JP5784476B2 - Uranium recovery method - Google Patents

Description

  The present invention relates to a method for precipitating and recovering uranium contained in an aqueous solution.

Spent nuclear fuel generated at nuclear power plants contains alkali metal, alkaline earth metal, rare earth metal, etc. as fission product FP in addition to uranium U, plutonium Pu, minor actinide MA, and reprocessing process After that, it is reused as fuel.
There are two reprocessing methods: a wet method such as the conventional Purex method, and a dry method using a molten salt electrolysis method in which spent nuclear fuel is processed as it is at a high temperature.

Then, a hybrid process that separates most of the uranium from the spent fuel solution by a wet method and collects high-purity uranium oxide, and collects Pu and MA (Np, Am, Cm, etc.) in a batch by a dry method. A processing method has been developed (for example, Patent Documents 1-4).
According to this hybrid reprocessing method, it is possible to realize a reprocessing process with high nuclear non-proliferation, that is, Pu is not recovered alone.

Japanese Patent No. 2880919 Japanese Patent No. 3319657 Japanese Patent No. 3100022 JP 2002-236195 A

By the way, in the hybrid reprocessing method, a step of adding oxalic acid to the uranium solution and precipitating and recovering uranium is adopted, and uranium is adjusted from hexavalent to tetravalent in order to improve the recovery rate of this uranium. A valence adjustment step is provided.
However, since this tetravalent uranium is oxidized with time in the nitric acid solution and returns to hexavalent uranium, there is a problem that the recovery rate does not improve as expected.

  The present invention has been made in view of such circumstances, and an object of the present invention is to provide a uranium recovery method in which an improvement in recovery rate is achieved.

In the method for recovering uranium , a dissolving step in which spent fuel is dissolved in nitric acid to form an aqueous solution, and a first electrolytic reduction valence adjustment for reducing the uranium valence by electrolytic reduction of the aqueous solution after passing through the dissolving step A separation step of mixing an organic extraction solvent with the aqueous solution after the first electrolytic reduction valence adjustment step and separating the mixture into an organic layer and an aqueous layer, and the water after the separation step An oxidation inhibitor is introduced into the uranium solution of the layer to suppress an increase in the valence of uranium, and a valence is obtained by electrolytic reduction of the uranium solution in the water tank after the oxidation inhibitor is introduced. A second electrolytic reduction valence adjusting step for reducing the uranium valence increased, and an oxalic acid to be precipitated by adding oxalic acid to the uranium solution in the water tank after the second electrolytic reduction valence adjusting step An acid precipitation step, and That.

  The present invention provides a method for recovering uranium in which an improvement in recovery rate is achieved.

FIG. 3 is a process diagram showing a first embodiment of a uranium recovery method according to the present invention. Process drawing which shows 2nd Embodiment of the collection | recovery method of uranium based on this invention.

(First embodiment)
Hereinafter, embodiments of the present invention will be described with reference to the accompanying drawings.
The uranium recovery method according to the first embodiment includes a dissolution step (S12) in which spent fuel is dissolved in nitric acid to form an aqueous solution, and electrolytic reduction valence adjustment for reducing the uranium valence by electrolytic reduction of the aqueous solution. A step (S14), a liquid separation step (S15) in which an organic extraction solvent is mixed with the electrolytically reduced aqueous solution and separated into an organic layer and an aqueous layer, and an oxidation inhibitor is added to the separated aqueous layer to add uranium. The oxidation inhibitor addition process (S21) which suppresses the increase in the valence of, and the oxalic acid precipitation process (S23) which precipitates uranium by introducing oxalic acid into the water layer into which the oxidation inhibitor is introduced.

  Moreover, the electrolytic reduction valence adjustment process (S22) which decreases the valence of the uranium in which the valence increased is included after the oxidation inhibitor injection process (S21) which injected the oxidation inhibitor into the water layer.

The dismantling / shearing step (S11) is a step of disassembling and shearing spent nuclear fuel that has been taken out of the reactor, stored and cooled in the storage pool until it reaches the specified radioactivity level, into small pieces of several centimeters is there.
In addition to uranium U, plutonium Pu, and minor actinides MA (Np, Am, Cm, etc.), spent fuel includes alkali metal (AM) elements, alkaline earth metal (AEM) elements as fission products (FP), Contains platinum group elements. Note that the FP gas released from the fuel rod at the time of cutting is subjected to waste gas treatment.

