JP2001174590A - Treating method for radioactive waste - Google Patents

Abstract

PROBLEM TO BE SOLVED: To facilitate decontamination of a carbon fiber reinforced resin (CFRP) material, to eliminate the discharge of waste water, to reduce the volume of the CFRP, to recover a nuclear fuel material, or a deposit, by metal or oxide, and reuse it. SOLUTION: In this method for removing and recovering uranium or transuranic element from the CFRP material with nuclear fuel materials stuck thereto from nuclear fuel handling facilities, the contaminated CFRP 1 is reduced in a reduction process 5 and the reduced tetrafluoride CFRP material 6 is electrolyzed in an electrolysis process 7. In the electrolysis process 7, an organic substance dipped in molten salt and comprising CFRP is decomposed, the electrolysis is performed using a carbon fiber comprising the CFRP as a positive electrode in the molten salt to deposite and recovere the metallic uranium or the transuranic element metal 10 in the negative electrode.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for decontaminating a nuclear fuel material discharged from a nuclear fuel handling facility and a carbon fiber reinforced resin (hereinafter referred to as CFRP) material to which a mixture thereof has adhered by a molten salt electrolysis method. The present invention relates to a method for treating radioactive waste.

[0002]

2. Description of the Related Art Conventionally, as a method for decontaminating CFRP to which a nuclear fuel substance or a mixture thereof discharged from a nuclear fuel handling facility has been attached, for example, a method of removing the attached substance by washing the surface with an acid or alkali solution is known. It has been known. In addition, a method of burning in order to reduce the volume of CFRP material is known.

[0003]

However, FIG.
As shown in (a), on the surface of the target contaminated CFRP material 1, the deposit a composed of nuclear fuel material is attached.
FIG. 7 (a) shows the carbon fiber b constituting the CFRP material 1.
As is clear from FIG.
The deposit d has entered between the carbon fiber near the surface and the CFRP substrate c.

[0004] It is difficult to remove the attached deposit d by washing. That is, as shown in FIG. 8, on the surface of the CFRP material 1a after cleaning, there is a problem that the deposit d that has entered between the carbon fiber b near the surface and the CFRP base material c remains. is there.

Another problem is that a large amount of waste liquid is generated after decontamination. Furthermore, the method of incinerating and reducing the volume of the CFRP material 1 while contaminating it with adhering substances has a problem that it is difficult to recover nuclear fuel material from incinerated ash.

SUMMARY OF THE INVENTION The present invention has been made to solve the above-mentioned problems. For example, it is possible to easily decontaminate CFRP material discharged from a nuclear fuel handling facility, and there is no discharge of waste liquid. At the same time, the volume of CFRP material can be reduced,
It is an object of the present invention to provide a method for treating radioactive waste in which nuclear fuel material, which is a deposit, can be recovered and reused in the form of metal or oxide.

[0007]

The invention according to claim 1 is a method for treating a radioactive waste for removing and recovering uranium or transuranium element from a CFRP material to which a nuclear fuel substance or a mixture thereof is attached, CFRP immersed in molten salt
Decomposing the organic matter that constitutes the material and electrolyzing the carbon fiber that constitutes the CFRP material in the molten salt as an anode, thereby precipitating and collecting metal of uranium metal or transuranium element on the cathode. It is characterized by. According to the present invention, nuclear fuel substances such as uranium and transuranium attached to the CFRP material can be easily recovered in the form of metal or oxide.

[0008] The invention of claim 2 is to provide a nuclear fuel material adhered to the CFRP material, uranium hexafluoride or transuranium hexafluoride at room temperature hydrogen gas or an inert gas containing hydrogen gas or phosgene. Or the like, and is reduced to uranium tetrafluoride or transuranium element tetrafluoride by contact with a reducing gas such as

According to the present invention, the CFRP material to be decontaminated can be reduced to uranium tetrafluoride by a reducing agent at a temperature at which uranium hexafluoride does not evaporate. The invention according to claim 3 is that, after immersing the container loaded with the CFRP material to which the nuclear fuel substance is attached in a molten salt, the container is electrically connected to an anode terminal, and a metal solid cathode is used as a cathode for depositing and recovering the uranium. A liquid metal cathode is used as a cathode for recovering the transuranium element, and these cathodes are immersed in the molten salt, and a current is applied between the anode and the cathode to cause uranium adhered to the carbon fiber reinforced resin. Alternatively, after dissolving the transuranium element in the molten salt by an anodic dissolution reaction to form chloride ions, reducing the dissolved uranium or the chloride ions of the transuranium element at a cathode to precipitate and recover as a metal. Features.

