JP3910605B2 - Method for electrolytic reprocessing of spent salt from spent fuel - Google Patents
Method for electrolytic reprocessing of spent salt from spent fuel Download PDFInfo
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Description
本発明は、使用済み燃料の溶融塩電解再処理方法に係わり、さらに詳しくは、原子力発電所で使用済みの燃料を溶融塩電解により再処理する方法に関する。 The present invention relates to a molten salt electrolysis reprocessing method of spent fuel, and more particularly to a method of reprocessing spent fuel in a nuclear power plant by molten salt electrolysis.
従来から、原子力発電所で使用済みの、酸化ウラン、アルカリ金属元素、貴金属元素、希土類元素、および原子炉内で生成したプルトニウム等の超ウラン元素(TRU)を含む燃料を再処理するには、せん断し脱被覆した後、NaClやKClのような溶融塩中で、陽極と陰極との間に電圧をかけて電解を行なう溶融塩電解方法が行なわれている。 To reprocess fuel that has traditionally been used in nuclear power plants and contains uranium oxide, alkali metal elements, noble metal elements, rare earth elements, and transuranium elements (TRU) such as plutonium produced in the reactor, A molten salt electrolysis method in which electrolysis is performed by applying a voltage between an anode and a cathode in a molten salt such as NaCl or KCl after shearing and decoating.
この方法においては、陽極で塩の電解により発生した塩素ガスによって、使用済み燃料の各成分が溶融塩中に溶解し、一方陰極で、電気的に還元されて酸化ウラン(UO2)等が顆粒状酸化物として回収される。そして、この溶解および電解回収工程で残った溶融塩に、塩素と酸素の混合ガスを吹き込むことにより、プルトニウムとネプツニウムのイオンがそれぞれ酸化され、かつ電解によりPuO2、およびNpO2が回収されるように構成されている。 In this method, each component of the spent fuel is dissolved in the molten salt by chlorine gas generated by electrolysis of the salt at the anode, while uranium oxide (UO 2 ) and the like are granulated by being electrically reduced at the cathode. It is recovered as a state oxide. Then, a mixed gas of chlorine and oxygen is blown into the molten salt remaining in this dissolution and electrolytic recovery process, so that plutonium and neptunium ions are oxidized, and PuO 2 and NpO 2 are recovered by electrolysis. It is configured.
しかしながら、このような従来からの溶融塩電解再処理方法においては、以下に示すような種々の問題があった。すなわち、
1)陽極溶解工程で、使用済み燃料に含まれている貴金属元素が溶融塩中に溶解するが、これら貴金属イオンが析出する電位(酸化還元電位)がウラニルイオンのそれと極めて近いため、陰極で貴金属元素も同時に析出してしまい、回収したUO2中に貴金属が混入してしまう。
2)パイログラファイトのような電解槽(電解るつぼ)を構成する材料が、ウラニルイオンによって腐食されやすいため、るつぼの寿命が1000〜2000時間と短い。
3)ウラナスイオンやプルトナスイオンの酸化処理に時間がかかる。
4)溶融塩から発生する揮発性成分が、電解槽内の各部に付着し、配管等の目詰まりを引き起こしやすい。
などの問題があった。
However, such conventional molten salt electrolytic reprocessing methods have various problems as described below. That is,
1) In the anodic dissolution process, the noble metal element contained in the spent fuel dissolves in the molten salt, but the potential at which these noble metal ions precipitate (oxidation-reduction potential) is very close to that of the uranyl ion, so the noble metal at the cathode Elements are also precipitated at the same time, and noble metals are mixed in the recovered UO 2 .
2) Since the material constituting the electrolytic cell (electrolytic crucible) such as pyrographite is easily corroded by uranyl ions, the life of the crucible is as short as 1000 to 2000 hours.
3) It takes time to oxidize uranas ions and plutonas ions.
4) Volatile components generated from the molten salt adhere to each part in the electrolytic cell and easily cause clogging of piping and the like.
There were problems such as.
本発明は、これらの問題を解決するためになされたもので、使用済み燃料の溶融塩電解による再処理において、貴金属元素等の不純物の混入のない、より高品質の燃料成分を回収することができ、また電解槽を構成するるつぼ基材の腐食を抑制し、寿命の延長が可能な溶融塩電解再処理方法を提供することを目的とする。 The present invention has been made to solve these problems, and can recover higher-quality fuel components free from impurities such as precious metal elements in reprocessing of spent fuel by molten salt electrolysis. Another object of the present invention is to provide a molten salt electrolytic reprocessing method capable of suppressing corrosion of the crucible base material constituting the electrolytic cell and extending the life.
