JP5193687B2 - Spent fuel reprocessing method - Google Patents

Description

  The present invention relates to a spent fuel reprocessing method including a step of recovering uranium (U), plutonium (Pu), and minor actinides (MA) from spent oxide nuclear fuel.

  This is a typical technology for reprocessing spent fuel generated from nuclear power plants, refining and recovering useful substances contained in spent fuel, separating unnecessary fission products, and reusing them as fuel. There is a Purex method as a process. Spent fuel contains alkali metal (AM) element, alkaline earth metal (AEM) element and platinum group element as fission product (FP) in addition to transuranium element (TRU) such as uranium and plutonium. .

The purex method is adopted at the reprocessing plant of Japan Nuclear Fuel Co., Ltd. in Rokkasho. That is, after the spent fuel is dissolved in the nitric acid solution, the fission product is separated in the co-decontamination step, U and Pu are separated in the U and Pu distribution step, and U and Pu are respectively U purification step, After purification in the Pu purification step, the Pu solution is mixed with the U solution and mixed and denitrated, whereby Pu cannot be recovered alone.
Japanese Patent No. 2880919 Japanese Patent No. 3319657

  In the conventional Purex method, since U and Pu are once separated in the distribution step, it cannot be said that there is absolute nuclear non-proliferation.

  Therefore, a part of the process of the PUREX method is changed, and a reprocessing process with high nuclear non-proliferation property, that is, cannot recover Pu alone is desired.

  By the way, the high-level waste liquid of the Purex method contains a small amount of U and Pu and most of the minor actinides (Np, Am, Cm, etc.). As a process for collectively collecting these transuranium elements (Pu, minor actinides), there is an aqua pyro method in which oxalic acid precipitation-chloride conversion-molten salt electrolysis is applied to a high-level waste liquid (Patent Documents 1 and 2). In the Aqua Pyro method, Pu accompanies U and minor actinides and is collected in a batch. That is, Pu is not recovered alone.

  The present invention has been made in view of the problems of the background art, so that most of uranium can be separated from spent fuel solution and recovered as light water reactor fuel, while Pu and minor actinides are recovered together with U. Thus, an object of the present invention is to provide a method for reprocessing spent fuel that has high nuclear non-proliferation and can be used for metal fuel in fast reactors.

  In order to achieve the above object, one aspect of a spent fuel reprocessing method according to the present invention comprises a dismantling / shearing step of disassembling and shearing spent oxide nuclear fuel, and a fuel that has undergone the dismantling / shearing step. A dissolving step for dissolving in a nitric acid solution, an electrolytic valence adjusting step for reducing plutonium to trivalent while maintaining neptunium to be pentavalent with respect to the fuel that has undergone the dissolving step, and a fuel that has undergone the electrolytic valence adjusting step Is contacted with an organic solvent, and hexavalent uranium is extracted into the extractant to extract uranium oxide, and the minor actinides and fission products remaining in the nitric acid solution in the uranium extraction step are precipitated with oxalic acid. Oxalic acid precipitation step that precipitates together as an oxalic acid precipitate by the method, and a chlorination step that converts the oxalic acid precipitate to hydrochloric acid by adding hydrochloric acid to the oxalic acid precipitate A dehydration step of synthesizing anhydrous chloride by dehydrating the chloride in a reducing inert gas stream; and dissolving the anhydrous chloride in a molten salt and electrolyzing uranium, plutonium and And a molten salt electrolysis step for recovering the minor actinide.