According to the present invention, uranium tetrafluoride adhering to the CFRP material is electrolyzed in the molten salt, and uranium metal can be deposited in the cathode. Further, uranium in a metal state can be taken out from the molten salt and recovered.

According to a fourth aspect of the present invention, the cathode material for precipitating and recovering uranium is solid at the temperature at the time of electrolysis and made of a metal such as iron or molybdenum which does not react with molten salt or uranium. Alternatively, the cathode material for recovering uranium and transuranium elements is characterized by being composed of a liquid metal that is liquid at the temperature during electrolysis. According to the present invention, CFRP
Easily remove uranium and transuranium elements from materials
It can be recovered and gasified organic matter of CFRP material, and its waste can be reduced in volume.

According to a fifth aspect of the present invention, the chemical form of the molten salt is lithium chloride, sodium chloride, potassium chloride, magnesium chloride, calcium chloride, aluminum chloride,
Chlorides such as tin chloride or their mixed chlorides, lithium fluoride, sodium fluoride, potassium fluoride, magnesium fluoride, calcium fluoride, aluminum fluoride,
Fluoride or mixed fluoride such as tin fluoride, lithium hydroxide, sodium hydroxide, potassium hydroxide, magnesium hydroxide, calcium hydroxide, aluminum hydroxide, tin hydroxide or other hydroxide or mixed hydroxide, carbonic acid lithium,
Carbonates or mixed carbonates such as sodium carbonate, potassium carbonate, magnesium carbonate, calcium carbonate, aluminum carbonate and tin carbonate; or nitrates such as lithium nitrate, sodium nitrate, potassium nitrate, magnesium nitrate, calcium nitrate, aluminum nitrate, tin nitrate and the like It is characterized by a mixed nitrate thereof. According to the present invention, the operation temperature of the process can be lowered by lowering the melting point temperature of the molten salt, and the operation can be facilitated.

According to a sixth aspect of the present invention, the molten salt is 300 ° C.
It is characterized by having a wide potential range in which it can be stably electrolyzed without being decomposed when an electric current is applied by immersing an anode and a cathode by melting at ~ 1200 ° C. According to the present invention, it is possible to stably electrolyze the anode and the cathode without being decomposed when the electric current flows by immersing the anode and the cathode in the molten salt, and to use a molten salt having a wide potential range.

According to a seventh aspect of the present invention, the oxidizing gas is blown into the molten salt to oxidatively decompose organic substances reinforcing the carbon fiber reinforced resin material into carbon dioxide and water. Is used as an anode for electrolysis.

According to the present invention, all the nuclear fuel substances adhering to the CFRP material in the oxidation step can be converted into the chemical form of oxide. The conversion of the oxide allows the oxide to be easily dissolved in the molten salt as an acid chloride.

[0016]

DESCRIPTION OF THE PREFERRED EMBODIMENTS (First Embodiment) (Claim 1)
Fig. 1 is a flowchart (flow chart) showing a method for treating CFRP in radioactive waste discharged from a nuclear fuel handling facility.
It is. CFRP discharged from a nuclear fuel handling facility will adhere to nuclear fuel materials and their mixtures in the chemical form of oxides, chlorides, fluorides, hydrides, or nitrates or sulfates.

In the cutting step 2, the CFRP material 1 to which the nuclear fuel material adheres and is contaminated is cut into a size which can be easily processed. Among the adhered substances of the cut CFRP material 3, hexafluoride is particularly volatile, and it is difficult to perform off-gas treatment and recovery of hexafluoride itself in the treatment process. The reaction reduces to non-volatile tetrafluoride.

When hydrogen is used as the reducing agent 4, the excess fluorine due to the reaction becomes HF30. Uranium and transuranium elements (T
RU) -adhered CFRP material 6 adhering to electrolysis step 7
Sent to In the electrolysis step 7, first, the organic material of the constituent material is immersed in a high-temperature molten salt and thermally decomposed into an organic substance decomposition gas 9.

