JP4787998B2 - Solidification method for radioactive waste - Google Patents

Description

  The present invention relates to a solidification method for radioactive waste particularly suitable for use as incinerated ash-like radioactive waste reduced in volume by incineration.

  It is necessary to dispose of radioactive waste generated in facilities that handle nuclear fuel such as nuclear power plants by finally burying it underground. In order to make radioactive waste into a state suitable for such disposal, various treatments for volume reduction and stabilization are performed. For example, after reducing the volume of combustible waste by incineration in a dedicated incinerator, the radioactive elements do not diffuse to the outside even after a long period of time, and various types of materials can be maintained stably. Processing is performed.

  As various treatment methods for such radioactive waste, a cement solidification method for solidifying radioactive waste using a cement-based solidifying material has been put into practical use. Cement is a widely used material, which is inexpensive and easy to handle, and has many advantages such as high strength. Therefore, the cement solidification method is a particularly useful method as a method for solidifying and fixing radioactive waste.

  This conventional cement solidification method is a technology that can stabilize waste, but incineration ash often contains amphoteric metals such as aluminum, and when incineration ash is solidified with cement, these amphoteric metals become cement. Hydrogen gas may be generated by reacting with alkali components therein, particularly calcium hydroxide. In this case, the generated hydrogen gas may cause expansion, cracking, and voids of the solidified body, and further improvement of the solidification method for fixing cement as a solidifying material is desired.

  As one of such improvement measures, Patent Document 1 (Japanese Patent Publication No. 2-62200) discloses that incineration ash and alkali are mixed in advance and generation of hydrogen gas is terminated to some extent before solidification. A method for suppressing hydrogen generation after mixing is disclosed.

Patent Document 2 (Japanese Patent Laid-Open No. 6-102397) discloses (1) a protective film is formed on the surface of the amphoteric metal on the solidified material in order to suppress the reaction between the amphoteric metal and the solidified material (2) solidified. There has been proposed a method of implementing any one or a plurality of three means for promoting the hydration reaction of cement to the material (3) reducing the alkalinity of the solidified material.
Japanese Examined Patent Publication No. 2-62200 Japanese Patent Laid-Open No. 6-102397

  In the method described in Patent Document 1, a pretreatment process is required in cementing incinerated ash, and when the content of a metal such as aluminum is large, a long time is required for the pretreatment, which causes an increase in treatment costs. There was a problem. In addition, since the metal is dissolved in the pretreatment, there is a possibility that a radionuclide such as iodine may diffuse, and it is necessary to take measures for managing this. .

  Also in the method described in Patent Document 2, a plurality of pretreatments are required for the waste or cement material before the cement material is added, resulting in a problem that the treatment cost increases with the complexity of the equipment.

  In order to solve the above-described problems, the invention according to claim 1 is directed to a step of stirring radioactive waste and magnesium oxide, and the stirred radioactive waste and magnesium oxide to sodium dihydrogen phosphate. And a step of kneading and mixing the radioactive waste.

  The invention according to claim 2 is characterized in that, in the method for solidifying radioactive waste according to claim 1, the radioactive waste is incinerated ash-like radioactive waste.

  The invention according to claim 3 is the radioactive waste solidification method according to claim 1, wherein the radioactive waste contains an amphoteric metal.

  The invention according to claim 4 is the method for solidifying radioactive waste according to any one of claims 1 to 3, wherein the magnesium oxide stirred together with the radioactive waste is light magnesium oxide. Features.

  The invention according to claim 5 is the radioactive waste solidification method according to any one of claims 1 to 3, wherein the magnesium oxide stirred together with the radioactive waste is heavy magnesium oxide. It is characterized by.

The invention according to claim 6 is the method for solidifying radioactive waste according to any one of claims 1 to 5, wherein the ratio of magnesium oxide to sodium dihydrogen phosphate is converted into a molar ratio ( MgO / NaH ₂ PO ₄ ), which is 0.5 to 1.0.

The invention according to claim 7 is the method for solidifying radioactive waste according to any one of claims 1 to 6 , wherein dibasic phosphate obtained from radioactive waste liquid as sodium dihydrogen phosphate to be added. It is characterized by using sodium hydride.

  According to the method for solidifying radioactive waste according to the present invention, a pretreatment step for cement solidification is not required, and only a simple process such as a kneading step is performed. It becomes. Further, since there is no process for dissolving the metal, problems such as radionuclide diffusion do not occur.

  Moreover, according to the solidification processing method of the radioactive waste of this invention, generation | occurrence | production of hydrogen gas is relieved even when filling and solidifying the radioactive waste containing an amphoteric metal.

