JP4533980B2 - High volume reduction vitrification treatment method of high level radioactive liquid waste - Google Patents

High volume reduction vitrification treatment method of high level radioactive liquid waste Download PDF

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JP4533980B2
JP4533980B2 JP2006085976A JP2006085976A JP4533980B2 JP 4533980 B2 JP4533980 B2 JP 4533980B2 JP 2006085976 A JP2006085976 A JP 2006085976A JP 2006085976 A JP2006085976 A JP 2006085976A JP 4533980 B2 JP4533980 B2 JP 4533980B2
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厚 青嶋
孝治 藤原
盛弘 新原
秀和 小林
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独立行政法人 日本原子力研究開発機構
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本発明は、使用済燃料の再処理工程で発生する高レベル放射性廃液をガラス溶融炉へ供給しガラス固化処理する方法に関し、更に詳しく述べると、高レベル放射性廃液に含まれているモリブデン及びセシウム、ストロンチウムを供給経路内での一連の工程で分離除去処理してガラス溶融炉に供給することにより、安定な高減容ガラス固化体を得ることができるようにする高レベル放射性廃液の高減容ガラス固化処理方法に関するものである。   The present invention relates to a method of supplying a high-level radioactive waste liquid generated in a spent fuel reprocessing step to a glass melting furnace and vitrifying, and more specifically, molybdenum and cesium contained in the high-level radioactive waste liquid, High-level radioactive waste liquid volume-reduction glass that enables to obtain a stable high-volume-reduced vitrified body by separating and removing strontium through a series of steps in the supply path and supplying it to the glass melting furnace The present invention relates to a solidification processing method.

再処理工場で使用済核燃料を再処理する過程では高レベル放射性廃液が発生する。この高レベル放射性廃液は、濃縮した後、ガラス溶融炉へ供給し、ガラス原料と混合して溶融することで固化処理される。得られたガラス固化体は、貯蔵により短寿命核種の減衰を行った後、地層処分されることになる。   In the process of reprocessing spent nuclear fuel at the reprocessing plant, high-level radioactive liquid waste is generated. This high-level radioactive liquid waste is concentrated, then supplied to a glass melting furnace, mixed with a glass raw material, and melted to be solidified. The obtained vitrified body is subjected to geological disposal after attenuation of short-lived nuclides by storage.

このような地層処分を行う際には、ガラス固化体は、均質で且つ長期間にわたって化学的に安定でなければならない。このため、従来技術では、廃棄物量に対するガラス原料の割合を多くすることで安定化を図っている。ガラス原料の割合を低くして廃棄物含有量を多くすると、ガラスに対する溶解度が低く（３〜４ｗｔ％）、溶解度を超えると水溶性の化合物として析出する高放射性廃液中のモリブデンが、ガラス中に析出するためである。そこで具体的には、廃棄物濃度を３０ｗｔ％以下に制限し、固化ガラスによる廃棄物閉じ込め性能を保持する必要があった。しかし、ガラス固化体中の廃棄物濃度が低いと、ガラス固化体の発生本数が多くなって貯蔵・地層処分のコストアップとなる問題が生じる。   When performing such geological disposal, the vitrified body must be homogeneous and chemically stable over a long period of time. For this reason, in the prior art, stabilization is achieved by increasing the ratio of the glass raw material to the amount of waste. When the percentage of glass raw material is reduced and the waste content is increased, the solubility in glass is low (3 to 4 wt%), and when the solubility is exceeded, molybdenum in the highly radioactive waste liquid that precipitates as a water-soluble compound is contained in the glass. It is for precipitation. Therefore, specifically, it was necessary to limit the waste concentration to 30 wt% or less and maintain the waste confinement performance by the solidified glass. However, if the concentration of waste in the vitrified body is low, the number of vitrified bodies generated increases, resulting in an increase in costs for storage and geological disposal.

このような問題を解決できる技術として、高レベル放射性廃液からＭｏ（モリブデン）及びＺｒ（ジルコニウム）を主成分とする沈殿物を除去した後、ガラス溶融炉に供給するガラス固化処理方法が提案されている（特許文献１参照）。これは、高レベル放射性廃液中の沈殿物の主成分がＭｏとＺｒであることに着目し、固化処理前に沈殿物（モリブデン酸ジルコニウム）を分離除去するものである。しかし、ガラス固化処理前に単に沈殿物を分離除去しただけではイエローフェーズと呼ばれる水溶性の析出物が固化ガラス中に析出する現象を防止することはできず、そのため沈殿物の除去操作のみならず、使用するガラス原料の組成を特定し従来のガラス原料とは異なる組成にしなければならない。そのために、使用実績がありデータが豊富な従来から用いられてきたガラス原料が使用できないという問題がある。しかも、固化処理前に沈殿物（モリブデン酸ジルコニウム）を分離除去し、ガラス組成を変更しても、ガラス固化体中の廃棄物濃度は４５ｗｔ％までにとどまっている。   As a technology that can solve such problems, a vitrification method for supplying a glass melting furnace after removing precipitates mainly composed of Mo (molybdenum) and Zr (zirconium) from a high-level radioactive liquid waste has been proposed. (See Patent Document 1). This focuses on the fact that the main components of the precipitate in the high-level radioactive liquid waste are Mo and Zr, and separates and removes the precipitate (zirconium molybdate) before the solidification treatment. However, simply separating and removing precipitates before vitrification treatment cannot prevent the phenomenon of water-soluble precipitates called yellow phase, which precipitates in the solidified glass. The composition of the glass raw material to be used must be specified and the composition different from that of the conventional glass raw material. For this reason, there is a problem that glass materials that have been used and have abundant data cannot be used. In addition, even if the precipitate (zirconium molybdate) is separated and removed before the solidification treatment and the glass composition is changed, the waste concentration in the glass solidified body is limited to 45 wt%.

