JP2017026536A - Radioactive waste solidification method - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a radioactive waste solidification method capable of shortening a storage period of a vitrified substance.SOLUTION: In S1A, an acid aqueous solution is injected into contaminated water containing a long-half-life radionuclide (U-238) and a short-half-life radionuclide (Cs-137) within a radioactive waste vessel, thereby reducing a pH of the contaminated water and ionizing the long-half-life radionuclide. In S1B, the pH of the contaminated water is increased by injecting an alkali aqueous solution into the contaminated water, and the long-half-life radionuclide is separated from the contaminated water by electrodeposition. In S2, the contaminated water within the radioactive waste vessel is supplied to a solidification vessel, and the short-half-life radionuclide in the contaminated water is adsorbed by an adsorbent within the solidification vessel. After discharging the contaminated water, low-melting glass is supplied into the solidification vessel. In S3, the solidification vessel filled with adsorbent adsorbing the short-half-life radionuclide and the low-melting glass is stored in a heat insulation vessel, and low-melting glass is molten with heat generated by a radiation from the short-half-life radionuclide and a vitrified substance is produced.SELECTED DRAWING: Figure 2

Description

  The present invention relates to a radioactive waste solidification processing method, and more particularly to a radioactive waste solidification processing method suitable for processing high-dose radioactive waste with a high radioactivity level.

  Radioactive waste generated from nuclear facilities and the like is solidified with cement or glass and converted into a form suitable for storage, transportation and disposal. Among these solidification treatments, cement solidification is a method of solidifying radioactive waste using cement and water, and is characterized by being inexpensive and easy to treat. However, when radioactive waste with a high radioactivity level is solidified with cement, the water contained in the produced solidified material is radioactively decomposed to generate hydrogen gas, and this hydrogen is embedded in the solidified material itself or embedded in it. There is a possibility that the facility after disposal may be affected (see JP 2007-132787 A).

  In vitrification without using water, there is no concern that hydrogen gas is generated even with radioactive waste having a high radioactivity level. However, as shown in, for example, Japanese Patent Application Laid-Open No. 2011-46996, vitrification requires processing at a high temperature and requires a large-scale melting facility.

  In the method for solidifying radioactive waste described in Japanese Patent Application Laid-Open No. 62-124499, solid or liquid radioactive waste is mixed with low-melting glass (softening point is 400 ° C. to 800 ° C.). Molded and fired or melted to produce a solidified body. Japanese Patent Application Laid-Open No. 62-165198 also describes a hydrothermal solidification treatment method for high-level radioactive waste in which high-level radioactive waste is solidified using low-melting glass. In JP-A-62-165198, a low-melting glass is melted by the decay heat of high-level radioactive waste to solidify the high-level radioactive waste.

  When low-melting glass is used as the solidifying material, the low-melting glass is easily melted at a low temperature, so a large-scale melting facility is unnecessary, and it is considered to be the most suitable method for solidifying high-dose radioactive waste. .

  Japanese Patent Laid-Open No. 8-271692 removes radionuclides from radioactive liquid waste containing sodium nitrate generated in a spent fuel reprocessing facility, and reuses acid (nitric acid) and alkali (sodium hydroxide) from the radioactive liquid waste The method for treating the radioactive liquid waste collected for the purpose is described. In JP-A-8-271692, as a technique for removing radionuclides, coprecipitation removal in which radionuclides contained in radioactive liquid waste are coprecipitated with a coprecipitation agent added to the radioactive liquid waste and removed as a precipitate, And radionuclide adsorption / removal in which the radionuclide contained in the radioactive liquid waste is adsorbed and removed by an ion exchanger. The recovery of acid and alkali is performed by decomposing neutral salt contained in the waste liquid after removing the radionuclide into acid and alkali by electrodialysis.

  In the method for purifying spent nuclear fuel containing uranium and plutonium described in JP-A-1-237497, a cadmium molten bath is present in the lower part of the electrolytic cell, and an electrolyte bath is present on the cadmium molten bath. An anode basket filled with spent nuclear fuel (uranium and plutonium) to be refined is inserted in the electrolyte layer into the electrolyte bath and further toward the cadmium melting bath. In the electrolyte bath, uranium and plutonium are dissolved from the anode basket into the cadmium melt bath. The cathode is immersed in an electrolyte bath, a voltage is applied between the anode basket and the cathode, and uranium and plutonium dissolved in the cadmium molten bath are deposited on the cathode.

  As an adsorbent that adsorbs both cesium and strontium, a titanium silicate compound, for example, crystallized silicotitanate (CST) is known (see Japanese Patent Application Laid-Open No. 2014-29269 and Japanese Patent No. 5285183).

JP 2007-132787 A JP 2011-46996 A JP 62-124499 A JP-A-62-165198 JP-A-8-271692 JP-A-1-237497 JP 2014-29269 A Japanese Patent No. 5285183

  As shown in Japanese Patent Application Laid-Open No. 2011-46996, vitrification requires high-temperature treatment using a large-scale melting facility, so when targeting the vitrification of short half-life nuclides contained in spent nuclear fuel, It is considered that the vitrification method using the low melting point glass that can be maintained for about 300 years and can be melted with a simple facility is suitable in terms of cost because of the nuclide.

  However, radioactive liquid waste containing radioactive nuclides generated at nuclear facilities, for example, contaminated water containing radioactive nuclides, is not only high-dose, short half-life nuclide such as Cs, but also high-dose, long half-life such as uranium. May contain radionuclides. When long-lived nuclides (for example, uranium and plutonium) are mixed in vitrified solids of radioactive waste produced using low-melting glass, the solidified performance cannot be maintained for tens of thousands or hundreds of millions of years. The inventors thought. Therefore, the inventors have come to recognize that long half-life nuclides need not be mixed into the low melting point glass solidification process in order to establish solidification with the low melting point glass.

  The objective of this invention is providing the solidification processing method of the radioactive waste which can shorten the storage period of a vitrified body.

The feature of the present invention that achieves the above-described object is to separate the long half-life nuclide from the radioactive liquid waste containing the long half-life nuclide and the short half-life nuclide,
The long half-life nuclide is removed, the radioactive waste liquid containing the short half-life nuclide is brought into contact with the adsorbent, the short half-life nuclide is adsorbed on the adsorbent, and the short half-life nuclide is separated from the radioactive liquid waste.
Supply the glass raw material into the first container filled with the adsorbent, which is a radioactive waste, adsorbing the short half-life nuclide,
The glass raw material in the first container is melted by heating to produce a vitrified body of the adsorbent.

  Long half-life nuclides were removed from radioactive liquid waste containing long half-life nuclides and short half-life nuclides, and then short half-life nuclides contained in radioactive liquid waste were adsorbed and removed to adsorb short half-life nuclides. Since the adsorbent is solidified using the glass raw material, the storage period of the vitrified glass produced using the glass raw material can be shortened.

  According to the present invention, the storage period of the vitrified body can be shortened.

It is a flowchart which shows the process sequence in the solidification processing method of the radioactive waste of Example 1 which is one suitable Example of this invention. It is a flowchart which shows the detailed process sequence of the process sequence of the solidification processing method of the radioactive waste of Example 1 shown in FIG. It is explanatory drawing which shows the specific content of each process of removing the long half life nuclide and short half life nuclide contained in a radioactive waste liquid in the process sequence shown in FIG. It is explanatory drawing which shows the solidification process by the low melting glass of the adsorbent which adsorb | sucked the short half life nuclide in the process sequence shown in FIG. It is the electric potential-pH figure of uranium which is a long half life nuclide. It is an electric potential-pH figure of cesium which is a short half life nuclide. It is a characteristic view which shows a time-dependent change of the radioactivity in each radionuclide. It is explanatory drawing which shows each process of isolation | separation of a long half life nuclide and adsorption | suction of a short half life nuclide in the solidification processing method of the radioactive waste of Example 2 which is another suitable Example of this invention. It is explanatory drawing which shows the solidification process by the low melting glass of the adsorbent which adsorb | sucked the short half life nuclide in the solidification processing method of the radioactive waste of Example 2 of this invention. It is a flowchart which shows the process sequence in the solidification processing method of the radioactive waste of Example 3 which is another suitable Example of this invention. It is a flowchart which shows the process sequence in the solidification processing method of the radioactive waste of Example 4 which is another suitable Example of this invention. It is an electric potential-pH figure of plutonium which is a long half life nuclide. It is a flowchart which shows the process sequence in the solidification processing method of the radioactive waste of Example 5 which is another suitable Example of this invention.

  Inventors arise from reprocessing spent nuclear fuel in a reprocessing facility. In the treatment of radioactive liquid waste containing long half-life nuclides such as U-238 and Pu-239 and short half-life nuclides such as Cs-137 (for example, U-238 and Pu-239 etc.), long-lived nuclides are not included. The method of solidifying radioactive waste containing short half-life nuclides with low-melting glass was investigated. Here, the long half-life nuclide is a radionuclide having a half-life of 40 years or more, such as U-235, U-238, Pu-239, Pu-242, Am-241, Np-236, and Cu-246. . The short half-life nuclide is a radionuclide having a half-life of less than 40 years, such as Cs-137 and Sr-90.

