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    • 专利摘要:
    The invention relates to a process for separating actinides (12) from a material containing the components of a spent nuclear fuel (7) by means of an ionic liquid (5), comprising the dissolution steps (S1) of the nuclear fuel used in a first solution (8) containing a complexing agent (6) and the ionic liquid (5), and for recovering actinides in an extraction solvent (9), by bringing the first solution (8) containing the spent nuclear fuel dissolved in contact with the extraction solvent (9). The invention also relates to a separation apparatus for separating the actinides (12) from a material containing the components of a spent nuclear fuel (7) by means of an ionic liquid (5).
    公开号:FR3023052A1
    申请号:FR1555803
    申请日:2015-06-24
    公开日:2016-01-01
    发明作者:Daisuke Watanabe；Yuko Kani；Akira Sasahira
    申请人:Hitachi Ltd；
    IPC主号:
    专利说明:

    [0001] The present invention relates to a method and an apparatus for separating the components of a spent nuclear fuel into uranium, plutonium, minor actinides and fission products, namely, a process for separating the components of a spent fuel from uranium, plutonium, minor actinides and fission products. actinides and a device for separating actinides.
    [0002] It is expected that nuclear fuel materials from spent nuclear fuel, discharged from nuclear power plants, will be recovered through reprocessing of the fuel, converted into a vitrified waste, and then stored in geological layers. In reprocessing with the current PUREX process, it is considered that the geological disposal of vitrified waste requires stable confinement for several hundreds of thousands of years because the vitrified waste contains minor actinides which are radionuclides with very long half-lives. lives. Actinides refer here to the elements of atomic numbers between 89 and 103, and correspond to uranium (U) and plutonium (Pu) used as nuclear fuel, and to groups of elements called minor actinides. Minor actinides refer to transuranic elements other than Pu among actinides, and correspond to neptunium, americium, curium and other elements. The following technique is being studied to reduce the confinement period of vitrified waste. It includes the steps of separation of minor actinides from highly radioactive liquid waste produced by the reprocessing of spent nuclear fuel; and conversion of minor actinides to radionuclides with short half-lives through neutron irradiation. Highly radioactive liquid waste refers here to the liquid residues obtained after the separation of uranium (U) and plutonium (Pu) from a spent nuclear fuel solution in nitric acid. Fission products and minor actinides are mainly dissolved in highly radioactive liquid waste. In short, the separation of uranium and plutonium through the reprocessing of spent nuclear fuel and the separation of minor actinides from high-level radioactive liquid waste in the minor actinide separation step allow spent nuclear fuel components to be separated into respective groups of U and Pu; minor actinides; and fission products. Various methods of separating minor actinides from highly radioactive liquid wastes have been investigated. For example, there is a method for selectively separating minor actinides from highly radioactive liquid waste by means of a complexing agent. On the other hand, materials called ionic liquids have recently attracted attention, ionic liquids being composed of ions and likely to be in a liquid state at a temperature of about room temperature. Ionic liquids can include various combinations of cations and anions, depending on the different applications. Typical cations include the imidazolium, pyridinium, pyrrolidinium, piperidinium, ammonium, phosphonium cations. Typical anions include halide ions (Cl-, Br-, I-), tetrafluoroborate (BF4-), hexafluorophosphate (PF6), bis (trifluoromethylsulfonyl) imide (C2F6NO4S2-), trifluoromethanesulfonate (CF303S- ) and trifluoroacetate (CF3C00-). The separation of elements in spent nuclear fuel using these ionic liquids is currently being studied. For example, JP-A-2002-503820 discloses a method of separating U and Pu by dissolving spent nuclear fuel or material containing spent nuclear fuel components in an ionic liquid. JP-A-2001-516871 discloses a method of regenerating a spent metal salt with an ionic liquid, the metal salt being produced during the reprocessing of spent fuel using a molten salt. When reprocessing spent nuclear fuel and handling radionuclide-containing materials, as described above, the use of ionic liquid as a solvent may advantageously facilitate operation management to prevent criticality because ionic liquids contain fewer hydrogen atoms that readily modulate neutrons and tend to contribute to criticality. Solvents such as a nitric acid solution consisting essentially of water contain more hydrogen atoms than ionic liquids. There is a chemical reaction between spent nuclear fuel and fluorine to separate uranium which is a major component of spent nuclear fuel as uranium hexafluoride gas. As examples of devices that can be used for such a reaction, flame-type fluorination devices are described in JP-A-2004-233066 and JP-A-2012-47546 and a fluorinator discontinuous type is described