更に、ガラス固化体は、その発熱量が約０．３５ｋｗ以下となるまで冷却しながら貯蔵し、その後、地層処分することになるが、廃棄物濃度を高めると発熱量も上昇するため、貯蔵時の冷却や貯蔵期問の長期化に伴うコストが増加するという問題が生じる。
特開平８−２３３９９３号公報 Furthermore, the vitrified body is stored while cooling until its calorific value is about 0.35 kw or less, and then disposed of in the geological layer. However, if the waste concentration is increased, the calorific value also increases. There arises a problem that the costs associated with the cooling and storage period increase.
JP-A-8-233993

本発明が解決しようとする課題は、ガラス固化体中の廃棄物濃度を４５〜５５ｗｔ％まで高め、従来のガラス原料を用いても均質で且つ長期間にわたって化学的に安定なガラス固化体を得ることができ、貯蔵・地層処分のコストダウンを図ることができるようにすることである。本発明が解決しようとする他の課題は、ガラス固化体の発熱を抑制し、貯蔵時の冷却や貯蔵期間の短縮化を図り、コスト削減を実現できるようにすることである。   The problem to be solved by the present invention is to increase the waste concentration in the vitrified body to 45 to 55 wt%, and to obtain a vitrified body that is homogeneous and chemically stable over a long period of time even when conventional glass raw materials are used. It is possible to reduce the cost of storage and geological disposal. Another problem to be solved by the present invention is to suppress heat generation of the vitrified body, to reduce the cooling during storage and the storage period, and to realize cost reduction.

本発明は、高レベル放射性廃液をガラス溶融炉に供給する経路の途中に、高レベル放射性廃液から固体として存在するモリブデン酸塩及びイオンとして溶解しているモリブデンを順次分離するモリブデン除去ユニット、次いで発熱元素でありイオンとして溶解しているセシウム及びストロンチウムを分離する発熱元素除去ユニットを配置し、高レベル放射性廃液からモリブデン及びセシウム、ストロンチウムを供給経路内での一連の工程で分離除去処理し、それらが含まれていない廃液をガラス溶融炉に供給することで、ガラス原料との混合・溶融固化処理により廃棄物濃度４５〜５５ｗｔ％の高減容ガラス固化体にすることを特徴とする高レベル放射性廃液の高減容ガラス固化処理方法である。   The present invention is a molybdenum removal unit that sequentially separates molybdate present as a solid and molybdenum dissolved as ions from the high-level radioactive waste liquid in the middle of a path for supplying the high-level radioactive waste liquid to the glass melting furnace, and then generates heat. An exothermic element removal unit that separates cesium and strontium dissolved as ions and elements is arranged, and molybdenum, cesium, and strontium are separated and removed from the high-level radioactive liquid waste through a series of steps in the supply path. A high-level radioactive waste liquid characterized in that a waste liquid concentration of 45 to 55 wt% is obtained by mixing and melting and solidifying with a glass raw material by supplying waste liquid not contained in the glass melting furnace. This is a high volume reduction vitrification method.

前記モリブデン除去ユニットは、典型的な例としては、上流側に位置し固体として存在するモリブデン酸塩を沈殿除去する沈降分離器と、下流側に位置しイオンとして溶解しているモリブデンを析出除去する電解析出器とを具備している。固体として存在するモリブデン酸塩の分離除去は、廃液とモリブデン酸塩との比重差を利用した沈降分離器が簡便で最適であるが、メンブレンフィルタなどを用いた濾過や圧搾脱水、遠心清澄、廃液に膜等の多孔性物質を通して電位を与えた場合に生じる電気浸透による溶媒移動現象や電気泳動による粒子移動現象等を利用した電気浸透脱水なども利用可能である。イオンとして溶解しているモリブデンの分離除去は、電解析出器による析出除去が、除去性能の観点から好ましいが、アルミナなどの吸着材に通過させることで選択的に分離除去することも可能である。   The molybdenum removal unit typically has a sedimentation separator for precipitating and removing molybdate present as a solid located upstream and a molybdenum which is located downstream and dissolved as ions. And an electrolytic depositor. For separation and removal of molybdate present as a solid, a sedimentation separator that uses the difference in specific gravity between waste liquid and molybdate is convenient and optimal, but filtration using membrane filters, pressure dehydration, centrifugal clarification, waste liquid In addition, electroosmosis dehydration using a solvent movement phenomenon due to electroosmosis or a particle movement phenomenon due to electrophoresis, which occurs when a potential is applied through a porous material such as a membrane, can be used. Separation and removal of molybdenum dissolved as ions is preferably performed by electrolytic deposition from the viewpoint of removal performance, but can be selectively separated and removed by passing it through an adsorbent such as alumina. .