  Contaminated water, which is a radioactive liquid waste containing the above-mentioned long half-life nuclides and short half-life nuclides, may also be generated by cooling the molten nuclear fuel generated by the melting of the reactor core.

  The inventors have examined the treatment of radioactive liquid waste that is a radioactive waste containing long half-life nuclides (eg, U-238 and Pu-239) and short half-life nuclides (eg, Cs-137 and Sr-90). went. Radioactive waste generated at nuclear facilities can be used as high-dose radionuclides, from short-lived radionuclides of about 30 years such as Cs-137 to long-lived radionuclides of about 4.5 billion years such as U-238. Including low-melting-point glass (softening point of 800 ° C or lower) when long-half-life nuclide and short-half-life nuclide are solidified together, low-melting-point glass solidifies low-melting-point glass over tens of thousands of years and hundreds of millions of years The body performance may not be maintained.

  The inventors examined measures for improving such problems. As a result, the long half-life nuclide contained in the radioactive liquid waste is removed, and then the short half-life nuclide contained in the radioactive liquid waste is removed by adsorbing to the adsorbent, and the adsorbent adsorbing the short half-life nuclide is vitrified. Thus, the inventors have newly found that the amount of radioactive waste buried and disposed of over a long period of time can be reduced. The removal of the long half-life nuclide and the removal of the short half-life nuclide using the adsorbent are performed by solidifying the adsorbent adsorbing the short half-life nuclide with a low melting glass (softening point is 800 ° C. or lower). Pre-processing. In addition, by removing the long half-life nuclide in advance, the adsorbent that has adsorbed the short half-life nuclide can be solidified using a low-melting glass, and the above-mentioned problems can be improved. Waste can be vitrified.

  An embodiment of the solidification method for radioactive waste according to the present invention reflecting the above examination results will be described below.

  A method for solidifying radioactive waste according to embodiment 1, which is a preferred embodiment of the present invention, will be described below with reference to FIGS.

  For example, in the unlikely event that a nuclear power plant accident occurs due to power loss, the core in the reactor may melt, and fuel debris may fall to the bottom of the reactor pressure vessel or the bottom of the reactor containment vessel. In order to cool the fuel debris present at the bottom of the reactor pressure vessel or the fuel debris present at the bottom of the reactor containment vessel, cooling water is sprayed on the fuel debris. The cooling water sprayed for cooling the fuel debris is stored at the bottom in the reactor building.

  It is assumed that the cooling water stored at the bottom in the reactor building contains Cs-137 (half-life: 30.1 years) which is a short half-life nuclide having a half-life of less than 40 years. Cooling water containing this short half-life nuclide is called contaminated water.

  In order to perform decommissioning work at a nuclear power plant where a severe accident has occurred, fuel debris is taken out of the reactor containment vessel. By taking out the fuel debris, the contaminated water in the reactor building may contain a long half-life nuclide such as U-238 (half-life: about 4,468 million years). In the fuel debris retrieval operation, it is assumed that this contaminated water contains U-238, which is a long half-life nuclide having a half-life of 40 years or more.

  The method for solidifying radioactive waste according to the first embodiment will be described with respect to contaminated water that is radioactive waste liquid that is radioactive waste liquid in a reactor building.

  First, an outline of the radioactive waste solidification method of the present embodiment will be described with reference to FIG. As shown in FIG. 1, the solidification method of the radioactive waste of the present embodiment includes separation of the long half-life nuclide from the radioactive waste liquid (contaminated water) (step S1), adsorption of the short half-life nuclide by the adsorbent (step S2) and vitrification of the adsorbent adsorbing the short half-life nuclide (step S3) are included.

  A long half-life nuclide is separated from the radioactive liquid waste (step S1). Separating long half-life nuclides from contaminated water, including long half-life nuclides (eg, U-238) and short half-life nuclides (eg, Cs-137), which are radioactive effluents present in reactor buildings, Remove.

  The short half-life nuclide is removed by adsorbing to the adsorbent (step S2). By bringing the contaminated water from which the long half-life nuclide has been removed into contact with the adsorbent, the short half-life nuclide contained in the contaminated water is adsorbed on the adsorbent. The short half-life nuclide contained in the contaminated water is removed by the adsorbent.

  The adsorbent is vitrified (step S3). The adsorbent adsorbing the short half-life nuclide is filled into a solidification container together with a glass raw material, specifically, a low melting glass raw material (softening point is 800 ° C. or lower). The gap between the adsorbents that adsorb the short half-life nuclide in the solidification container in which the low-melting glass raw material is melted by utilizing the heat generated by the radiation emitted from the short half-life nuclide Flow into. A glass solidified body is produced by solidifying the melt of the low melting point glass in the solidifying container.

  The method for solidifying radioactive waste according to this embodiment will be described in detail below with reference to FIGS.

  A long half-life nuclide is separated from the radioactive liquid waste (step S1). Separation of long half-life nuclides from contaminated water, which is radioactive liquid waste, is removal of long half-life nuclides from contaminated water. As shown in FIG. 2, the process of step S1 for contaminated water containing U-238, which is a long half-life nuclide, includes steps S1A and S1B.

  The radioactive waste solidification processing apparatus used in the radioactive waste solidification processing method of the present embodiment includes a long half-life nuclide removal apparatus and a short half-life nuclide used in the implementation of steps S1A and S1B and step S2 described later. It has an adsorption device and will be described with reference to FIG.

  The long half-life nuclide removing device includes an acid supply device 3, a radioactive waste liquid container 8, an alkali supply device 9, an anode electrode 11, a cathode electrode 12, and a variable resistor 23. The acid supply device 3 is connected to the radioactive liquid waste container 8 by an acid supply pipe 4 provided with a valve 36. Further, the alkali supply device 9 is communicated with the radioactive liquid waste container 8 by an alkali supply pipe 10 provided with a valve 38. A radioactive liquid waste supply pipe 5 provided with a valve 7 is connected to a radioactive liquid waste container 8. A pH meter 37 is installed in the radioactive liquid waste container 8. An anode electrode 11 and a cathode electrode 12 are disposed in the radioactive liquid waste container 8. The anode electrode 11 is connected to the power source 12 by the wiring 14A, and the cathode electrode 12 is connected to the power source 12 by the wiring 14B. The open / close switch 13 and the variable resistor 23 are connected to the wiring 14A.

  The short half-life nuclide adsorption apparatus has a solidification container 19 in which an adsorbent (for example, zeolite) 20 is filled. A pipe 16 connected to the bottom of the radioactive liquid waste container 8 is inserted into the solidification container 19 from above and communicated with the solidification container 19. A pump 18 and a valve 17 are provided in the pipe 16. A waste liquid discharge pipe 21 provided with a valve 22 is detachably connected to the bottom of the solidification container 19. A flange (not shown) attached to one end of the waste liquid discharge pipe 21 is provided on the outer surface of the bottom of the solidification container 19 and attached to a connecting member facing the flange by screws. In this connection member, a through hole (waste liquid discharge port) that is communicated with the solidification container 19 and communicated with the waste liquid discharge pipe 21 is formed. A wire mesh (not shown) for preventing the adsorbent 20 from being discharged from the solidification container 19 is attached to the bottom surface of the solidification container 19 so as to cover the through hole.

An acid is injected into the radioactive liquid waste to ionize the long half-life nuclide contained in the radioactive liquid waste (step S1A). Contaminated water 2 which is radioactive waste liquid in the reactor building is opened to a predetermined amount into the radioactive waste liquid container 8 through the radiation waste liquid supply pipe 5 by opening the valve 7. At this time, the pump 18 is stopped and the valve 17 is closed. The contaminated water 2 supplied to the radioactive liquid waste container 8 contains U-238 which is a long half-life nuclide and Cs-137 which is a short half-life nuclide. After the supply of the contaminated water 2 to the radioactive liquid waste container 8 is stopped, the valve 36 is opened, and an acid aqueous solution, for example, a sulfuric acid aqueous solution is supplied from the acid supply device 3 through the acid supply pipe 4 to the contaminated water 2 in the radioactive liquid waste container 8. (See FIG. 3A). As the acid, hydrochloric acid may be used instead of sulfuric acid. The pH of the contaminated water 2 in the radioactive liquid waste container 8 into which the sulfuric acid aqueous solution is injected is measured by a pH meter 37. The injection of the sulfuric acid aqueous solution is for ionizing U-238, which is a long half-life nuclide contained in the contaminated water 2. In order to ionize U-238, it is necessary to adjust the pH and the potential Eh of the contaminated water 2 in the radioactive liquid waste container 8. The potential Eh is adjusted by closing the open / close switch 13 and operating the variable resistor 23 to adjust the voltage applied between the anode electrode 11 and the cathode electrode 12 to adjust the potential Eh on the surface of the cathode electrode 12. Is done by. As shown in FIG. 5, U-238 has a potential Eh of, for example, 0.0 Vvs. SHE is ionized in region 2 (pH 1.5 or lower) to U ⁴⁺ . Region 2 is a region where U-238, which is a long half-life nuclide, is ionized. The surface potential Eh of the cathode electrode 12 is set to 0.0 Vvs. In the state adjusted to SHE, the sulfuric acid aqueous solution is injected from the acid supply device 3 into the contaminated water 2 in the radioactive liquid waste container 8 until the pH of the contaminated water 2 measured by the pH meter 37 becomes 1.0, for example. Is done. The surface potential Eh of the cathode electrode 12 is set to 0.0 Vvs. When it is SHE, the potential Eh of the contaminated water 2 in the radioactive liquid waste container 8 is also 0.0 Vvs. It is SHE. When the pH measured by the pH meter 37 reaches 1.0, the valve 36 is closed, and the supply of the sulfuric acid aqueous solution from the acid supply device 3 to the radioactive liquid waste container 8 is stopped. U-238 contained in the contaminated water 2, that is, UO ₂ , causes the reaction of the formula (1) by injecting the sulfuric acid aqueous solution to generate U ⁴⁺ .

