JPH0755991A - Reprocessing method for spent fuel - Google Patents

Abstract

PURPOSE:To safely and easily process radioactive waste products by reducing fission products in salt after electrolytic refining to metal compound, contacting with molten metal and extracting actinide mixture. CONSTITUTION:Spent oxide fuel after chrolination dissolution is precipitated as oxide of U and Pu from a salt mixture on a cathode by electrolysis refining and recovered as reactor fuel. As molten salt electrolytic solution after the electrolysis includes actinide, such reducer as Ca is added to this molten salt electrolytic solution to change the solved oxide ion to metal ion or metal. Then, reduction-processed molten salt electrolytic solution 24 is filled in a crucible 23 and such a molten metal 25 as Cd is contacted to this, the actinide and rare-earth are exchange-reacted with the metal 25 by stiring with an impeller 26 and the actinide is separated from the salt waste. The separation-recovered actinide is recycled in a reactor to be consumed by fission.

Description

Detailed Description of the Invention

[0001]

The present invention relates to reprocessing spent fuel generated from a nuclear power plant, refining and recovering useful metals contained in the spent fuel, and separating unnecessary fission products. The present invention relates to a method for reprocessing spent fuel by the molten salt electrolytic refining method of.

[0002]

2. Description of the Related Art Conventionally, as a technique for reprocessing spent oxide fuel generated from a nuclear power plant to purify and recover useful metals such as nuclear fuel components from the spent oxide fuel as oxides. , For example, the salt cycle process developed at the Hanford Institute is known.

This salt cycle process involves mechanically decoating spent oxide fuel and then in a suitable molten salt electrolyte.
React with gas to deoxidize spent oxide fuel
It is oxidized as in formula (1) and dissolved in the molten salt electrolyte as oxide ions. _{UO 2 + Cl 2 → UO 2} 2+ + 2Cl - ... (1)

By electrolyzing this, the substance deposited as an oxide on the cathode of uranium and plutonium can be recycled as shown in the formula (2). _{^{UO 2 2+ + 2e - → UO}} 2 PuO 2 2+ + 2e - → PuO 2 ... (2)

[0005]

However, according to this process, since the molten salt electrolyte after electrolytic refining contains actinide, the radioactive waste treatment of the molten salt electrolyte is extremely long due to its long half-life. There is a problem that it becomes difficult because of the need for dynamic monitoring. Therefore, it is desired to develop a method for separating and collecting actinide from the molten salt electrolyte.

The present invention has been made to solve the above-mentioned problems, and after electrolytic refining, the actinide contained in the molten salt electrolyte is separated and recovered from the molten salt electrolyte to treat the radioactive waste of the molten salt electrolyte. It is an object of the present invention to provide a method for reprocessing spent fuel that can facilitate the process.

[0007]

DISCLOSURE OF THE INVENTION The present invention comprises a decoating step of mechanically decoating a spent oxide fuel, a chlorination dissolution step of chlorinating and dissolving the fuel oxide in a molten salt, and this chlorine Electrolysis step of electrodepositing uranium and plutonium as oxides on the cathode from the salt mixture of the chemical dissolution step, oxide fuel production step of regenerating the electrolytic deposit of this electrorefining step as fuel, and the electrorefining step The method is characterized by further comprising a reduction step of reducing fission products in the salt to a metal compound, and an extraction step of bringing the salt mixture into contact with molten metal and extracting the actinide mixture with metal.

[0008]

In the present invention, the spent oxide fuel is chlorinated and dissolved to reduce the spent molten salt generated after electrolytic refining with a reducing agent,
Molten metal extraction and electrolytic deposition. In this way, the molten salt electrolyte after the electrolytic refining step can be deactinized to facilitate the treatment of radioactive waste.

[0009]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the spent fuel reprocessing method according to the present invention will be described with reference to FIGS. 1. FIG. 1 shows a block flow chart of the spent fuel reprocessing method of the present invention. 2 schematically shows the electrolytic cell in FIG. 1, and FIG. 3 schematically shows the extraction vessel in FIG.