前記発熱元素除去ユニットは、主にランタノイド及びアクチノイドを析出させる脱硝器と、その析出物を分離する濾過器と、セシウム吸着カラム及びストロンチウム吸着カラムと、廃液の濃度調整を行う組成調整槽からなり、前記濾過器による堆積物を組成調整槽に戻すようにする構成が好ましい。例えばセシウム吸着カラムは吸着材にフェリフェライトを使用したカラムであり、ストロンチウム吸着カラムは吸着材にＡ型ゼオライトを使用したカラムである。   The exothermic element removal unit mainly comprises a denitration device for precipitating lanthanoids and actinides, a filter for separating the precipitates, a cesium adsorption column and a strontium adsorption column, and a composition adjustment tank for adjusting the concentration of waste liquid, It is preferable that the deposit by the filter be returned to the composition adjustment tank. For example, a cesium adsorption column is a column using ferriferrite as an adsorbent, and a strontium adsorption column is a column using A-type zeolite as an adsorbent.

本発明に係る高レベル放射性廃液の高減容ガラス固化処理方法は、高レベル放射性廃液をガラス溶融炉に供給する経路の途中に、モリブデン除去ユニット、次いで発熱元素除去ユニットを配置し、高レベル放射性廃液から固体として存在しているモリブデン酸塩のみならず、イオンとして溶解しているモリブデン、セシウム、ストロンチウムを供給経路内での一連の工程で分離除去処理し、それらが含まれていない廃液をガラス溶融炉に供給するように構成されているので、ガラス固化体中の廃棄物濃度を４５〜５５ｗｔ％まで高め、均質で且つ長期間にわたって化学的に安定な、しかも発熱の少ないガラス固化体を得ることができる。そのため、貯蔵時の冷却や貯蔵期間の短縮化を図ることができ、貯蔵・地層処分のコストダウンを図ることもできる。   The high volume reduction vitrification processing method of the high level radioactive liquid waste according to the present invention includes a molybdenum removal unit and then a heat generating element removal unit in the middle of the path for supplying the high level radioactive liquid waste to the glass melting furnace. Not only the molybdate present as a solid from the waste liquid, but also molybdenum, cesium, and strontium dissolved as ions are separated and removed in a series of steps in the supply path, and the waste liquid that does not contain them is made into glass. Since it is configured to be supplied to the melting furnace, the waste concentration in the vitrified body is increased to 45 to 55 wt%, and a vitrified body that is homogeneous, chemically stable over a long period of time, and generates less heat is obtained. be able to. As a result, cooling during storage and shortening of the storage period can be achieved, and costs for storage and geological disposal can be reduced.

本発明に係る高レベル放射性廃液の高減容ガラス固化処理方法の典型例を図１に示す。再処理工場で使用済核燃料を再処理する過程で発生する高レベル放射性廃液１０を、ガラス溶融炉１２に供給する経路の途中に、高レベル放射性廃液から固体として存在するモリブデン酸塩及びイオンとして溶解しているモリブデンを順次分離するモリブデン除去ユニット１４、次いで発熱元素でありイオンとして溶解しているセシウム及びストロンチウムを分離する発熱元素除去ユニット１６を配置し、高レベル放射性廃液からモリブデン及びセシウム、ストロンチウムを供給経路内での一連の工程で分離除去処理する。そして、それらが含まれていない廃液を、廃液供給槽１８に貯留し、その後、ガラス原料供給系２０から送られるガラス原料に染み込ませ、ガラス溶融炉１２に供給して混合・溶融し、排出してガラス固化体とする。なお、分離除去したモリブデン、及びセシウム、ストロンチウムは、モリブデン回収槽２２及び発熱元素回収槽２４へ回収する。また、各工程で発生するガスは、オフガス処理系２６で処理する。   A typical example of the high-volume radioactive waste liquid high-volume reduction vitrification method according to the present invention is shown in FIG. The high-level radioactive liquid waste 10 generated in the process of reprocessing spent nuclear fuel at the reprocessing plant is dissolved as molybdate and ions present as solids from the high-level radioactive liquid waste in the course of supplying the glass melting furnace 12. A molybdenum removal unit 14 for sequentially separating molybdenum, and then a heat generation element removal unit 16 for separating cesium and strontium dissolved as ions as exothermic elements, so that molybdenum, cesium and strontium are removed from the high-level radioactive liquid waste. Separation and removal are performed in a series of steps in the supply path. And the waste liquid which does not contain them is stored in the waste liquid supply tank 18, and then impregnated into the glass raw material sent from the glass raw material supply system 20, supplied to the glass melting furnace 12, mixed and melted, and discharged. To make a vitrified body. The separated molybdenum, cesium, and strontium are recovered in the molybdenum recovery tank 22 and the exothermic element recovery tank 24. Further, the gas generated in each process is processed by the off-gas processing system 26.