in JP-A-2013-101066. The present invention relates to a process for separating actinides from a material containing the components of a spent nuclear fuel by means of an ionic liquid, the process comprising the following steps of: dissolving spent nuclear fuel in a first solution containing a complexing agent and the ionic liquid; and recovering the actinides in an extraction solvent, bringing the first solution containing the spent spent nuclear fuel into contact with the extraction solvent. Advantageously, the process according to the invention comprises one or more of the following characteristics, taken alone or in combination: The process further comprises a step of: reacting the spent nuclear fuel with fluorine to give solid fluorides, dissolution step involving the dissolution of solid fluorides in the first solution; The complexing agent includes a phosphorus-oxygen double bond in its composite structure; the complexing agent includes an amide bond in its complex structure; the complexing agent includes at least one of octyl (phenyl) -N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO), tributylphosphate (TBP), trioctylphosphine oxide (TOPO), di-2-ethylhexyl-phosphoric acid (D2EHPA), N, N, N ', N'-tetrakis (2-pyridylmethyl) ethane-1,2-diamine (TPEN), 2-ethylhexyl hydrogen-2-ethylhexylphosphonate (PC -88A), N, N, N ', N'-tetraoctyl-diglycolamide (TODGA), N, N-dioctyldiglycol-amic acid (DODGAA) and dicyclohexyl-18-crown-6 (DC18C6); and the extraction solvent includes at least one of alkanes and compounds having an alkyl group. In another aspect, the invention relates to a separation apparatus for separating actinides from a material containing the components of a spent nuclear fuel by means of an ionic liquid, the apparatus comprising: a dissolving tank configured to dissolve spent nuclear fuel in a first solution containing a complexing agent and the ionic liquid; and a separation tank configured to separate actinides from the first solution by extraction. In yet another aspect, the invention relates to a separation apparatus for separating actinides from a material containing the components of spent nuclear fuel, using an ionic liquid, the apparatus comprising: a reservoir dissolver configured to dissolve the spent nuclear fuel in a first solution containing a complexing agent and the ionic liquid, the dissolving tank having a separating reservoir function configured to separate the actinides from the first solution by extraction. The invention will be better understood on reading the description which follows, given with reference to the appended figures in which: FIG. 1 is a flowchart illustrating a process for separating the actinides from a spent nuclear fuel according to one embodiment of the present invention; Figure 2 is a schematic representation of an actinide separation apparatus according to a first embodiment of the present invention; Figure 3 is a schematic representation of an actinide separation apparatus according to a second embodiment of the present invention; and Fig. 4 is a flowchart of a process for separating the actinides from spent nuclear fuel according to a fourth embodiment of the present invention. The separation of actinides from spent nuclear fuel by means of an ionic liquid requires the dissolution of spent nuclear fuel components in the ionic liquid, and then the separation of the actinides by an ordinal separation process, such as solvent extraction. Thus, effective dissolution of spent nuclear fuel components in the ionic liquid facilitates the process of separating the elements.
    [0003] Although JP-A-2002-503820 discloses the process of dissolving spent nuclear fuel or material containing the spent nuclear fuel components in the ionic liquid, it does not display data including the amount of the components of the spent fuel. spent nuclear fuel dissolved in the ionic liquid. The present inventors have experimentally measured the solubilities of a fluoride compound in ionic liquids. The solubilities of fluoride (cerium fluoride) in ionic liquids are shown in Table 1, the solubilities being obtained in this experiment. In the experiment carried out by the present inventors, the solubilities of cerium fluoride in ionic liquids are examined. These are combinations of anions and cations, the values of which are given in Table 1. Specifically, the cations of the ionic liquids are the respective cations of an imidazolium group, a pyridinium group and a pyrrolidinium group. . The column of Table 1 lists the cationic species in the order of Lewis acidity of the cations, as well as a cation of an ammonium group as a reference. The row of Table 1 lists the tetrafluoroborate ion (BF4), the hexafluorophosphate ion (PF6), the bis (trifluoromethylsulfonyl) amine anion (C2F6NO4S2), the trifluoromethanesulfonate anion (CF303S) and the chloride ion ( Cl-), as anions used in the experiment. As for the cations, Table 1 lists the anionic species in the order of Lewis basicity of the anions, as well as the trifluoroacetate ion (CF3C00-) as a reference. The experimental method includes: providing cerium fluoride as a typical example to each of the ionic liquids; stirring the mixture for one hour at 100 ° C to dissolve the cerium fluoride; then measure the amount of cerium