  Further, another aspect of the spent fuel reprocessing method according to the present invention includes a dismantling / shearing step of disassembling and shearing spent oxide nuclear fuel, and dissolving the fuel that has undergone the dismantling / shearing step in a nitric acid solution. A dissolving step, an electrolytic valence adjusting step of reducing plutonium to trivalent and reducing neptunium to pentavalent with respect to the fuel that has undergone the dissolving step, and an organic solvent for the fuel that has undergone the electrolytic valence adjusting step The uranium extraction process for recovering uranium oxide by contacting and extracting hexavalent uranium into the extractant, and the minor actinides and fission products remaining in the nitric acid solution in the uranium extraction process together with the oxalic acid precipitation method. An oxalic acid precipitation step for precipitation as an acid precipitate, an oxidation / dehydration step for dehydrating the oxalic acid precipitate and then converting it to an oxide precipitate in an oxidizing atmosphere, The precipitate in a mixed molten salt in which an alkali metal oxide is dissolved in a metal chloride molten salt or in a mixed molten salt in which an alkaline earth metal oxide is dissolved in a chloride molten salt of an alkaline earth metal An oxide is immersed, the precipitate oxide is brought into contact with the cathode, oxygen ions in the precipitate oxide are extracted, and removed as oxygen gas or carbon dioxide gas on the anode side in the molten salt, and the cathode And an electrolytic reduction process for recovering uranium, plutonium and minor actinides in the precipitate oxide.

  According to the present invention, most uranium can be separated from spent fuel solution and recovered as light water reactor fuel, and Pu and minor actinides can be recovered together with U to be used as metal fuel for fast reactors. it can. Since Pu cannot be recovered alone and Pu and minor actinides are recovered together with U, nuclear non-proliferation is high.

  Embodiments of a spent fuel reprocessing method according to the present invention will be described below with reference to the drawings.

[First Embodiment]
First, a first embodiment of a spent fuel reprocessing method according to the present invention will be described with reference to FIGS. 1 and 2.

  FIG. 1 is a flowchart showing a first embodiment of a spent fuel reprocessing method according to the present invention. In FIG. 1, first, in a dismantling / shearing step 2, the spent oxide fuel 1 is disassembled and sheared. Thereafter, in dissolution step 3, the entire amount is dissolved with nitric acid. At this time, U exists in a hexavalent state and Pu exists in a tetravalent state.

  Next, in electrolytic valence adjustment step 4, Pu is electrolytically reduced to trivalent. FIG. 2 is a schematic vertical sectional view showing an example of an apparatus used in the electrolytic valence adjusting step 4 in the first embodiment. That is, in this apparatus, the cathode chamber 27 and the anode chamber 28 are separated via the diaphragm 50. A catholyte 24 is stored in the cathode chamber 27, and a cathode 25 and a reference electrode 30 are inserted into the catholyte 24. An anolyte 51 is stored in the anodic chamber 28, and an anode 26 is inserted into the anolyte 28. The cathode 25 and the anode 26 are connected to a power source 29. Further, the cathode 25 and the reference electrode 30 are connected to a potentiometer 31. As the reference electrode 30, for example, a silver / silver chloride electrode is used. The cathode chamber 27 is provided with a stirrer 52 for stirring the catholyte 24.

At this time, by reducing the cathode potential to −100 mV or less, or by setting the cathode current density in the range of 20 mA / cm ² to 40 mA / cm ² , Pu is reduced to trivalent while Np is maintained at pentavalent. be able to. The partially tetravalent reduced U is also used to reduce Pu from tetravalent to trivalent, whereas U itself is oxidized to hexavalent.

FIG. 3 is a graph of experimental results showing the correlation between the cathode potential and the current density in the electrolytic valence adjusting step 4. Experiments have shown that the cathode potential is -0.1 V (-100 mV) by setting it to about 20 mA / cm ² or more.

  Since U is mostly hexavalent, when it is extracted with TBP (tributyl phosphate) -30% dodecane in U extraction step 5, only the hexavalent U is extracted into a TBP-30% dodecane solution. The trivalent ion of Pu and the pentavalent ion of Np remain in the aqueous solution together with some U tetravalent ions.

FIG. 4 is a graph showing an example of measurement results of changes in current density over time when the electrolytic potential is held at −100 mV based on a silver / silver chloride electrode standard as a reference electrode in the electrolytic valence adjustment step 4 and the U extraction step 5. It is. At this time, it is shown that the cathode current density is in the range of 20 mA / cm ² to 40 mA / cm ² with respect to the cathode potential of −100 mV.

  Next, oxalic acid is added to the aqueous solution remaining in the U extraction process 5 in the oxalic acid precipitation process 6 to produce an oxalic acid precipitation 7. The oxalic acid precipitate 7 contains Pu and minor actinides such as Np, Am, and Cm, rare earth elements (RE), and a part of alkaline earth metal elements. Among the fission products, alkali metal elements and platinum group elements are dissolved in the filtrate without being precipitated.