モリブデン除去ユニットの例を図２に示す。このモリブデン除去ユニットは、上流側に位置し、固体として存在するモリブデン酸塩（主にモリブデン酸ジルコニウム）を廃液との比重差を利用して沈殿除去する沈降分離器３０と、その下流側に位置し、イオンとして溶解しているモリブデンを直流通電により電極に選択的に析出させて分離除去する電解析出器３２などからなる。   An example of the molybdenum removal unit is shown in FIG. This molybdenum removal unit is located on the upstream side, a sedimentation separator 30 for precipitating and removing molybdate (mainly zirconium molybdate) present as a solid by utilizing the specific gravity difference with the waste liquid, and located on the downstream side thereof. In addition, it includes an electrolytic deposition device 32 that selectively separates molybdenum dissolved as ions on the electrode by direct current application and separates and removes it.

高レベル放射性廃液を沈降分離器３０内の沈殿槽に導き一時貯留し、比重の大きいモリブデン酸ジルコニウムを底部に沈降させ、上澄み液と分離することにより除去する。沈殿槽に沈降したモリブデン酸ジルコニウムは、洗浄液で洗浄後、モリブデン回収槽２２へ回収する。次に、上澄み液を電解析出器３２へ導く。ここで電解析出器３２は、電極棒が挿入された貯槽と直流電源ユニットなどから構成される。電解析出器３２では、廃液を貯槽に一時貯留し、電極へ直流通電し電位を調節することで、廃液中にイオンとして溶解しているモリブデンを電極棒表面に選択的に析出させ、廃液から分離する。電解析出器３２の電極を、定期的に交換し、電極棒に析出したモリブデンを回収する。このようなプロセスによりモリブデンを除去した廃液は、オーバーフロー方式により発熱元素除去ユニット１６へ送られる。   The high-level radioactive waste liquid is guided to a sedimentation tank in the sedimentation separator 30 and temporarily stored, and zirconium molybdate having a large specific gravity is settled at the bottom and removed by separating it from the supernatant. The zirconium molybdate that has settled in the settling tank is collected in the molybdenum collection tank 22 after being washed with the washing liquid. Next, the supernatant liquid is guided to the electrolytic depositor 32. Here, the electrolytic depositor 32 includes a storage tank in which electrode bars are inserted, a DC power supply unit, and the like. In the electrolytic depositor 32, the waste liquid is temporarily stored in a storage tank, and direct current is applied to the electrode to adjust the potential, thereby selectively depositing molybdenum dissolved as ions in the waste liquid on the surface of the electrode rod. To separate. The electrode of the electrolytic depositor 32 is periodically replaced, and the molybdenum deposited on the electrode rod is collected. The waste liquid from which molybdenum has been removed by such a process is sent to the exothermic element removal unit 16 by the overflow method.

発熱元素除去ユニットの例を図３に示す。この発熱元素除去ユニット１６は、脱硝器４０と、その析出物を分離する濾過器４２と、セシウム吸着カラム４４及びストロンチウム吸着カラム４６と、廃液の濃度調整を行う組成調整槽４８などからなる。モリブデンを除去した廃液を脱硝器４０内の沈殿槽に一時貯留する。脱硝器４０では、廃液へギ酸を添加してｐＨを調整し、加熱することで廃液中のセシウムとその同属元素であるナトリウム、ストロンチウムとその同属元素であるカルシウム、バリウム以外の元素を析出させる。特にランタノイド及びアクチノイドを析出させることができる。この廃液を濾過器４２で濾過し、析出物を分離除去する。濾過器４２に残った堆積物を硝酸で溶解し、組成調整槽４８へ移送する。濾過器４２を通過した廃液をセシウム吸着カラム４４に通してセシウムを吸着分離する。吸着材であるフェリフェライトは定期的に硝酸アンモニウムで洗浄し、吸着したセシウムを溶解して発熱元素回収槽２４へ回収する。更に、セシウム吸着カラムを通過した廃液をストロンチウム吸着カラム４６に通してストロンチウムを吸着分離する。吸着材であるＡ型ゼオライトは定期的にＥＤＴＡで洗浄し、吸着したストロンチウムを溶解して発熱元素回収槽２４へ回収する。前記脱硝器と濾過器により、セシウム、ストロンチウムの吸着と関係のない元素を予め取り除くことで、その後の吸着カラムでの吸着効率を上げることができる。   An example of the exothermic element removal unit is shown in FIG. The exothermic element removing unit 16 includes a denitrator 40, a filter 42 for separating the precipitate, a cesium adsorption column 44 and a strontium adsorption column 46, a composition adjustment tank 48 for adjusting the concentration of the waste liquid, and the like. The waste liquid from which the molybdenum has been removed is temporarily stored in a settling tank in the denitrator 40. In the denitration device 40, formic acid is added to the waste liquid to adjust the pH, and heating is performed to precipitate elements other than cesium and its related elements sodium, strontium and its related elements such as calcium and barium in the waste liquid. In particular, lanthanoids and actinides can be precipitated. This waste liquid is filtered with a filter 42 to separate and remove the precipitate. The deposit remaining in the filter 42 is dissolved with nitric acid and transferred to the composition adjustment tank 48. The waste liquid that has passed through the filter 42 is passed through a cesium adsorption column 44 to adsorb and separate cesium. The ferriferrite as an adsorbent is periodically washed with ammonium nitrate, and the adsorbed cesium is dissolved and recovered in the exothermic element recovery tank 24. Furthermore, the waste liquid that has passed through the cesium adsorption column is passed through a strontium adsorption column 46 to adsorb and separate strontium. The adsorbent A-type zeolite is periodically washed with EDTA, and the adsorbed strontium is dissolved and recovered in the exothermic element recovery tank 24. By removing elements that are not related to the adsorption of cesium and strontium in advance by the denitrator and the filter, the adsorption efficiency in the subsequent adsorption column can be increased.

そして、以上のプロセスにより発熱元素であるセシウム、ストロンチウムを除去した廃液を、組成調整槽４８で濾過器４２の洗浄液と混合し、硝酸、純水により組成調整を行いオーバーフロー方式により、廃液供給槽１８に貯留する。廃液供給槽１８の廃液をガラス供給系２０からのガラス原料に供給し、廃液が染み込んだガラス原料をガラス溶融炉１２内へ供給する。ガラス溶融炉１２は、耐火物の炉体５０の側壁に電極５２を対向配置し、ガラス原料を通電溶融するものであり、溶融ガラスを下方の排出口から容器内に排出し、冷却することでガラス固化体となる。このとき、ガラス原料に対して廃棄物の混合割合を高め、廃棄物濃度４５〜５５ｗｔ％のガラス固化体とする。   The waste liquid from which the exothermic elements cesium and strontium are removed by the above process is mixed with the cleaning liquid of the filter 42 in the composition adjustment tank 48, the composition is adjusted with nitric acid and pure water, and the waste liquid supply tank 18 is obtained by the overflow method. Store in. The waste liquid in the waste liquid supply tank 18 is supplied to the glass raw material from the glass supply system 20, and the glass raw material soaked with the waste liquid is supplied into the glass melting furnace 12. In the glass melting furnace 12, an electrode 52 is disposed opposite to the side wall of a refractory furnace body 50, and the glass raw material is electrically melted. The molten glass is discharged into a container from a lower discharge port and cooled. It becomes a vitrified body. At this time, the mixing ratio of the waste is increased with respect to the glass raw material to obtain a glass solidified body having a waste concentration of 45 to 55 wt%.

図４は、ガラス固化体へのモリブデン溶解度を表している。一般にガラス固化体へのモリブデンの溶解は三酸化モリブデン（ＭｏＯ3 ）の状態で存在するため、モリブデンの濃度をＭｏＯ3 濃度としてグラフに表している。ここで、ＭｏＯ3 濃度が３〜４ｗｔ％以上で、モリブデンはセシウム等のアルカリ金属、アルカリ土類金属を同伴し、水溶性のモリブデン酸塩としてガラス中に析出することが知られている。現状廃液組成（例）の直線では廃棄物濃度が４５％でＭｏＯ3 析出領域となる。モリブデン酸塩の析出を防止して均質なガラスとするために、現状のガラス固化処理においては、工程変動を考慮して、廃棄物含有率を２０〜３０ｗｔ％となるように設定している。このときのＭｏＯ3 濃度は２％である。それに対して本発明においては、ガラス溶融炉への廃液供給過程にモリブデン除去ユニットを設けることで、廃液中のＭｏＯ3 の６０％以上を除去し、ガラス固化可能な廃棄物含有率約５５％においても、ＭｏＯ3 濃度を２ｗｔ％以下に抑えつつ、モリブデン酸塩の析出のない均質なガラスの製造が可能となる。本発明における廃液組成（例）の直線では、廃棄物濃度が５５ｗｔ％でもＭｏＯ3 濃度が２ｗｔ％以下となる。 FIG. 4 shows the solubility of molybdenum in the vitrified body. In general, the dissolution of molybdenum in the vitrified body exists in the form of molybdenum trioxide (MoO 3 ), so the concentration of molybdenum is shown in the graph as the MoO 3 concentration. Here, it is known that when the MoO 3 concentration is 3 to 4 wt% or more, molybdenum accompanies an alkali metal such as cesium or an alkaline earth metal and precipitates in the glass as a water-soluble molybdate. In the straight line of the current waste liquid composition (example), the waste concentration is 45%, which becomes the MoO 3 precipitation region. In order to prevent the precipitation of molybdate and to obtain a homogeneous glass, in the current vitrification treatment, the waste content rate is set to 20 to 30 wt% in consideration of process variations. At this time, the MoO 3 concentration is 2%. On the other hand, in the present invention, by providing a molybdenum removal unit in the waste liquid supply process to the glass melting furnace, 60% or more of MoO 3 in the waste liquid is removed, and the waste content rate that can be vitrified is about 55%. However, it is possible to produce a homogeneous glass without molybdate precipitation while suppressing the MoO 3 concentration to 2 wt% or less. In the straight line of the waste liquid composition (example) in the present invention, even if the waste concentration is 55 wt%, the MoO 3 concentration is 2 wt% or less.

処理手順の一実施例の詳細について説明する。図５はモリブデン除去ユニットの部分を示し、図６は発熱元素除去ユニットの前段部分、図７は発熱元素除去ユニットの後段部分を示している。   Details of one embodiment of the processing procedure will be described. 5 shows a portion of the molybdenum removal unit, FIG. 6 shows a front portion of the exothermic element removal unit, and FIG. 7 shows a rear portion of the exothermic element removal unit.

モリブデン除去ユニット１４では、まず沈降分離器３０に、一定の液位となるまで廃液を受け入れる。所定時間放置して固体成分を沈降分離し、上澄み液のみを全量次の電解析出器３２に移送する。洗浄液（２．５Ｎ硝酸及び純水）を沈降分離器３０に受け入れ、所定時間放置して固体成分を沈降分離し、上澄み液を次の電解析出器３２に移送する。再び洗浄液（硝酸及び純水）を沈降分離器３０に受け入れ、洗浄液の水流によって堆積物（主にモリブデン酸ジルコニウム）をモリブデン回収槽２２に移送し回収する。   In the molybdenum removing unit 14, first, the waste liquid is received in the sedimentation separator 30 until a certain liquid level is reached. The solid component is settled and separated by leaving for a predetermined time, and only the supernatant liquid is transferred to the next electrolytic depositing device 32. The cleaning liquid (2.5N nitric acid and pure water) is received in the sedimentation separator 30 and allowed to stand for a predetermined time to settle and separate the solid components, and the supernatant is transferred to the next electrolytic depositor 32. The cleaning liquid (nitric acid and pure water) is again received by the sedimentation separator 30, and the deposit (mainly zirconium molybdate) is transferred to the molybdenum recovery tank 22 and recovered by the water flow of the cleaning liquid.

電解析出器３２に、一定の液位となるまで沈降分離器３０の上澄み液を受け入れる。電極間に直流通電し電位を調節することでモリブデンを電極表面上に析出させる。電極間の抵抗値を測定し、所定の抵抗値になるまで通電を続ける。所定の抵抗値に達したならば、次に通電電気量を求め、所定の値に達したか否かを判定する。所定の通電電気量に達していなければ、通電を停止し、廃液を発熱元素除去ユニット１６に移送する。所定の通電電気量に達したならば、通電を停止し、電極の交換によりモリブデンを回収する。   The electrolytic separator 32 receives the supernatant liquid of the sedimentation separator 30 until a certain liquid level is reached. Molybdenum is deposited on the surface of the electrode by applying a direct current between the electrodes and adjusting the potential. The resistance value between the electrodes is measured, and energization is continued until a predetermined resistance value is reached. When the predetermined resistance value is reached, the energized electricity amount is then obtained to determine whether or not the predetermined value has been reached. If the predetermined energized electricity amount has not been reached, the energization is stopped and the waste liquid is transferred to the exothermic element removing unit 16. When the predetermined amount of energized electricity is reached, the energization is stopped and molybdenum is recovered by exchanging the electrodes.

発熱元素除去ユニット１６では、まず脱硝器４０に、一定の液位となるまでモリブデン除去済廃液を受け入れる。攪拌しながら、ｐＨが６．０〜７．０になるまでギ酸を添加調整する。また攪拌しながら加熱蒸気により９５℃に加熱し、その後、室温まで冷却する。脱硝器４０内の廃液全量を濾過器４２に移送する。濾液は、組成調整槽まで連続的に通過させる。濾過器４２ではメンブレンフィルタ（ここでは０．４５ＨＶを使用）で濾過処理を行い、濾液を順次セシウム吸着カラム４４に移送する。メンブレンフィルタ前後の差圧を測定し、所定値以上に達した場合、脱硝器４０からセシウム吸着カラム４４へのバルブを閉じ、フィルタ堆積物を硝酸によって溶解し、組成調整槽４８に移送する。   In the exothermic element removal unit 16, first, the molybdenum-removed waste liquid is received in the denitrator 40 until a certain liquid level is reached. While stirring, formic acid is added and adjusted until the pH is 6.0-7.0. Moreover, it heats to 95 degreeC with heating steam, stirring, and cools to room temperature after that. The total amount of waste liquid in the denitration device 40 is transferred to the filter 42. The filtrate is continuously passed to the composition adjustment tank. In the filter 42, filtration is performed with a membrane filter (0.45 HV is used here), and the filtrate is sequentially transferred to the cesium adsorption column 44. When the differential pressure before and after the membrane filter is measured and reaches a predetermined value or more, the valve from the denitrator 40 to the cesium adsorption column 44 is closed, and the filter deposit is dissolved with nitric acid and transferred to the composition adjustment tank 48.

濾液を、セシウム吸着カラム４４を通じてストロンチウム吸着カラム４６に移送する。セシウム吸着カラム４４での処理量が所定量以上になった場合は、脱硝器４０からの廃液供給を停止し、セシウム吸着カラム４４からストロンチウム吸着カラム４６へのバルブを閉じ、５Ｎ硝酸アンモニウムをセシウム吸着カラム４４に流し、吸着物を溶離する。そして、連続的に溶離液を発熱元素回収槽２４へ移送する。フェリフェライト吸着材を用いることによって、廃液中のセシウムの９８％を除去することができる。セシウム吸着カラム４４を通過した濾液を、ストロンチウム吸着カラム４６を通して組成調整槽４８に移送する。ストロンチウム吸着カラム４６での処理量が所定量以上になった場合は、脱硝器４０からの廃液供給を停止し、セシウム吸着カラム４４からストロンチウム吸着カラム４６へのバルブを閉じ、０．０５ＮのＥＤＴＡをストロンチウム吸着カラム４６に流し、吸着物を溶離する。そして、連続的に溶離液を発熱元素回収槽２４へ移送する。Ａ型ゼオライト吸着材を用いることによって、廃液中のストロンチウムの９７．４％を除去することができる。   The filtrate is transferred to the strontium adsorption column 46 through the cesium adsorption column 44. When the processing amount in the cesium adsorption column 44 exceeds a predetermined amount, the waste liquid supply from the denitrator 40 is stopped, the valve from the cesium adsorption column 44 to the strontium adsorption column 46 is closed, and 5N ammonium nitrate is added to the cesium adsorption column. 44 to elute the adsorbate. Then, the eluent is continuously transferred to the exothermic element recovery tank 24. By using the ferriferrite adsorbent, 98% of the cesium in the waste liquid can be removed. The filtrate that has passed through the cesium adsorption column 44 is transferred to the composition adjustment tank 48 through the strontium adsorption column 46. When the processing amount in the strontium adsorption column 46 exceeds a predetermined amount, the waste liquid supply from the denitrator 40 is stopped, the valve from the cesium adsorption column 44 to the strontium adsorption column 46 is closed, and 0.05 N EDTA is added. The strontium adsorption column 46 is flowed to elute the adsorbate. Then, the eluent is continuously transferred to the exothermic element recovery tank 24. By using the A-type zeolite adsorbent, 97.4% of strontium in the waste liquid can be removed.

組成調整槽４８に、ストロンチウム吸着カラム４６を通過した濾液及び濾過器４２の堆積物を硝酸によって溶解した廃液を一定液位まで受け入れる。攪拌しながら硝酸を加え、固体成分を完全に溶解し、廃液供給槽１８に移送する。廃液供給槽１８からオーバーフローによって、ガラス固化体中の廃棄物濃度が４５〜５５ｗｔ％となるように、ガラス原料に対して廃棄物量を多くしてガラス溶融炉１２に廃液を供給する。   The composition adjustment tank 48 receives the filtrate that has passed through the strontium adsorption column 46 and the waste liquid obtained by dissolving the deposits of the filter 42 with nitric acid to a certain liquid level. Nitric acid is added with stirring to completely dissolve the solid components and transfer to the waste liquid supply tank 18. The waste liquid is supplied to the glass melting furnace 12 by increasing the amount of waste with respect to the glass raw material so that the waste concentration in the glass solidified body becomes 45 to 55 wt% due to overflow from the waste liquid supply tank 18.

このようにして、ガラスに対する溶解度が低く溶解度を超えると水溶性の化合物として析出する高放射性廃液中のモリブデン、及び発熱量の大きい元素を、廃液供給段階において連続的に分離除去した後、溶融炉へ供給することで、廃棄物濃度を約５５ｗｔ％まで高くしてもモリブデン酸塩の析出を防止できるとともに、発熱量を低い状態に維持でき、貯蔵・地層処分コストの大幅な低減を図ることが可能となる。   In this way, the molybdenum in the highly radioactive waste liquid that precipitates as a water-soluble compound when the solubility in glass is low and exceeds the solubility, and the element with a large calorific value are continuously separated and removed in the waste liquid supply stage, and then the melting furnace Can prevent the precipitation of molybdate even when the waste concentration is increased to about 55 wt%, and can maintain the calorific value at a low level, thereby significantly reducing storage and geological disposal costs. It becomes possible.

本発明に係る高レベル放射性廃液の高減容ガラス固化処理方法の典型例を示す全体説明図。The whole explanatory drawing which shows the typical example of the high volume reduction vitrification processing method of the high level radioactive waste liquid which concerns on this invention. そのモリブデン除去ユニットの一例を示す説明図。Explanatory drawing which shows an example of the molybdenum removal unit. その発熱元素除去ユニットの一例を示す説明図。An explanatory view showing an example of the exothermic element removal unit. モリブデン溶解度と廃棄物濃度の関係を示すグラフ。The graph which shows the relationship between molybdenum solubility and waste concentration. モリブデン除去ユニットの処理手順の説明図。Explanatory drawing of the process sequence of a molybdenum removal unit. 発熱元素除去ユニットの前段での処理手順の説明図。Explanatory drawing of the process sequence in the front | former stage of a heat generating element removal unit. 発熱元素除去ユニットの後段での処理手順の説明図。Explanatory drawing of the process sequence in the back | latter stage of a heat generating element removal unit.