In the spent oxide fuel 1 shown in FIG.
Besides depleted uranium, there are fission products such as alkali metals, alkaline earth metals, rare earths, and actinides.

The spent oxide fuel 1 is mechanically decoated in the decoating step 2, charged into a molten salt electrolyte, and reacted with a gas such as chlorine to chlorinate and dissolve 3 in the form of oxide ions. To do. The electrolytic cell used in this chlorination dissolution step is schematically shown in FIG.

Reference numeral 18 in FIG. 2 is a crucible, which serves as an anode when the circuit is closed and is used in the electrolytic refining step 4. In the chlorination dissolution step 3, electrolysis is not performed, so the circuit is performed in an open state.

A molten salt electrolyte bath 19 is filled in the crucible 18, and a gas blowing pipe 20 and a cathode 21 used in the electrolytic refining step 4 shown in FIG. 1 are inserted into the molten salt electrolyte bath 19. . Further, a power source 22 for electrolysis in the electrolytic refining step 4 is provided.

Following the decoating step 2 and the chlorination dissolution step 3,
The next step is electrorefining process 4, which is carried out electrolytically and purified as-oxide.

Electrolysis is started by closing the circuit of the electrolytic cell of FIG. 2, and uranyl (UO ₂ ) is used in the spent oxide fuel 1 dissolved in the molten salt electrolyte bath 19 in the chlorination dissolution step 3.
²⁺ ) and plutonium (PuO ₂ ²⁺ ) are electrochemically reduced on the cathode 21 and co-deposited in the form of uranium dioxide (UO ₂ ), plutonium dioxide (PuO ₂ ), other spent oxidation Fission products in solid fuel 1 are molten salt electrolyte baths
It remains dissolved in 19.

UO ₂ and PuO of this precipitate (electrodeposit) 5
₂ is transferred to the oxide fuel manufacturing process (system) 6 where the recycled fuel for the nuclear reactor is manufactured.

In the reduction step 8 which is a step following the electrolytic refining step 4, by adding a reducing agent to the spent molten salt bath 7 containing the actinide generated in the electrolytic refining step 4,
Dissolved oxide ions are metal ions or metals, or metal ions are chemically reduced to metals. For example, when Ca is used as the reducing agent, at this time, Ca is oxidized and exists in the molten salt electrolyte as CaO (Ca ²⁺ ).

Therefore, in this reduction step 8, oxide ions are converted into metal ions or metals. This C
The main reaction of the reduction step when a is used as the reducing agent is shown in the following formula. (FP oxide ion) + Ca → (FP metal ion) + CaO (3)

The next step following the reduction step 8 is an extraction step 9, in which the alkali metal and alkaline earth metal 12 in the molten salt bath 10 and the actinide and rare earth 13 in the molten metal bath 11 are extracted and separated. The extraction tank used at this time is schematically shown in FIG.

Reference numeral 23 in FIG. 3 is a crucible, and the crucible 23 is filled with the molten salt electrolyte bath treated in the reduction step 8, and the molten metal electrolyte 24 is brought into contact with the molten metal electrolyte bath 24. Use a different impeller to increase the contact area.

Taking Cd as an example of the molten metal 25, the absolute value of the free energy for chloride formation is smaller than that of Cd chloride due to the contact between the molten salt electrolyte 24 and Cd (molten metal 25). Substances that are less likely to exist as chlorides than substances, such as actinides and rare earths, are reduced in the form of exchange reaction with Cd as shown in the following formula, and are transferred from the molten salt electrolyte 24 to the Cd bath (molten metal 25) in the metallic state. , Dissolve and disperse. (FP oxide ion) · Cl ⁻ + Cd → Fp metal + Cd 2 ⁺ · Cl ⁻ (4)

Therefore, since actinide is transferred from the spent molten salt bath 7 to the molten metal bath 11 by chemical reduction, the actinide can be separated from the salt waste, and the radioactive waste treatment of the salt waste can be facilitated. Can be Here, the actinides separated and collected are recycled to the nuclear reactor and consumed by nuclear fission.

On the other hand, a substance having an absolute value of chloride formation free energy larger than that of Cd chloride, that is, a substance which is more liable to become a chloride than Cd chloride, such as an alkali metal or an alkaline earth metal, is not reduced and a molten salt is obtained. Chlorides, i.e., metal ions, remain in the bath 10, but the alkali metal and alkaline earth metal 12 and the like can be selectively removed by an electric floating method or the like, so that the salt can be regenerated and reprocessed. it can.

Therefore, the molten salt bath 10 can reduce the amount of salt waste as compared with the conventional untreated used molten salt bath 7.

Although Cd was used as an example of molten metal in the extraction step in this example, molten Mg, Cd-Li, Cd-Mg, Zn- capable of dissolving actinides.
It is also possible to use a molten metal such as Mg or Cu-Mg.

In this extraction step, electrochemical reduction can be used instead of the above-mentioned chemical reduction, and the electrolytic cell used at this time is schematically shown in FIG.

In FIG. 4, reference numeral 14 is an electrolytic cell. The electrolytic cell 14 is filled with a molten salt electrolyte 15 as a salt bath, and a Cd cathode 16 is inserted in the salt bath. Further, a power supply 17 is provided for electrolysis. Furthermore, the order of the reduction step 8 and the extraction step 9 can be exchanged.

[0028]

EFFECTS OF THE INVENTION According to the present invention, actinides are separated and recovered from the salt waste generated by electrolytic refining, so that these waste products can be treated safely and easily, and actinides, rare earths, etc. By separating the used salt, it is possible to regenerate the used salt and reduce the amount of waste.

[Brief description of drawings]

FIG. 1 is a block flow diagram showing an embodiment of a spent fuel reprocessing method according to the present invention.

FIG. 2 is a vertical sectional view schematically showing the electrolytic cell in FIG.

FIG. 3 is a vertical sectional view schematically showing the extraction tank in FIG.

FIG. 4 is a schematic cross-sectional view for explaining an application example of the spent fuel reprocessing method according to the present invention.

[Explanation of symbols]

DESCRIPTION OF SYMBOLS 1 ... Spent oxide fuel, 2 ... Decoating process, 3 ... Chlorination | dissolution process, 4 ... Electrorefining process, 5 ... Electrolytic deposit, 6 ... Oxide fuel manufacturing process, 7 ... Spent molten salt bath, 8 ... Reduction process,
9 ... Extraction step, 10 ... Molten salt bath, 11 ... Molten metal bath, 12 ... Alkali metal and alkaline earth metal, 13 ... Actinide and rare earth, 14 ... Electrolyzer, 15 ... Molten salt electrolyte, 16 ... Cd
Cathode, 17 ... Power supply, 18 ... Crucible (anode), 19 ... Molten salt electrolyte bath, 20 ... Gas blowing tube, 21 ... Cathode, 22
… Power supply, 23… Crucible, 24… Molten salt electrolyte, 25… Molten metal, 26… Impeller.

Claims (4)

[Claims] 1. A decladding step of mechanically decoating a spent oxide fuel, a chlorination dissolution step of chlorinating and dissolving the fuel oxide in a molten salt, and a salt mixture of this chlorination dissolution step. An electrolytic refining step of electrodepositing uranium and plutonium as oxides on the cathode, and an oxide fuel manufacturing step of regenerating electrolytic deposits of this electrolytic refining step as fuel,
Use comprising a reduction step of reducing fission products in the salt after the electrolytic refining step to a metal compound, and an extraction step of contacting the salt mixture with molten metal and extracting the actinide mixture with a metal. Reprocessing method of spent fuel. 2. The molten metal is Cd, Mg, Cd-L
The method for reprocessing spent fuel according to claim 1, wherein the method is selected from i, Cd-Mg, Zn-Mg, and Cu-Mg. 3. The reduction step of reducing the oxide in the salt to a metal compound is performed by chemical reduction, and the salt is chloride, fluoride or a mixed salt of chloride and fluoride. The method for reprocessing spent fuel according to claim 1, 4. The method for reprocessing spent fuel according to claim 3, wherein electrochemical reduction is used instead of the chemical reduction.