JP2004361417A - Melted salt electrolytic re-treatment method of spent fuel - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide a melted salt electrolytic re-treatment method capable of recovering a higher-quality fuel component having no contamination of impurities such as a noble metal element, and prolonging the service life by suppressing corrosion of a crucible base material constituting an electrolytic cell, concerning re-treatment of a spent fuel by melted salt electrolysis. <P>SOLUTION: In this re-treatment method, a process of melting uranium oxide in the spent fuel at a positive electrode and depositing/recovering it at a solid negative electrode is repeated in a plurality of times, and the concentration of plutonium ions in the melted salt is raised, and then the plutonium ions are oxidized into pultonyl ions. Then, a liquid metal is inputted into the melted salt, and the plutonium oxide is deposited and recovered by an exchange reaction between the pultonyl ions and the liquid metal. <P>COPYRIGHT: (C)2005,JPO&NCIPI

Description

  The present invention relates to a method for reprocessing spent fuel by molten salt electrolysis, and more particularly to a method for reprocessing spent fuel in a nuclear power plant by molten salt electrolysis.

  Conventionally, to reprocess fuel containing uranium oxide, alkali metal elements, precious metal elements, rare earth elements, and transuranium elements (TRU) such as plutonium produced in nuclear reactors, used in nuclear power plants, A molten salt electrolysis method in which a voltage is applied between an anode and a cathode to perform electrolysis in a molten salt such as NaCl or KCl after shearing and decoating is performed.

In this method, each component of spent fuel is dissolved in a molten salt by chlorine gas generated by electrolysis of a salt at an anode, while uranium oxide (UO ₂ ) and the like are electrically reduced at a cathode to form granules. It is recovered as a state oxide. Then, by blowing a mixed gas of chlorine and oxygen into the molten salt remaining in the melting and electrolytic recovery steps, plutonium and neptunium ions are oxidized, respectively, and PuO ₂ and NpO ₂ are recovered by electrolysis. Is configured.

However, such a conventional molten salt electrolytic reprocessing method has various problems as described below. That is,
1) In the anode dissolving step, the noble metal element contained in the spent fuel dissolves in the molten salt. However, the potential (oxidation-reduction potential) at which these noble metal ions precipitate is very close to that of uranyl ion. The element is also precipitated at the same time, and the noble metal is mixed into the recovered UO ₂ .
2) Since the material constituting the electrolytic cell (electrolytic crucible) such as pyrographite is easily corroded by uranyl ions, the crucible has a short life of 1000 to 2000 hours.
3) It takes time to oxidize uranas ions or plutonas ions.
4) Volatile components generated from the molten salt adhere to various parts in the electrolytic cell, and easily cause clogging of pipes and the like.
There was such a problem.

  The present invention has been made to solve these problems, and in the reprocessing of spent fuel by molten salt electrolysis, it is possible to collect higher quality fuel components without contamination of impurities such as noble metal elements. It is another object of the present invention to provide a molten salt electrolysis reprocessing method capable of suppressing corrosion of a crucible base material constituting an electrolytic cell and extending the life.

  In the method for reprocessing spent fuel for molten salt of the present invention, after the spent fuel is sheared and uncoated, electrolysis is performed in a molten salt, the anode is dissolved at the same time, and the required components are deposited on the cathode and recovered. In the reprocessing method, the step of dissolving uranium oxide in the fuel at the anode and depositing and recovering it at the solid cathode is repeated a plurality of times to increase the concentration of plutonium ions in the molten salt, and then oxidize the plutonium ions. Then, a liquid metal is introduced into the molten salt, and a plutonium oxide is precipitated and recovered by an exchange reaction between the plutony ion and the liquid metal.

  In the reprocessing method of the present invention, the plutonium ion in the molten salt after anodic dissolution simultaneous electrolysis is oxidized, and then the liquid metal is charged into the molten salt, and the plutonium ion is exchanged with the liquid metal to convert the plutonium oxide. Since it is deposited and collected, the operation time of the process can be significantly reduced compared to the conventional collection method using the precipitation method.In addition, the collection is easy and no special collection device is required. It can be simplified.