It is a figure which shows typically the outline of the stirring process in the solidification processing method of the radioactive waste which concerns on embodiment of this invention. It is a figure which shows typically the outline of the kneading | mixing process in the solidification processing method of the radioactive waste which concerns on embodiment of this invention. It is the figure which put together the outline of each process of the solidification processing method of the radioactive waste which concerns on embodiment of this invention in the flow. It is a figure which shows typically the outline of the stirring process in the solidification processing method of the radioactive waste which concerns on other embodiment of this invention. It is the figure which put together the outline of each process of the solidification processing method of the radioactive waste which concerns on other embodiment of this invention in the flow.

Hereinafter, the method for solidifying radioactive waste according to the present invention will be described with reference to the drawings as appropriate. FIG. 1 is a diagram schematically showing an outline of a stirring step in a solidification processing method for radioactive waste according to an embodiment of the present invention, and FIG. 2 is kneading in a solidification processing method for radioactive waste according to an embodiment of the present invention. It is a figure which shows the outline of a process typically. Moreover, FIG. 3 is the figure which put together the outline of each process of the solidification processing method of the radioactive waste which concerns on embodiment of this invention in the flow. 1 and 2, 10 is a solidification vessel, 11 is a stirring device, 12 is a magnesium oxide supply pipe, 13 is a magnesium oxide storage tank, 15 is a magnesium oxide metering device, 16 is a magnesium oxide supply device, 17 is a kneading device, 18 Denotes a phosphoric acid solution supply device, 19 denotes a scatter prevention hood, and 20 denotes a motor.

  In the method for solidifying radioactive waste according to the present invention, first, the radioactive waste incinerated in an incinerator and incinerated ash is charged into the solidification container 1 (step S201).

  In the stirring step in the next step S202, first, magnesium oxide (MgO) is added into the solidification container 10 from the magnesium oxide supply pipe 12. This magnesium oxide is supplied from a magnesium oxide storage tank 13 such as a hopper provided with a magnesium oxide metering device 15. A predetermined amount of magnesium oxide supplied from the magnesium oxide supply pipe 12 is added to the solidification container 10 by a magnesium oxide supply device 16 such as a feeder.

  Magnesium oxide powder quantitatively collected by the magnesium oxide metering device 15 is supplied from the magnesium oxide supply pipe 12 into the solidification container 10. As the magnesium oxide powder, either heavy magnesium oxide or light magnesium oxide can be used. In this embodiment, powdered magnesium oxide is added from the magnesium oxide supply pipe 12. However, a small amount of water is added to the magnesium oxide in advance, and the magnesium oxide in a slurry state is added. Also good.

  The agitator 11 agitates the contents in the solidification container 10 by the rotation of the agitating blade. In the agitation process in step S202, the incinerated ash-like radioactive waste and the magnesium oxide powder added as described above are used. By stirring with the stirring device 11, each is uniformly dispersed.

  The solidification container 10 filled with radioactive waste in which magnesium oxide is uniformly dispersed by the stirring process as described above includes a stirring device 11, a magnesium oxide supply pipe 12, a magnesium oxide storage tank 13, a magnesium oxide metering device 15, and an oxidation device. The system as shown in FIG. 1 comprising the magnesium supply device 16 is transferred to the system as shown in FIG. 2 and undergoes a kneading process.

As shown in FIG. 2, the solidification container 10 is covered with a hood 19 for preventing scattering. Such a scatter prevention hood 19 is to prevent radioactive waste from overflowing from the solidification container 10 during kneading. The scattering prevention hood 19 is inserted with the rotating shaft of the kneading device 17 and the supply of the phosphoric acid solution supply device 18. The kneading device 17 kneads the contents in the solidification vessel 10 with its rotary blades, and a sodium dihydrogen phosphate (NaH ₂ PO ₄ ) solution is added to the solidification vessel 10 from the phosphoric acid solution supply device 18. It has become so.

  The sodium dihydrogen phosphate supplied from the phosphoric acid solution supply device 18 preferably has a concentration of about 30 wt%. This is because the concentration of 30 wt% is the upper limit concentration at which the sodium dihydrogen phosphate salt does not precipitate.

  Next, the kneading process by the system of FIG. 2 configured as described above will be described. First, the temperature of the sodium dihydrogen phosphate solution supplied from the phosphoric acid solution supply device 18 is about 20 ° C to 40 ° C. If the temperature of the sodium dihydrogen phosphate solution is too high, it will solidify rapidly without sufficient kneading in the kneading step. It is desirable that the temperature be about room temperature.