本発明の使用済み燃料の溶融塩電解再処理方法は、使用済み燃料をせん断し脱被覆した後、溶融塩中で電解を行ない、陽極で溶解するとともに、所要の成分を陰極に析出させて回収する再処理方法において、前記燃料中の酸化ウランを陽極で溶解し固体陰極で析出・回収する工程を複数回繰り返し、前記溶融塩中のプルトニウムイオンの濃度を上昇させた後、該プルトニウムイオンを酸化してプルトニルイオンとし、次いで前記溶融塩中に液体金属を投入し、前記プルトニルイオンと該液体金属との交換反応により、酸化プルトニウムを析出させて回収することを特徴とする。 The method for reprocessing spent salt electrolysis of spent fuel according to the present invention is to recover spent components by shearing and decoating spent fuel, performing electrolysis in molten salt, dissolving at the anode, and depositing the required components on the cathode. In this reprocessing method, the process of dissolving uranium oxide in the fuel at the anode and precipitating and collecting at the solid cathode is repeated a plurality of times, and after increasing the concentration of plutonium ions in the molten salt, the plutonium ions are oxidized. Then, plutonium ions are obtained, and then a liquid metal is introduced into the molten salt, and plutonium oxide is precipitated and recovered by an exchange reaction between the plutonyl ions and the liquid metal.
本発明の再処理方法においては、陽極溶解同時電解後の溶融塩中のプルトニウムイオンを酸化し、次いで溶融塩中に液体金属を投入しプルトニルイオンと液体金属との交換反応により、酸化プルトニウムを析出させて回収しているので、従来の沈殿方式での回収方法に比べてプロセスの操業時間を大幅に短縮することができるうえに、回収が容易で特別な回収装置を必要とせず、装置を簡素化することができる。 In the reprocessing method of the present invention, the plutonium ion in the molten salt after the anodic dissolution simultaneous electrolysis is oxidized, and then the liquid metal is introduced into the molten salt, and the plutonium oxide is converted by the exchange reaction between the plutonium ion and the liquid metal. Since it is collected by precipitation, the operation time of the process can be greatly shortened compared to the collection method of the conventional precipitation method, and it is easy to collect and no special collection device is required. It can be simplified.
そして、回収する燃料中への貴金属元素等の不純物の混入を低減することができ、より高品質の燃料を得ることができる。また、使用済み燃料の陽極側での溶解と核燃料物質の陰極側での析出を同時に効率よく行なうことができ、使用済み燃料の処理効率を向上させることができる。さらに、アメリシウムやキュウリウムのようなマイナーアクチノイドとわずかに残るウラン、プルトニウムをともに回収することができ、これらの超半減期核種を再び炉心で消滅処理することができる。またさらに、従来からの沈殿法の処理時間と比べて、30分程度と極めて短時間で処理を行なうことができ、プロセスの操業時間が大幅に短縮される。 Further, impurities such as noble metal elements can be reduced in the fuel to be recovered, and a higher quality fuel can be obtained. Moreover, the spent fuel can be efficiently dissolved on the anode side and the nuclear fuel material can be deposited on the cathode side at the same time, and the processing efficiency of the spent fuel can be improved. Furthermore, minor actinides such as americium and cucurium can be recovered together with a little remaining uranium and plutonium, and these ultra-half-life nuclides can be extinguished again in the core. Furthermore, the treatment can be performed in an extremely short time of about 30 minutes as compared with the treatment time of the conventional precipitation method, and the operation time of the process is greatly shortened.
以下、本発明の実施例を図面に基づいて説明する。 Embodiments of the present invention will be described below with reference to the drawings.
本発明の実施例である、使用済み燃料からプルトニウムを回収する方法について説明する。この実施例のフローチャートを、図1に示す。 A method for recovering plutonium from spent fuel, which is an embodiment of the present invention, will be described. A flowchart of this embodiment is shown in FIG.
実施例では、まず使用済み燃料を電解槽の溶融塩中に投入し、陽極溶解同時陰極析出法により、陰極に顆粒状の酸化ウラン(UO2)を析出させて回収する。この操作を10回程度繰り返すと、使用済み燃料中に含まれるウラン、プルトニウム、マイナーアクチノイドの各成分は、溶融塩中で、図2に示すような濃度変化をする。 In the examples, first, spent fuel is put into a molten salt of an electrolytic cell, and granular uranium oxide (UO 2 ) is deposited on the cathode and collected by an anodic dissolution simultaneous cathode deposition method. When this operation is repeated about 10 times, the uranium, plutonium and minor actinoid components contained in the spent fuel change in concentration as shown in FIG. 2 in the molten salt.