The dissolution step (S12) is a step in which the spent nuclear fuel that has become a cut piece is placed in a stainless dissolution tank and dissolved in an aqueous nitric acid solution to elute U, Pu, FP, and MA into the aqueous nitric acid solution. This eluted U exists in the liquid phase in a hexavalent (U ⁺⁶ ) state and Pu in a tetravalent (Pu ⁺⁴ ) state.
The waste treatment / disposal process (S13) treats and disposes the fuel cladding tube made of Zircaloy alloy or stainless steel, which is insoluble in nitric acid aqueous solution, and other fuel assembly structure as solid radioactive waste. It is a process to do.

The electrolytic reduction valence adjustment step (S14) is a step in which an electrode is put into a nitric acid aqueous solution from which U, Pu, FP, and MA are eluted and a DC voltage is applied to electrolytically reduce these metal ions.
By this electrolytic reduction, each metal ion exchanges electrons, and the valence is adjusted such that a part of U ⁺⁶ is U ⁺⁴ and most of Pu ⁺⁴ is Pu ⁺³ .
In addition, by adjusting the production rate of U ⁺⁴ , it is possible to adjust the U composition in the aqueous layer in the next liquid separation step (S15).

The liquid separation step (S15) is a step of mixing and separating TBP (tributyl phosphate) / n-dodecane as an organic extraction solvent. Metal ions are distributed into the organic extraction solvent and water based on their respective partition coefficients. Almost all U ⁺⁶ and a predetermined proportion of U ⁺⁴ are extracted into the TBP / n-dodecane solution, and the remainder of U ⁺⁴ , Pu ⁺³ , MA, and FP remain in the aqueous layer.
This aqueous layer is a 1-5 M nitric acid aqueous solution containing 0.1-3 M (mol / L) uranium.

In the U purification step (S16), U ⁺⁶ and U ⁺⁴ extracted in the organic layer in the liquid separation step (S15) are back-extracted into nitric acid by washing with nitric acid. The back-extracted U ⁺⁶ and U ⁺⁴ are converted into oxides in the denitration step (S17), recovered as high-purity UO _{2 and} used as fuel for light water reactors (S18).

In the oxidation inhibitor charging step (S21), since U ⁺⁶ is more stable than U ^{+4 in the} aqueous nitric acid solution, an oxidation inhibitor that prevents the reaction of U ⁺⁴ oxidizing and returning to U ⁺⁶ is added. This is a step of charging the water layer. As this oxidation inhibitor, hydrazine can be suitably used, but any hydrazine having an action of inhibiting U ⁺⁴ oxidation can be used as appropriate.

The amount of hydrazine introduced into the aqueous layer is such that the molar ratio to uranium contained in this aqueous layer is 1 or more. If the molar ratio of hydrazine to uranium in the aqueous layer is less than 1, the oxidation inhibiting action of U ⁺⁴ is insufficient and the ratio of returning to U ⁺⁶ increases.
In the electrolytic reduction valence adjustment step (S22), what has been oxidized in the aqueous layer and changed to U ⁺⁶ in spite of passing through the oxidation inhibitor addition step (S21) is electrolytically reduced to U ⁺⁴ It is the process of returning to.

  In the oxalic acid precipitation step (S23), oxalic acid is added to the aqueous layer, and metal ions contained in the aqueous layer are precipitated as oxalate. The amount of oxalic acid added is such that the molar ratio to uranium contained in the aqueous layer is 2 or more. If the molar ratio of oxalic acid to uranium in this aqueous layer is less than 2, the rate at which metal ions are converted to oxalate is reduced, and the rate of precipitate recovery is reduced. And the collection rate of a precipitate is improved by stirring the water layer which added the oxalic acid to normal temperature or heating, and standing still at normal temperature or cooling.

In this oxalic acid precipitation step (S23), U ⁺⁴ and Pu ⁺³ generate and precipitate oxalate at a high rate. In addition, some of fission products (FP) such as minor actinides (MA) such as Np, Am, and Cm, rare earth elements (RE), and alkaline earth metal elements are contained in the oxalate precipitate.
In the fission product (FP), alkali metal elements and platinum group elements are dissolved in the filtrate without being precipitated.

  In the oxide conversion step (S24), ozone or oxidizing gas is blown into the oxalate precipitate recovered in the oxalic acid precipitation step (S23), and moisture is removed while heating. U, Pu, MA And a step of producing an oxide of a rare earth element.