  Next, in the kneading step in step S203, the radioactive waste is started to be stirred by the rotary blades of the kneading device 17 inserted into the solidification container 10. Then, as shown in step S <b> 204, the sodium dihydrogen phosphate solution is supplied into the solidification container 10 by the phosphoric acid solution supply device 18.

Here, the amount of magnesium oxide and sodium dihydrogen phosphate solution added to the radioactive waste will be described. In the solidification processing method for radioactive waste according to the present invention, the ratio of magnesium oxide to sodium dihydrogen phosphate is 0.5 to 1.0 in terms of molar ratio (MgO / NaH ₂ PO ₄ ). One characteristic is preferably about 0.5.

The reason for setting as above is that when the ratio of magnesium oxide and sodium dihydrogen phosphate is less than 0.5 in terms of molar ratio (MgO / NaH ₂ PO ₄ ), floating water remains on the solidified body. In addition, if the ratio exceeds 1.0 in terms of molar ratio, the viscosity of the kneaded product in the kneading step of step S203 increases rapidly, making it difficult to produce a homogeneous solidified body.

  As shown in step S204, when a sodium dihydrogen phosphate solution is added to the solidification container 10 filled with radioactive waste in which magnesium oxide is uniformly dispersed, a reaction represented by the following equation occurs, Crystals of sodium magnesium acid and water are produced.

NaH ₂ PO ₄ + MgO + 5H ₂ O → MgNaPO ₄ .6H ₂ O
In the kneading step of step S203, as such crystals are generated, the radioactive waste in the solidification container 10 gradually increases in viscosity and becomes a solidified body. In the kneading step, it is preferable to further knead for about 10 minutes after adding a prescribed amount of sodium dihydrogen phosphate solution to the radioactive waste.

  The radioactive waste is cured until a sufficient strength is developed although setting starts within one hour after the kneading step (step S205). Compared with the solidification method using a cement-based material, the solidification method for radioactive waste according to the present invention has a remarkably short setting time, and after the addition of magnesium oxide, the curing reaction starts within 10 to 60 minutes and 4 after mixing. It has a significant feature that radioactive waste is completely solidified in about time.

Moreover, the pH of the kneaded mixture of sodium magnesium phosphate obtained by the above reaction formula is in the range of 6.70 to 8.39, and the pH range in which amphoteric metals such as aluminum are stably present is pH 4.1 to 8. 4 ([Non-Patent Document] Nori Seiri: “Aluminum Corrosion,“ Potential-pH Diagram and Polarization Curve in Corrosion Protection Research ”: Light Metal, 57 pp. 371-380
(2007)), the method for solidifying radioactive waste according to the present invention reduces the generation of hydrogen gas even when filling and solidifying radioactive waste containing amphoteric metals such as aluminum and lead. It has a significant feature.

  As described above, according to the method for solidifying radioactive waste according to the present invention, a pretreatment step for cement solidification is unnecessary, and only a simple process such as a kneading step is performed. It becomes possible to suppress. Further, since there is no process for dissolving the metal, problems such as radionuclide diffusion do not occur.

  In the kneading step, a borate (for example, sodium tetraborate) may be added in order to suppress the rapid setting phenomenon in which the viscosity of the kneaded material increases in a short time.

In the solidification method for radioactive waste according to the present invention, the proportion of NaH ₂ PO ₄ in the produced solidified body is in the range of 20 to 50% by weight, more preferably 45 to 50%. Ideally. This is because when the ratio of NaH ₂ PO _{4 in} the solidified body exceeds 50% by weight, the viscosity of the kneaded product increases in the kneading step.

  In the above-described embodiment, the solidified body is prepared in the solidified container 1, but the filled waste body in which sodium magnesium phosphate is shaped into a pellet, filled into the container, and cement mortar is poured therein. It is also possible to make it into a waste.

  Next, another embodiment of the present invention will be described. In the previous embodiment, a reagent was used as the sodium dihydrogen phosphate solution supplied into the solidification container 10, but in this embodiment, the radioactive waste liquid is also used in the sodium dihydrogen phosphate solution supplied into the solidification container 10. To use. The background of the generation of such radioactive waste liquid containing sodium dihydrogen phosphate solution will be described. In the reprocessing of nuclear fuel, uranium, plutonium, etc. are recovered after the spent fuel is dissolved with nitric acid. As a treatment method for this purpose, the PUREX method has been put into practical use. In this PUREX method, tributyl phosphate (TBP) is used as a solvent. When an organic solvent in which TBP is dissolved in dodecane (diluent) is mixed with a nitric acid solution in which spent fuel is dissolved, uranium and plutonium move to the solvent side, and most of fission products remain in the nitric acid solution. This can be used to separate uranium and plutonium from fission products.

