JP2015094611A - Treatment method of radioactive contaminated water and treatment device thereof - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a treatment method of radioactive waste, which can efficiently convert sludge containing radionuclides in high concentration into a solidified substance excellent in safety by using a general heating furnace.SOLUTION: A treatment method of radioactive waste includes a process of making dry sludge by removing water from sludge which is generated at a treatment facility of radioactive contaminated water and which contains radionuclides in high concentration, and a process of vitrifying the dry sludge by adding a melting agent made of glass materials including sodium and applying the heat of a melting temperature of 1100°C or lower to the dry sludge.

Description

  Embodiments described herein relate generally to a method for treating radioactive contaminated water and an apparatus for treating radioactive contaminated water.

  If an accident occurs in the nuclear reactor due to an unforeseen natural disaster or the like, the core may be exposed from the reactor water, and the core melt may melt due to the decay heat of the nuclear fuel.

  In this case, there is a possibility that the high-temperature core melt (corium) melts down to the lower part of the reactor pressure vessel, further melts and penetrates the lower part of the reactor pressure vessel, and falls onto the floor in the containment vessel.

  In such an emergency situation, a large amount of cooling water is supplied from the outside to cool the core melt that has fallen into the containment vessel, and as a result, the underground floors of the turbine building, reactor building, etc. A large amount of radioactive polluted water becomes accumulated.

  The radioactive polluted water staying in the reactor facility can be used for cooling the reactor by removing and purifying the radionuclide. For this reason, in the nuclear reactor facility where radioactive polluted water stays, it is composed of a processing device composed of a cesium adsorption device, a second cesium adsorption device, a decontamination device, a reverse osmosis membrane device, an evaporator device, etc. Contaminated water purification system consisting of desalination equipment will be installed.

  However, in the above-described contaminated water purification system, it is difficult to remove all radionuclides other than cesium, and some radionuclides such as α nuclides remain in the treated water.

  Therefore, in order to remove the multiple types of radionuclides remaining in the radioactively contaminated water with a higher degree of accuracy and purify them to a level below the ocean release level, a separate multinuclide removal device may be installed after the contaminated water purification system. It is being considered.

  Multi-nuclide removal equipment is an iron coprecipitation treatment facility that removes radionuclides such as α-nuclide, Co-60, and Mn-60 contained in radioactive polluted water, and carbonate precipitation that removes adsorption-inhibiting ions such as Mg and Ca. A multi-nuclide that adsorbs and removes radionuclides remaining in a small amount in the pre-treatment device treated water by passing the pre-treatment device consisting of treatment equipment and the treated water of this pre-treatment device through an adsorbent such as activated carbon, zeolite, chelate resin It is comprised by the removal apparatus. According to such a multi-nuclide removal apparatus, it is possible to remove the radionuclide contained in the radioactively contaminated water with high accuracy.

  In such a pretreatment device of the multi-nuclide removal device, precipitates (hereinafter referred to as sludge) are generated in the iron coprecipitation treatment facility and the carbonate precipitation treatment facility, but these sludges have a relatively high dose. Therefore, a technique for solidifying these as a stable storage body is required for intermediate storage.

  As a method for solidifying radioactive waste, for example, solid waste such as high-level resin or liquid waste is put into a kneading machine with a hydraulic solidifying material and added water, and a stirring blade provided in the kneading machine is used. A method is disclosed in which a mixed paste is produced by rotation and poured into a solidification container to solidify. However, since sludge generally contains an extremely high concentration of radionuclide, when moisture is contained in the solidified body, moisture is radiolyzed by the high dose of radiation released from the sludge, There is a risk of generating flammable gases such as hydrogen. Moreover, since sludge is hard and is easily crushed, it is necessary to suppress scattering and the like during handling.

  On the other hand, a method in which the obtained sludge is mixed with a glass raw material and heated at a high temperature to be vitrified has been studied. In this method, since sludge is heated at a high temperature, it is possible to form a solidified body without containing moisture that causes flammable gas generation by radiolysis, and there is a problem of scattering due to crushing during handling. Is also resolved.