At this time, CFRP is only carbon fiber to which tetrafluoride is adhered. Therefore, the carbon fiber is used as an anode, and the tetrafluoride is electrochemically dissolved in a molten salt as a cation, and then the electrochemically To precipitate and recover as metal uranium or metal TRU10. In addition, the carbon fiber 8 in a state where the nuclear fuel substance attached to the surface has been removed remains on the anode.

According to the present embodiment, the contaminated CFRP material 1
Uranium and transuranium elements adhering to the surface can be easily removed and recovered, and the organic matter in the CFRP material is an organic matter decomposition gas 9
Therefore, the carbon fiber 8 is the only decontaminated waste, and the volume of CFRP material waste can be reduced. The recovered uranium or transuranium element can be recovered in the form of a metal that can be easily reused.

(Second Embodiment) (corresponding to claim 2) A second embodiment according to the present invention will be described. In the present embodiment, in the reduction step 5 in the first embodiment, CFRP to which uranium or transuranium element to be decontaminated is attached
Uranium hexafluoride is reduced to uranium tetrafluoride by flowing an inert gas such as hydrogen gas, hydrogen-containing argon gas or the like, or a reducing gas such as phosgene to the material 1.
It is important that the reaction is performed at a temperature of 70 ° C. or lower, which is a temperature at which uranium hexafluoride does not evaporate.

The reactions occurring in these processes are as follows. UF + H ₂ = UF ₄ + 2HF After the reduction is completed, the uranium or transuranium element is attached to the tetrafluoride-adhered CFRP material 6 in the form of UF ₄ or TRUF ₄ . According to this embodiment, tetrafluoride can be generated at a low temperature.

(Third Embodiment) (corresponding to claim 3) A third embodiment of the present invention will be described with reference to FIGS. FIG. 2 shows an example of the electrolysis step 7 in the first embodiment.

In FIG. 2, reference numeral 11 denotes an electrolytic cell.
A molten salt 12 is accommodated in 11. An anode basket 13 and a solid cathode 14 connected to a DC power supply 16 are immersed in the molten salt 12. An exhaust line 15 is connected to the upper end of the electrolytic cell 11, and the exhaust line 15 communicates with a gas space 31 in the electrolytic cell 11.
In FIG. 2, reference numeral 10 denotes metal uranium or metal TRU (transuranium element) deposited on the solid cathode 14.

Referring to FIG. 1, uranium or transuranium element (T
The tetrafluoride-adhered CFRP material 6 to which RU is attached is loaded in the anode basket 13 shown in FIG. 2 made of graphite such as pyrographite, tantalum, tungsten, carbon steel or stainless steel.

In this state, at least one of an alkali metal chloride, an alkaline earth metal chloride, an alkali metal fluoride or an alkaline earth metal fluoride in a molten state,
Alternatively, the constituent elements of the nuclear fuel material compound to be treated, the chloride or fluoride of the attached nuclear fuel material,
Alternatively, a high-temperature molten salt electrolyte 12 comprising a mixture thereof
(Hereinafter, referred to as a molten salt or simply a salt).

Since the temperature of the molten salt 12 is as high as 500 ° C. or more, the organic substance constituting the CFRP material is thermally decomposed into an organic substance decomposed gas 9, from which dry air in the electrolytic cell 11 or non-decomposed gas is removed. The gas is discharged from the exhaust means 15 connected to the electrolytic cell 11 through the gas space 31 filled with the active gas.

Carbon fiber, which is a constituent material of CFRP, to which the nuclear fuel material which has become tetrafluoride adheres remains in the anode basket 13. At this time, the anode basket 13 is used as an anode, and a current is supplied from the DC power supply 16 to the solid cathode 14 to perform electrolysis. According to the present embodiment, the nuclear tetrafluoride nuclear fuel material adhering to the carbon fiber surface is dissolved in the molten salt 12 electrochemically. Uranium of the nuclear fuel material dissolved in the molten salt is precipitated as the metallic uranium 10 on the solid cathode 14. After passing a predetermined current, the solid cathode 14 on which the metal uranium 10 is deposited
Can be taken out from the molten salt 12 in a molten state and recovered.

(Fourth Embodiment) (corresponding to claim 4) Next, a fourth embodiment of the present invention will be described with reference to FIG. In the present embodiment, the cathode material for depositing and recovering uranium is solid at the temperature at the time of electrolysis and made of a metal such as iron or molybdenum that does not react with molten salt or uranium. Further, the cathode material for recovering uranium or transuranium element is made of liquid metal which is liquid at the temperature during electrolysis.

