JP2004069383A - Method and apparatus for processing waste from nuclear fuel cycle facility - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To reduce an amount of a secondary waste generated in processing a waste such as uranium adsorbent, etc. generated from a nuclear fuel cycle facility, and to reduce an operation cost. <P>SOLUTION: The method includes steps of reducing a radioactive material (UF<SB>6</SB>) adhered to a used uranium adsorbent (NaF), electrolyzing a fused salt in which a chloride (NaCl) having a common cation is added to the reduced uranium adsorbent and depositing uranium metal on a cathode, distilling and separating a salt mixture obtained after the electrolysis, returning the separated chloride to the fused salt in the electrolysis process, removing the salt mixture adhered to the deposited uranium metal, and processing a volatile gas generated in the reduction and electrolysis processes. <P>COPYRIGHT: (C)2004,JPO

Description

[0001]
[Industrial applications]
The present invention relates to a method and an apparatus for treating waste from a nuclear fuel cycle facility, and in particular, radioactive substances such as uranium and uranium compounds generated from a nuclear fuel cycle facility such as a uranium handling facility and a transfer facility between facilities are attached. The present invention relates to a method and an apparatus for separating and recovering uranium and the like from a chemical trap agent.
[0002]
[Prior art]
Generally, uranium, transuranium elements or radionuclide elements or compounds (hereinafter referred to as radioactive substances) generated from a nuclear fuel cycle facility are adsorbed on an adsorbent (chemical trapping agent) such as NaF, CaF ₂ or MgF _2. Collected. The nuclear fuel cycle facility includes a uranium mining facility, a smelting facility, a conversion facility, an enrichment facility, a processing facility, a nuclear reactor, a reprocessing facility, a waste facility, and a facility for transferring between these facilities.
[0003]
As a method for separating and recovering radioactive substances such as uranium and transuranium elements attached from the adsorbent, a method disclosed in JP-A-2000-88991 has been conventionally used.
[0004]
As shown in FIG. 7, this method includes a reducing step 52 of reducing UF ₆ adhering to the used uranium adsorbent 51 to UF ₄ by a reducing gas such as hydrogen gas, There is a molten salt electrolysis step 53 in which a chloride having a common cation is added and heated and melted, an anode and a cathode are placed in the molten salt, and a voltage is applied to perform electrolysis. In the molten salt electrolysis step 53, metallic uranium, which is a radioactive waste 54, is deposited and collected on the cathode, and in the electrolytic cell, a decontaminated uranium adsorbent and a chloride 55 having a common cation added thereto are added. Remains.
[0005]
[Problems to be solved by the invention]
However, in this method, the uranium adsorbent after the molten salt electrolysis and the chloride having the same cation as the uranium adsorbent become waste. A larger amount of secondary waste was generated than the amount of the adsorbent 51.
[0006]
In addition, since it is necessary to add a chloride having a cation common to the uranium adsorbent every time the molten salt electrolysis step 53 is performed, there is a problem that the processing cost is increased. Furthermore, volatile gas generated in the reduction step 52 and the molten salt electrolysis step 53 has not been sufficiently treated.
[0007]
The present invention has been made in order to solve these problems, and in the treatment of waste generated from a nuclear fuel cycle facility, it is possible to reduce the amount of secondary waste generated and to reduce operating costs. It is an object of the present invention to provide a waste treatment method and a waste treatment apparatus that can perform the treatment.
[0008]
[Means for Solving the Problems]
The method for treating waste from a nuclear fuel cycle facility according to the present invention is a method for treating a waste in which the chemical form to which the radioactive substance discharged from the nuclear fuel cycle facility is attached is a salt of fluoride, wherein the radioactive substance attached to the waste is A reduction step of reducing the substance, and to the waste to which the radioactive substance reduced in the reduction step adheres, electrolyze a molten salt of a salt mixture obtained by adding and mixing chloride having a cation common to the waste, and forming a cathode. A molten salt electrolysis step of precipitating and recovering the metal of the radioactive substance, and distilling the salt mixture after the molten salt electrolysis step, and having the same waste and the same cation as the waste as a fluoride salt A distillation step of separating the chloride, a step of returning the chloride separated in the distillation step to the molten salt in the molten salt electrolysis step, and attaching to the metal of the radioactive substance deposited and recovered. And the adhered salt-removing step of removing the salt mixture is characterized by comprising the off-gas treatment step of treating the volatile gas generated in the reduction step and the molten salt electrolysis process.