  However, if the sludge is simply mixed with a general-purpose glass raw material such as silicon oxide, boric acid, or aluminum oxide, it is difficult to vitrify uniformly unless the temperature is considerably increased. In particular, when the processing temperature for vitrification exceeds 1200 ° C., the required performance for the heat resistance of the heating furnace is further increased, such as the need to use a special refractory material as the furnace wall material of the heating furnace. There was a possibility that the cost of waste disposal would increase.

Patent No. 3048806

  The present invention has been made in order to solve the above-described problem, and sludge containing a high concentration of radionuclide is efficiently processed as a solidified body having excellent safety using a general heating furnace. An object of the present invention is to provide a method for treating radioactive waste.

  The method for treating radioactive waste according to the embodiment includes a step of removing moisture from sludge containing radioactive nuclides generated in a radioactive polluted water treatment facility at a high concentration to obtain dry sludge. Further, the method for treating radioactive waste includes a step of adding a melting aid made of a glass raw material containing sodium to the dried sludge and heating the dried sludge at a melting temperature of 1100 ° C. or less for vitrification.

  According to the present invention, sludge containing a radionuclide at a high concentration can be efficiently processed at low cost as a solidified body excellent in safety.

Schematic of a radioactive polluted water treatment facility (multi-nuclide removal device). Schematic of the radioactive waste processing apparatus which concerns on embodiment. Schematic of the radioactive waste processing apparatus which concerns on embodiment. The figure which shows the measurement result by X-ray diffraction of the vitrified body obtained in Example 2. FIG.

  The processing method of the radioactive waste which concerns on embodiment is demonstrated.

  Specifically, as the sludge applied to the radioactive waste processing method of the embodiment, for example, it is generated at a radioactive contaminated water treatment facility (hereinafter, referred to as a multi-nuclide removal apparatus 1) shown in FIG. Sludge.

  The multi-nuclide removal device 1 is a device for removing radio nuclides from radioactive polluted water containing radioactive nuclides (hereinafter simply referred to as contaminated water). For example, radio nuclides dissolve in rainwater, groundwater, seawater, etc. Then, various radioactive nuclides are generated from the treated water after removing cesium from the contaminated water staying in the reactor building with a cesium removal device (not shown), and concentrated water and permeated water by the reverse osmosis membrane of this treated water. Removed.

  In FIG. 1, the multi-nuclide removal apparatus 1 includes a pre-treatment having an iron coprecipitation processing unit 2 that removes α-nuclide and the like, and a carbonate precipitation processing unit 3 that removes adsorption-inhibiting ions such as Mg and Ca. An apparatus 4, an adsorption tower 6 having an adsorbent 5 such as zeolite or chelate resin, and a storage tank 7 for storing treated water treated in the adsorption tower 6 are provided.

  Specifically, in the multi-nuclide removal apparatus 1, for example, radioactive contaminated water is supplied to a coprecipitation tank (not shown) of the iron coprecipitation processing unit 2 via a pipe 8. In the coprecipitation tank (not shown) of the iron coprecipitation processing unit 2, for example, ferric chloride is added as a coprecipitation agent, and sodium hydroxide is further added as a pH adjuster to produce iron hydroxide. . Radionuclides such as α nuclides in radioactively contaminated water co-precipitate with iron hydroxide and are radioactive as sludge containing iron as a main component (hereinafter, sludge containing iron as a main component is simply referred to as iron-containing sludge). Removed from contaminated water (iron coprecipitation method).

  Examples of the radionuclide contained in the sludge generated in the iron coprecipitation processing unit 2 include Pu-238, Pu-239, Pu-240, Am-241, Am-242m, Am-243, Cm-242, and Cm-243. , Cm-244 and other α-ray emitting nuclides (hereinafter referred to as α nuclides), and Co-60 and Mn-54 and other γ-ray emitting nuclides (hereinafter referred to as γ nuclides).