  In addition, the contamination of impurities such as noble metal elements into the recovered fuel can be reduced, and a higher quality fuel can be obtained. Further, the dissolving of the spent fuel on the anode side and the precipitation of the nuclear fuel material on the cathode side can be performed efficiently at the same time, and the processing efficiency of the spent fuel can be improved. Furthermore, minor actinoids such as americium and cuurium, as well as slightly remaining uranium and plutonium, can be recovered, and these ultra-half-life nuclides can be extinguished again in the core. Furthermore, the processing can be performed in an extremely short time of about 30 minutes as compared with the processing time of the conventional precipitation method, and the operation time of the process is greatly reduced.

  Hereinafter, embodiments of the present invention will be described with reference to the drawings.

  A method for recovering plutonium from spent fuel, which is an example of the present invention, will be described. FIG. 1 shows a flowchart of this embodiment.

In the embodiment, first, spent fuel is put into a molten salt in an electrolytic cell, and uranium oxide (UO ₂ ) is precipitated and recovered on a cathode by an anodic dissolution simultaneous cathodic deposition method. When this operation is repeated about ten times, the components of uranium, plutonium, and minor actinoids contained in the spent fuel change in the molten salt as shown in FIG.

That is, since only uranium oxide is recovered at the cathode, the concentration of uranyl ion shows a constant value, and the concentrations of plutonium ion and minor actinoid ion increase about 10 times. Thereafter, when the electrolysis is further continued (squeezing electrolysis) and UO ₂ is further recovered, the uranyl ion concentration in the molten salt decreases, reaching a level lower than the concentrations of plutonium ions and minor actinoid ions. At this stage, a mixed gas of oxygen and chlorine is blown into the molten salt to oxidize the plutonium ions (Pu ⁴⁺ , Pu ³⁺ ) to the state of acid chloride ions (Plutonyl ions: PuO ₂ ²⁺ ).

  Furthermore, when a metal such as bismuth placed in a ceramic container (crucible) is thrown into a molten salt and melted, the molten metal reacts with plutonium ions, and plutonium becomes liquid metal (plutonium oxide). (Bismuth) in the form of a powder. Thus, plutony ions are collected in the liquid metal.

Further, the treatment after the precipitation and recovery of plutonyl ion is performed as described below. That is, as shown in FIG. 3, the liquid metal 2 placed in the ceramic container 1 is agitated by the stirrer 3 to generate convection 5 in the direction of the wall surface on which the baffle plate 4 is mounted. PuO ₂ deposited on the surface of the container 1 is moved toward the inner wall of the container 1 and deposited. Then, the surface of the liquid metal 2 is constantly renewed, and electrolysis is performed using the liquid metal 2 as a cathode. In this electrolysis step, the potential is further lowered than in the squeezing electrolysis of UO ₂ , and the remaining plutonium ions and minor actinoid ions that cannot be collected by the squeezing electrolysis are collected as metal on the surface of the liquid metal 2. After recovering these components using the liquid metal 2 cathode, the ceramic container 1 is pulled out of the molten salt 6 and cooled, and the solid metal is recovered. Then, the liquid metal 2 and the recovered metal particles of the plutonium oxide and the minor actinoid are separated by distillation or filtration, and the fuel component is returned to the nuclear reactor and used again.

  In addition, in order to obtain a fuel with higher decontamination, the liquid metal 2 in which plutonium oxide and minor actinoid were recovered was brought into contact with a molten salt, and bismuth chloride was introduced. It is also possible to extract and decontaminate the water.

  Such a method for recovering plutonium oxide has the following advantages as compared with 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 chemical reaction between the liquid metal and the plutony ion was performed in a short time of about 30 minutes. As the reaction proceeds, the operation time of the process can be greatly reduced.

In addition, in the precipitation method, the particle size of the deposited PuO ₂ is small, so that not only the recovery device becomes complicated, but also because the operation is performed in a high-temperature, high-humidity atmosphere, the life of the device is short and the reliability is poor. However, in the method of the embodiment, the cathode of the liquid metal 2 is pulled up together with the ceramic container 1, whereby the liquid metal 2 can be easily recovered, no special recovery device is required, and the entire device can be simplified. Furthermore, conventionally, it was difficult to recover plutonium oxide having a small particle diameter at a high rate. As a result, the recovery rate is extremely high at almost 100%.

  Furthermore, in the precipitation method, the minor actinoids could not be collectively recovered. However, according to the embodiment, in addition to the plutonium oxide, the minor actinoids were simultaneously recovered by using electrolysis together, and used as fuel. be able to. Since these are ultra-half-life nuclides, they can be annihilated again in the core, which is very advantageous in disposing of radioactive waste.