[0009]
In the waste treatment method of the present invention, in the reduction step, the waste to which the radioactive substance has adhered can be heated stepwise while flowing the reducing gas. Further, the method may have a solid salt transfer step of cooling and solidifying the molten salt after the molten salt electrolysis step and sending the solidified salt to the distillation step. Further, the method may include a molten salt transfer step of transferring the molten salt after the molten salt electrolysis step to a distillation step in a molten state.
[0010]
Further, the attached salt removing step may include a heating and evaporating step of heating and evaporating and removing the salt mixture attached to the metal of the radioactive substance deposited and recovered. Further, the attached salt removing step is a washing step of washing and removing a salt mixture attached to the metal of the radioactive substance deposited and recovered with water, and the aqueous solution containing the salt mixture from the washing step is treated with the radioactive substance. And an evaporating and drying step of evaporating and drying the separated aqueous solution. Then, the salt mixture removed in the attached salt removing step can be supplied to the distillation step. In addition, the heating and evaporating step is performed simultaneously with the distillation step, and the salt mixture after the molten salt electrolysis step is distilled and separated, and the salt mixture attached to the metal of the radioactive substance can be removed by distillation.
[0011]
The waste treatment apparatus from a nuclear fuel cycle facility of the present invention heats, while flowing a reducing gas, waste in which a chemical form to which a radioactive substance discharged from the nuclear fuel cycle facility adheres is a salt of fluoride, A reduction device for reducing the substance, and a waste to which a radioactive substance reduced by the reduction device is attached, and a molten salt of a salt mixture obtained by adding / mixing a chloride having a cation common to the waste, and electrolyzing the cathode A molten salt electrolyzer for precipitating and recovering the metal of the radioactive substance, a distillation apparatus for distilling the salt mixture decontaminated by the molten electrolysis apparatus, and a waste that has been distilled and separated by the distillation apparatus. A recycling line for returning chlorides having a common cation to the molten salt electrolyzer, and attaching to the metal of the radioactive substance deposited and collected on the cathode in the molten salt electrolyzer; Characterized by comprising the deposited salt removal device for removing the salt mixture, and off-gas treatment apparatus for treating volatile gas generated in the reduction device and the molten salt electrolysis apparatus.
[0012]
In the waste treatment apparatus of the present invention, the attached salt removing apparatus is a washing apparatus for washing / removing a salt mixture attached to a radioactive metal with water, and an aqueous solution containing the salt mixture discharged from the washing apparatus. And a solid-liquid separation device for separating the aqueous solution from the metal of the radioactive substance, and an evaporating and drying device for evaporating and drying the separated aqueous solution. Also, a reducing device supplies a reducing gas to the waste to which the radioactive substance is attached, a reducing gas supply device, a heating device for heating the waste, and a temperature for controlling the heating temperature of the heating device in a stepwise manner. And a control device.
[0013]
ADVANTAGE OF THE INVENTION According to the method and apparatus of this invention, in the processing of the waste which is a fluoride salt to which radioactive substances such as uranium discharged from a nuclear fuel cycle facility are attached, the same cation as the waste added in the molten salt electrolysis step is used. Is separated by distillation and reused in the molten salt electrolysis step, it is possible to reduce the amount of secondary waste generated, and it is necessary to add the chloride for each molten salt electrolysis step Because there is no operation cost, the operation cost can be reduced.
[0014]
Further, since volatile gases such as HF gas, chlorine gas, and fluorine gas generated in the reduction step and the molten salt electrolysis step are collected and treated in the off-gas treatment step, a method for collecting and treating waste including gas components is included. And as a device, it is fully satisfactory.
[0015]
The method and apparatus for treating waste from a nuclear fuel cycle facility according to the present invention includes uranium, a uranium compound, a transuranium element, a compound of a transuranium element, or a radioactive substance generated from a uranium handling facility and a facility for transferring between the uranium handling facilities. It is suitable for treating chemical trapping agents such as NaF, CaF ₂ , and MgF ₂ to which radioactive substances such as nuclides are adsorbed.
[0016]
BEST MODE FOR CARRYING OUT THE INVENTION
Hereinafter, an embodiment of a method for treating waste from a nuclear fuel cycle facility according to the present invention will be described with reference to the drawings. In the following embodiment, NaF is used as a uranium chemical trapping agent (uranium adsorbent), and UF ₆ which is a radioactive substance attached to the uranium adsorbent is removed.