  The treated water of the iron coprecipitation treatment unit 2 is supplied as treated water to a precipitation tank (not shown) of the carbonate precipitation treatment unit 3 via a pipe (not shown). In the carbonate precipitation processing unit 3, for example, sodium carbonate and sodium hydroxide are added to a coprecipitation tank (not shown) as a precipitation forming agent, and Mg, Ca, and Sr that are adsorption inhibitors in the adsorbent 5 in the subsequent stage. -90 grade (β-ray emitting nuclide) radionuclide is precipitated and precipitated as carbonate, and sludge containing carbonate as a main component (hereinafter, sludge containing carbonate as a main component is simply referred to as carbonate-containing sludge. ) To be removed from the water to be treated (precipitation precipitation method).

  The treated water of the carbonate precipitation treatment unit 3 is supplied as treated water to the adsorption tower 6 via the pipe 9. The water to be treated supplied to the adsorption tower 6 is passed through the adsorbent 5 such as zeolite, and radionuclides such as Sr and Cs are adsorbed and removed, and then passed through the pipe 10 to the storage tank 7. Supplied.

  The radioactive waste processing method according to the embodiment can be applied to, for example, iron-containing sludge and carbonate-containing sludge generated by the pretreatment device 4.

  FIG. 2 is a schematic configuration diagram of a radioactive waste treatment apparatus applied to the radioactive waste treatment method of the present invention.

  In FIG. 2, a radioactive waste treatment apparatus 20 includes a sludge receiving tank 21 that contains sludge, a sludge heating and dehydrating apparatus 22 that contains sludge supplied from the sludge receiving tank 21 and heat-dehydrates to produce dry sludge. The sludge measuring tank 23 is sequentially installed.

  Examples of the sludge stored in the sludge receiving tank 21 include iron-containing sludge and carbonate-containing sludge generated by the pretreatment device 4 of the above-described multi-nuclide removal device 1.

  The sludge receiving tank 21 and the sludge heating / dehydrating device 22 and the sludge heating / dehydrating device 22 and the sludge metering tank 23 are connected by a pipe 24 and a pipe 25, respectively.

  On the other hand, the radioactive waste treatment apparatus 20 includes a melting aid tank 26 for storing a melting aid made of a glass raw material containing sodium, and a melting aid measuring tank 27, a sludge receiving tank 21, a sludge heating and dehydrating device 22. And it is installed in parallel with the line of the sludge measuring tank 23. The melting aid tank 26 and the melting aid measuring tank 27 are connected by a pipe 28.

  In the subsequent stage of the sludge metering tank 23 and the melting auxiliary agent measuring tank 27, a dry sludge / melting auxiliary agent mixing device 29 for mixing the dry sludge and the melting auxiliary agent at a predetermined ratio, and a dry sludge / melting auxiliary agent mixing device 29 are mixed. And a treatment tank 30 to which a mixture of the dried sludge obtained in step 1 and the melting aid is supplied. The dry sludge / melting aid mixing device 29 and the treatment tank 30 are connected by a pipe 31. The processing tank 30 is provided with a heating device 32 that heats and melts the mixture of the dried sludge and the melting aid accommodated in the processing tank 30 so as to surround the outer peripheral surface of the processing tank 30. .

  The processing tank 30 is connected with an exhaust pipe 33 provided with, for example, a bag filter and the like for ensuring the safety of an off-gas system. The treatment tank 30 may be appropriately equipped with a stirring device.

  The dry sludge / melting aid mixing device 29 may be provided with a stirring blade, or may be capable of mixing by a blender (vibration method).

  As a material which comprises the processing tank 30, if it has normal heat resistance and corrosion resistance in 800-1500 degreeC, it can apply, without specifically limiting.

  Although it does not specifically limit as a size of the processing tank 30, The size of the storage container of the vitrified body for high level radioactive waste, ie, the processing tank of the size of internal diameter 430mm x height 1340mm can be used conveniently.

  A pipe 34 and a pipe 35 are connected to the sludge metering tank 23 and the melting auxiliary agent measuring tank 27, respectively, and are connected at a point P to a pipe 36 connected to the dry sludge / melting auxiliary agent mixing device 29. .