すなわち、陰極では酸化ウランのみを回収しているので、ウラニルイオンの濃度は一定値を示し、プルトニウムイオンおよびマイナーアクチノイドイオンの濃度は、10倍程度上昇する。その後さらに電解を継続(しぼり電解)し、UO2をさらに回収すると、溶融塩中のウラニルイオン濃度は低下し、プルトニウムイオンおよびマイナーアクチノイドイオンの濃度よりも低いレベルに達する。この段階で、溶融塩中に酸素と塩素との混合ガスを吹き込み、プルトニウムイオン(Pu4+,Pu3+)を酸塩化物イオン(プルトニルイオン:PuO2 2+)の状態まで酸化する。 That is, since only uranium oxide is recovered at the cathode, the concentration of uranyl ions shows a constant value, and the concentration of plutonium ions and minor actinide ions increases about 10 times. When electrolysis is further continued (squeezing electrolysis) and UO 2 is further recovered, the uranyl ion concentration in the molten salt decreases and reaches a level lower than the concentrations of plutonium ions and minor actinide ions. At this stage, a mixed gas of oxygen and chlorine is blown into the molten salt to oxidize plutonium ions (Pu 4+ , Pu 3+ ) to the state of acid chloride ions (plutonyl ions: PuO 2 2+ ).
さらに、セラミック製の容器(るつぼ)に入れたビスマスのような金属を、溶融塩中に投入して溶融させると、溶融した金属とプルトニルイオンとが反応し、プルトニウムは酸化プルトニウムとして液体金属(ビスマス)中に粉末状で析出する。こうして、プルトニルイオンが液体金属中に回収される。 Furthermore, when a metal such as bismuth placed in a ceramic container (crucible) is poured into a molten salt and melted, the molten metal reacts with plutonyl ions, and plutonium is converted into plutonium oxide as a liquid metal ( The powder precipitates in (bismuth). Thus, plutonyl ions are recovered in the liquid metal.
さらに、プルトニルイオンの析出・回収後の処理は、以下に示すように行なわれる。すなわち、図3に示すように、セラミック製の容器1に入れられた液体金属2を撹拌機3により撹拌して、邪魔板4が取付けられた壁面方向への対流5を生じさせ、液体金属2の表面に析出したPuO2を容器1の内壁方向に移動させて堆積させる。そして、常に液体金属2の表面が更新されるようにし、液体金属2を陰極として電解を行なう。この電解工程では、UO2のしぼり電解よりもさらに電位を下げ、しぼり電解では回収できずにわずかに残ったプルトニウムイオンやマイナーアクチノイドイオンを、金属として、液体金属2の表面に回収する。こうしてこれらの成分を液体金属2の陰極で回収した後、セラミック製の容器1を溶融塩6中から引き上げて冷却し、固体となった金属を回収する。そして、蒸留またはろ過により、液体金属2と回収された酸化プルトニウムやマイナーアクチノイドの金属粒子とを分離し、燃料成分を再度原子炉に戻して使用する。
Further, the treatment after the precipitation and recovery of plutonyl ions is performed as follows. That is, as shown in FIG. 3, the
なお、さらに高除染の燃料を得るためには、酸化プルトニウムとマイナーアクチノイドを回収した液体金属2を、溶融塩と接触させて塩化ビスマスを投入し、塩化物になりやすい希土類元素を溶融塩中に抽出して除染することも可能である。
In order to obtain a highly decontaminated fuel,
このような酸化プルトニウムの回収方法においては、従来の沈殿方式での回収方法に比べて、以下に示す利点を有している。すなわち、従来の沈殿法では、酸化プルトニウムの沈殿が完了するのに24時間程度かかっていたが、前記した実施例の回収方法では、30分程度の短時間で液体金属とプルトニルイオンとの化学反応が進行するので、プロセスの操業時間を大幅に短縮することができる。 Such a method for recovering plutonium oxide has the following advantages as compared to a conventional recovery method using a precipitation method. That is, in the conventional precipitation method, it took about 24 hours to complete the precipitation of plutonium oxide. However, in the recovery method of the above-described embodiment, the chemistry of the liquid metal and the plutonyl ion is completed in a short time of about 30 minutes. Since the reaction proceeds, the operation time of the process can be greatly shortened.