dissolved in the solution. Ionic liquids are considered to have been developed in the 1990s or later, and no report on the amount of organic compounds dissolved in these ionic liquids has been published. According to Table 1, the ionic liquids having the cation of the imidazolium group and having a higher Lewis basicity tend to dissolve a greater quantity of the fluoride, whereas the ionic liquids possessing the anions such as BF4 or C2F6NO4S2 and having an acidity of Weaker Lewis tend to dissolve a larger amount of fluoride. Thus, the fluoride should further dissolve in an ionic liquid composed of a cation having the lowest Lewis acidity and an anion having the strongest Lewis basicity. According to Table 1, the fluoride can be dissolved even in the ionic liquid composed of the cation of the imidazolium group having the strongest Lewis acidity and BF4- which is an anion having the lowest Lewis basicity. This suggests that the fluoride can be dissolved in any ionic liquid composed of the anion and cation shown in Table 1. As described above, the present inventors have experimentally revealed that any fluoride compound is dissolved in ionic liquids. However, the PUREX process, which is one of the reprocessing processes for spent nuclear fuel, uses nitric acid as a solvent to dissolve spent nuclear fuel components. Methods involving the dissolution of spent nuclear fuel components in an ionic liquid require a reduced amount of the ionic liquid used because ionic liquid is more expensive than nitric acid in terms of volume cost. One solution to this problem is, for example, improving the solubility of spent nuclear fuel components in the ionic liquid. The present inventors have conducted numerous studies and have recently discovered an actinide separation process including a new step of dissolving components of a spent nuclear fuel, in an ionic liquid, in which a complexing agent has been formed. dissolved in advance. Specifically, the separation process of the present invention includes dissolving the spent nuclear fuel components containing actinides in an ionic liquid containing a dissolved complexing agent, and separating the dissolved components by means of extraction by solvent. This process may benefit from a reaction between spent nuclear fuel and fluorine to separate uranium, which is a major component of spent nuclear fuel, in the form of a uranium hexafluoride gas with a chemical conversion of the others. fluoride components. Thus, this process does not require dissolution of all spent nuclear fuel components in the ionic liquid, or may further reduce the amount of dissolved material in the ionic liquid. The use of fluorination devices disclosed in JP-A-2004-233066, JP-A-2012-47546 and JP-A-2013-101066 allows reactions of spent nuclear fuel with the fluorinated gas to convert nuclear fuel used in fluorides. Such an actinide separation process includes: reacting the spent nuclear fuel with the fluorinated gas to convert spent nuclear fuel to fluoride and to separate uranium, which is the main component of spent nuclear fuel; dissolving the components of the spent nuclear fuel containing the actinides after the separation of the uranium in the ionic liquid, the ionic liquid containing the dissolved complexing agent; and separating the dissolved components by solvent extraction. It should be noted that the separation of uranium (U), plutonium (Pu), fission products (PF) and minor actinides is referred to as the separation of actinides. According to one embodiment of the present invention, there is provided a process for separating actinides from a material containing the components of a spent nuclear fuel, using an ionic liquid, the process including: dissolving the fuel spent nuclear fuel in a first solution containing a complexing agent and the ionic liquid; and recovering the actinides in an extraction solvent by bringing the first solution containing the spent spent nuclear fuel into contact with the extraction solvent. More preferably, a complexing agent including a phosphorus-oxygen double bond in its compound structure is used as a complexing agent. More preferably, a complexing agent including an amide bond in its compound structure is also used as a complexing agent. More preferably, a complexing agent including at least one of octyl (phenyl) -N, N-diisobutylcarbamoyl-methylphosphine oxide (CMPO), tributyl phosphate (TBP), trioctylphosphine oxide (TOPO) ), di-2-ethylhexylphosphoric acid (D2EHPA), N, N, N ', N'-tetrakis (2-pyridylmethyl) ethane-1,2-diamine (TPEN), 2-ethylhexylhydrogen 2-ethylhexylphosphonate (PC-88A), N, N, N ', N'-tetraoctyl diglycolamide (TODGA), N, N-dioctyldiglycol-amic acid (DODGAA) and dicyclohexyl-18-crown-6 ( DC18C6) is also used as a complexing agent. More preferably, a separation apparatus for separating the actinides from the spent nuclear fuel by means of the ionic liquid includes: a dissolution tank configured to dissolve spent nuclear fuel in a first solution containing the complexing agent and the ionic liquid; and a separation tank configured to separate the actinides from the first solution by extraction, thus solving the aforementioned problem. The present invention can reduce the amount of ionic liquid needed to dissolve spent nuclear fuel components.