UO ₂ + 4H ⁺ → U ⁴⁺ + 2H ₂ O (1)
For this reason, the contaminated water 2 in the radioactive liquid waste container 8 contains U ⁴⁺ generated from U-238.

An alkali is injected into the radioactive liquid waste, and the long half-life nuclide in the radioactive liquid waste is separated (step S1B). The potential Eh on the surface of the cathode electrode 12 is set to 0.0 Vvs. In the SHE, the valve 38 is opened, and an alkaline aqueous solution, for example, a sodium hydroxide aqueous solution is injected from the alkali supply device 9 through the alkali supply pipe 10 into the contaminated water 2 in the radioactive liquid waste container 8 ((A in FIG. 3). )reference). As the alkali, potassium hydroxide may be used instead of sodium hydroxide. The pH of the contaminated water 2 in the radioactive liquid waste container 8 into which the sodium hydroxide aqueous solution is injected is measured by a pH meter 37B. The pH of the contaminated water 2 is increased by injecting the alkaline aqueous solution, specifically, the sodium hydroxide aqueous solution into the contaminated water 2 in the radioactive liquid waste container 8. The pH of the contaminated water 2 is 0 Vvs. A sodium hydroxide aqueous solution is injected into the contaminated water 2 in the radioactive liquid waste container 8 until the pH reaches 1.5 or more, for example, 5.0, at which U-238 is precipitated as UO ₂ when it is SHE. When the pH measured by the pH meter 37A reaches 5.0, the valve 38 is closed and the supply of the aqueous sodium hydroxide solution from the alkali supply device 9 to the radioactive liquid waste container 8 is stopped.

The potential Eh on the surface of the cathode electrode 12 is 0.0 Vvs. Since the pH of the contaminated water 2 is held at SHE and the pH of the contaminated water 2 is 5.0, U-238 contained in the contaminated water 2 is converted to the surface of the cathode electrode 12 as UO ₂ as the precipitated substance 15 by the reaction of the equation (2). (See FIG. 3B). This is because the condition of region 1 shown in FIG. 5 is satisfied on the surface of the cathode electrode 12. Region 1 is a region where U-238 is deposited.

U ⁴⁺ + 2H ₂ O → UO ₂ + 4H ⁺ (2)
In each step of step S1A (ionization of long half-life nuclide) and step S1B (separation of long half-life nuclide), the voltage application between the anode electrode 11 and the cathode electrode 12 is the potential at the surface of the cathode electrode 12. Is the potential of the water stabilization region (see FIG. 5) existing between the lower limit of the water stabilization region indicated by the thin broken line and the upper limit of the water stabilization region indicated by the thin broken line, Done. In particular, when U-238 is ionized in step S1A (ionization of long half-life nuclide), the pH of the contaminated water is the lower limit of pH and the upper limit of pH in region 2 shown in FIG. And the potential of the surface of the cathode electrode 12 may be adjusted to a potential within the range between the lower limit of the potential of the region 2 and the upper limit of the potential (potential within the water stabilization region). When U-238 is separated in step S1B (separation of long half-life nuclide), the pH of the contaminated water is the lower limit of pH and the upper limit of pH in region 1 shown in FIG. The pH of the cathode electrode 12 and the surface potential of the cathode electrode 12 may be adjusted to a potential within the range of the lower limit of the potential and the upper limit of the potential shown in FIG.

When UO ₂ is deposited on the surface of the cathode electrode 12, Cs-137 contained in the contaminated water 2 is in an ionic state regardless of the pH of the contaminated water in the water stabilization region, as shown in FIG. It does not deposit on the surface of the cathode electrode 12.

When the deposited material 15 (UO ₂ ) deposited on the surface of the cathode electrode 12 reaches a predetermined thickness, the open / close switch 13 is opened and the wiring 14B is removed from the cathode electrode 12, and then the cathode electrode 12 to which the deposited material 15 is attached. Is taken out from the radioactive liquid waste container 8. Thereafter, a new cathode electrode 12 is inserted into the radioactive liquid waste container 8 and immersed in the contaminated water 2 in the radioactive liquid waste container 8. The wiring 14B is connected to the new cathode electrode 12, and the open / close switch 13 is closed. U-238 contained in the contaminated water 2 in the radioactive liquid waste container 8 is deposited on the surface of the new cathode electrode 12 as UO ₂ which is the deposition substance 15. The replacement of the cathode electrode 12 immersed in the contaminated water 2 in the radioactive liquid waste container 8 is continuously performed until the U-238 contained in the contaminated water 2 is reduced to a predetermined concentration or less, preferably until it disappears. . As described above, U-238 contained in the contaminated water 2 can be deposited on the surface of the cathode electrode 12 by electrodeposition and separated from the contaminated water 2.

The cathode electrode 12 taken out from the radioactive liquid waste container 8 and attached with UO ₂ is a high-level waste and is a glass having a high softening point that is not a low melting point glass, or storage for a predetermined period in a predetermined storage place. The conventional vitrification using is performed (step S5).

When the contaminated water 2 supplied to the radioactive liquid waste container 8 contains U-235 or the like that is a uranium isotope, this uranium isotope is ionized to U ⁴⁺ in step S1A, and in step S1B. In the same manner as U-238, it is deposited on the surface of the cathode electrode 12 as UO ₂ which is the deposited substance 15.

  The short half-life nuclide is adsorbed on the adsorbent (step S2). The contaminated water 2 in the radioactive liquid waste container 8 from which the long half-life nuclide (U-238) has been separated is filled with the adsorbent (for example, zeolite) 20 by opening the valve 17 and driving the pump 18. It is supplied to the solidification container 19. As the adsorbent 20 that adsorbs cesium, calix may be used instead of zeolite. At this time, the valve 22 is open. For example, 100 kg of the adsorbent 20 is filled in the solidification container 19. The contaminated water 2 supplied to the solidification container 19 comes into contact with the adsorbent 20 when passing through the layer of the adsorbent 20 filled in the solidification container 19. At this time, Cs-137 which is a short half-life nuclide contained in the contaminated water 2 is adsorbed by the adsorbent 20 and removed from the contaminated water 2 (see FIG. 3C). The contaminated water 2 that has passed through the adsorbent 20 layer is discharged as a waste liquid from the radioactive waste liquid container 8 to the waste liquid discharge pipe 21. The waste liquid is separated from the long half-life nuclide and the short half-life nuclide to reduce the radioactivity level, and is led to and stored in a storage tank (not shown) through the waste liquid discharge pipe 21 (step S4). Since the waste liquid discharged from the radioactive liquid waste container 8 contains tritium (half-life: 12.32 years), it is stored in the storage tank for a predetermined period.

  When the adsorbent 20 in the solidification container 19 has a reduced ability to adsorb Cs-137, the solidification container 19 is replaced with a new solidification container 19 filled with a predetermined amount of new adsorbent 20. A decrease in the adsorption performance of the adsorbent 20 is determined based on a dose rate measured by a radiation detector (not shown) installed outside the solidification container 19. When the solidification container 19 having the adsorbent 20 whose Cs-137 adsorption performance has deteriorated is replaced with a new solidification container 19, the pump 18 is stopped, the valve 17 is closed, and the contaminated water from the radioactive liquid waste container 8 to the solidification container 19 2 supply is stopped. Even after the supply of the contaminated water 2 into the solidification container 19 having the adsorbent 20 with reduced adsorption performance is stopped, the waste water, which is the contaminated water 2 purified by the adsorbent 20 in the solidification container 19, is discharged as waste liquid. It is discharged to the tube 21. After all the waste water is discharged from the solidification container 19, the flange attached to one end of the waste liquid discharge pipe 21 is removed from the outer surface of the bottom of the solidification container 19. When the flange is removed, a sealing plate is attached to the outer surface of the bottom portion of the solidification container 19, and the waste liquid discharge port formed at the bottom portion of the solidification container 19 is closed (see FIG. 3D). ).