[0017]
FIG. 1 is a flowchart (flow chart) showing a first embodiment of the method for treating waste from a nuclear fuel cycle facility of the present invention.
[0018]
The treatment method of the first embodiment includes a reduction step 2 for reducing UF ₆ attached to the used uranium adsorbent 1, a uranium adsorbent to which the reduced radioactive substance is attached, and a cation common to the uranium adsorbent and the cation. A molten salt electrolysis step 3 for electrolyzing a molten salt mixed with a certain chloride, a solid salt transporting step 4 for cooling and solidifying the molten salt after electrolysis and transporting and supplying a solidified salt to the next step; A distillation step 5 of distilling and separating the uranium adsorbent, a chloride 6 having a cation common to the distilled and separated uranium adsorbent, into the molten salt of the molten salt electrolysis step 3, and a molten salt electrolysis step 3. It has an attached salt heating and removing step 8 for heating and removing the salt mixture attached to the deposited metal uranium, and an off-gas treating step 9 for treating volatile gas generated in the reduction step 2 and the molten salt electrolysis step 3. ing. In addition, the code | symbol 10 in a figure shows the decontaminated uranium adsorbent, 11 shows the radioactive waste metal uranium, 12 shows the adhesion salt removed from the metal uranium, respectively.
[0019]
In the reduction step 2, the used uranium adsorbent 1 is heated while flowing a reducing gas, and uranium hexafluoride UF ₆ which is a radioactive substance attached to the uranium adsorbent 1 is reduced to uranium tetrafluoride UF ₄ . As the reducing gas, hydrogen gas or a mixed gas of hydrogen gas and an inert gas such as argon gas can be used.
[0020]
Next, in the molten salt electrolysis step 3, a chloride having a common cation, such as NaCl, is added to NaF, which is a uranium adsorbent, so that a molar ratio of NaF: NaCl = 0.35: 0.65 is obtained. An anode and a cathode are placed in a molten salt heated and melted by an electric furnace or the like. For example, a solid electrode of glassy carbon for the anode and a low carbon steel for the cathode can be used.
[0021]
Then, electrolysis is performed by applying a DC voltage between the anode and the cathode. Uranium tetrafluoride in the molten salt is reduced by the electrolysis, and uranium metal is deposited on the cathode.
[0022]
In the solid salt transfer step 4, for example, a graphite pulling rod is inserted into the molten salt after the molten salt electrolysis step 3, and the molten salt is cooled and solidified as it is. When the molten salt is solidified, the pulling rod is integrally fixed, so that the solidified material can be pulled up by grasping the pulling rod. Thus, the solidified salt mixture can be pulled up and transferred to the next step.
[0023]
Next, the salt mixture thus transferred is distilled at normal pressure or reduced pressure in a distillation step 5, and NaF as a uranium adsorbent and chloride (NaCl) having a common cation added in the molten salt electrolysis step are used. And separated into
[0024]
Since the boiling point of NaF is 1704 ° C. and the boiling point of NaCl is 1413 ° C., NaCl is distilled and separated from NaF due to the difference between these boiling points. The separated NaCl is returned to the molten salt in the molten salt electrolysis step 3 and reused. Thus, NaCl is distilled and separated, and the decontaminated uranium adsorbent 10 is obtained.
[0025]
In the attached salt heating and removing step 8, the metallic uranium deposited and collected on the cathode in the molten salt electrolysis step 3 is heated, and the attached salt mixture is evaporated and removed. Thus, metal uranium is obtained as the radioactive waste 11. The salt mixture removed from the metal uranium is sent to the distillation step 5 as a deposit salt 12. Then, it is distilled in the same manner as the salt mixture sent through the solid salt transfer step 4 to be separated into NaF and NaCl.
[0026]
In the off-gas treatment step 9, volatile gas generated in the reduction step 2 and the molten salt electrolysis step 3, for example, hydrogen fluoride, vapor of molten salt (NaF—NaCl), vapor of Na, fluorine gas, chlorine gas, etc. It is collected and treated with a chemical adsorbent such as activated carbon, a high-temperature vapor trap, a reflux vapor trap, or the like.