    [0004] In the remainder of the description, the elements that are identical or of identical function bear the same reference sign. A first embodiment of the present invention is hereinafter described with reference to Fig. 1. Fig. 1 is a flowchart of this embodiment illustrating the steps until the actinides are separated from a fuel. nuclear power. This embodiment includes a dissolution step S1 of dissolving the spent nuclear fuel components 7 in an ionic liquid containing a complexing agent 6; an extraction step S2 consisting in obtaining an extraction liquid containing actinides 12 and a second solution 11 separated from the extraction liquid 10 by bringing a first solution 8 obtained in the dissolution step S1 into contact with an extraction solvent 9; a second regeneration step S4 consisting in separating the extraction liquid 10 into extraction solvent 9, complexing agent 6 and actinides 12; and a first regeneration step S3 of separating the second solution 11 into ionic liquid 5 and fission products (PF) 13, the second solution 11 being obtained after separation of the actinides 12 from the first solution 8. The ionic liquid 5 and The complexing agent 6 is used again in the dissolution step S1. The extraction solvent 9 is used again in the extraction step S2. In this embodiment, a complexing agent including a phosphorus-oxygen double bond in its compound structure is used as a complexing agent 6. The use of such a complexing agent can particularly improve the solubility of actinides and can reduce the amount of ionic liquid required for dissolution. In this embodiment, an extraction solvent which includes at least one member selected from the group consisting of alkanes and compounds having alkyl groups, respectively, is used as the extraction solvent. such an extraction solvent can minimize the losses of the ionic liquid and the extraction solvent in the extraction operation due to the phase separation between these liquids. In this embodiment, a case of using dodecane as extraction solvent 9 will be illustrated. The actinide separation method according to this embodiment is described in detail below. Since spent nuclear fuel, discharged from nuclear power plants, is stored in cladding tubes, shearing and stripping are performed by existing processes used in reprocessing facilities. The components 7 of the spent nuclear fuel thus obtained are recovered and dissolved in the ionic liquid containing the complexing agent 6 dissolved in the dissolution step S1 to provide the first solution 8. Any ionic liquid 25 composed of the cation and the anion shown in Table 1 can be used as ionic liquid 5.
    [0005] Cations (Low Lewis I-Acidity) Pyrrolidinium Ammonium Pyridinium Imidazolium (Low BF4 14 1.8 T Basin o H Lewis <1 Strong) C2F6N04S2 1.4 0.12 CF303S-4.6 CF3C00- Cl- 46 Table 1: Amount of cesium fluoride dissolved in the ionic liquids (ppm) Next, the first solution 8 is brought into contact with the extraction solvent 9 in the extraction step S2. The requirements for the extraction solvent 9 include a weak dissolution in the ionic liquid and the possibility for the actinides 12 to be transferred to the extraction solvent 9 in contact between the first solution 8 and the extraction solvent. 9. As such solvents, hydrocarbon solvents such as mainly alkanes are used particularly effectively. Among them, dodecane is a solvent used in reprocessing plants, and is suitable for use in this embodiment. The extraction liquid 10 contains the extraction solvent 9, the actinides 12 and the complexing agent 6. These components are obtained by a separation operation such as a stripping, an electrolysis and an adsorption in a second step of S4 regeneration. Here, the complexing agent 6 is transferred to the extraction liquid 10 together with the actinides 12 in the extraction step S2. The extraction solvent 9 and the complexing agent 6 recovered thus separately are used in the extraction step S2 and the dissolution step S1, respectively. On the other hand, the first solution 8 from which the actinides 12 are separated in the extraction step S2 is sent to the first regeneration step S3 as a second solution 11, and separated into ionic liquid 5 and PF 13 by a separation operation such as stripping, electrolysis and adsorption. The recovered ionic liquid 5 is recycled to the dissolution step S1. After separating the elements according to the respective groups in this manner, the PFs 13 are vitrified, and the actinides 12 are converted to nuclides having short half-lives through nuclear transmutation with neutron irradiation or the like. The nuclides are then subjected to vitrification in the same way as for the fission products. Nuclear fuel materials such as uranium (U) and plutonium (Pu) are recycled as nuclear fuel. In this embodiment, the complexing agent including the phosphorus-oxygen double bond in the compound structure is used as complexing agent 6, but a complexing agent including an amide bond in its compound structure can be used. The use of the complexing agent including the amide bond in the compounded structure as complexing agent 6 allows easy disposal of waste because the complexing agent does not contain phosphorus.
    [0006] As complexing agent 6, the following compounds can be used. These are octyl (phenyl) -N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO), tributylphosphate (TBP), trioctylphosphine oxide (TOPO), di-2-ethylhexylphosphoric acid (D2EHPA) N, N, N ', N'-tetrakis (2-pyridylmethyl) ethane-1,2-diamine (TPEN), 2-ethylhexyl hydrogen-2-ethylhexylphosphonate (PC-88A), N, N, N N, N'-tetraoctyl-diglycolamide (TODGA), N, N-dioctyldiglycol-amic acid (DODGAA) and dicyclohexyl-18-crown-6 (DC18C6). Two or more compounds selected from these compounds may be used. The use of the complexing agent 6 can particularly improve the solubility of actinides and can reduce the amount of ionic liquid required for dissolution. In this embodiment, spent nuclear fuel components are ionized when dissolved in the ionic liquid. The formation of complexes through the coordination of these ions with a complexing agent allows the spent nuclear fuel components to be dissolved more stably in the ionic liquid. Such formation of more soluble complexes in the ionic liquid can improve the solubility of spent nuclear fuel components. According to this embodiment, the dissolution of spent nuclear fuel components in the ionic liquid containing the dissolved complexing agent may have the effect of improving the solubility of spent nuclear fuel components in the ionic liquid to reduce the amount of the spent fuel. ionic liquid necessary for dissolution.