また沈殿法では、析出するPuO2の粒径が小さいため、回収装置が複雑になるばかりでなく、高温高湿雰囲気での操作なので装置の寿命が短く、信頼性に乏しい等の問題があったが、実施例の方法では、液体金属2の陰極をセラミック製容器1ごと引き上げることで、容易に回収することができ、特別な回収装置を必要とせず、装置全体を簡素化することができる。さらに従来は、小粒径の酸化プルトニウムを高率で回収することが難しかったが、実施例では、液体金属2界面の反応により生じた酸化プルトニウムの全量を、液体金属2中に回収することができるので、回収率はほぼ100%と極めて高くなる。
In addition, in the precipitation method, the particle size of the precipitated PuO 2 is small, so that not only the recovery apparatus becomes complicated, but also the operation is performed in a high-temperature and high-humidity atmosphere, so that the life of the apparatus is short and the reliability is poor. However, in the method of the embodiment, the cathode of the
またさらに、沈殿法では、マイナーアクチノイドを一括して回収することができなかったが、実施例によれば、電解を併用することで酸化プルトニウムの他にマイナーアクチノイドも同時に回収して、燃料とすることができる。これらは超半減期核種であるため、再び炉心で消滅処理を行なうことができ、放射性廃棄物の処分の上で非常に有利である。 Furthermore, in the precipitation method, minor actinides could not be collected at once. However, according to the examples, in addition to plutonium oxide, minor actinides can also be collected and used as fuel by using electrolysis together. be able to. Since these are ultra-half-life nuclides, they can be extinguished again in the core, which is very advantageous for the disposal of radioactive waste.
次に、析出・回収されたUO2と貴金属元素とを分離する方法について説明する。 Next, a method for separating the precipitated / recovered UO 2 and the noble metal element will be described.
使用済み燃料を溶融塩電解により再処理し、UO2を析出させる場合には、図4に示すように、ウラニルイオン(UO2 2+)と同じ酸化還元電位を有する貴金属元素(モリブデン(Mo)、ルテニウム(Ru)、ジルコニウム(Ζr)、ニッケル(Νi)、銀(Ag))が析出し、UO2との分離が悪くなるが、以下に示すようにして、両者を効率よく分離することができる。 When reprocessing spent fuel by molten salt electrolysis and precipitating UO 2 , as shown in FIG. 4, a noble metal element (molybdenum (Mo), having the same oxidation-reduction potential as uranyl ions (UO 2 2+ )). Ruthenium (Ru), zirconium (Ζr), nickel (Νi), silver (Ag)) precipitates and the separation from UO 2 is poor, but both can be separated efficiently as shown below. .
UO2は酸化物であるが、図5に示すように、高温においては電気伝導性(導電性)を有し、溶融塩を用いた比較的高い温度(550から700℃)で電解精製して回収できることが知られている。ここで、UO2を電解で回収する際の温度は、UO2の導電性を確保する上からも、最低550℃程度が必要である。そして、一旦回収したUO2を、次に 500℃以下の溶融塩中で電解しても、導電性が著しく低下しているため、UO2にはほとんど電流が流れない。しかし、UO2と共に析出する貴金属元素は、金属であるため、温度を500℃以下に保持しても電流が流れ、陽極から貴金属元素がイオンとして溶出するので、このような溶出量の差を利用して、UO2と貴金属元素とを分離し回収することができる。 Although UO 2 is an oxide, as shown in FIG. 5, it has electrical conductivity (conductivity) at a high temperature and is electrolytically purified at a relatively high temperature (550 to 700 ° C.) using a molten salt. It is known that it can be recovered. Here, the temperature at which the recovery of UO 2 in electrolysis, from top to ensure conductivity of UO 2, it is required to be about a minimum 550 ° C.. And even if the UO 2 once collected is electrolyzed in a molten salt at 500 ° C. or less, the conductivity is remarkably lowered, so that almost no current flows through the UO 2 . However, since the noble metal element precipitated together with UO 2 is a metal, current flows even if the temperature is kept below 500 ° C., and the noble metal element is eluted as ions from the anode. Thus, UO 2 and the noble metal element can be separated and recovered.
