JP3910605B2 - Method for electrolytic reprocessing of spent salt from spent fuel - Google Patents

Description

  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.

  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.

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.

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.

  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.

  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.

  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.

  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.

  Embodiments of the present invention will be described below with reference to the drawings.

  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.

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.

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⁺ ).

  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.

Further, the treatment after the precipitation and recovery of plutonyl ions is performed as follows. That is, as shown in FIG. 3, the liquid metal 2 contained in the ceramic container 1 is stirred by the stirrer 3 to generate convection 5 in the direction of the wall surface to which the baffle plate 4 is attached. The PuO ₂ deposited on the surface is moved toward the inner wall of the container 1 and deposited. Then, the surface of the liquid metal 2 is always renewed, and electrolysis is performed using the liquid metal 2 as a cathode. In this electrolysis process, the potential is further lowered than that of the UO ₂ squeezing electrolysis, and the plutonium ions and minor actinide ions which are not recovered by the squeezing electrolysis and are slightly remaining are recovered on the surface of the liquid metal 2 as metals. After these components are collected at the cathode of the liquid metal 2 in this way, the ceramic container 1 is pulled up from the molten salt 6 and cooled to collect the solid metal. Then, the liquid metal 2 and the recovered metal particles of plutonium oxide and minor actinoids are separated by distillation or filtration, and the fuel component is returned to the reactor for use again.

  In order to obtain a highly decontaminated fuel, liquid metal 2 from which plutonium oxide and minor actinides are recovered is brought into contact with the molten salt and bismuth chloride is introduced, and rare earth elements that tend to become chlorides are contained in the molten salt. It is also possible to extract and decontaminate.

  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.

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 liquid metal 2 can be easily recovered by pulling up the ceramic container 1 together, and no special recovery device is required, and the entire device can be simplified. Further, conventionally, it has been difficult to recover plutonium oxide having a small particle diameter at a high rate. However, in the embodiment, the total amount of plutonium oxide generated by the reaction at the liquid metal 2 interface can be recovered in the liquid metal 2. As a result, the recovery rate is as high as almost 100%.

  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.

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

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. .

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.

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.

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.

  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.

It is a flowchart which shows the method of collect | recovering plutonium from the spent fuel which is an Example of this invention. It is a graph which shows the density | concentration change in molten salt of uranium, plutonium, and a minor actinoid. We are a sectional view showing a behavior of stirring PuO ₂ produced in Example. It is a figure which shows the oxidation-reduction potential of the typical nuclide concerning the Example of this invention. Is a graph showing the temperature dependence of the conductivity of UO _2.

Explanation of symbols

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

Claims (2)

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. . 2. The molten salt electrolysis reprocessing method for spent fuel according to claim 1 , 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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