Next, a method of separating the precipitated and recovered UO ₂ from the noble metal element will be described.

When the spent fuel is reprocessed by molten salt electrolysis to precipitate UO ₂ , as shown in FIG. 4, a noble metal element (molybdenum (Mo), which has the same oxidation-reduction potential as uranyl ion (UO ₂ ²⁺ ), Ruthenium (Ru), zirconium (Ζr), nickel (Νi), silver (Ag)) are precipitated and the separation from UO ₂ is deteriorated, but both can be efficiently separated as shown below. .

Although UO ₂ is an oxide, as shown in FIG. 5, it has electrical conductivity (conductivity) at a high temperature, and is subjected to electrolytic purification 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.. Then, even if the recovered UO ₂ is subsequently electrolyzed in a molten salt at 500 ° C. or lower, almost no current flows through the UO ₂ because the conductivity is significantly reduced. However, since the noble metal element precipitated together with UO ₂ is a metal, a current flows even if the temperature is kept at 500 ° C. or less, 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.

As a specific example, UO ₂ electrolytically deposited on a graphite cathode in 650 ° C. molten salt (NaCl—KCl) and a noble metal element (Mo, Ru, Rh, Pd, Ag) are kept at 500 ° C. put each electrode, in LiCl-KCl molten salt was carried out electrolytic held in -0.3V against chlorine reference electrode, UO ₂ is hardly eluted, other metals eluted. After recovering the eluted noble metal element, the temperature was raised to 600 ° C. and electrolysis was performed, whereby UO ₂ could be recovered.

Further, UO ₂ and a noble metal element electrolytically deposited on a graphite cathode in a molten salt (NaCl-KCl) at 650 ° C. are once separated from the electrode after cooling, coarsely pulverized, and then granulated. Was mixed with graphite as a conductive material, and used as a positive electrode in a molten salt. When eluting while maintaining the same potential and temperature as in the above specific example, the noble metal element is easily eluted in the molten salt because the current flows more easily by the action of conductive graphite, and only the noble metal is recovered on the cathode. I was able to. After the completion of this operation, the temperature was raised to 600 ° C., and electrolysis was carried out, whereby UO ₂ was deposited on the cathode and could be recovered.

  ADVANTAGE OF THE INVENTION According to the molten salt electrolysis reprocessing method of the used fuel of this invention, contamination of impurities, such as a noble metal element, in the collect | recovered fuel can be reduced, and a higher quality fuel can be obtained. Further, the dissolving of the spent fuel on the anode side and the precipitation of the nuclear fuel material on the cathode side can be performed efficiently at the same time, and the processing efficiency of the spent fuel can be improved. Furthermore, minor actinoids such as americium and cuurium, as well as slightly remaining uranium and plutonium, can be recovered, and these ultra-half-life nuclides can be extinguished again in the core. Furthermore, the processing can be performed in an extremely short time of about 30 minutes as compared with the processing time of the conventional precipitation method, and the operation time of the process is greatly reduced.

4 is a flowchart illustrating a method for recovering plutonium from spent fuel, which is an embodiment of the present invention. It is a graph which shows the concentration change in the molten salt of uranium, plutonium, and a minor actinoid. It is a sectional view showing a behavior of stirring PuO ₂ produced in Example. FIG. 3 is a diagram showing oxidation-reduction potentials of typical nuclides according to the example of the present invention. Is a graph showing the temperature dependence of the conductivity of UO _2.

Explanation of reference numerals

  DESCRIPTION OF SYMBOLS 1 ... Ceramic container, 2 ... Liquid metal, 3 ... Stirrer, 4 ... Baffle plate, 6 ... Molten salt.

Claims (2)

After the spent fuel is sheared and uncoated, electrolysis is performed in a molten salt, and the anode is dissolved at the same time.
The process of dissolving uranium oxide in the fuel at the anode and depositing and recovering it 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. And then subjecting the molten salt to a liquid metal, exchanging plutonium ions with the liquid metal to precipitate and recover plutonium oxide, and the molten salt electrolytic reprocessing method for a spent fuel. .   The method according to claim 2, wherein after the plutonium oxide is recovered, minor actinide ions remaining in the molten salt are recovered by electrolysis using a liquid metal as a cathode.

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