  In the oxalic acid precipitation step 6, U, Pu, minor actinides, rare earth elements and the like are recovered as oxalic acid precipitation 7.

  In the chlorination step 8, hydrochloric acid is added to the oxalic acid precipitate 7 and dissolved at 100 ° C. or lower, and then hydrogen peroxide is added to decompose the oxalic acid into water and carbon dioxide. U, Pu and minor actinides in oxalic acid precipitate 7 are converted to chloride 9 in this chlorination step 8.

  Next, in the dehydration step 40, the water in the hydrochloric acid solution is removed by evaporation, and then the water is completely removed at about 200 ° C. in a stream of a reducing inert gas (for example, argon or nitrogen). This produces anhydrous U, Pu and minor actinide chlorides (anhydrous chlorides) 41.

  By electrolyzing the produced anhydrous chloride 41 in the molten salt electrolysis step 10, U, Pu, and minor actinide metals that can be used as fast reactor fuel can be collectively recovered.

  Next, a platinum group fission product recovery step 14 for recovering a platinum group fission product from the oxalic acid precipitation 7 obtained in the oxalic acid precipitation step 6 will be described with reference to FIGS. 1 and 2. Here, the structure of the device used in the platinum group fission product recovery step 14 may be the same as the device shown in FIG. 2 used in the electrolytic valence adjustment step and the U extraction step. For example, the same device may be used, or another device of the same or similar structure may be used.

  The oxalic acid precipitate 7 contains Pu, minor actinides such as Np, Am, and Cm, rare earth elements, and part of alkaline earth metal elements. Among the fission products, alkali metal elements and platinum group elements do not precipitate oxalic acid and are dissolved in the filtrate (catholyte) 24. In the platinum group fission product recovery step 14, the filtrate 24 in which the fission product is dissolved is placed in a cathode chamber 27, and an insoluble cathode 25 is immersed therein to perform electrolysis.

  When a voltage is applied from the power source 29 to the anode 26 and the cathode 25, among the fission products contained in the filtrate 24 of the cathode chamber 27, platinum group fission products Pd, Ru, Rh, Mo picture and Tc Is deposited and collected on the cathode 25. On the other hand, an acid anolyte 51 is placed in the anode chamber 28. At this time, alkali metal elements such as Cs and alkaline earth elements such as Sr in the filtrate which is the catholyte 24 remain in the filtrate and can be separated from the platinum group element fission products.

  The applied voltage is determined by measuring the potential difference between the reference electrode 30 immersed in the cathode chamber 27 and the cathode 25 with a potentiometer 31, and the platinum group fission products Pd, Ru, Rh, Mo, and Tc are not generated with hydrogen. It is important to control the potential at 25.

  Since platinum group fission products Pd, Ru, Rh, Mo and Tc do not migrate into the high-level waste, the burden on the production of the vitrified body can be reduced. Furthermore, the amount of high-level waste generated can be reduced.

The hexavalent U extracted with TBP-30% dodecane in the U extraction step 5 is washed with nitric acid in the U purification step 11 and then converted into an oxide in the denitration step 12 to obtain high purity UO ₂ 13. As recovered. High purity UO ₂ 13 can be used as an oxide fuel for light water reactors.

[Second Embodiment]
Next, a second embodiment of the spent fuel reprocessing method according to the present invention will be described with reference to FIGS. Here, the same or similar parts as those in the first embodiment are denoted by the same reference numerals, and redundant description is omitted.

  FIG. 5 is a flowchart showing a second embodiment of the spent fuel reprocessing method according to the present invention. FIG. 6 is a schematic sectional elevation view showing an example of an apparatus used in the electrolytic reduction process in the second embodiment.

  The procedures until the oxalic acid precipitation 7 of U, Pu, minor actinides, and rare earth elements is recovered in the oxalic acid precipitation step 6 are the same as those in the first embodiment.