  A flange attached to one end of the waste liquid discharge pipe 21 is attached to the bottom of the new solidification container 19, and the waste liquid discharge pipe 21 is connected to the new solidification container 19. The pipe 16 is also connected to the solidification container 19. Then, the valve 17 is opened and the pump 18 is driven, and the contaminated water 2 containing Cs-137 in the radioactive liquid waste container 8 is supplied into the new solidification container 19. This Cs-137 is adsorbed and removed by the adsorbent 20 in the solidification container 19. The solidification container 19 is replaced as many times as necessary until the contaminated water 2 from which the long half-life nuclide has been removed is removed from the radioactive liquid waste container 8.

  After all the contaminated water 2 in the radioactive liquid waste container 8 is discharged, the valve 17 is closed and the valve 7 is opened. New contaminated water 2 is supplied into the radioactive liquid waste container 8. And each process of step S1A, S1, and BS2 is implemented.

  A vitrified body of the adsorbent is produced (step S3). The adsorbent 20 that has adsorbed the short half-life nuclide in the solidification container 19 in which the waste liquid discharge pipe 21 is removed and the sealing plate is attached to the bottom is solidified by the low melting point glass. The adsorbent 20 that adsorbs the short half-life nuclide is radioactive waste. The production of such a vitrified body will be described with reference to FIG. In the process of step S3, the glass raw material of the low melting glass is supplied into the solidification container 19 filled with the adsorbent which is radioactive waste, and the glass raw material is melted by the heat insulation treatment of the solidification container 19.

  In step S <b> 2, 100 kg of the adsorbent 20 having adsorbed Cs-137 is filled, and the solidification container 19 from which the internal waste water is discharged is moved to a position immediately below the glass raw material tank 29. 100 kg of soda lime glass (softening point: about 700 ° C.), which is the glass raw material 32 in the glass raw material tank 29, is supplied into the solidification container (first container) 19 through the glass raw material supply pipe 30 connected to the glass raw material tank 29. To do. The soda-lime glass is supplied into the solidification container 19 by opening a valve 31 provided in the glass raw material supply pipe 30. As a glass raw material 32 of low melting point glass (softening point is 800 ° C. or lower), borosilicate glass (softening point: about 800 ° C.), vanadium glass (softening point: about 300 ° C.), phosphoric acid instead of soda lime glass Any one of glass (softening point: 330 ° C.) and titanium silicate glass (softening point: 530 ° C.) may be used.

  Next, melting of soda lime glass that is the glass raw material 32 by the heat insulation treatment of the solidification container 19 will be described.

  The adsorbent 20 adsorbing Cs-137, which is radioactive waste, and the solidification container 19 filled with the glass raw material 32 are stored in a heat insulating container (second container) 34 with the upper end opened. The heat insulating container 34 has a detachable lid 34A at the upper end, and when the solidification container 19 is stored in the heat insulating container 34, the cover 34A is removed and the heat insulating container 34 is opened upward. . In this state, the solidification container 19 is put into the heat insulating container 7 from above. Thereafter, the lid 34A is attached to the upper end of the heat insulating container 34, and the heat insulating container 34 housing the solidification container 19 is sealed. The heat insulating container 34 and the lid 34A have, for example, glass wool, which is a heat insulating material. A heat insulating region insulated by the heat insulating container 34 and the lid 34A is formed in the sealed heat insulating container 34, and the solidification container 19 filled with the adsorbent 20 and the glass raw material 32 is disposed in the heat insulating region.

  The heat insulating container 34 has a double structure of a metal outer container (not shown) and an inner container (not shown) disposed in the outer container, and glass wool is placed between the outer container and the inner container. The annular region and the bottom of the outer container and the bottom of the inner container are filled. The upper end of the annular region between the outer container and the inner container is sealed by a ring-shaped sealing plate attached to the respective upper ends of the outer container and the inner container. The lid 34A is configured by filling glass wool in a metal hollow casing (not shown).

  The glass wool used for the heat insulation container 34 and the lid 34A used in the heat insulation treatment in this embodiment is made of cellulose fiber, carbonized cork, urethane foam, phenol foam and polystyrene foam, and porous heat insulation material, which are fiber heat insulation materials. It may be changed to any one of the calcium silicate boards.

  Then, the heat insulation process of the solidification container 19 accommodated in the sealed heat insulation container 34 is implemented. The heat insulation process means a process for suppressing the heat of the solidification container 19 from escaping to the outside. Heat (decay heat) is generated in the solidified container 19 that has been heat-insulated, based on radiation emitted from Cs-137 adsorbed by the adsorbent 20 filled therein. The decay heat is suppressed from being radiated to the outside by the heat insulating container 34 sealed by the lid 34A, and is accumulated in the heat insulating container 34 sealed by the lid 34A. With this decay heat, the glass raw material 32 (soda lime glass) in the solidification container 19 is heated and melted.

  Since the solidification container 19 is surrounded by the heat insulating container 34 and the lid 34A, the solidification container 19, and the adsorbent 20 and the glass raw material 32 in the solidification container 19 are heated by the decay heat. The temperatures of the heated adsorbent 20 and the glass raw material 32 in the solidification container 19 are not uniform depending on the positions in the solidification container 19 but are almost uniform.

For example, Cs-137, which is a radionuclide adsorbed on the adsorbent 20, which is a radioactive waste, emits radiation (gamma rays) having an energy of about 1.15 MeV per decay. This radiation is absorbed by the adsorbent 20 and the glass raw material (soda lime glass) 32 in the solidification container 19, and changes into thermal energy (decay heat). Since the adsorbent 20 filled in the solidification container 19 contains 10 ¹⁶ Bq of Cs-137, all of the radiation emitted from each Cs-137 is adsorbent 20 in the solidification container 19 and the glass raw material. When absorbed by 32, 1.15 MeV × 10 ¹⁶ Bq = 1.15E22 eV / s, that is, heat energy with a heat generation rate of 1840 J / s is obtained. When the specific heat of the mixture of the adsorbent (zeolite adsorbing Cs-137) 20 and the glass raw material (soda lime glass) 32 is 0.5 J / (g · K), the adsorbent 20 and the glass raw material 32 respectively. The temperature rises about 66 ° C. in 1 hour. Due to this temperature rise, the soda-lime glass that is the glass raw material 32 melts and flows into the gaps between the adsorbents 20. Preferably, after the glass raw material 32 is filled in the solidification container 19 filled with the adsorbent 20, the adsorbent 20 and the glass raw material 32 are mixed with a stirrer.

  Since some heat escapes to the outside through the heat insulating container 34 and the lid 34A, the actual temperature increase rate inside the solidification container 19 is lower than 66 ° C./h. However, the temperature rise is continued with the passage of time, and all the glass raw material 32 is melted. As a result, the soda-lime glass melt as the glass raw material 32 is filled in the gaps between the adsorbents 20. A glass solidified body 35 in which a glass solidified product 33 integrated with the glass raw material 32 in which the adsorbent 20 is melted is present in the solidification container 19 is produced. After the vitrified body 35 is produced, the lid 34A is removed, so that the vitrified body 35 is taken out from the heat insulating container 34 and sealed with a lid (not shown), and a predetermined storage place as a waste body. (Not shown) temporarily stored.

The vitrified body 35 produced by solidifying soda-lime glass, which is a low-melting glass, is stored in its storage location until this dose is reduced to a general waste dose (about 10 ¹² Bq). As shown in FIG. 7, it takes about 300 years for the vitrified body 35 containing the adsorbent adsorbing Cs-137 to be reduced to the dose of general waste. The vitrified body 35 is stored in the storage place for about 300 years.

  In this embodiment, since the solidification container 19 filled with the adsorbent 20 and the glass raw material 32 is surrounded by the heat insulating container 34 sealed with the lid 34A, the radionuclide adsorbed on the adsorbent 20 (Cs-137). The adsorbent 20 and the glass raw material 32 are heated by the heat energy (decay heat) generated by the radiation released by the absorption of the adsorbent 20 and the glass raw material 32 disposed in the heat insulating region in the heat insulating container 34. Is done. For this reason, heating unevenness does not occur in the adsorbent 20 and the glass raw material 32 existing in the heat insulating region, and the adsorbent 20 and the glass raw material 32 in the solidification container 19 have a more uniform temperature. For this reason, the high temperature corrosion of the solidification container 19 is suppressed, and the uniform vitrified body 35 of the adsorbent 20 which is radioactive waste is obtained. This vitrified body 35 is stable.