[0027]
According to the first embodiment, in the distillation step 5 after the molten salt electrolysis step 3, chloride (NaCl) having a common cation added to NaF as a uranium adsorbent is separated, and this chloride is melted. It is recycled to the salt electrolysis step 3 and used again. In addition, the salt mixture attached to the metal uranium, which is a cathode precipitate, is also removed from the metal uranium, then distilled and separated in a distillation step, and chloride (NaCl) is reused in a molten salt electrolysis step. The amount of secondary waste generated can be significantly reduced. And since there is no need to add the chloride for each step of the molten salt electrolysis step, the operation cost can be reduced.
[0028]
Further, in the off-gas treatment step 9, volatile gases such as HF gas, chlorine gas and fluorine gas generated in the reduction step 2 and the molten salt electrolysis step 3 are treated, so that waste including gas components is collected and treated. This is a satisfactory method.
[0029]
Next, second to fourth embodiments of a method for treating waste from a nuclear fuel cycle facility will be described with reference to FIGS. In these figures, the same parts as those in FIG. 1 are denoted by the same reference numerals, and the description of the overlapping parts will be omitted.
[0030]
In the second embodiment, as shown in FIG. 2, a molten salt transfer step 13 is provided instead of the solid salt transfer step in the first embodiment. In the molten salt transfer step 13, the molten salt after the molten salt electrolysis step 3 is transferred to the next step (distillation step 5) in a molten state without cooling. It is possible to adopt a method of pumping through a transfer pipe provided with a heating means such as a heater and using a pump.
[0031]
According to the second embodiment, since the salt mixture after the electrolysis of the molten salt is transferred in a molten salt state, there is no need to cool and solidify the molten salt. Therefore, the time required for cooling and solidifying the molten salt can be reduced, and the speed of waste disposal can be increased.
[0032]
In the third embodiment, as shown in FIG. 3, instead of the attached salt heating and removing step in the first embodiment, a water washing step 14, a solid-liquid separation step 15, and an evaporation to dryness step 16 are provided in order. .
[0033]
In the water washing step 14, the salt mixture (NaF-NaCl) attached to the metal uranium deposited on the cathode in the molten salt electrolysis step 3 is dissolved in water and removed. In addition to water, an aqueous solution of an acid such as nitric acid, sulfuric acid, or hydrochloric acid can be used.
[0034]
Next, in a solid-liquid separation step 15, after the solid metal uranium and the washing liquid are separated by filtration, the filtrate is heated to evaporate water to precipitate and recover a salt mixture (evaporation and drying step 16). ). The recovered salt mixture is sent to the distillation step 5 as the attached salt 12 removed from the uranium metal, and is distilled in the same manner as the salt mixture sent through the solid salt transfer step 4 to be separated into NaF and NaCl. .
[0035]
Since the amount of the salt mixture adhering to the metallic uranium, which is the cathode deposit after the electrolysis of the molten salt, is very small, it can be removed by washing such as washing with water. According to the third embodiment, it is possible to remove the salt mixture adhering to the metal uranium by a simpler device and operation than by heating and evaporating, and to recover only the radioactive waste 11, that is, the metal uranium.
[0036]
In the fourth embodiment, as shown in FIG. 4, the cathode deposit 17 deposited on the cathode in the molten salt electrolysis step 3 is directly sent to the distillation step 5, where it is distilled and separated. The cathode deposit 17 is metallic uranium, on which the salt mixture used in the molten salt electrolysis step 3 is attached. In the distillation step 5, the cathode precipitate 17 is distilled in the same manner as the salt mixture fed through the solid salt transfer step 4 to remove the salt mixture adhering to the metal uranium, and at the same time, the salt mixture is adsorbed on the uranium. The agent NaF and the chloride (NaCl) added to it are separated.
[0037]
According to the fourth embodiment, the removal of the attached salt from the metal uranium can be performed in the same step and the same apparatus as the distillation step, and the economic efficiency is improved.
[0038]
Next, an embodiment of a device for treating waste from a nuclear fuel cycle facility will be described with reference to the drawings. FIG. 5 is a block diagram showing a first embodiment of the apparatus for treating waste from a nuclear fuel cycle facility according to the present invention.