    [0007] According to this embodiment, a more compact installation is provided for separating spent nuclear fuel, uranium (U), plutonium (Pu), fission products (PF) and minor actinides.
    [0008] Referring to Figure 2, there is described below an actinide separation apparatus according to this embodiment for separating actinides from spent nuclear fuel components. The actinide separation apparatus of this embodiment includes a dissolution tank 20, a heat exchanger 21, a pump 22, a separation tank 23, a pump 24, a pump 25, a valve 26, a valve 27 , a valve 28, a valve 29, a valve 30, a valve 31, a valve 32 and a valve 33. When the valves 26 to 28 are open, the components 7 of the spent nuclear fuel, an ionic liquid 5 and a complexing agent 6 are supplied to the dissolving tank 20 through the inlets. The dissolution tank 20 includes the heat exchanger 21 and can heat the ionic liquid 5. Once the dissolution of the spent nuclear fuel components 7 has been completed, the valves 29-30 are opened to transfer a first solution 8 to the separation tank 23 by means of the pump 22. Once the transfer is complete, the valves 29-30 are closed. The valve 33 is then opened to introduce an extraction solvent 9 into the separation tank 23 through an inlet. The first solution 8 and the extraction solvent 9 cause a phase separation without being mixed with each other, and the actinides 12 and the complexing agent 6 are transferred to the extraction solvent 9 from the first solution 8. Once this extraction separation operation is completed, the pump 31 is first opened to send the total amount of a second solution 11 to a first regeneration step S3 by means of the pump 24. The second Solution 11 is separated into fission products (PF) 13 and ionic liquid 5 in the first regeneration step S3. Then, the valve 31 is closed and the valve 32 is opened to send an extraction liquid 10 to a second regeneration step S4 by means of the pump 25. In the second regeneration step S4, the extraction liquid 10 is separated from actinides 12, complexing agent 6 and extraction solvent 9. It should be noted that the first regeneration step S3 and the second regeneration step S4 signify a first regeneration unit and the second regeneration unit in FIG. , respectively. The use of the actinide separation apparatus according to this embodiment allows easy separation of the actinides from the spent nuclear fuel components. A second embodiment of the present invention is hereinafter described with reference to FIG. 3. In this embodiment, the steps until the moment when the actinides are separated from the spent nuclear fuel are the same as in the first one. Embodiment (FIG. 1), but the actinide separation apparatus for separating the actinides from spent nuclear fuel is different from that of the first embodiment (FIG. 2). With reference to FIG. 3, it is described, with emphasis on constituents different from those of the first embodiment, the actinide separation apparatus according to this embodiment for separating actinides from used nuclear fuel components. . The actinide separation apparatus according to this embodiment includes a dissolution and separation tank 130, a heat exchanger 21, a stirring unit 131, a pump 24, a pump 25, a valve 134, a valve 135 a valve 136, a valve 137, a valve 138, and a valve 139. When the valves 134 to 136 are open, the spent nuclear fuel components 7, an ionic liquid 5 and a complexing agent 6 are supplied to the reservoir 130. through the entrances. The reservoir 130 includes the heat exchanger 21 and can heat the ionic liquid 5. When the dissolution of the spent nuclear fuel components 7 has been completed, the valve 137 is opened to introduce an extraction solvent 9 into the reservoir 130 to through an entrance. The ionic liquid 20 and the extraction solvent 9 cause a phase separation without being mixed with each other, and the actinides 12 and the complexing agent 6 are transferred to the extraction solvent 9. vigorous solution in the reservoir 130 by means of the stirring unit 131 can accelerate here the transfer of actinides 12 and the complexing agent 6 to the extraction solvent 9. Once this separation operation by extraction At the end, the valve 139 is first opened to send the total amount of a second solution 11 to a first regeneration step S3 by means of the pump 24. The second solution 11 is separated into fission products (PF) 13 and ionic liquid 5 in the first regeneration step S3. Then, the valve 139 is closed and the valve 140 is opened to send an extraction liquid 10, to a second regeneration step S4, by means of the pump 25. In the second regeneration step S4, the extraction liquid 10 is separated into actinides 12, complexing agent 6 and extraction solvent 9. It should be noted that the first regeneration step S3 and the second regeneration step S4 signify a first regeneration unit and the second regeneration unit of the second regeneration unit S3. Figure 3, respectively. In this embodiment, spent nuclear fuel components are ionized when dissolved in the ionic liquid. The formation of complexes through the coordination of these ions with a complexing agent allows the spent nuclear fuel components to be dissolved more stably in the ionic liquid. Such formation of more soluble complexes in the ionic liquid can improve the solubility of spent nuclear fuel components. According