具体的例としては、650℃の溶融塩(NaCl−KCl)中でグラファイトの陰極上に電解析出したUO2と貴金属元素(Mo、Ru、Rh、Pd、Ag)とを、500℃に保持したLiCl−KCl溶融塩中に電極ごと入れ、塩素基準電極に対して−0.3Vに保持して電解を行なったところ、UO2はほとんど溶出しなかったが、他の金属は溶出した。溶出した貴金属元素を回収した後、温度を600℃に上げて電解を行ない、UO2を回収することができた。 As a specific example, UO 2 and noble metal elements (Mo, Ru, Rh, Pd, Ag) electrolytically deposited on a graphite cathode in a molten salt (NaCl—KCl) at 650 ° C. are maintained at 500 ° C. When the entire electrode was placed in the molten LiCl-KCl molten salt and electrolysis was performed while maintaining the chlorine reference electrode at -0.3 V, UO 2 was hardly eluted, but other metals were eluted. After recovering the eluted noble metal element, the temperature was raised to 600 ° C. to conduct electrolysis, and UO 2 could be recovered.
また、650℃の溶融塩(NaCl−KCl)中でグラファイトの陰極上に電解析出したUO2と貴金属元素とを、冷却後一旦電極から剥脱して粗粉砕し、顆粒状にした後、これに導電性物質であるグラファイトを混合し、陽極として溶融塩中で使用した。前記具体例と同じ電位、温度に保持して溶出させると、貴金属元素は導電性グラファイトの働きでより電流が通じやすくなるため、容易に溶融塩中に溶出し、陰極上に貴金属のみを回収することができた。この操作の終了後、温度を600℃に上げて電解を行なうことで、UO2が陰極上に析出しこれを回収することができた。 Further, after cooling, UO 2 and noble metal elements that were electrolytically deposited on the graphite cathode in a molten salt (NaCl—KCl) at 650 ° C. were peeled off from the electrode and then coarsely pulverized to form granules. Was mixed with graphite, which is a conductive material, and used as an anode in molten salt. When elution is carried out while maintaining the same potential and temperature as in the above specific example, the noble metal element becomes easier to conduct current due to the action of the conductive graphite, so it easily elutes in the molten salt and collects only the noble metal on the cathode. I was able to. After completion of this operation, the temperature was raised to 600 ° C. and electrolysis was performed, so that UO 2 was deposited on the cathode and could be recovered.
本発明の使用済み燃料の溶融塩電解再処理方法によれば、回収する燃料中への貴金属元素等の不純物の混入を低減することができ、より高品質の燃料を得ることができる。また、使用済み燃料の陽極側での溶解と核燃料物質の陰極側での析出を同時に効率よく行なうことができ、使用済み燃料の処理効率を向上させることができる。さらに、アメリシウムやキュウリウムのようなマイナーアクチノイドとわずかに残るウラン、プルトニウムをともに回収することができ、これらの超半減期核種を再び炉心で消滅処理することができる。またさらに、従来からの沈殿法の処理時間と比べて、30分程度と極めて短時間で処理を行なうことができ、プロセスの操業時間が大幅に短縮される。 According to the molten salt electrolysis reprocessing method of spent fuel of the present invention, it is possible to reduce the contamination of impurities such as noble metal elements into the recovered fuel, and to obtain a higher quality fuel. Moreover, the spent fuel can be efficiently dissolved on the anode side and the nuclear fuel material can be deposited on the cathode side at the same time, and the processing efficiency of the spent fuel can be improved. Furthermore, minor actinides such as americium and cucurium can be recovered together with a little remaining uranium and plutonium, and these ultra-half-life nuclides can be extinguished again in the core. Furthermore, the treatment can be performed in an extremely short time of about 30 minutes as compared with the treatment time of the conventional precipitation method, and the operation time of the process is greatly shortened.
1…セラミック製の容器、2…液体金属、3…撹拌機、4…邪魔板、6…溶融塩。
DESCRIPTION OF
Claims (2)
前記燃料中の酸化ウランを陽極で溶解し固体陰極で析出・回収する工程を複数回繰り返し、前記溶融塩中のプルトニウムイオンの濃度を上昇させた後、該プルトニウムイオンを酸化してプルトニルイオンとし、次いで前記溶融塩中に液体金属を投入し、前記プルトニルイオンと該液体金属との交換反応により、酸化プルトニウムを析出させて回収することを特徴とする使用済み燃料の溶融塩電解再処理方法。 In a reprocessing method in which spent fuel is sheared and decoated, then electrolyzed in molten salt, dissolved at the anode, and the required components are deposited and recovered at the cathode.
The process of dissolving uranium oxide in the fuel at the anode and precipitating and collecting at the solid cathode is repeated a plurality of times, and after increasing the concentration of plutonium ions in the molten salt, the plutonium ions are oxidized to plutonium ions. Then, a molten metal electrolytic reprocessing method for spent fuel, wherein a liquid metal is introduced into the molten salt, and plutonium oxide is deposited and recovered by an exchange reaction between the plutonyl ion and the liquid metal. .
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