  In this second embodiment, in order to obtain metals U, Pu and minor actinides, instead of the chlorination step 8, the dehydration step 40 and the molten salt electrolysis step 10 of the first embodiment, an oxidation / dehydration step 15 and It has an electrolytic reduction step 17.

  That is, in the oxidation / dehydration step 15, when moisture is removed while heating ozone or an oxidizing gas into the oxalic acid precipitate 7 collected in the oxalic acid precipitation step 6, U, Pu, minor actinides and Oxides (precipitate oxides) 16 of rare earth elements are generated.

  Next, the moisture of the oxide 16 is completely removed while pulling oxygen to a vacuum. Thereafter, the oxide 16 is put into a cathode basket 19 made of stainless steel and loaded into a molten salt electrolysis tank 22. The cathode basket containing the U, Pu, minor actinide and rare earth element oxide 16 is connected to the cathode of the power source 23, and an insoluble anode 20 made of platinum or glassy carbon is installed. A voltage is applied to the cathode basket 19 and the anode 20 in the molten salt 21, and oxygen ions in the U, Pu and minor actinide oxides in the cathode basket 19 are extracted and reduced to metal, so U, Pu and minor The actinide metal 18 can be recovered.

Oxide 16 is placed in a cathode basket 19 made of stainless steel in a mixed molten salt. The mixed molten salt is preferably one in which an alkali metal or alkaline earth metal oxide is dissolved in a molten salt of an alkali metal or alkaline earth metal chloride. More specifically, for example, a mixed molten salt obtained by dissolving Li ₂ O in a molten salt of LiCl, a mixed molten salt obtained by dissolving MgO in a molten salt of MgCl ₂ , and CaO dissolved in a molten salt of CaCl ₂ . Any of the mixed molten salts is preferred.

After putting the oxide 16 into the cathode basket 19 in the mixed molten salt, the oxygen ions in the oxide 16 are extracted, and the oxygen ions are removed as oxygen gas or CO ₂ gas at the anode. Since alkali metal elements such as Cs and alkaline earth metal elements such as Sr, which are fission products, and rare earth elements such as Ce and Nd are dissolved in the molten salt from the cathode basket 19, U, Pu and minor actinide metals. 18 and can be separated.

  At this time, reduction to the metal represented by the following formula occurs at the cathode.

UO ₂ + 4e− → U + 2O ²⁻
PuO ₂ + 4e− → Pu + 2O ²⁻
Further, oxygen gas is generated at the anode as represented by the following formula.

2O ²⁻ → O ₂ + 4e−

The flowchart which shows 1st Embodiment of the spent fuel reprocessing method which concerns on this invention. The typical sectional view showing the example of the device used in the electrolytic valence adjustment process in the 1st embodiment of the spent fuel reprocessing method concerning the present invention, and the platinum group fission product recovery process. The graph which shows the example of a measurement result of the electrode potential and the initial value of a current density in the electrolytic valence adjustment process of 1st Embodiment of the spent fuel reprocessing method which concerns on this invention. In the electrolytic valency adjustment step of the first embodiment of the spent fuel reprocessing method according to the present invention, the change in current density with time when the electrolytic potential is held at −100 mV based on a silver / silver chloride electrode standard as a reference electrode. The graph which shows the example of a measurement result. The flowchart which shows 2nd Embodiment of the spent fuel reprocessing method which concerns on this invention. The typical sectional view showing the example of the device used at the electrolytic reduction process in the 2nd embodiment of the spent fuel reprocessing method concerning the present invention.

Explanation of symbols

1: spent oxide fuel 2: dismantling / shearing process 3: dissolution process 4: electrolytic valence adjustment process 5: U extraction process 6: oxalic acid precipitation process 7: oxalic acid precipitation 8: chlorination process 9: chloride 10 : Molten salt electrolysis step 11: U purification step 12: Denitration step 13: High purity UO ₂
14: Platinum group fission product recovery step 15: Oxidation / dehydration step 16: Oxide (precipitate oxide)
17: Electrolytic reduction process 18: U, Pu and minor actinide metal 19: Cathode basket 20, 26: Anode 21: Molten salt 22: Molten salt electrolytic cell 23, 29: Power supply 24: Catholyte (filtrate)
25: cathode 27: cathode chamber 28: anode chamber 30: reference electrode 31: potentiometer 40: dehydration step 41: anhydrous chloride 50: diaphragm 51: anolyte