  The treatment tank 30 is not limited to the configuration shown in FIG. 2. For example, as shown in FIG. 3, a flow-down valve 37 is provided in the treatment tank 30, and a storage container 38 is installed at the subsequent stage of the treatment tank 30. A configuration may be adopted.

  When the sludge supplied to the sludge receiving tank 21 is a carbonate-containing sludge, as the melting aid, one containing sodium, boron, aluminum and silicon can be used. By mixing a melting aid containing sodium, boron, aluminum and silicon with a carbonate-containing sludge, the carbonate-containing sludge can be added at 1100 ° C. or less without excessively increasing the addition amount of the melting aid. It can be solidified as a vitrified glass integrated as a matrix at the heating temperature.

As the glass raw material constituting the melting aid added to the carbonate-containing sludge, a compound containing at least one element selected from sodium, boron, aluminum and silicon as a main component can be used. Specifically, as the glass raw material, for example, those composed of the following components (A) to (C) can be used.
(A) At least one selected from (a) sodium borate or (b) a combination of a sodium compound and boric acid.
(B) (c) At least one selected from a combination of sodium aluminate or (d) a sodium compound and aluminum oxide.
(C) At least one selected from (e) sodium silicate or (f) a combination of a sodium compound and silicon oxide.

  In addition, although it does not specifically limit as a sodium compound contained in (A) component, (B) component, and (C) component, For example, sodium carbonate, sodium hydroxide, etc. can be used.

  As a melting aid to be added to the carbonate-containing sludge, by using one containing the above-mentioned components (A), (B) and (C), for example, high concentrations of adsorption-inhibiting ions such as Mg and Ca. The sludge contained can be solidified at a heating temperature of 1100 ° C. or less as a solidified body that is well integrated with few voids and is fully vitrified. For this reason, carbonate containing sludge can be processed as a vitrified body with high density and excellent water resistance in the temperature range described above.

  (A) It is preferable to mix | blend sodium borate with a fusion aid so that it may become 0.1 times or more by mass ratio with respect to dry sludge. When the amount of (a) sodium borate blended in the melting aid is less than 0.1 times by mass with respect to the dry sludge, a sufficiently integrated matrix at a melting temperature of 1100 ° C. or less There is a possibility that the solidified body cannot be obtained, and the porosity in the obtained glass solidified body increases.

  More preferably, (a) sodium borate is blended in the melting aid so that the mass ratio is 0.8 times or more with respect to the dry sludge. When the amount of (a) sodium borate blended in the melting aid is less than 0.8 times by mass with respect to the dry sludge, when the mixture of the sludge and the melting aid is heated and melted, The whole is vitrified before it is vitrified, so that crystals are likely to remain in the obtained vitrified body, and devitrification may occur in a part of the vitrified body. In this case, sufficient water resistance cannot be obtained in the obtained vitrified body, and problems such as radionuclide outflowing easily occur.

  (C) It is preferable to mix | blend sodium aluminate with a fusion aid so that it may become 0.4 times or more by mass ratio with respect to dry sludge. If the amount of (c) sodium aluminate blended in the melting aid is less than 0.4 times by mass with respect to the dry sludge, a stable vitrified product cannot be obtained, and the obtained vitrified product There is a risk that the weather resistance of the body will be reduced. For this reason, the vitrified body is easily eroded by contact with moisture or the like, and problems such as radionuclide outflowing easily occur. The amount of (c) sodium aluminate blended in the melting aid is preferably 0.5 times or more by mass ratio with respect to the dry sludge.

  (E) It is preferable to mix | blend sodium silicate with a fusion aid so that it may become 0.5 time or more by mass ratio with respect to dry sludge. If the amount of (e) sodium silicate blended in the melting aid is less than 0.5 times by mass with respect to the dry sludge, a stable vitrified product cannot be obtained, and the resulting glass There is a possibility that the porosity in the solidified body increases. (E) The amount of sodium silicate is more preferably 0.65 times or more by mass ratio with respect to the dry sludge.