That is, this embodiment uses a liquid metal cathode 33 instead of the solid cathode 14 installed in the electrolytic cell 11 described in the third embodiment, and uses only a transuranium element or uranium and transuranium. The purpose is to deposit and collect elements on the cathode as metals. In FIG. 3, the same portions as those in FIG. 2 are denoted by the same reference numerals, and the description of the overlapping portions will be omitted.

In the present embodiment, the liquid metal 34 is immersed in the high-temperature molten salt 12 while being loaded in the cathode crucible 17, and is held in the molten salt by the liquid metal cathode support mechanism 35. A cathode conductor 32 electrically insulated from the molten salt 12 by an insulating material 19 such as ceramics is immersed in the liquid metal.

After a predetermined current is supplied from the DC power supply 16 to the anode basket 13 and the liquid metal cathode 33, the liquid metal 34 is taken out of the molten salt 12 together with the cathode crucible 17 and collected. It is important that the liquid metal 34 be a substance that melts at the temperature of the molten salt 12, and cadmium, bismuth, or zinc is suitable.

According to this embodiment, uranium and transuranium elements adhering to the surface of the CFRP material can be easily removed and recovered, and the organic matter of the CFRP material is gasified. Is only carbon fiber and can reduce the volume of CFRP waste. The recovered uranium and transuranium elements can be recovered in the form of a metal that can be easily reused.

(Fifth Embodiment) (corresponding to claims 5 and 6) Next, a fifth embodiment of the present invention will be described with reference to FIGS. In this embodiment, FIGS.
As the molten salt 12, a mixed salt of sodium chloride and sodium fluoride is used. Further, the following mixed salt can be used instead of the mixed salt of sodium chloride and sodium fluoride.

That is, chlorides or mixed chlorides such as lithium chloride, sodium chloride, potassium chloride, magnesium chloride, calcium chloride, aluminum chloride, tin chloride, lithium fluoride, sodium fluoride, potassium fluoride, magnesium fluoride; Fluoride or mixed fluoride such as calcium fluoride, aluminum fluoride and tin fluoride, water such as lithium hydroxide, sodium hydroxide, potassium hydroxide, magnesium hydroxide, calcium hydroxide, aluminum hydroxide and tin hydroxide Oxides or mixed hydroxides, lithium carbonate, sodium carbonate, potassium carbonate, magnesium carbonate, calcium carbonate, aluminum carbonate,
Carbonate or mixed carbonate such as tin carbonate, or nitrate or mixed nitrate such as lithium nitrate, sodium nitrate, potassium nitrate, magnesium nitrate, calcium nitrate, aluminum nitrate, tin nitrate and the like.

These mixed salts are melted at 300 to 1200 ° C., and the anode and the cathode are immersed in the molten salt to allow stable electrolysis without decomposition when an electric current is applied (potential window).
However, mixed salts having a wide range can be used.

With respect to combinations of these molten salts, chloride + fluoride, chloride + carbonate, fluoride + carbonate, hydroxide + carbonate, nitrate + carbonate, chloride + nitrate, fluoride + Nitrate can be applied. Note that combinations of chloride + hydroxide, fluoride + hydroxide, nitrate + hydroxide cannot be performed.

The reason for selecting the nitrate is as follows. That is, since the nitrate has a low melting point and can lower the operating temperature of the process using the molten salt, there is an advantage that the operation becomes easy and the range of selection of the structural material of the apparatus is widened.

The reason why the temperature range of the molten salt was selected to be 300 to 1200 ° C. is that if the temperature is lower than 300 ° C., there is no molten salt having a wide potential window when practical, and if it exceeds 1200 ° C.
This is because there is no device material capable of containing a molten salt of 1200 ° C. or higher.

(Sixth Embodiment) (corresponding to claim 7) A sixth embodiment of the present invention will be described with reference to FIG. The present embodiment is a method for treating radioactive waste in which organic matter in CRFP, which is radioactive waste, is oxidatively decomposed in a molten salt, and then carbon fibers are decontaminated. In FIG. 4, the same portions as those in FIG. 1 are denoted by the same reference numerals, and the description of the overlapping portions will be omitted.