  In this embodiment, the adsorbent 20 and the glass raw material 32 are heated by the thermal energy generated by the absorption of the radiation generated by the decay of the radionuclide (Cs-137) adsorbed by the adsorbent 20 in the solidification container 19. A melting facility for melting the glass raw material 32 is not required as in the case of vitrification of radioactive waste described in Japanese Unexamined Patent Publication No. 2011-46996. That is, the adsorbent 20 which is a radioactive waste can be solidified by the glass raw material 32 with a simple system.

  In this embodiment, since the contaminated water 2 from which the long half-life nuclide has been removed is supplied from the radioactive liquid waste container 8 to the solidification container 19, the long half-life nuclide is not adsorbed by the adsorbent 20 in the solidification container 19. The short half-life nuclide is adsorbed on the adsorbent 20. Thus, by removing the long half-life nuclide from the contaminated water 2, the vitrified product 35 produced by solidifying the adsorbent 20 with soda-lime glass does not contain the long half-life nuclide. The storage period of the glass solid body 35 produced using the melting point glass can be remarkably shortened to about 300 years, for example. Furthermore, according to the present Example, the vitrified body 35 can be disposed of as general waste after the storage period has elapsed. According to the vitrified body 35 produced in the present example, the storage period for the vitrified body 35 required to reduce the radioactivity level of the vitrified body 35 to a radioactive level that can be disposed as general waste is reduced. In addition, it can be remarkably shortened than the embedding period of the vitrified body containing the long half-life nuclide.

  According to the present embodiment, since the vitrified body 35 does not contain a long half-life nuclide, the soundness of the vitrified body 35 containing soda-lime glass which is a low melting point glass can be maintained over its storage period. . For this reason, in the storage period (about 300 years), release of radionuclides contained in the vitrified body 35, specifically, short half-life nuclides (for example, Cs-137) can be avoided. That is, according to the present embodiment, the vitrified body 35 can be produced using low-melting glass, and the vitrified body 35 produced using low-melting glass until it can be disposed as general waste. It can be stored in a healthy state.

  In this example, since the adsorbent 20 that adsorbs the short half-life nuclide and the glass solidified body 35 including the low melting point glass are produced, it is necessary to bury deeply in the ground for 10,000 years or more. The amount of vitrified body containing the phase nuclide is reduced.

  There is a possibility that U-238, which is a long half-life nuclide contained in the contaminated water 2 supplied to the radioactive liquid container 8, is present as a stable oxide. For this reason, in this example, since the long half-life nuclide contained in the contaminated water 2 is ionized by lowering the pH of the contaminated water 2, it is possible to completely ionize the long half-life nuclide to be separated. By increasing the pH of the contaminated water 2, a long half-life nuclide (for example, U-238) can be deposited on the surface of the cathode electrode 12, and the long half-life nuclide can be separated from the contaminated water 2. .

  In this embodiment, the contaminated water 2 from which the long half-life nuclide has been removed is supplied to the solidification container 19 filled with the adsorbent 20, and the short half-life nuclide contained in the contaminated water 2 in the solidification container 19 (for example, Since Cs-137) is adsorbed on the adsorbent 20, as in Example 2, the adsorbent 20 having adsorbed the short half-life nuclide is transferred into the waste tank 25, and the solidification container 28 is further transferred from the waste tank 25. The operation of filling the inside becomes unnecessary, and the production of the vitrified body 35 becomes easy.

In this embodiment, the potential on the surface of the cathode electrode 12 is adjusted to the potential in the water stabilization region in each step of ionization of the long half-life nuclide (step S1A) and separation of the long half-life nuclide (step S1B). Therefore, generation of hydrogen can be suppressed, and there is no possibility that the pH of the contaminated water 2 is changed, and separation of the long half-life nuclide from the contaminated water 2 can be easily performed. When the potential at the surface of the cathode electrode 12 is lower than the lower limit potential of the water stabilization region, hydrogen is generated from the contaminated water 2, and the potential at the surface of the cathode electrode 12 is higher than the upper limit potential of the water stabilization region. If it is high, a reduction reaction of oxygen occurs, OH ⁻ is generated in the contaminated water 2, and the pH of the contaminated water 2 may change. In this embodiment, since the potential on the surface of the cathode electrode 12 is adjusted to the potential in the water stabilization region, those problems do not occur.

  The method for solidifying radioactive waste according to embodiment 2, which is another preferred embodiment of the present invention, will be described below with reference to FIGS.

  Similarly to the radioactive waste solidification processing method of the first embodiment, the radioactive waste solidification processing method of the present embodiment is implemented by steps S1A, S1B, S2, S3, S4 and S5 shown in FIG. Is done. However, the radioactive waste solidification processing method of the present embodiment is different from the radioactive waste solidification processing method of the first embodiment and the method of creating the glass solidified body 35 using the low-melting glass in step S3.

  The radioactive waste solidification processing apparatus (see FIG. 8) used in the radioactive waste solidification processing method of this embodiment is different from the radioactive waste solidification processing apparatus used in Example 1 and the short half-life nuclide adsorption apparatus. . As shown in FIG. 8, the long-lived nuclide removal device of the radioactive waste solidification processing apparatus used in the solidification processing method of the radioactive waste of this embodiment is the same as that of the radioactive waste solidification processing apparatus used in the first embodiment. Have the same configuration.

  As shown in FIG. 8, the short half-life nuclide adsorption apparatus used in the present embodiment has a configuration in which the solidification container 19 is replaced with a nuclide adsorption tank 24 in the short half-life nuclide adsorption apparatus used in Example 1. The volume of the nuclide adsorption tank 24 is larger than the volume of the solidification container 19, and the adsorbent 20 filled in the nuclide adsorption tank 24 is larger than the amount filled in the solidification container 19. A pipe 16 provided with a pump 18 and a valve 17 is connected to the bottom of the radioactive liquid waste container 8, inserted into the nuclide adsorption tank 24 from above, and communicated with the nuclide adsorption tank 24. A waste liquid discharge pipe 21 provided with a valve 22 is connected to the bottom of the nuclide adsorption tank 24.

In the present embodiment, the contaminated water 2 supplied to the radioactive liquid waste container 8 contains U-238, which is a long half-life nuclide, and Cs-137, which is a short half-life nuclide. Also in this example, each of the step S1A (ionization of long half-life nuclide) and step S1B (separation of long half-life nuclide) is performed using the long half-life nuclide removal apparatus including the radioactive liquid waste container 8 as in Example 1. A process is performed. The cathode electrode 12 to which UO ₂ is adhered by the execution of step S1B is taken out from the radioactive waste liquid container 8, and this cathode electrode 12, which is a high-level waste, is subjected to conventional vitrification (step S5).

  In step S2, the contaminated water 2 in the radioactive liquid waste container 8 from which the long half-life nuclide (U-238) has been separated passes through the pipe 16 into the nuclide adsorption tank 24 filled with the adsorbent (for example, zeolite) 20. Supplied. At this time, the valve 22 of the waste liquid discharge pipe 21 is open. The contaminated water 2 flowing in the layer of the adsorbent 20 in the nuclide adsorption tank 24 contacts the adsorbent 20 in the layer, and the short half-life nuclide (Cs-137) contained in the contaminated water 2 is adsorbed on the adsorbent 20. To be removed. The waste liquid from which the short half-life nuclide has been removed and discharged from the nuclide adsorption tank 24 is guided to a storage tank (not shown) through the waste liquid discharge pipe 21 and stored (step S4). When the performance of the adsorbent 20 in the nuclide adsorption tank 24 to adsorb Cs-137 decreases, the pump 18 is stopped and the valve 17 is closed, and the contaminated water 2 from the radioactive liquid waste container 8 to the nuclide adsorption tank 24 is reduced. Supply is stopped. After the drainage of the wastewater from the nuclide adsorption tank 24 to the waste liquid discharge pipe 21 is completely completed, the valve 22 is closed.

  A vitrified body of the adsorbent is produced (step S3). After the supply of the contaminated water 2 to the nuclide adsorption tank 24 is stopped and the drainage of the waste water from the nuclide adsorption tank 24 to the waste liquid discharge pipe 21 is completed, the adsorbent 20 that adsorbs the short half-life nuclide becomes the nuclide adsorption tank 24. And is transferred into a waste tank 25 (see FIG. 9). The vitrified body 35 is produced using the adsorbent 20 transferred into the waste tank 25 and soda lime glass which is a low melting point glass.

The production of the vitrified body 35 will be specifically described with reference to FIG. An empty solidification container 28 is disposed below the waste tank 1 in which the adsorbent 20 that adsorbs Cs-137, which is radioactive waste, is stored. 100 kg of high-dose adsorbent 20 containing 10 ¹⁶ Bq of Cs-137 in the waste tank 1 is supplied into the solidification container 28 through the waste supply pipe 26 with the gate valve 27 opened. Thereafter, as in Example 1, 100 kg of soda-lime glass as the glass raw material 32 is supplied from the crow raw material tank 29 to the solidification container 28 filled with the adsorbent 20. The solidification container 28 filled with the adsorbent 20 and the glass raw material 32 is stored in a heat insulating container 34, and the heat insulating container 34 is sealed with a lid 34A. The heat treatment of the solidification container 28 is performed in the heat insulation container 34, and the glass material in the solidification container 28 is generated by the heat (decay heat) generated based on the radiation emitted from Cs-137 adsorbed by the adsorbent 20. 32 (soda lime glass) is melted. The molten glass material 32 is filled in the gaps between the adsorbents 20, and the glass material 35 is produced by solidifying the glass material 32.

