| Shadrin et al. | РH process as a technology for reprocessing mixed uranium–plutonium fuel from BREST-OD-300 reactor | |
| US5809394A (en) | Methods of separating short half-life radionuclides from a mixture of radionuclides | |
| KR102416164B1 (en) | In particular a method for producing an iodine radioisotope fraction of I-131, in particular an iodine radioisotope fraction of I-131 | |
| US5852786A (en) | Process for decontaminating radioactive materials | |
| JPH10213697A (en) | Decontamination method of radioactive waste | |
| JP2013127437A (en) | Method and device for treating radioactive cesium-containing substance | |
| JP3342968B2 (en) | Reprocessing of spent fuel | |
| JP2016114414A (en) | Separation removal method for toxic elements contained in radioactive waste liquid | |
| EP0347255B1 (en) | Method of treating high-level radioactive waste liquid | |
| JP6253046B2 (en) | Cesium extraction method | |
| US5340506A (en) | Method to synthesize dense crystallized sodalite pellet for immobilizing halide salt radioactive waste | |
| US5252258A (en) | Method of recovering and storing radioactive iodine by freeze vacuum drying process | |
| KR102549477B1 (en) | Decontamination method for radioactive-contaminated soil | |
| JP2016176742A (en) | Radioactive cesium immobilization method and radioactive cesium absorbing inorganic mineral | |
| KR101369123B1 (en) | Method for separation of group Ⅱ nuclides and recovery of pure LiCl by using Li2O | |
| JP7692603B2 (en) | Methods for removing radioactive cesium from contaminated soil | |
| USH1227H (en) | Method for immobilizing mixed waste chloride salts containing radionuclides and other hazardous wastes | |
| JPH01316695A (en) | Reprocessing of nuclear fuel by using vacuum freeze drying method | |
| EP0361773B1 (en) | Method of recovering radioactive iodine in a spent nuclear fuel retreatment process | |
| JP3058705B2 (en) | Radioactive waste treatment method and pretreatment equipment | |
| JP2005331276A (en) | Uranium recovery method and decontamination method for uranium-containing alumina particles | |
| JP2003294888A (en) | Method and apparatus for processing radioactive waste, and its solidifying method | |
| JP3058854B2 (en) | Radioactive waste treatment method | |
| JP7298052B2 (en) | Method for separating cesium and technetium | |
| JPH10206594A (en) | Radioactive waste treatment method and apparatus |