to this embodiment, the dissolution of spent nuclear fuel components in the ionic liquid containing the dissolved complexing agent may have the effect of improving the solubility of spent nuclear fuel components in the ionic liquid to reduce the amount of the liquid. ionic needed for dissolution. According to this embodiment, a more compact installation is provided for separating spent nuclear fuel from uranium (U), plutonium (Pu), fission products (PF) and minor actinides. The use of the actinide separation apparatus according to this embodiment allows easy separation of the actinides from spent nuclear fuel components. The actinide separation apparatus according to this embodiment incorporates a reservoir for dissolving spent nuclear fuel components and a reservoir for separating actinides 12, and thus advantageously eliminates the need to recover the ionic liquid containing the dissolved components. 7 spent nuclear fuel from the dissolution tank and transfer the ionic liquid 5 to a different separation tank. A third embodiment of the present invention is hereinafter described with reference to Fig. 1. In the first embodiment, there is disclosed a method of dissolving spent nuclear fuel components in an ionic liquid, the components 7 being obtained by stripping the spent nuclear fuel. In this embodiment, spent nuclear fuel components are reacted with fluorine-containing gas and thereby converted to solid fluorides (also referred to as spent nuclear fuel components). These spent nuclear fuel components 7 are dissolved in an ionic liquid. In this embodiment, the use of the separating apparatus (FIG. 2) according to the first embodiment is illustrated as an apparatus for separating actinides for separating the actinides from the spent nuclear fuel, but the separation apparatus (Figure 3) according to the second embodiment can be used. A method of separating actinides according to this embodiment is described below with reference to Figure 1, with emphasis on operations different from those of the first embodiment. A process for separating actinides according to this embodiment includes: a dissolution step S1 consisting in dissolving the components 7 (solid fluorides) of a spent nuclear fuel in an ionic liquid 5, the components 7 being obtained by a fluorination of the nuclear fuel worn; an extraction step S2 consisting of recovering the actinides 12 in an extraction liquid 10 by bringing a first solution 8 obtained in the dissolution step S1 into contact with an extraction solvent 9; a second regeneration step S4 consisting in separating the extraction liquid 10 into extraction solvent 9, complexing agent 6 and actinides 12; and a first regeneration step S3 consisting in separating a second solution 11 in ionic liquid 5 and fission products (PF) 13, the second solution 11 being obtained after separation of the actinides 12 from the first solution 8. The separation process actinides according to this embodiment is hereinafter described in detail. Since spent nuclear fuel discharged from nuclear power plants is stored in cladding tubes, shearing and stripping are performed by existing methods used in reprocessing facilities. The spent nuclear fuel thus obtained is reacted with fluorinated gas to provide spent nuclear fuel components. In the fluorination step, a flame-type fluorination device may be used. Since the fluorinated gas is highly reactive, substantially all spent nuclear fuel components are reacted with the fluorinated gas to convert the oxides to fluorides. The spent nuclear fuel contains various elements whose fluorides have significantly different boiling points. In the fluorination step, the fluorides are broadly divided into volatile uranium hexafluoride (UF6 gas) and residual components 7 of the spent nuclear fuel. Components 7 of this spent nuclear fuel include plutonium (Pu), fission products (PF), minor actinides, and a small amount of uranium (U). The spent nuclear fuel components 7 are recovered and dissolved in the ionic liquid 5 containing the complexing agent 6 dissolved in the dissolution step S1 to provide the first solution 8. As the ionic liquid 5, any ionic liquid composed of the cation and the anion shown in Table 1 can be used. Table 1 indicates the amount of cerium fluoride dissolved in the respective ionic liquids, with cerium fluoride being a typical example. Since the actinides that are elements to be separated in this embodiment belong to the same group as cerium, actinide fluorides should also be dissolved in the ionic liquids. Then, the first solution 8 is brought into contact with the extraction solvent 9 in the extraction step S2. The requirements for the extraction solvent 9 include difficult dissolution in the ionic liquid and the possibility for the actinides 12 to be transferred to the extraction solvent 9 in contact between the first solution 8 and the extraction solvent 9 As such solvents, hydrocarbon solvents such as mainly alkanes are used particularly effectively. Among them, dodecane, which is a solvent used in reprocessing plants, is suitable for use in this embodiment.
    [0009] The extraction liquid 10 contains the extraction solvent 9, the actinides 12 and the complexing agent 6. These components are obtained by a separation operation such as stripping, electrolysis and adsorption in a second regeneration step. S4.