  In this example, each effect produced in Example 1 except for the effect obtained by adsorbing the short half-life nuclide (for example, Cs-137) contained in the contaminated water 2 to the adsorbent 20 in the solidification container 19. Can be obtained.

  The method for solidifying radioactive waste according to embodiment 3, which is another preferred embodiment of the present invention, will be described below with reference to FIG.

  The radioactive waste solidification processing method of the present embodiment is a solidification processing method in which step S1B is replaced with step S1B ′ in the radioactive waste solidification processing method of the first embodiment. In the radioactive waste solidification processing method of the present embodiment, steps S1A, S2, S3, S4 and S5, which are performed in the radioactive waste solidification processing method of the first embodiment, are performed. The step S1 performed in the present embodiment includes the steps S1A and S1B ′. In this embodiment, the radioactive waste solidification processing apparatus having the long half-life nuclide removing apparatus and the short half-life nuclide adsorption apparatus shown in FIG.

  Step S1B ′ that is different from the radioactive waste solidification method according to the first embodiment, which is performed in the radioactive waste solidification method according to the present embodiment, will be described in detail.

In step S1A (ionization of long half-life nuclide), the potential Eh of the surface of the cathode electrode 12 is set to 0.0 Vvs. The pH of the contaminated water 2 in the radioactive liquid waste container 8 is adjusted to 1.0 by injection of an aqueous sulfuric acid solution. U-238 contained in the contaminated water 2 is ionized into U ⁴⁺ .

By controlling the potential, the long half-life nuclide in the radioactive liquid waste is separated (step S1B ′). The pH of the contaminated water 2 is maintained at 1.0, the voltage applied between the anode electrode 11 and the cathode electrode 12 is controlled by operating the variable resistor 23, and the potential Eh on the surface of the cathode electrode 12 is adjusted to pH1. 0 in the stabilization region of water at 0, for example 0.5 V vs. Adjust to SHE. This 0.5 Vvs. SHE is about 0.15 Vvs., The upper limit potential of region 2 at pH 1.0. The potential is higher than SHE. As a result, U ⁴⁺ contained in the contaminated water 2 becomes UO ₂ [2+] and is deposited as UO _{2 on} the surface of the cathode electrode 12. When the deposited material 15 (UO ₂ ) deposited on the surface of the cathode electrode 12 reaches a predetermined thickness, the cathode electrode 12 is replaced with a new cathode electrode 12 as described above, and U from contaminated water 2 is replaced. Separation continues. In the step S1B ′, the alkaline aqueous solution (for example, sodium hydroxide aqueous solution) is not injected into the contaminated water 2 in the radioactive liquid waste container 8 as in the step S1B of the first embodiment.

  After the separation of U from the contaminated water 2 is completed, the adsorption of the short half-life nuclide (Cs-137) to the adsorbent 20 in the solidification container 19 (step S2) and the adsorbent Each process of preparation (step S3) of a glass solidification body is implemented.

  In the present embodiment, each effect produced in the first embodiment can be obtained. In this embodiment, the deposition of the long half-life nuclide on the surface of the cathode electrode 12 is performed not by increasing the pH of the contaminated water 2 but by increasing the potential of the surface of the cathode electrode 12. The long half-life nuclide can be deposited in a shorter time than in Example 1. That is, the time required for removing the long half-life nuclide from the contaminated water 2 can be shortened.

  A method for solidifying radioactive waste according to embodiment 4, which is another preferred embodiment of the present invention, will be described below with reference to FIG.

  The radioactive waste solidification processing method of the present embodiment is a solidification processing method in which step S1A is replaced with step S1A 'in the radioactive waste solidification processing method of embodiment 1, and further, step S1C is added. In the radioactive waste solidification processing method of the present embodiment, steps S1B, S2, S3, S4, and S5, which are performed in the radioactive waste solidification processing method of the first embodiment, are also performed. The step S1 performed in this embodiment includes steps S1A ', S1B, and S1C. In this embodiment, the radioactive waste solidification processing apparatus having the long half-life nuclide removing apparatus and the short half-life nuclide adsorption apparatus shown in FIG.

  The solidification processing method of the radioactive waste of a present Example is demonstrated concretely.

  An acid is injected into the radioactive liquid waste to ionize the long half-life nuclide contained in the radioactive liquid waste (step S1A '). In this embodiment, the contaminated water 2 supplied to the radioactive liquid waste container 8 through the radioactive liquid waste supply pipe 5 contains long half-life nuclides such as U-238 and Pu-239 (half life: about 24000 years). -238 and Pu-239 are each ionized.

The potential-pH diagram of U shown in FIG. In the case of SHE, U-238 is ionized to U ⁴⁺ at pH 1.5 or lower, and the potential-pH diagram of Pu shown in FIG. In the case of SHE, Pu-239 is ionized to Pu ³⁺ at pH 5.0 or lower. Therefore, in order to ionize both U-238 and Pu-239, the potential is 0.0 Vvs. What is necessary is just to make the pH of contaminated water 1.5 or less when it is SHE.

The voltage applied between the anode electrode 11 and the cathode electrode 12 is adjusted, and the potential of the surface of the cathode electrode 12 is set to 0.0 Vvs. Adjust to SHE. And sulfuric acid aqueous solution is inject | poured into the contaminated water 2 containing U-238 and Pu-239 in the radioactive waste liquid container 8 from the acid supply apparatus 3, and pH of the contaminated water 2 is 1.5 or less, for example, 1.0 To. As a result, each of U-238 and Pu-239 is ionized, and U-238 becomes U ⁴⁺ and Pu-239 becomes Pu ³⁺ in the contaminated water 2. The pH of 1.5 or lower is the first pH at which each of the long half-nuclides U-238 and Pu-239 is ionized.

Thereafter, similarly to Example 1, the step S1B is performed, and U-238, which is a long half-life nuclide, is separated from the contaminated water 2. In Step S1B, in order to deposit U-238 on the surface of the cathode electrode 12, the surface potential of the cathode electrode 12 is set to 0.0 Vvs. The pH of the contaminated water 2 is adjusted to SHE, and U-238 is precipitated as UO ₂ and Pu-239 is not precipitated, that is, a pH within a range of more than 1.5 to 5.0 or less, For example, adjust to 3.0. The pH 3.0 of the contaminated water 2 is adjusted by controlling the injection amount of the aqueous sodium hydroxide solution injected from the alkali supply device 9 into the contaminated water 2 in the radioactive waste liquid container 8. When the pH of the water 2 reaches 3.0, the injection of the aqueous sodium hydroxide solution into the contaminated water 2 is stopped. As a result, UO ₂ is deposited on the surface of the cathode electrode 12 immersed in the contaminated water 2, and U-238 is separated from the contaminated water 2. At this time, Pu does not precipitate on the surface of the cathode electrode 12. While replacing the cathode electrode 12, UO ₂ is deposited on the surface of the cathode electrode 12, that is, U-238 is continuously separated.

By the way, one kind of long half-life nuclide (for example, U-238) contained in the contaminated water 2 is precipitated, and another kind of long half-life nuclide (for example, Pu-239) contained in the contaminated water 2 is not precipitated. The pH range, that is, the pH in the range of more than 1.5 and not more than 5.0 is the pH of the contaminated water 2 when the former half-life nuclide (U-238) is precipitated (second pH). And the second longest half-life nuclide (Pu-239) is the second lowest pH among the pH (second pH) of the contaminated water 2 when it precipitates. In this example, the separation of U-238 is performed by setting the pH of the contaminated water 2 to the lowest second pH among the second pHs, for example, 3.0, and the potential is 0.0 Vvs. This is performed by depositing UO ₂ on the surface of the cathode electrode 12 which is SHE.

Alkaline is injected into the radioactive liquid waste, and other long half-life nuclides in the radioactive liquid waste are separated (step S1C). After the deposition of UO _{2 on} the surface of the cathode electrode 12 is completed, a new cathode electrode 12 is immersed in the contaminated water 2 in the radioactive liquid waste container 8. By operating the variable resistor 23, the potential of the surface of the cathode electrode 12 is set to 0.0 Vvs. In the state of being held in SHE, injection of aqueous sodium hydroxide solution into the contaminated water 2 in the radioactive liquid waste container 8 is started, and the pH of the contaminated water 2 is larger than 5.0 in which Pu-239 precipitates as PuO _2. For example, it is increased to 7.0. When the pH of the contaminated water 2 reaches 7.0, the injection of the aqueous sodium hydroxide solution into the contaminated water 2 is stopped. Pu-239 contained in the contaminated water 2 becomes PuO ₂ and precipitates on the surface of the cathode electrode 12. In this way, Pu-239 is separated from the contaminated water 2. The deposition of PuO _{2 on} the surface of the cathode electrode 12, that is, the separation of Pu-239, is continued while replacing the cathode electrode 12 until the Pu-239 contained in the contaminated water 2 disappears.