This embodiment is different from the first embodiment in that an oxidation step 20 is provided instead of the reduction step 5, and a chlorination dissolution step 25 is provided after the oxidation step 20. That is, this embodiment is based on the first embodiment, and the contaminated CFRP material 1 discharged from the nuclear fuel handling facility
Has attached thereto a nuclear fuel material whose chemical form is oxide, chloride, fluoride, hydride, or nitrate or sulfate, and a mixture thereof, and cuts the contaminated CFRP material 1 by the attachment of these nuclear fuel materials. In step 2, it is cut into a size that can be easily processed.

In the oxidation step 20, the cut CFRP material 3 is
Low melting point molten salt such as sodium hydroxide (NaOH)
Or sodium hydroxide (NaOH) and sodium carbonate
(Na_TwoCO_ThreeD) Nuclear fuel
The CRFP structural material other than the carbon fiber
As an oxidizing agent in the mixed molten salt 22
Oxygen gas 21 is blown. The carbon dioxide generated by this,
NO_x, SO_xTo Na _TwoCO_Three, NaNO_Two, Na_TwoS
O_FourImmobilized in molten salt as 23.

When oxidatively decomposing organic substances in the oxidation step 20,
All chemical forms of the nuclear fuel material attached to the CFRP material 1 are converted to oxides. The oxide-adhered carbon fiber 24 to which the oxide of the nuclear fuel substance obtained from the oxidation step 20 is attached is sent to a chlorination dissolution step 25.

In the chlorination dissolution step 25, the molten salt 22 is immersed and dissolved in a mixed molten salt different from the mixed molten salt 22. That is, this mixed molten salt is sodium chloride and sodium fluoride, lithium chloride, sodium chloride, potassium chloride, magnesium chloride, calcium chloride, aluminum chloride, cesium chloride, tin chloride and the like or a mixed chloride, lithium fluoride, Sodium fluoride, potassium fluoride, magnesium fluoride, calcium fluoride, aluminum fluoride,
Fluoride or mixed fluoride such as tin fluoride, lithium hydroxide, sodium hydroxide, potassium hydroxide, magnesium hydroxide, calcium hydroxide, aluminum hydroxide, tin hydroxide or other hydroxide or mixed hydroxide, carbonic acid lithium,
Carbonate or mixed carbonate such as sodium carbonate, potassium carbonate, magnesium carbonate, calcium carbonate, aluminum carbonate, tin carbonate, or nitrate such as lithium nitrate, sodium nitrate, potassium nitrate, magnesium nitrate, calcium nitrate, calcium nitrate, aluminum nitrate, tin nitrate or It is a mixed nitrate that melts at 300-1200 ° C and does not decompose during electrolysis and has a wide potential window.

In the mixed molten salt, carbon oxide-adhered carbon fibers 24
After immersion, the chlorine gas 27 is reacted with the molten salt to dissolve the oxide as an acid chloride in the molten salt. The reaction at this time is as follows when the oxide of the nuclear fuel material is UO ₂ . UO ₂ + C (carbon fiber 24) + 2Cl ₂ → CO ₂ ↑ + UC
l ₄

The case where the oxide of the nuclear fuel material is TRUO ₂ is as follows. TRUO ₂ + C (carbon fiber 24) + 2Cl ₂ → CO ₂ ↑ +
TRUCl ₄ According to the present embodiment, tetravalent ions of the nuclear fuel substance dissolved in the molten salt are subjected to electrolysis in the same molten salt as in the chlorination dissolution step in the electrolysis step 7, also using the carbon fiber 24 itself as the anode. It can be recovered as metal uranium or metal TRU10 on the cathode. In this case, chlorine gas 29 is generated from the anode.

Next, FIG. 5 and FIG. 6 show the seventh embodiment of the present invention.
An embodiment will be described. In FIG. 5, reference numeral 36 denotes a chlorination dissolving tank, to which a chlorine gas introduction line 37 and an exhaust line 38 are connected, in which the molten salt 12 is housed. The basket 13 and the lower part of the chlorine gas introduction line 37 are immersed. In FIG. 5, reference numeral 27 denotes chlorine gas, 28 denotes CO ₂ , and 39 denotes a gas space. The present embodiment relates to an apparatus for performing the chlorination dissolution step 25 and the electrolysis step 7 in the sixth embodiment.

Mixed molten salt 12 used in chlorination dissolution step 25
Then, the carbon fibers 24 loaded with the oxide of the nuclear fuel material loaded on the anode basket 13 are immersed. Chlorine gas 27 is introduced into the molten salt 12 from a chlorine gas introduction means 37 connected to a chlorination dissolution tank 36, and the chlorine gas 27 is the carbon fiber 24 to which the oxide of the nuclear fuel substance is attached and the nuclear fuel to which the oxide is attached. Reacts with oxides of matter.