As described above, an 85% phosphoric acid (H ₃ PO ₄ ) solution is added to the organic solvent containing TBP used for the recovery of uranium and plutonium. Thereby, an adduct is formed by TBP and phosphoric acid, and this adduct can be separated from dodecane. The separated dodecane is reused in the above treatment as a diluent.

On the other hand, pure water is further added to the adduct formed of TBP and phosphoric acid, and separated into TBP and phosphoric acid. The former is melt-mixed with the plastic to solidify the plastic, and the latter is neutralized with NaOH added to form a sodium dihydrogen phosphate (NaH ₂ PO ₄ ) solution, and the water is evaporated for volume reduction. The sodium dihydrogen phosphate solution is about 30 wt%. Conventionally, this sodium dihydrogen phosphate solution has been studied for solidification by a method using a cement-based material or the like as a radioactive waste liquid. In this method, the weight ratio of sodium dihydrogen phosphate in the solidified product is The limit is about 6%, and a method for producing a solidified body with higher processing efficiency has been sought.

  In the present embodiment, a sodium dihydrogen phosphate solution as the radioactive waste liquid as described above is used, and this is solidified together with the incinerated ash-like radioactive waste to produce a solidified body having a high waste filling rate. And the processing cost can be suppressed.

Other embodiments are the same as the previous embodiment with respect to the stirring step, and therefore, the steps different from the previous embodiment will be described below. FIG. 4 is a view schematically showing an outline of a stirring step in a method for solidifying radioactive waste according to another embodiment of the present invention, and FIG. 5 is a diagram of solidifying radioactive waste according to another embodiment of the present invention. It is the figure which put together the outline of each process of a processing method in the flow.

  As shown in FIGS. 4 and 5, in this embodiment, during the stirring process, radioactive waste liquid is used for the sodium dihydrogen phosphate added in step S204 '.

  Here, the adjustment of the radioactive liquid waste used in step S204 'will be described. The concentration of sodium dihydrogen phosphate used in step S204 'is adjusted to be about 30 wt%. This concentration of 30 wt% is the upper limit concentration at which the sodium dihydrogen phosphate salt does not precipitate. Such a concentration of 30 wt% has been selected to treat the maximum concentration of sodium dihydrogen phosphate while avoiding the difficulty in handling the solution by salt precipitation.

  In addition, the temperature of the radioactive liquid waste used in step S204 'is preferably about 20 ° C to 80 ° C, more preferably about 20 ° C to 40 ° C. If the temperature of the radioactive liquid waste is too high, it will solidify rapidly without sufficient kneading in the kneading step, and thus the temperature of the radioactive liquid waste is preferably about room temperature.

  According to the embodiment as described above, in addition to the same effects as the previous embodiment, the radioactive waste liquid containing sodium dihydrogen phosphate solution together with the rejected ash-like radioactive waste is treated at the same time, so the processing efficiency is extremely high. A solidified body can be produced.

DESCRIPTION OF SYMBOLS 10 ... Solidification container 11 ... Stirrer 12 ... Magnesium oxide supply piping 13 ... Magnesium oxide storage tank 15 ... Magnesium oxide metering device 16 ... Magnesium oxide supply device 17 ... Kneading device 18 ... Phosphoric acid solution supply device 19 ... Scatter prevention hood 20 ... Motor

Claims (7)

A step of stirring the radioactive waste and magnesium oxide;
And a step of adding sodium dihydrogen phosphate to the stirred radioactive waste and magnesium oxide and kneading the mixture. 2. The radioactive waste solidification method according to claim 1, wherein the radioactive waste is incinerated ash-like radioactive waste. The method for solidifying radioactive waste according to claim 1, wherein the radioactive waste contains an amphoteric metal. The method for solidifying radioactive waste according to any one of claims 1 to 3, wherein the magnesium oxide stirred together with the radioactive waste is light magnesium oxide. The method for solidifying radioactive waste according to any one of claims 1 to 3, wherein the magnesium oxide stirred together with the radioactive waste is magnesium oxide heavy. The ratio between magnesium oxide and sodium dihydrogen phosphate is 0.5 to 1.0 in terms of molar ratio (MgO / NaH ₂ PO ₄ ), 6. Solidification method of radioactive waste as described in 1. The method for solidifying radioactive waste according to any one of claims 1 to 6 , wherein sodium dihydrogen phosphate obtained from radioactive waste liquid is used as sodium dihydrogen phosphate to be added.

Images