The pH range in which each of the two types of long half-life nuclides (U-238 and Pu-239) contained in the contaminated water precipitates, i.e., a pH within a range greater than 5.0, is included in the contaminated water. The pH (second pH) of contaminated water 2 when one long half-life nuclide (U-238) of the types of long half-life nuclides precipitates and the other long of the multiple types of long half-life nuclides The second pH is the second lowest among the pHs (second pH) of the contaminated water 2 when the half-life nuclide (Pu-239) is precipitated. In this example, the separation of PuO ₂ containing Pu-239 is the second pH that is the second lowest pH of the contaminated water 2, for example, 7.0 and the potential is 0.0 Vvs. This is performed by depositing PuO ₂ on the surface of the cathode electrode 12 which is SHE.

Each of the cathode electrode 12 with the UO ₂ generated in each of the steps S1B and S1C attached to the surface and the cathode electrode 12 with the PuO ₂ attached to the surface is a high-level waste, and has a predetermined period in a predetermined storage location. Storage or conventional glass solidification using a glass having a high softening point that is not a low-melting glass is performed (step S5).

  After the step S1C is completed, the steps S2, S3, and S4 are performed as in the first embodiment.

  In the present embodiment, each effect produced in the first embodiment can be obtained. Furthermore, this example can easily separate these long half-life nuclides from the contaminated water 2 even when the contaminated water 2 contains a plurality of types of long half-life nuclides. The glass solid body 35 including the adsorbent 20 that has been adsorbed can be easily produced. In this embodiment, since uranium and plutonium, which are long half-life nuclides, are separated from the contaminated water 2 separately, high-level disposal is performed with the separated uranium and plutonium adhered to the surface of the cathode electrode 12, respectively. Rather than being buried as a product, it can be reused to produce a fuel assembly with a burnup of 0 Gwd / t as nuclear fuel material.

When each of the separated uranium and plutonium is buried as high-level waste in a state of being attached to the surface of the cathode electrode 12, the steps S1B and S1C are performed as in this embodiment. For example, uranium and plutonium may be deposited on the surface of one cathode electrode 12 in the step S1B. That is, after the step S1A ′ is completed, the pH of the contaminated water 2 that is ionized by U-238 and Pu-239 and contained as U ⁴⁺ and Pu ³⁺ is set to the potential of the surface of the cathode electrode 12, respectively. 0.0Vvs. While being held in SHE, it is increased to, for example, 7.0 by injection of an aqueous sodium hydroxide solution. When the pH of the contaminated water 2 is 7.0, both UO ₂ and PuO ₂ substances are deposited as the precipitated substance 15 on the surface of one cathode electrode 12. Since both UO ₂ and PuO ₂ materials are deposited on the surface of the cathode electrode 12 as the deposited material 15, the procedure of the solidification method of radioactive waste shown in FIG. 11 of Example 4 can be simplified. it can. In this example, the precipitation of both UO ₂ and PuO ₂ on the surface of the cathode electrode 12 is performed by setting the highest second pH among the aforementioned second pHs, for example, 7.0. Yes.

  The present embodiment can be applied to solidification processing of waste liquid (including uranium and plutonium) generated by nuclear fuel reprocessing in a nuclear fuel reprocessing facility.

  Minor actinides (neptium, americium, and curium) other than uranium and plutonium contained in the fuel debris during the operation of removing the fuel debris existing at the bottom of the reactor pressure vessel or the fuel debris existing at the bottom of the containment vessel Uranium and plutonium can also be released into the contaminated water 2. Application of the radioactive waste solidification processing method of the present embodiment for contaminated water 2 containing uranium, plutonium, neptium, americium and curium will be described.

In step S1A ′, the potential Eh is set to 0.0 vs. The pH of the contaminated water 2 containing uranium, plutonium, neptinium, americium and curium is adjusted to 1.0 by injection of sulfuric acid aqueous solution. At this time, each of uranium, plutonium, neptonium, americium, and curium is ionized in the contaminated water 2. Uranium becomes U ⁴⁺ , plutonium becomes Pu ³⁺ , neptium becomes Np ³⁺ , americium becomes Am ³⁺ , and curium becomes Cm ³⁺ . Thereafter, when the pH of the contaminated water 2 is increased to 3.0 in step S1B, UO ₂ and NpO ₂ are deposited on the surface of the cathode electrode 12, respectively. When the pH of the contaminated water 2 is increased to 7.0 by performing Step S1C, PuO ₂ is deposited on the surface of the replaced cathode electrode 12. When the pH of the contaminated water 2 is raised to 12.0, Am (OH) ₃ and Cm (OH) ₃ are deposited on the surface of the cathode electrode 12 that has been replaced again. In this way, each of uranium, plutonium, neptium, americium and curium is separated. And each process of step S2 and S3 from contaminated water 2 is implemented.

  A method for solidifying radioactive waste according to embodiment 5, which is another preferred embodiment of the present invention, will be described below with reference to FIG.

  In the radioactive waste solidification processing method of the present embodiment, step S1B is replaced with step S1B ′ performed in the third embodiment in the radioactive waste solidification processing method of the fourth embodiment, and further, step SC1 is replaced with step S1C ′. This is a different solidification method. In the radioactive waste solidification method according to the present embodiment, steps S1A ′, S2, S3, S4, and S5 that are performed in the radioactive waste solidification method according to the fourth embodiment are performed. The step S1 performed in this embodiment includes the steps S1A ', S1B', and S1C '. The process of step S1 performed in the present embodiment is an application of the concept of the process of step S1 performed in the third embodiment to the contaminated water 2 containing uranium and plutonium. In this embodiment, the radioactive waste solidification processing apparatus having the long half-life nuclide removing apparatus and the short half-life nuclide adsorption apparatus shown in FIG.

In step S1A ′ of the radioactive waste solidification method according to the present embodiment, as in the fourth embodiment, uranium and plutonium contained in the contaminated water 2 are ionized into U ⁴⁺ and Pu ³⁺ , respectively. Thereafter, in the step S1B ′, the pH of the contaminated water 2 is maintained at 1.0, and the potential Eh on the surface of the cathode electrode 12 is set to 0.5 Vvs. Increase to SHE. Similarly to Example 3, U ⁴⁺ contained in the contaminated water 2 becomes UO ₂ [2+], and is deposited as UO _{2 on} the surface of the cathode electrode 12. UO ₂ is deposited on the surface of the cathode electrode 12 while replacing the cathode electrode 12 as necessary. Uranium is separated from the contaminated water 2.

Cathode electrode in which one kind of long half-life nuclide (for example, U-238) contained in the contaminated water 2 is deposited and another kind of long half-life nuclide (for example, Pu-239) contained in the contaminated water 2 is not deposited 12 potential range, that is, 0.15 V vs. 0.1 when the pH of the contaminated water 2 is 1.0. 0.95 Vvs. The potential within the range below SHE is the second potential of the cathode electrode 12 when the former long half-life nuclide (U-238) is deposited and when the latter long half-life nuclide (Pu-239) is deposited. This is the lowest second potential among the second potentials of the cathode electrode 12. In this embodiment, the separation of U-238 is performed by changing the first potential that is the actual potential of the cathode electrode 12 to the lowest second potential among the second potentials, for example, 0.5 V vs. SHE is performed by depositing UO ₂ on the surface of the cathode electrode 12.

Next, other long half-life nuclides in the radioactive liquid waste are separated by potential control (step S1C ′). The pH of the contaminated water 2 is maintained at 1.0 as in the steps S1A ′ and S1B ′, and the voltage applied between the anode electrode 11 and the cathode electrode 12 is controlled by operating the variable resistor 23. , The potential Eh of the surface of the cathode electrode 12 is the potential within the water stabilization region at pH 1.0, for example, 1.0 Vvs. Adjust to SHE. This 1.0 Vvs. SHE is about 0.9 Vvs., Which is the upper limit potential of region 2 shown in FIG. The potential is higher than SHE. As a result, Pu ³⁺ contained in the contaminated water 2 becomes PuO ₂ [2+] and precipitates as PuO _{2 on} the surface of the cathode electrode 12. When the deposited material 15 (PuO ₂ ) deposited on the surface of the cathode electrode 12 has a predetermined thickness, the cathode electrode 12 is replaced with a new cathode electrode 12 as described above, and Pu from the contaminated water 2 is replaced. Separation continues. In the step S1C ′, the pH is not increased by the injection of the alkaline aqueous solution (for example, sodium hydroxide aqueous solution) into the contaminated water 2 in the radioactive liquid waste container 8 as in the step S1B ′. In this way, Pu is separated from the contaminated water 2.