  In a melting aid containing (a) sodium borate, (c) sodium aluminate and (e) sodium silicate, the sum of the amount of (c) sodium aluminate and the amount of (e) sodium silicate Calculated by the ratio of the amount of sodium borate to the amount, that is, (a) the amount of sodium borate / (c) the amount of sodium aluminate + (e) the amount of sodium silicate) It is preferable that the value to be measured is 1.02-0.44 by mass ratio.

  When the value calculated by (a) sodium borate blending amount / ((c) sodium aluminate blending amount + (e) sodium silicate blending amount) is less than 0.44 in terms of mass ratio, glass When the softening temperature of the glass is increased and a melting temperature of 1100 ° C. or lower is obtained, a solidified body as a sufficiently integrated matrix cannot be obtained, and the porosity of the obtained glass solidified body may increase. On the other hand, when the value calculated by (a) the amount of sodium borate / ((c) the amount of sodium aluminate + (e) the amount of sodium silicate) exceeds 1.02, A stable vitrified body cannot be obtained. More preferably, the value calculated by (a) the amount of sodium borate / ((c) the amount of sodium aluminate + (e) the amount of sodium silicate) is 0.6 to 1 in terms of mass ratio. It is preferable that (A) Sodium borate compounding amount / ((c) Sodium aluminate compounding amount + (e) Sodium silicate compounding amount) The mixture with the auxiliary agent can be vitrified well by heating and melting, there is no devitrification due to the residual of the crystal, and a vitrified body excellent in water resistance can be obtained.

  When the sludge supplied to the sludge receiving tank 21 is iron-containing sludge, as the melting aid, one containing sodium and phosphorus can be used. By mixing the melting aid containing sodium and phosphorus with the iron-containing sludge, the iron-containing sludge can be used as a matrix at a heating temperature of 1100 ° C. or less without excessively increasing the addition amount of the melting aid. It can be solidified as an integrated vitrified body.

  As the glass raw material constituting the melting aid to be added to the iron-containing sludge, it is preferable to use a material composed of at least one selected from (g) sodium phosphate or (h) a combination of a sodium compound and phosphoric acid. .

  In addition, although it does not specifically limit as a sodium compound, For example, sodium carbonate, sodium hydroxide, etc. can be used.

  Among them, (g) sodium phosphate is a sludge containing iron in a high concentration at a heating temperature of 1100 ° C. or less, is well integrated with little porosity and is fully glassy and has excellent water resistance. Since a glass solid can be formed, it is preferable.

  When (g) sodium phosphate is used as a melting aid, the amount of (g) sodium phosphate used is preferably 0.5 times or more in terms of mass ratio with respect to dry sludge.

  Note that the radioactive waste treatment apparatus 20 according to the embodiment may be formed integrally with the above-described multi-nuclide removal apparatus 1 or may be formed separately from the multi-nuclide removal apparatus 1.

  In addition, the radioactive waste processing apparatus 20 according to the embodiment is not necessarily limited to the above-described configuration. For example, the drying sludge / melting aid mixing apparatus 29 is not provided, and the processing tank 30 has a dry sludge / melting assistance. It is good also as a structure which combines the function of the agent mixing apparatus 29. FIG.

  Next, based on FIG. 2, the radioactive waste processing method by the radioactive waste processing apparatus 20 of this invention is demonstrated.

  First, the sludge accommodated in the sludge receiving tank 21 is supplied into the sludge heating and dehydrating device 22 through the pipe 24.

  In the sludge heating and dehydrating device 22, moisture is removed from the sludge to obtain dry sludge. The dry sludge generated by the sludge heating and dehydrating device 22 is supplied to the sludge metering tank 23 through the pipe 25 and a predetermined amount is measured, and is supplied to the dry sludge / melting aid mixing device 29 through the pipe 34 and the pipe 36.

  On the other hand, the melting auxiliary agent stored in the melting auxiliary agent tank 26 is supplied to the melting auxiliary agent measuring tank 27 through the pipe 28, a predetermined amount is measured, and the dry sludge / melting auxiliary agent is mixed through the pipes 35 and 36. Supplied to the device 29.

  The molten auxiliary agent and the dried sludge supplied to the dry sludge / melt auxiliary agent mixing device 29 are mixed by stirring with a stirring blade or a vibration method (blender), and supplied to the treatment tank 30 as a mixture through the pipe 31. The

  The mixture of the dried sludge and the melting aid accommodated in the processing tank 30 is heated and melted by the heating device 32 while exhaust gas in the processing tank 30 is exhausted by the exhaust pipe 33.

  The heating and melting temperature of the mixture of the dried sludge and the melting aid is performed at a temperature of 1100 ° C. or lower. When the heating and melting temperature exceeds 1100 ° C., the heat resistance required for the treatment tank 30 becomes excessively high, which may increase the manufacturing cost. On the other hand, the heating and melting temperature of the mixture of dry sludge and melting aid is preferably 900 ° C. or higher. When the heating and melting temperature of the mixture of the dry sludge and the melting aid is less than 900 ° C., the mixture of the dry sludge and the melting aid cannot obtain a sufficient molten state, and the obtained solidified body is sufficiently integrated. Otherwise, the component added as a melting aid or the component of the dry sludge is not sufficiently amorphized, and the obtained solidified body may be devitrified.

  The heating and melting temperature of the mixture of dry sludge and melting aid is more preferably 930 ° C or higher and 1050 ° C or lower.

  The heating and melting time can be appropriately adjusted according to the size of the treatment tank 30, the amount of the mixture of dry sludge and melting aid, and the like.

  Since the volume of the mixture of dry sludge and melting aid supplied into the processing tank 30 is reduced by heating and melting, after the mixture is supplied to the processing tank 30, there is a space above the processing tank 30. When formed, it is preferable to supply a new mixture to this space and perform heat melting treatment.

  The sludge heated and melted in the treatment tank 30 is allowed to cool for a predetermined time to form a vitrified body. The processing tank 30 filled with the glass solidified body is closed by closing the opening, and transferred to an intermediate storage facility.

  In the case where the radioactive waste processing apparatus 20 has a configuration in which the flow valve 37 is disposed in the processing tank 30, the mixture of the dried sludge and the melting aid is heated and melted in the processing tank 30. Thereafter, the flow valve 37 is opened and supplied into the storage container 38. The melt supplied into the storage container 38 is gradually cooled in the storage container 37 to form a glass solid.

  As mentioned above, although embodiment of this invention was described referring a specific example, said Example is mentioned as an example of this invention and does not limit this invention. Further, in the description of each of the above embodiments, in the radioactive waste processing method and the radioactive waste processing apparatus, description of parts and the like that are not directly required for the description of the present invention is omitted. Each element can be appropriately selected and used.

  In addition, all the radioactive waste processing methods and radioactive waste processing apparatuses that are provided with the elements of the present invention and whose design can be changed as appropriate by those skilled in the art without departing from the spirit of the present invention are included in the scope of the present invention. Is done. The scope of the present invention is defined by the appended claims and equivalents thereof.

  Hereinafter, the present invention will be described in more detail with reference to Examples and Comparative Examples, but the present invention is not limited to these Examples. Examples 1 to 7 and Examples 8 to 9 are examples.

(Examples 1-7)
First, 10 g of dried sludge (1) obtained by drying carbonate-containing sludge is put into an alumina crucible, and then Na ₂ Al ₂ O ₄ (sodium aluminate), Na ₂ SiO ₃ (silica) as a melting aid. Acid sodium) and Na ₂ B ₄ O ₇ (sodium borate) were mixed in the proportions shown in Table 1 and charged into an alumina crucible. The mixture of the melting aid and the dried sludge (1) was heated and melted at 1000 ° C. for 3 hours in an alumina crucible in an air atmosphere. The obtained melt was left to cool at room temperature in an alumina crucible to obtain a glass solid.

  The appearance of each vitrified body obtained in Examples 1 to 7 was visually observed. Table 1 shows the evaluation results of the appearances of the respective vitrified bodies together with the blending amounts, melting temperatures, and melting times of the components of the dry sludge and melting aid.