    [0010] Here, the complexing agent 6 is transferred to the extraction liquid 10 together with the actinides 12 in the extraction step S2. The extraction solvent 9 and the complexing agent 6 thus recovered separately are used in the extraction step S2 and the dissolution step S1, respectively. On the other hand, the first solution 8 from which the actinides 12 are separated in the extraction step S2 is sent to the first regeneration step S3 as a second solution 11, and separated into ionic liquid 5 and products fission (PF) 13 by a separation operation such as stripping, electrolysis and adsorption. The recovered ionic liquid 5 is recycled to the dissolution step S1. After separating the elements according to the respective groups, in this way, the fission products (PF) 13 are vitrified, and the actinides 12 are converted into nuclides having short half-lives by means of a nuclear transmutation with neutron irradiation or the like. The nuclides are then subjected to vitrification in the same way as for the fission products. Nuclear fuel materials such as uranium (U) and plutonium (Pu) are recycled as nuclear fuel. In this embodiment, spent nuclear fuel components are ionized when dissolved in the ionic liquid. The formation of complexes through the coordination of these ions with a complexing agent allows the spent nuclear fuel components to be dissolved more stably in the ionic liquid. Such formation of more soluble complexes in the ionic liquid can improve the solubility of spent nuclear fuel components. According to this embodiment, the dissolution of spent nuclear fuel components in the ionic liquid containing the dissolved complexing agent may have the effect of improving the solubility of spent nuclear fuel components in the ionic liquid to reduce the amount of the ionic liquid. necessary for dissolution.
    [0011] According to this embodiment, a more compact installation is provided for separating the spent nuclear fuel with uranium (U), plutonium (Pu), fission products (PF) and minor actinides. According to this embodiment, spent nuclear fuel components obtained by stripping off spent nuclear fuel are reacted with a fluorine-containing gas, and uranium, which is a major component of spent nuclear fuel, is separated in advance. This operation can significantly reduce the amount of solid fluoride to be dissolved in an ionic liquid and consequently reduce the amount of ionic liquid needed to dissolve the total amount of solid fluorides.
    [0012] A fourth embodiment of the present invention is hereinafter described with reference to Fig. 4. Fig. 4 is a flowchart of this embodiment which illustrates the steps until the actinides are separated from the spent nuclear fuel.
    [0013] A method of separating actinides according to this embodiment includes: a dissolution step S1 consisting of dissolving the components 7 (solid fluorides) of a spent nuclear fuel, in an ionic liquid 5; an extraction step S2 consisting of recovering the actinides 12 in an extraction liquid 10 by bringing a first solution 8 obtained in the dissolution step S1 into contact with an extraction solvent 9; a second regeneration step S4 consisting in separating the extraction liquid 10 into extraction solvent 9, complexing agent 6 and actinides 12; and a first regeneration step S3 consisting in separating a second solution 11 into ionic liquid 5 and fission products (PF) 13, the second solution 11 being obtained after separation of the actinides 12 from the first solution 8.
    [0014] This embodiment is different from the first embodiment in that the complexing agent 6 is not added to the dissolution step S1 and the extraction solvent 9 and the complexing agent 6 are recovered together and not separately. in the second regeneration step S4.
    [0015] The actinide separation method according to this embodiment is described in detail below, with emphasis on operations different from those of the first embodiment. The case of the reaction of spent nuclear fuel with the fluorinated gas to give spent nuclear fuel components in a manner similar to the first embodiment is described. Components 7 of spent nuclear fuel include plutonium (Pu), fission products (PF), minor actinides, and a small amount of uranium (U). These spent nuclear fuel components 7 are recovered and dissolved in the ionic liquid 5 containing the complexing agent 6 dissolved in the dissolution step S1 to provide the first solution 8. As the ionic liquid 5, the same ionic liquid as in the first embodiment is used. Then, the first solution 8 is brought into contact with the extraction solvent 9 containing the complexing agent dissolved in the extraction step S2. As extraction solvent 9, the same solvent as in the first embodiment is used. This operation transfers the actinides 12 dissolved in the first solution 8 to the extraction solvent 9. The formation of complexes between the complexing agent 6 and the actinides 12 here facilitates the dissolution of the actinides 12 in the extraction solvent 9, which which allows easy transfer of the actinides 12 to the extraction solvent 9.
    [0016] The extraction liquid 10 is a solution containing the extraction solvent 9, the actinides 12 and the complexing agent 6. The actinides 12 in the extraction liquid 10 are separated by a separation operation such as a stripping, electrolysis and adsorption in the second regeneration step S4. The recovered extraction solvent 9 and the complexing agent 6 are both recycled in the extraction step S2. On the other hand, the first solution 8, from which the actinides 12 are separated in the extraction step S2, is sent to the first regeneration step S3 as a second solution 11, and separated into ionic liquid 5 and products fission (PF) 13 by a separation operation such as stripping, electrolysis and adsorption. The recovered ionic liquid 5 is recycled to the dissolution step S1. After separating the elements according to the respective groups, in this manner, the fission products 13 are vitrified, and the actinides 12 are converted into nuclides having short half-lives through nuclear transmutation with neutron irradiation. or similar. The nuclides are then subjected to vitrification in the same way as for the fission products. Nuclear fuel materials such as uranium (U) and plutonium (Pu) are recycled as nuclear fuel. In this embodiment, spent nuclear fuel components are ionized when dissolved in the ionic liquid. The formation of complexes through the coordination of these ions with the complexing agent allows the spent nuclear fuel components to be dissolved more stably in the ionic liquid. Such formation of more soluble complexes in the ionic liquid can improve the solubility of spent nuclear fuel components.