[0039]
This processing apparatus heats a used uranium adsorbent while flowing a reducing gas to reduce the attached UF ₆ to UF ₄ , and a uranium adsorbent to which the UF ₄ reduced by the reduction apparatus 18 is attached. (NaF) is electrolyzed with a molten salt obtained by adding and mixing a chloride (NaCl) having a common cation with the molten salt, and a molten salt electrolysis apparatus 19 for precipitating and recovering metallic uranium at the cathode is provided. A distillation device 20 for distilling the decontaminated salt mixture to separate NaF and chloride NaCl as a uranium adsorbent, and a recycling line 21 for returning the NaCl distilled and separated by the distillation device 20 to the molten salt electrolysis device 19 And an attached salt removing device 22 for removing a salt mixture (NaF—NaCl) attached to metal uranium deposited on the cathode in the molten salt electrolyzing device 19, and a reducing device 1 And it has an off-gas treatment apparatus 23 for collecting and processing volatile gas generated in the molten salt electrolysis apparatus 19. Further, the apparatus has an attached salt recycling line 24 for supplying the salt mixture attached to the metal uranium removed by the attached salt removing device 22 to the distillation device 20.
[0040]
As shown in FIG. 6, the reducing device 18 includes a NaF storage container 25 that stores a used uranium adsorbent (NaF), and a reducing gas (H ₂ gas or H ₂ + Ar gas) in the NaF storage container 25. , A heater 27 for heating the NaF storage container 25, a temperature control device 28 for adjusting the heating temperature of the heater 27, and a gas evaporating and discharged from the NaF storage container 25. A trap NaF storage container 29 for absorbing and recovering volatile uranium fluoride (UF ₆ ) is provided. In the drawing, reference numeral 30 indicates a reducing gas cylinder, 31 indicates a reducing gas supply port, and 32 indicates a reducing gas discharge port.
[0041]
This reduction device 18 is used as described below. First, the used uranium adsorbent (NaF) is stored in the NaF storage container 25 and heated by the heater 27. At this time, the operation of the heater 27 is adjusted by the temperature control device 28, and the heating temperature is increased stepwise.
[0042]
The reducing gas (H ₂ gas or H ₂ + Ar gas) in the cylinder 30 is reduced while heating at 200 ° C., 300 ° C., 350 ° C., 400 ° C., 450 ° C., and 500 ° C. for 2 hours. The UF ₆ adsorbed on NaF is reduced to UF ₄ as shown in the following chemical reaction formula (1) by flowing the gas from the reactive gas supply port 31 into the NaF storage container 25. The generated hydrogen fluoride gas is discharged from the reducing gas discharge port 32 together with the unused H ₂ gas and Ar gas.
UF ₆ + H ₂ → UF ₄ + 2HF (1)
[0043]
In this way, the heating temperature is increased stepwise because the reduction of UF ₆ is only partially performed at a low temperature, and the reduction reaction does not proceed even if the heating time is increased. This is because UF ₆ sublimates before being reduced. At a low heating temperature, if the heating temperature is increased when the chemical reaction represented by the above formula (1) does not proceed, the chemical reaction of the formula (1) occurs again. First, the chemical reaction of the formula (1) is caused from a low heating temperature, and when the reaction does not proceed, the heating temperature is increased stepwise, whereby UF ₆ is reduced to UF ₄ without sublimation.
[0044]
UF _{6 has} a sublimation point of 56 ° C., whereas UF _{4 has} a very high boiling point of 1100 ° C., so that when UF ₆ is not reduced to UF ₄ , UF ₆ is volatilized (sublimated) by heating at the time of reduction. I do. The volatilized UF ₆ is adsorbed by the NaF in the NaF storage container 29 for trapping.
[0045]
When the UF ₆ adsorbed on the used uranium adsorbent (NaF) is reduced by the reducing device 18 and the amount of uranium in the NaF for trap is measured, the uranium amount is measured both before and after the reduction. It was below the limit. Thus, it was found that UF ₆ adsorbed on the used uranium adsorbent (NaF) was reduced to UF ₄ at a reduction rate of 100%.
[0046]
The molten salt electrolysis apparatus 19 includes, for example, an electrolytic cell made of, for example, graphite, in which a uranium adsorbent (NaF) and a chloride (NaCl) having a cation in common with the uranium adsorbent are mixed, and the molten salt is heated and melted. It has a solid electrode (anode and cathode) installed so as to be immersed in a molten salt therein, and a DC power supply for applying a DC voltage between these electrodes. When electricity is supplied by a DC power supply, the molten salt is electrolyzed, and uranium tetrafluoride (UF ₄ ) in the molten salt is reduced by the electrolysis, and is precipitated and recovered as uranium metal on the cathode.
[0047]
The distillation apparatus 20 distills the solidified salt mixture or the molten salt mixture transferred from the molten salt electrolysis apparatus 19 under normal pressure or reduced pressure to form a uranium adsorbent, NaF and chloride (NaCl). It works to separate. You can select and use those used in chemical engineering unit operations.
[0048]
The recycle line 21 has a function of returning the chloride (NaCl) distilled and separated by the distillation apparatus 20 to the molten salt electrolysis apparatus 19, and can use a transfer pipe or a conveyor type.
[0049]
The attached salt removing device 22 has a function of removing the salt mixture attached to the uranium deposited on the cathode of the molten salt electrolyzer 19 by, for example, heating and evaporating the salt mixture. The attached salt removing device 22 is provided with an attached salt recycling line 24 for feeding the removed salt mixture (adhered salt) to the distillation device 20.
[0050]
In the off-gas treatment device 23, volatile gas generated in the reduction device 18 and the molten salt electrolysis device 19, for example, hydrogen fluoride, vapor of molten salt (NaF—NaCl), vapor of Na, fluorine gas, chlorine gas and the like are collected. And processed. As the off-gas treatment device 23, a chemical adsorbent such as activated carbon, a high-temperature vapor trap, a reflux vapor trap, or the like can be used.
[0051]
According to the waste treatment apparatus of this embodiment, the chloride (NaCl) added to the uranium adsorbent (NaF) is distilled and separated by the distillation device 20, and the separated chloride passes through the recycling line 21. Since the molten salt is supplied to the molten salt electrolysis device 19 and reused, the amount of secondary waste generated can be significantly reduced. In addition, since it is not necessary to add the chloride for each step of the molten salt electrolysis, the operation cost can be reduced.
[0052]
In addition, since volatile gases such as HF gas, chlorine gas, and fluorine gas generated in the reduction device 18 and the molten salt electrolysis device 19 are treated by the off-gas treatment device 23, comprehensive waste including gas components is treated. Collection and processing can be performed sufficiently.
[0053]
【The invention's effect】
As is apparent from the above description, according to the method and apparatus of the present invention, in the treatment of fluoride salt waste to which radioactive substances such as uranium discharged from a nuclear fuel cycle facility are attached, the radioactive substances are added in the molten salt electrolysis step. Since the chloride having the same cation as the waste waste is separated in the distillation step and reused in the molten salt electrolysis step, the amount of secondary waste generated can be reduced. Further, since it is not necessary to add the chloride for each step of the molten salt electrolysis step, the operation cost can be reduced.
[0054]
Furthermore, since volatile gas generated in the reduction step and the molten salt electrolysis step is collected and processed in the off-gas processing step, collection and processing of waste including gas components is sufficiently performed.
[Brief description of the drawings]
FIG. 1 is a flowchart showing a first embodiment of a method for treating waste from a nuclear fuel cycle facility according to the present invention.
FIG. 2 is a flowchart showing a waste disposal method according to a second embodiment of the present invention.
FIG. 3 is a flowchart illustrating a waste disposal method according to a third embodiment of the present invention.
FIG. 4 is a flowchart illustrating a waste disposal method according to a fourth embodiment of the present invention.
FIG. 5 is a block diagram showing a first embodiment of a device for treating waste from a nuclear fuel cycle facility according to the present invention.
FIG. 6 is a diagram showing a structure of a reduction device used in the first embodiment of the waste treatment device.
FIG. 7 is a flowchart illustrating a conventional method of treating waste from a nuclear fuel cycle facility.
[Explanation of symbols]
1 ... used uranium adsorbent, 2 ... reduction step, 3 ... molten salt electrolysis step, 4 ... solid salt transfer step, 5 ... distillation step, 6 ... uranium adsorbent Chloride having the same cation as above, 7: Step of returning to molten salt, 8: Step of removing adhered salt by heating, 9: Off-gas treatment step, 10: Decontaminated uranium adsorbent, 11 ... radioactive waste, 12 ... attached salt, 13 ... molten salt transfer step, 14 ... washing step, 15 ... solid-liquid separation step, 16 ... evaporative drying step, 17 ... ... Cathode deposit, 18 ... Reduction device, 19 ... Molten salt electrolysis device, 20 ... Distillation device, 21 ... Recycling line, 22 ... Deposited salt removal device, 23 ... Offgas Processing equipment, 24: Adhered salt recycling line, 25: NaF storage container, 26: Supply of reducing gas Device 27 Heater 28 Temperature control device 29 NaF storage container for trap 30 Reducing gas cylinder 31 Reducing gas supply port 32 Reducing Gas outlet

Claims (11)

In a method for treating waste in which a chemical form to which radioactive substances emitted from a nuclear fuel cycle facility are attached is a fluoride salt,
A reduction step of reducing the radioactive material attached to the waste,
To the waste to which the radioactive substance reduced in the reduction step is attached, a molten salt of a salt mixture obtained by adding and mixing chloride having the same cation as the waste is electrolyzed, and the metal of the radioactive substance is deposited on a cathode. A molten salt electrolysis process for recovery;
A distillation step of distilling the salt mixture after the molten salt electrolysis step to separate the waste having a fluoride salt and the chloride having a common cation with the waste,
Returning the chloride separated in the distillation step into the molten salt in the molten salt electrolysis step,
An attached salt removing step of removing the salt mixture attached to the deposited and recovered metal of the radioactive substance,
A method for treating waste from a nuclear fuel cycle facility, comprising: an off-gas treatment step of treating volatile gas generated in the reduction step and the molten salt electrolysis step. 2. The method for treating waste from a nuclear fuel cycle facility according to claim 1, wherein in the reducing step, the waste to which the radioactive substance is attached is heated stepwise while flowing a reducing gas. 2. The method for treating waste from a nuclear fuel cycle facility according to claim 1, further comprising a solid salt transfer step of cooling and solidifying the molten salt after the molten salt electrolysis step and sending the solidified salt to the distillation step. The method for treating waste from a nuclear fuel cycle facility according to claim 1, further comprising a molten salt transfer step of transferring the molten salt after the molten salt electrolysis step to the distillation step in a molten state. 2. The nuclear fuel according to claim 1, wherein the attached salt removing step includes a heating / evaporating step of heating and evaporating and removing the salt mixture attached to the metal of the radioactive substance deposited and recovered. How to process waste from cycle facilities. A washing step of washing and removing the salt mixture adhering to the metal of the radioactive substance deposited and recovered by water, and removing the aqueous solution containing the salt mixture from the washing step with the radioactive salt; The method for treating waste from a nuclear fuel cycle facility according to claim 1, further comprising a solid-liquid separation step of separating the separated aqueous solution from the metal, and an evaporation-drying step of evaporating and drying the separated aqueous solution. The method for treating waste from a nuclear fuel cycle facility according to claim 5 or 6, wherein the salt mixture attached to the metal of the radioactive substance removed in the attached salt removing step is supplied to the distillation step. The heat evaporation step is performed simultaneously with the distillation step, and the salt mixture after the molten salt electrolysis step is distilled and separated, and the salt mixture attached to the metal of the radioactive substance is removed by distillation. Item 6. The method for treating waste from a nuclear fuel cycle facility according to Item 5. A reduction device that heats a waste in which the chemical form to which the radioactive material discharged from the nuclear fuel cycle facility adheres is a salt of fluoride while flowing a reducing gas, and reduces the radioactive material,
To the waste to which the radioactive substance reduced by the reducing device is attached, a molten salt of a salt mixture obtained by adding and mixing a chloride having the same cation as the waste is electrolyzed, and the metal of the radioactive substance is deposited on a cathode. A molten salt electrolysis device for recovery;
A distillation device for distilling the salt mixture decontaminated by the molten electrolysis device,
A recycling line that is distilled / separated by the distillation apparatus and has a chloride having a common cation with the waste, and is returned to the molten salt electrolysis apparatus.
An attached salt removal device that removes a salt mixture attached to the metal of the radioactive substance deposited and collected on the cathode in the molten salt electrolysis device,
An apparatus for treating waste from a nuclear fuel cycle facility, comprising: an apparatus for treating volatile gas generated by the reducing apparatus and the molten salt electrolysis apparatus. A washing apparatus for washing and removing the salt mixture adhering to the metal of the radioactive substance with water, and an aqueous solution containing the salt mixture discharged from the washing apparatus, the metal of the radioactive substance, 10. The apparatus for treating waste from a nuclear fuel cycle facility according to claim 9, further comprising: a solid-liquid separation device for separating the aqueous solution, and an evaporating and drying device for evaporating and drying the separated aqueous solution. The reducing device controls the heating temperature by the reducing gas supply device that supplies the reducing gas to the waste to which the radioactive substance is attached, the heating device that heats the waste, and the heating device. The apparatus for treating waste from nuclear fuel cycle facilities according to claim 9, further comprising a temperature controller.

Images