The CO ₂ gas 28 generated by the reaction is discharged through the exhaust line 38 to the outside of the chlorination dissolution tank 36. The gas space 39 in the chlorination dissolution tank 36 has a dry air atmosphere or an inert gas atmosphere.

FIG. 6 shows an apparatus for performing the electrolysis step 7. 6, the same portions as those in FIG. 2 are denoted by the same reference numerals, and the description of the overlapping portions will be omitted. The electrolyzer may be the same as the chlorination and dissolution apparatus.

That is, in FIG. 6, the anode basket 13 immersed in the molten salt 12 is used as an anode, and a current is supplied from a DC power supply 16 to the solid cathode 14 to perform electrolysis. The uranium metal 10 is precipitated and recovered in the form of metal. When recovering TRU or uranium and TRU, a liquid metal cathode 33 having the same structure as that shown in FIG.

In each of the above embodiments, when the CFRP material to which the nuclear fuel material such as uranium or transuranium element is attached is immersed in the molten salt, the oxide of the nuclear fuel material is dissolved in the molten salt. And may precipitate in the form of oxides. Therefore, a container for collecting oxides of nuclear fuel material that can be taken out of the molten salt is provided at the bottom of the container holding the molten salt. The collection container may be, for example, one having a large number of holes on the side surface, or a mesh on the side surface.

[0053]

According to the present invention, the decontamination of CFRP material, which has been difficult to decontaminate, can be facilitated, and no or little waste liquid is discharged, and the CFRP material is also decomposed. It is possible to reduce waste volume. Further, the nuclear fuel material as the deposit can be recovered in the form of metal or oxide, and can be reused.

[Brief description of the drawings]

FIG. 1 is a flow chart for explaining first to third embodiments of the method for treating radioactive waste according to the present invention.

FIG. 2 is a longitudinal sectional view schematically showing a part of an apparatus for explaining a third and fifth embodiments of the radioactive waste treatment method according to the present invention.

FIG. 3 is a longitudinal sectional view schematically showing a part of an apparatus for explaining fourth and fifth embodiments of the radioactive waste treatment method according to the present invention.

FIG. 4 is a flow chart for explaining a sixth embodiment of the method for treating radioactive waste according to the present invention.

FIG. 5 is a longitudinal sectional view schematically showing a part of an apparatus for explaining a seventh embodiment of the method for treating radioactive waste according to the present invention.

FIG. 6 is a longitudinal sectional view schematically showing an apparatus during an electrolysis step in a seventh embodiment.

FIG. 7 (a) is a longitudinal sectional view conceptually showing a CRFP material in a state where a nuclear fuel substance is attached, for explaining a problem of a conventional example;
FIG. 2B is an enlarged partial cross-sectional view illustrating a portion A of FIG.

FIG. 8 is a longitudinal sectional view conceptually showing the CFRP material in a state where the nuclear fuel material after cleaning with an acid or the like is attached in FIG.

[Explanation of symbols]

1 ... contaminated CFRP material, 2 ... cutting process, 3 ... cut CFRP
Material, 4 ... Reducing agent, 5 ... Reducing process, 6 ... CF4 with CF
RP material, 7: electrolysis process, 8: carbon fiber, 9: organic matter decomposition gas, 10: metal uranium or metal TRU, 11: electrolytic cell, 12 ...
Molten salt, 13… Anode basket, 14… Solid cathode, 15… Exhaust line, 16… DC power supply, 17… Cathode crucible, 19… Insulation material,
20: oxidation step, 21: oxidizing agent, 22: mixed molten salt (NaO
H, NaCO ₃ ), 23 ... NaCO ₃ , NaNO ₂ , Na
₂ SO ₄ , 24: carbon fiber to which oxide of nuclear fuel substance is attached, 25: chlorination dissolution step, 26: carbon fiber, 27: chlorine gas, 28: CO ₂ , 29: chlorine gas, 30: hydrogen fluoride, 31…
Gas space, 32: Cathode conductor, 33: Liquid metal cathode, 34: Liquid metal, 35: Liquid cathode support mechanism, 36: Chlorination dissolving tank, 37 ...
Chlorine gas introduction line, 38… exhaust line, 39… gas space,
a: deposits composed of nuclear fuel material, b: carbon fiber, c: CF
RP substrate, d: attached matter.