In Examples 1 to 6 in which Na ₂ Al ₂ O ₄ , Na ₂ SiO ₃ , and Na ₂ B ₄ O ₇ were blended as melting aids to the dry sludge (1), the melting temperature was 1000 ° C. A solidified body comprising a mixture of the dried sludge (1) and the melting aid could be obtained. In particular, in Examples 2 to 3 and Examples 5 to 6 in which the added amount of Na ₂ B ₄ O ₇ is 8 to 10 g with respect to 10 g of the dried sludge (1), the vitrified body is well integrated without devitrification. It was confirmed that a solidified body was formed. On the other hand, in Example 7 in which Na ₂ B ₄ O ₇ was not added, a solidified body composed of a mixture of the dried sludge (1) and the melting aid was obtained, but the overall appearance was powdery. It was confirmed that the integration as a solidified body did not proceed sufficiently. Further, in Examples 1 and 4 in which the blending amount of Na ₂ B ₄ O ₇ is 6 g with respect to 10 g of the dried sludge (1), the obtained solidified body was integrally formed, but vitrification was sufficient. It was confirmed that a part of the crystalline material remained in the solidified body and devitrification occurred.

When the density of the vitrified body of Example 2 was measured, it was 3.35 g / cm ³ , and the bulk density increased as compared with the sludge after drying. It was confirmed that a volume reduction effect was obtained. Further, when the normalized leaching rate by the PCT-B method was measured for the vitrified body of Example 2, it was about 0.04 g · m ⁻² · d ⁻¹ in the Si component, and sufficient water resistance was recognized. .

(Examples 8 to 9)
First, 10 g of dried sludge (2) obtained by drying iron-containing sludge was charged into an alumina crucible, and then Na ₃ PO ₄ (sodium phosphate) was added as a melting aid in the proportions shown in Table 2. The mixture of the dried sludge (2) and the melting aid was heated and melted in an alumina crucible in an air atmosphere at 1000 ° C. for 3 hours, and then allowed to cool at room temperature to obtain a solidified glass. .

  The appearance of the vitrified bodies obtained in Examples 8 to 9 was visually observed. Table 2 shows the evaluation results of the appearance of the respective vitrified bodies, together with the blending amounts of dry sludge and melting aid, melting temperature, and melting time.

In Example 9 in which 10 g of Na ₃ PO ₄ was added as a melting aid to 10 g of dry sludge, the dry sludge containing iron as a main component was integrated with the melting aid under the condition of a melting temperature of 1000 ° C. It was confirmed that a solidified body could be obtained. On the other hand, in Example 8 where the blending amount of Na ₃ PO _{4 with} respect to 10 g of dry sludge was 5 g, the obtained solidified body was integrally formed, but vitrification was not sufficiently progressed and partly lost. It was confirmed that transparency was generated.

When the density of the vitrified body of Example 9 described above was measured, it was 2.7 g / cm ³ , and the bulk density was increased as compared with the sludge after drying. By melt mixing with the melting aid, A sufficient volume reduction effect was observed. Moreover, when the normalized leaching rate by the PCT-B method was measured for the vitrified body of Example 9, sufficient water resistance was recognized.

(Example 10)
First, 10 g of dried sludge (1) is put into an alumina crucible, and then Na ₂ Al ₂ O ₄ , Na ₂ SiO ₃ , and Na ₂ B ₄ O ₇ are used in the same proportion as in Example 1 as melting aids. After mixing, the mixture was put into an alumina crucible. The mixture of the melting aid and the dried sludge (1) was heated and melted at 800 ° C. for 3 hours in an alumina crucible in an air atmosphere, and then allowed to cool at room temperature to obtain a solidified glass. . When the obtained vitrified body was visually observed, it was confirmed that the entire appearance was powdery and integration as a solidified body did not sufficiently proceed.

  The vitrified body obtained in Example 2 was subjected to X-ray diffraction using an X-ray diffractometer. The measurement result by X-ray diffraction is shown in FIG.

  As is clear from FIG. 4, in the vitrified body obtained in Example 2, the peaks of the crystal components contained in the sludge and the heat aid before the heat treatment disappear, and the vitrified body is made amorphous as a vitrified body. It was confirmed that

  According to at least one embodiment described above, sludge containing a high concentration of radionuclide can be efficiently processed at low cost as a solidified body having excellent safety.

  Although several embodiments of the present invention have been described, these embodiments are presented by way of example and are not intended to limit the scope of the invention. These novel embodiments can be implemented in various other forms, and various omissions, replacements, and changes can be made without departing from the scope of the invention. These embodiments and modifications thereof are included in the scope and gist of the invention, and are included in the invention described in the claims and the equivalents thereof.

DESCRIPTION OF SYMBOLS 1 ... Multi-nuclide removal apparatus, 2 ... Iron coprecipitation processing part, 3 ... Carbonate precipitation processing part, 4 ... Pretreatment apparatus, 5 ... Adsorbent, 6 ... Adsorption tower, 7 ... Storage tank, 8-10, 24- 25, 28, 34 to 36 ... piping, 20 ... radioactive waste processing device, 21 ... sludge receiving tank, 22 ... sludge heating and dewatering device, 23 ... sludge metering tank, 26 ... melting aid tank, 27 ... melting aid Weighing tank,
29 ... Dry sludge / melting aid mixing device, 30 ... Treatment tank, 32 ... Heating device, 33 ... Exhaust piping, 37 ... Flow valve, 38 ... Storage container

Claims (11)

  A process of removing moisture from sludge containing radioactive nuclides generated at a radioactive polluted water treatment facility at a high concentration to make dry sludge, and a melting aid made of glass raw material containing sodium is added to this dry sludge to 1100 ° C or less And a vitrification step by heating at the melting temperature of the radioactive waste.   The method for treating radioactive waste according to claim 1, wherein the melting aid contains sodium, boron, aluminum and silicon.   The method for treating radioactive waste according to claim 1 or 2, wherein the glass raw material comprises a compound containing as a main component at least one element selected from sodium, boron, aluminum and silicon. The said glass raw material consists of the following (A) component-(C) component, The processing method of the radioactive waste of any one of the Claims 1 thru | or 3 characterized by the above-mentioned.
(A) At least one selected from (a) sodium borate or (b) a combination of a sodium compound and boric acid.
(B) (c) At least one selected from a combination of sodium aluminate or (d) a sodium compound and aluminum oxide.
(C) At least one selected from (e) sodium silicate or (f) a combination of a sodium compound and silicon oxide.   The method for treating radioactive waste according to claim 1, wherein the melting aid contains sodium and phosphorus.   The said glass raw material consists of at least one chosen from the combination of (g) sodium phosphate or (h) sodium compound and phosphoric acid, The processing method of the radioactive waste of Claim 1 or 5 characterized by the above-mentioned. .   The sludge containing the radionuclide at a high concentration is sludge formed by adding a precipitation-forming agent or a coprecipitation agent to the radioactively contaminated water. Radioactive waste disposal method.   The radioactive waste processing method according to any one of claims 1 to 4 and 7, wherein the sludge containing the radionuclide at a high concentration is a sludge mainly composed of carbonate.   The method for treating radioactive waste according to any one of claims 1 and 5 to 7, wherein the sludge containing the radionuclide at a high concentration is a sludge mainly composed of iron. A sludge heating and dehydrating device that contains sludge containing the radionuclide according to any one of claims 1 to 9 in a high concentration and that heats and dehydrates to produce dry sludge;
A dry sludge / melting aid mixing device that mixes dry sludge produced by the sludge heating and dehydrating device and a melting aid made of a glass raw material containing sodium at a predetermined ratio;
An apparatus for treating radioactive waste, comprising: a heating device that heats and melts a mixture of dry sludge and melting aid produced by the dry sludge / melting aid mixing device to vitrify the mixture.   11. The radioactive waste treatment apparatus according to claim 10, further comprising a pretreatment facility for generating sludge containing a high concentration of radionuclides from the radioactive contaminated water by iron coprecipitation method or precipitation precipitation method.

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