  The molten salt electrolysis step (S25) is a step in which moisture is completely removed from the oxide converted in S24, held in a cathode basket, and subjected to molten salt electrolysis using an anode made of platinum or glassy carbon. Then, oxygen ions in the oxides of U, Pu, and MA in the cathode basket are extracted and reduced to metals, so that the metals of U, Pu, and MA are recovered and used as fuel for the fast reactor (S26). ).

The platinum group FP recovery step (S31) is a step of recovering platinum group fission products (FP) from the filtrate after oxalate is precipitated in the oxalic acid precipitation step (S23). When a voltage is applied by inserting an anode and a cathode into the filtrate, platinum group FP (Pd, Ru, Rh, Mo, Tc) is deposited and collected on the cathode.
At this time, alkali metal elements such as Cs and alkaline earth elements such as Sr remaining in the filtrate are vitrified as high-level radioactive waste (S32).

As a comparative example, when the oxidation inhibitor charging step (S21) and the electrolytic reduction valence adjustment step (S22) are omitted from the steps shown in FIG. 1, the uranium element precipitates in the oxalic acid precipitation step (S23). The amount that shifts to is about 80%, and the rest shifts to the filtrate.
For this reason, in a comparative example, the process of collect | recovering uranium from a filtrate and the melt | dissolution process of melt | dissolving the collect | recovered uranium are needed, The process will become complicated and the problem which radioactive waste will increase will generate | occur | produce.

(Second Embodiment)
In the above description, focusing on the filtrate after solvent extraction with TBP in the liquid separation step (S15), the embodiment prevents U ^{+4 in} this uranium solution from oxidizing and returning to U ^+6. showed that.
However, the uranium solution to which the embodiments of the invention are applied is not limited to the filtrate separated by such TBP.

  That is, as shown in FIG. 2, an oxidation inhibitor charging step (S21) in which an oxidation inhibitor (for example, hydrazine) is added to the adjusted uranium solution (S20) to suppress an increase in the valence of uranium, The present invention can be applied to a technique including an oxalic acid precipitation step (S23) in which oxalic acid is added to a uranium solution charged with an oxidation inhibitor to precipitate the uranium, and a uranium recovery step (S24). It should be noted that an electrolytic reduction valence adjusting step (S22) for decreasing the valence of uranium having an increased valence may be provided after the oxidation inhibitor adding step (S21) as appropriate.

According to the uranium recovery method of at least one embodiment described above, by adding a U ⁺⁴ oxidation inhibitor to the uranium solution, it is prevented from being oxidized to U ⁺⁶ and uranium precipitates due to oxalic acid. It is possible to improve the recovery rate. Further, the process can be simplified and radioactive waste can be reduced.

  Although several embodiments of the present invention have been described, these embodiments are presented by way of example and are not intended to limit the scope of the invention. These embodiments can be implemented in various other forms, and various omissions, replacements, changes, and combinations can be made without departing from the scope of the invention. These embodiments and their modifications are included in the scope and gist of the invention, and are also included in the invention described in the claims and the equivalents thereof.

Claims (4)

A dissolving step of dissolving spent fuel in nitric acid to make an aqueous solution;
A first electrolytic reduction valence adjustment step of reducing the uranium valence by electrolytic reduction of the aqueous solution after the dissolution step;
A liquid separation step of mixing an organic extraction solvent into the aqueous solution after the first electrolytic reduction valence adjustment step and separating the mixture into an organic layer and an aqueous layer;
An oxidation inhibitor charging step for suppressing an increase in the valence of uranium by adding an oxidation inhibitor to the uranium solution in the aqueous layer after the liquid separation step ;
A second electrolytic reduction valence adjustment step of reducing the valence of uranium having an increased valence by electrolytic reduction of the uranium solution in the water tank after the oxidation inhibitor charging step;
An oxalic acid precipitation step of adding oxalic acid to the uranium solution in the water tank after the second electrolytic reduction valence adjustment step to precipitate the uranium. The uranium recovery method according to claim 1 ,
The method for recovering uranium, wherein the oxidation inhibitor contains hydrazine. The method for recovering uranium according to claim 2 ,
The method for recovering uranium, wherein a molar ratio of the hydrazine to uranium contained in the aqueous layer is 1 or more. In the recovery method of uranium according to any one of claims 1 to 3 ,
A method for recovering uranium, wherein a molar ratio of the oxalic acid to uranium contained in the uranium solution is 2 or more.

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