符号の説明Explanation of symbols

１０ 高レベル放射性廃液
１２ ガラス溶融炉
１４ モリブデン除去ユニット
１６ 発熱元素除去ユニット
１８ 廃液供給槽
２０ ガラス原料供給系
２２ モリブデン回収槽
２４ 発熱元素回収槽
２６ オフガス処理系
DESCRIPTION OF SYMBOLS 10 High level radioactive waste liquid 12 Glass melting furnace 14 Molybdenum removal unit 16 Exothermic element removal unit 18 Waste liquid supply tank 20 Glass raw material supply system 22 Molybdenum collection tank 24 Exothermic element collection tank 26 Off-gas processing system

Claims (4)

高レベル放射性廃液をガラス溶融炉に供給する経路の途中に、高レベル放射性廃液から固体として存在するモリブデン酸塩及びイオンとして溶解しているモリブデンを順次分離するモリブデン除去ユニット、次いで発熱元素でありイオンとして溶解しているセシウム及びストロンチウムを分離する発熱元素除去ユニットを配置し、高レベル放射性廃液からモリブデン及びセシウム、ストロンチウムを供給経路内での一連の工程で分離除去処理し、それらが含まれていない廃液をガラス溶融炉に供給することで、ガラス原料との混合・溶融固化処理により廃棄物濃度４５〜５５ｗｔ％の高減容ガラス固化体にすることを特徴とする高レベル放射性廃液の高減容ガラス固化処理方法。   Molybdenum removal unit that sequentially separates molybdate present as solid and molybdenum dissolved as ions from high-level radioactive waste liquid in the middle of the path for supplying high-level radioactive liquid waste to the glass melting furnace, then ions that are exothermic elements and ions A pyrogen element removal unit that separates dissolved cesium and strontium is arranged, and molybdenum, cesium, and strontium are separated and removed from the high-level radioactive liquid waste in a series of steps in the supply path, and they are not included High volume reduction of high-level radioactive waste liquid, characterized by supplying waste liquid to a glass melting furnace to produce a high volume reduction glass solidified with a waste concentration of 45 to 55 wt% by mixing and melting and solidifying with glass raw materials Vitrification method. 前記モリブデン除去ユニットは、上流側に位置し固体として存在するモリブデン酸塩を沈殿除去する沈降分離器と、下流側に位置しイオンとして溶解しているモリブデンを析出除去する電解析出器とを具備し、前記発熱元素除去ユニットは、セシウム吸着カラムとストロンチウム吸着カラムを具備している請求項１記載の高レベル放射性廃液の高減容ガラス固化処理方法。   The molybdenum removal unit includes a sedimentation separator that precipitates and removes molybdate existing as a solid located upstream, and an electrolytic depositor that precipitates and removes molybdenum dissolved as ions located downstream. The high-volume radioactive waste liquid solidification treatment method according to claim 1, wherein the exothermic element removing unit includes a cesium adsorption column and a strontium adsorption column. 前記発熱元素除去ユニットは、主にランタノイド及びアクチノイドを析出させる脱硝器と、その析出物を分離する濾過器と、セシウム吸着カラム及びストロンチウム吸着カラムと、廃液の濃度調整を行う組成調整槽からなり、前記濾過器による堆積物を組成調整槽に戻すようにした請求項２記載の高レベル放射性廃液の高減容ガラス固化処理方法。   The exothermic element removal unit mainly comprises a denitration device for precipitating lanthanoids and actinides, a filter for separating the precipitates, a cesium adsorption column and a strontium adsorption column, and a composition adjustment tank for adjusting the concentration of waste liquid, The high-volume radioactive vitrification treatment method for high-level radioactive liquid waste according to claim 2, wherein the deposits by the filter are returned to the composition adjustment tank. セシウム吸着カラムは吸着材にフェリフェライトを使用したカラムであり、ストロンチウム吸着カラムは吸着材にＡ型ゼオライトを使用したカラムである請求項２又は３記載の高レベル放射性廃液の高減容ガラス固化処理方法。
4. A high volume reduction vitrification treatment of a high-level radioactive liquid waste according to claim 2, wherein the cesium adsorption column is a column using ferriferrite as an adsorbent, and the strontium adsorption column is a column using A-type zeolite as an adsorbent. Method.
JP2006085976A 2006-03-27 2006-03-27 High volume reduction vitrification treatment method of high level radioactive liquid waste Expired - Fee Related JP4533980B2 (en)

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JPH08105998A (en) * 1994-10-07 1996-04-23 Power Reactor & Nuclear Fuel Dev Corp High volume reduction solidification method for high level radioactive waste liquid
JPH08233993A (en) * 1995-02-28 1996-09-13 Power Reactor & Nuclear Fuel Dev Corp Glass solidification method for high level radioactive waste liquid
JP2003161798A (en) * 2001-11-28 2003-06-06 Japan Nuclear Cycle Development Inst States Of Projects Separation and recovery method for rare element fp in spent nuclear fuel, and nuclear power generation- fuel cell power generation coexistence system utilizing the same

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* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JPH08105998A (en) * 1994-10-07 1996-04-23 Power Reactor & Nuclear Fuel Dev Corp High volume reduction solidification method for high level radioactive waste liquid
JPH08233993A (en) * 1995-02-28 1996-09-13 Power Reactor & Nuclear Fuel Dev Corp Glass solidification method for high level radioactive waste liquid
JP2003161798A (en) * 2001-11-28 2003-06-06 Japan Nuclear Cycle Development Inst States Of Projects Separation and recovery method for rare element fp in spent nuclear fuel, and nuclear power generation- fuel cell power generation coexistence system utilizing the same

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