The potential range of the cathode electrode 12 where each of the two types of long half-life nuclides (U-238 and Pu-239) contained in the contaminated water is deposited, that is, 0 when the pH of the contaminated water 2 is 1.0. .95Vvs. Beyond SHE 1.16Vvs. The potential within the range below SHE is the second potential of the cathode electrode 12 when the former long half-life nuclide (U-238) is deposited and when the latter long half-life nuclide (Pu-239) is deposited. The second potential is the second lowest potential among the second potentials of the cathode electrode 12. In this example, the separation of Pu-239 is performed by changing the first potential that is the actual potential of the cathode electrode 12 to the second potential that is the second lowest among the second potentials, for example, 0.5 Vvs. SHE is performed by depositing PuO ₂ on the surface of the cathode electrode 12. In this embodiment, the surface potential of the cathode electrode 12 is adjusted to a potential within the range between the lower limit of the water stabilization region and the upper limit of the water stabilization region.

  After the separation of U and Pu from the contaminated water 2 is completed, the short half-life nuclide (Cs-137) is adsorbed on the adsorbent 20 in the solidification vessel 19 (step S2), as in Example 4. And each process of preparation of the vitrified body of an adsorbent (step S3) is implemented.

  In the present embodiment, each effect produced in the fourth embodiment can be obtained.

In this embodiment, in the step S1B ′, the pH of the contaminated water 2 is maintained at 1.0, and the surface potential of the cathode electrode 12 is set to 1.0 Vvs. You may adjust to SHE. As a result, each of UO ₂ and PuO ₂ is deposited on the surface of one cathode electrode 12. Since both UO ₂ and PuO ₂ materials are deposited on the surface of the cathode electrode 12 as the deposited material 15, the procedure of the radioactive waste solidification method shown in FIG. 13 of Example 5 can be simplified. it can. In this example, the deposition of both UO ₂ and PuO ₂ on the surface of the cathode electrode 12 is, for example, 1.0 Vvs. This is done by using SHE.

  The adsorption of Cs-137 by the adsorbent 20 filled in the nuclide adsorption tank 24 and the production of the glass solidified body 35 using the low-melting glass shown in FIG. Or may be applied to 5.

  When the contaminated water 2 contains Sr-90 (half-life: 28.90 years), manganese dioxide is used as the adsorbent 20 that adsorbs Sr. When contaminated water 2 contains Cs-137 and Sr-90, a titanium silicate compound that adsorbs Cs-137 and Sr-90, for example, crystallized silicotitanate (CST) is used as adsorbent 20. . When manganese dioxide is used as the adsorbent 20, Sr-90, which is a short half-life nuclide contained in the contaminated water 2, can be removed by adsorption in step S2. Further, when a titanium silicate compound is used as the adsorbent 20, each of Cs-137 and Sr-90, which are short half-life nuclides contained in the contaminated water 2, can be removed by adsorption in step S2. .

  2 ... contaminated water (radioactive waste liquid), 3 ... acid supply device, 8 ... radioactive waste liquid container, 9 ... alkali supply device, 1 ... anode electrode, 12 ... cathode electrode, 13 ... open / close switch, 15 ... depositing substance, 19, 28 ... solidification container (first container), 20 ... adsorbent, 24 ... nuclide adsorption tank, 25 ... waste tank, 29 ... glass raw material tank, 32 ... glass raw material, 34 ... heat insulation container (second container), 35 ... glass Solidified body.

Claims (15)

Separating the long half-life nuclide from the radioactive liquid waste containing the long half-life nuclide and the short half-life nuclide,
The radioactive liquid waste containing the short half-life nuclide is contacted with an adsorbent after the long half-life nuclide is removed, and the short half-life nuclide is adsorbed on the adsorbent to separate the short half-life nuclide from the radioactive liquid waste. And
Supplying the glass raw material into the first container filled with the adsorbent which is the radioactive waste adsorbing the short half-life nuclide;
A method for solidifying radioactive waste, wherein the glass raw material in the first container is melted by heating to produce a vitrified body of the adsorbent.   The separation of the long half-life nuclide is performed by lowering the pH of the radioactive liquid waste to ionize the long half-life nuclide contained in the radioactive liquid waste, and then increasing the pH of the radioactive liquid waste and immersing in the radioactive liquid waste. The solidification processing method of the radioactive waste of Claim 1 performed by depositing the said long half-life nuclide on the surface of the electrode.   When the radioactive waste liquid includes a plurality of types of the long half-life nuclide, the ionization of the long half-life nuclide is performed until the first pH at which all of the plurality of types of the long half-life nuclide is ionized. The deposition of the long half-life nuclide is carried out by lowering the pH, and the pH of the radioactive liquid waste is set to the second pH of the radioactive liquid waste when the long half-life nuclide is deposited on the surface of the electrode. The method for solidifying radioactive waste according to claim 2, wherein the radioactive waste is solidified by depositing the plurality of types of long half-life nuclides separately on the surface of the electrode by increasing from the lowest second pH.   When the radioactive waste liquid includes a plurality of types of the long half-life nuclide, the ionization of the long half-life nuclide is performed until the first pH at which all of the plurality of types of the long half-life nuclide is ionized. The deposition of the long half-life nuclide is carried out by lowering the pH, and the pH of the radioactive liquid waste is set to the second pH of the radioactive liquid waste when the long half-life nuclide is deposited on the surface of the electrode. The method for solidifying radioactive waste according to claim 2, wherein the method is carried out by increasing the highest pH to the second pH and precipitating the plural types of long half-life nuclides on the surface of the electrode.   The separation of the long half-life nuclide is performed by lowering the pH of the radioactive liquid waste to ionize the long half-life nuclide contained in the radioactive liquid waste, and then increasing the potential of the electrode immersed in the radioactive liquid waste. The solidification method of the radioactive waste of Claim 1 performed by depositing a long half life nuclide on the surface of the said electrode.   When a plurality of types of the long half-life nuclide are included in the radioactive liquid waste, the ionization of the long half-life nuclide reduces the pH of the radioactive liquid waste to a pH at which all of the plurality of types of long half-life nuclides are ionized. The deposition of the long half-life nuclide is performed by applying a first potential which is the actual potential of the electrode to the first potential of the electrode when the long half-life nuclide is deposited on the surface of the electrode. The method for solidifying radioactive waste according to claim 5, wherein the radioactive waste is solidified by depositing the plurality of types of long half-life nuclides separately on the surface of the electrode by increasing from the lowest second potential of the two potentials.   When a plurality of types of the long half-life nuclide are included in the radioactive liquid waste, the ionization of the long half-life nuclide reduces the pH of the radioactive liquid waste to a pH at which all of the plurality of types of long half-life nuclides are ionized. The electrode when the long half-life nuclide is deposited on the surface of the electrode by depositing the long half-life nuclide into the first potential that is the actual potential on the surface of the electrode. The method for solidifying radioactive waste according to claim 5, which is performed by increasing the second potential to the highest second potential and depositing the plurality of types of long half-life nuclides on the surface of the electrode.   The method for solidifying radioactive waste according to any one of claims 5 to 7, wherein the potential of the electrode is controlled within a range of a lower limit potential of the water stabilization region and an upper limit potential of the water stabilization region. .   The method for solidifying radioactive waste according to any one of claims 2 to 8, wherein the pH of the radioactive liquid waste is lowered by injecting an acid into the radioactive liquid waste.   The method for solidifying radioactive waste according to any one of claims 2 to 4, wherein the pH of the radioactive liquid waste is increased by injecting alkali into the radioactive liquid waste. The first container in which the adsorbent and the glass raw material are present is disposed in a heat insulating region in the second container,
The adsorbent in the first container existing in the heat insulation region is the heat generated by the radiation emitted from the short half-life nuclide in the heat insulation region when the glass raw material in the first container is melted. The solidification processing method of the radioactive waste of any one of Claim 1 thru | or 10 performed by heating the said glass raw material.   Contact between the radioactive liquid waste containing the short half-life nuclide and the adsorbent is performed by supplying the radioactive liquid waste containing the short half-life nuclide to the first container filled with the adsorbent. The radioactive waste solidification treatment according to any one of claims 1 to 11, wherein the raw material is supplied to the first container after the radioactive waste liquid is discharged from the first container filled with the adsorbent. Method.   The filling of the adsorbent adsorbing the short half-life nuclide into the first container is performed before supplying the glass raw material to the first container. Solidification method of radioactive waste.   The solidification processing method of the radioactive waste of any one of Claim 1 thru | or 13 using the low melting glass whose softening point is 800 degrees C or less as said glass raw material.   15. The radionuclide according to claim 1, wherein the short half-life nuclide is a radionuclide having a half-life of less than 40 years, and the long half-life nuclide is a radionuclide having a half-life of 40 years or more. Solidification method for radioactive waste.

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