    [0017] According to this embodiment, the dissolution of spent nuclear fuel components into the ionic liquid containing the dissolved complexing agent may have the effect of improving the solubility of spent nuclear fuel components in the ionic liquid to reduce the amount of the spent fuel. ionic liquid necessary for dissolution. According to this embodiment, a more compact installation is provided for separating spent nuclear fuel into uranium (U), plutonium (Pu), fission products (PF) and minor actinides. According to this embodiment, the solubility of the components of the spent nuclear fuel decreases with respect to the first embodiment, because the complexing agent is dissolved in advance in the ionic liquid.
    [0018] This embodiment does not, however, require recovery by separation of the extraction solvent and the complexing agent in the extraction liquid regeneration step, having the effect of simplifying the regeneration step of the liquid of the extraction liquid. 'extraction.
    [0019] According to this embodiment, with respect to the first embodiment, the complexing agent is dissolved in the extraction solvent in order to facilitate the formation of the complexes between the actinides and the complexing agent which are dissolved in the liquid. ionic, which allows easy transfer of the actinide complexes to the extraction solvent, having the effect of facilitating the separation of the actinides in the extraction solvent in the extraction step.
    权利要求:
    Claims (8)
    [0001]
    REVENDICATIONS1. A process for separating the actinides (12) from a material containing spent nuclear fuel components (7) by means of an ionic liquid (5), the process comprising the following steps of: dissolving (S1) the spent nuclear fuel in a first solution (8) containing a complexing agent (6) and the ionic liquid (5); and recovering the actinides in an extraction solvent (9), bringing the first solution (8) containing the spent spent nuclear fuel into contact with the extraction solvent (9). 15
    [0002]
    2. The process according to claim 1, further comprising a step of: reacting the spent nuclear fuel with fluorine to give solid fluorides, the dissolving step (S1) involving the dissolution of the solid fluorides in the first solution (8).
    [0003]
    3. The process according to claim 1, wherein the complexing agent (6) includes a phosphorus-oxygen double bond in its composite structure. 25
    [0004]
    The method of claim 1, wherein the complexing agent (6) includes an amide bond in its complex structure. 30
    [0005]
    The process according to claim 1, wherein the complexing agent (6) includes at least one of octyl (phenyl) -N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO), tributylphosphate (TBP), trioctylphosphine oxide (TOPO), di-2-ethylhexylphosphoric acid (D2EHPA), N, N, N ', N'-tetrakis (2-pyridylmethyl) ethane-1,2-diamine (TPEN), 2-ethylhexyl hydrogen-2-ethylhexylphosphonate (PC-88A), N, N, N ', N'-tetraoctyl-diglycolamide (TODGA), N, N-dioctyldiglycol-amic acid (DODGAA) and dicyclohexyl-18 - crown-6 (DC18C6).
    [0006]
    The process according to claim 1, wherein the extraction solvent (9) includes at least one of alkanes and compounds having an alkyl group.
    [0007]
    Separating apparatus for separating the actinides (12) from a material containing spent nuclear fuel components (7) by means of an ionic liquid (5), the apparatus comprising: a dissolution tank ( 20) configured to dissolve spent nuclear fuel in a first solution (8) containing a complexing agent (6) and the ionic liquid (5); and a separation tank (23) configured to separate the actinides (12) from the first solution (8) by extraction.
    [0008]
    8. Separation apparatus for separating the actinides (12) from a material containing spent nuclear fuel components (7) by means of an ionic liquid (5), the apparatus comprising: a dissolving tank (130) configured to dissolve the spent nuclear fuel in a first solution (8) containing a complexing agent (6) and the ionic liquid (5), the dissolution tank (130) having a separation tank function configured to separate the actinides (12) of the first solution (8) by extraction.
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    同族专利:
    公开号 | 公开日
    GB201510669D0|2015-07-29|
    JP2016008891A|2016-01-18|
    GB2534432A|2016-07-27|
    GB2534432B|2017-06-07|
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    WO2015059777A1|2013-10-23|2015-04-30|株式会社日立製作所|Method for separating actinide and device for treating spent fuel|
    FR3015760B1|2013-12-20|2016-01-29|Commissariat Energie Atomique|PROCESS FOR TREATING A USE NUCLEAR FUEL COMPRISING A DECONTAMINATION STEP OF URANIUM IN AT LEAST ONE ACTINIDE BY COMPLEXATION OF THIS ACTINIDE |WO2017145352A1|2016-02-26|2017-08-31|株式会社日立製作所|Method for separating minor actinides and apparatus for separating minor actinides|
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    2017-04-28| PLFP| Fee payment|Year of fee payment: 3 |
    优先权:
    申请号 | 申请日 | 专利标题
    JP2014129805A|JP2016008891A|2014-06-25|2014-06-25|Separation method of actinide and separation unit of actinide|
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