Claims (7)

Dismantling and shearing process to disassemble and shear spent oxide nuclear fuel;
A dissolution step of dissolving the fuel that has undergone the dismantling / shearing step in a nitric acid solution;
An electrolytic valence adjusting step for reducing plutonium to trivalent while maintaining neptunium at pentavalent for the fuel that has undergone the dissolution step;
Contacting the fuel that has undergone the electrolytic valence adjustment step with an organic solvent, and extracting hexavalent uranium into the extractant to extract uranium oxide; and
An oxalic acid precipitation step in which minor actinides and fission products remaining in the nitric acid solution in the uranium extraction step are precipitated together as an oxalic acid precipitate by an oxalic acid precipitation method;
A chlorination step for converting to a chloride by adding hydrochloric acid to the oxalic acid precipitate;
A dehydration step of synthesizing the anhydrous chloride by dehydrating the chloride in a reducing inert gas stream;
A molten salt electrolysis step of dissolving the anhydrous chloride in the molten salt and recovering uranium, plutonium and minor actinides at the cathode by electrolysis;
A spent fuel reprocessing method comprising: Dismantling and shearing process to disassemble and shear spent oxide nuclear fuel;
A dissolution step of dissolving the fuel that has undergone the dismantling / shearing step in a nitric acid solution;
An electrolytic valence adjusting step of reducing plutonium to trivalent and reducing neptunium to pentavalent for the fuel that has undergone the melting step
Contacting the fuel that has undergone the electrolytic valence adjustment step with an organic solvent, and extracting hexavalent uranium into the extractant to extract uranium oxide; and
An oxalic acid precipitation step in which minor actinides and fission products remaining in the nitric acid solution in the uranium extraction step are precipitated together as an oxalic acid precipitate by an oxalic acid precipitation method;
An oxidation / dehydration step in which the oxalic acid precipitate is dehydrated and then converted into a precipitate oxide in an oxidizing atmosphere;
In the mixed molten salt obtained by dissolving an alkali metal oxide in an alkali metal chloride molten salt, or in the mixed molten salt obtained by dissolving an alkaline earth metal oxide in an alkaline earth metal chloride molten salt. The product oxide is immersed, the precipitate oxide is brought into contact with the cathode to extract oxygen ions in the precipitate oxide, and removed as oxygen gas or carbon dioxide gas on the anode side in the molten salt, An electrolytic reduction step of recovering uranium, plutonium and minor actinides in the precipitate oxide at the cathode;
A spent fuel reprocessing method comprising:   The electrolytic reduction step is performed by accommodating the precipitate oxide in a cathode basket made of stainless steel, immersing the cathode basket in the molten salt, and connecting the cathode to the cathode basket. The spent fuel reprocessing method according to claim 2. The mixed molten salt is a mixed molten salt obtained by dissolving Li ₂ O in a molten salt of LiCl, a mixed molten salt obtained by dissolving MgO in a molten salt of MgCl ₂ , or a mixed molten salt obtained by dissolving CaO in a molten salt of CaCl _2. The spent fuel reprocessing method according to claim 2, wherein the spent fuel is reprocessed.   The filtrate remaining without being precipitated in the oxalic acid precipitation step is put into a cathode chamber, a cathode made of an insoluble material is inserted into the cathode chamber, and an acidic solution is put into an anode chamber separated from the cathode chamber by a partition wall. 5. The method further comprises a fission product recovery step of depositing and recovering platinum group fission products remaining in the filtrate by depositing and collecting on the cathode. The spent fuel reprocessing method as described in the paragraph.   The spent fuel reprocessing method according to any one of claims 1 to 5, wherein the electrolytic valence adjustment step is performed at -100 mV or less based on a silver / silver chloride electrode standard as a reference electrode. 7. The electrolytic valence adjustment step is performed at a cathode current density of 20 mA / cm ² to 40 mA / cm ^{2 based} on a silver / silver chloride electrode standard as a reference electrode. The spent fuel reprocessing method as described.

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