Continuing from the front page (72) Inventor Nakamura et al. 2-1 Ukishima-cho, Kawasaki-ku, Kawasaki-ku, Kawasaki, Kanagawa Prefecture Inside the Toshiba Hamakawasaki Plant (72) Inventor Kazuhiro Utsunomiya 8-8 Shinsugita-cho, Isogo-ku, Yokohama-shi, Kanagawa 4D004 AA50 AB09 AC05 BA05 CA02 CA27 CA29 CA36 CA37 CA44 CB04 CB12 CB31 CC01 CC11 DA06 4K058 AA21 BA01 CB03 CB04 CB05 CB06 CB12 CB17 EB13 EC07 FA21

Claims (7)

[Claims] 1. A method for treating a radioactive waste in which a uranium or transuranium element is removed and recovered from a carbon fiber reinforced resin material to which a nuclear fuel substance or a mixture thereof is attached, wherein the carbon fiber reinforced resin material is immersed in a molten salt. Decomposes the organic material constituting the carbon fiber reinforced resin material, and electrolyzes the carbon fiber constituting the carbon fiber reinforced resin material as an anode in the molten salt, so that metal uranium or transuranium is formed on the cathode. A method for treating radioactive waste, comprising depositing and recovering elemental metals. 2. Among the nuclear fuel substances attached to the carbon fiber reinforced resin material, uranium hexafluoride or transuranium element hexafluoride is converted to hydrogen gas or an inert gas containing hydrogen gas or phosgene at room temperature. 2. The method for treating radioactive waste according to claim 1, wherein the reduction is performed by contacting with uranium tetrafluoride or transuranium element by reducing gas. 3. A container loaded with the carbon fiber reinforced resin material to which the nuclear fuel material is attached is immersed in a molten salt, and then electrically connected to an anode terminal, and a metal solid is deposited on a cathode for precipitating and recovering the uranium. Using a cathode, a liquid metal cathode is used as the cathode for recovering the transuranium element, these cathodes are immersed in the molten salt, and a current is applied between the anode and the cathode to adhere to the carbon fiber reinforced resin material. After dissolving the uranium or transuranium element in the molten salt by anodic dissolution reaction to form chloride ions, the dissolved uranium or transuranium element chloride ions are reduced at the cathode to precipitate and recover as metal. The method for treating radioactive waste according to claim 1, wherein: 4. The cathode material for depositing and recovering uranium is solid at the temperature during electrolysis and made of a metal such as iron or molybdenum that does not react with molten salt or uranium. The method for treating radioactive waste according to claim 3, wherein the cathode material for recovering the element is made of a liquid metal which is a liquid at the temperature during electrolysis. 5. The chemical form of the molten salt is chloride such as lithium chloride, sodium chloride, potassium chloride, magnesium chloride, calcium chloride, aluminum chloride, tin chloride, or a mixture thereof, lithium fluoride, fluoride, and the like. Sodium, potassium fluoride, magnesium fluoride,
Fluoride or mixed fluoride such as calcium fluoride, aluminum fluoride, tin fluoride, lithium hydroxide, sodium hydroxide, potassium hydroxide, magnesium hydroxide, calcium hydroxide, aluminum hydroxide, tin hydroxide, etc. Or mixed hydroxide, carbonate or mixed carbonate such as lithium carbonate, sodium carbonate, potassium carbonate, magnesium carbonate, calcium carbonate, aluminum carbonate, tin carbonate, or lithium nitrate, sodium nitrate, potassium nitrate, magnesium nitrate, calcium nitrate 4. The method for treating radioactive waste according to claim 3, wherein the nitrate is a nitrate such as aluminum nitrate, tin nitrate or the like or a mixed nitrate thereof. 6. The molten salt melts at 300 ° C. to 1200 ° C., and has a wide potential range in which stable electrolysis can be performed without decomposition when an anode and a cathode are immersed and an electric current is applied. The method for treating radioactive waste according to claim 3. 7. An oxidizing gas is blown into the molten salt to oxidize and decompose organic substances reinforcing the carbon fiber reinforced resin material into carbon dioxide and water, and then perform electrolysis using the remaining carbon fibers as an anode. The method for treating radioactive waste according to claim 1, wherein: