JP2014098637A - Method and device for treating radioactive waste solution - Google Patents

Abstract

PROBLEM TO BE SOLVED: To treat a radioactive solution including radioactive waste without generating hydrogen system gas, such as tritium gas, by solidifying with a simplified method.SOLUTION: An oxidation reduction potential of a radioactive solution is set to 0 mV or higher in a reaction vessel. Then, the radioactive solution is mixed with cement in a cement solidification tub to make a solidified body. During the solidification the oxidation reduction potential of moisture content included in the solidified body is kept to 0 mV or higher.

Description

  The present invention relates to a processing method and a processing apparatus for radioactive liquid waste.

  The radioactive liquid waste generated from nuclear facilities contains various radioactive substances. Generally, such radioactive waste is assumed to be disposed of after being cemented and vitrified. The cement solidification method is cheap and easy to process, so it is applied to solidify many wastes. However, if the amount of radioactive material contained in the radioactive liquid waste is large, that is, if the radionuclide concentration is high, It is assumed that pore water, crystal water, and the like existing therein are radiolyzed, and as a result, hydrogen gas is generated.

  In order to suppress the release of hydrogen gas from the solidified container, it is conceivable that the container is sealed, but there is a concern about an increase in internal pressure during long-term storage. For this reason, it is assumed that the pressure in the container will increase after the landfill disposal, affecting the soundness of the disposal site. From this point of view, CSD-C Universal Canister is equipped with a filter on the top so that the generated gas can be released in the container that holds the compressed body of spent fuel cladding tube, which is a waste with high radionuclide concentration. Is being studied (Non-patent Document 1).

  However, when such a filter is attached, if water containing a radioactive substance such as tritium is present inside, there is a concern that a gas containing the radioactive substance is generated and the radioactive gas is released to the outside of the container. Moreover, although the method of arrange | positioning a fixed chemical | medical agent in a container is proposed (patent document 1), when a fixed chemical | medical agent does not function normally, possibility of causing a pressure rise cannot be denied.

  On the other hand, vitrification is also applied to the solidification of high-level radioactive waste and is considered to be a stable solidification method. Glass is considered to be suitable for solidification treatment of high-dose waste because it has high radiation resistance and no concern about the generation of hydrogen gas (Non-Patent Document 2). However, since the temperature required for producing the vitrified body is as high as about 1200 to 1300 ° C. and requires a large cost for maintenance of the melting furnace, it is difficult to apply the high-level radioactive waste to other than the solidified body. It is.

  From such a viewpoint, a vitrification method using a sol-gel method has been proposed as a vitrification method for radioactive waste (Patent Document 1). However, the glass produced by the sol-gel method is difficult to produce a bulk body. In Patent Document 2, a compression molding machine is required to perform compression molding and sintering after calcining the dried gel, In addition to the complicated manufacturing process, there is a problem that the manufacturing cost increases.

JP 2001-228296 A JP 2009-115490 A

F.Chotin, G.Limeui, G.Maurin., “Quality and Safety of COGEMA Universal Canister Filled with Compacted HULL, END-PIECES and Technological Waste”., The 5th International Conference on Recycling, Conditioning and Disposal (RECOD 98) Proceedings Vol.2 pp636-643, Oct.25-28 (1998), Nice-France JNC TN1400 99-020 Technical reliability of geological disposal of high-level radioactive waste in Japan-Secondary research and development of geological disposal

  An object of the present invention is to solidify and process a radioactive liquid waste containing a radioactive waste by a simple method without generating a hydrogen-based gas such as tritium gas.

  One aspect of the present invention relates to a method for treating a radioactive liquid waste, comprising the steps of: setting a redox potential of the radioactive liquid waste to 0 mV or more; and cementing the radioactive liquid waste. The “redox potential” in the present invention is a redox potential measured using a standard hydrogen electrode as a reference electrode.

  Moreover, one aspect of the present invention is a reaction vessel for setting the oxidation-reduction potential of radioactive waste liquid to 0 mV or more, a cement solidification tank for mixing the radioactive waste liquid and cement, and generating a solidified body, It is related with the processing apparatus of radioactive waste liquid characterized by comprising.

  According to the present invention, radioactive liquid waste containing radioactive waste can be solidified and processed by a simple method without generating hydrogen-based gas such as tritium gas.

It is a figure which shows schematic structure of the processing apparatus of the radioactive waste liquid in embodiment.

FIG. 1 is a diagram illustrating a schematic configuration of a radioactive liquid waste treatment apparatus according to the present embodiment.
A radioactive waste liquid treatment apparatus 10 shown in FIG. 1 is connected to a reaction container 11 containing a radioactive waste liquid S, and the reaction container 11 via a pipe 21 to solidify the radioactive waste liquid S treated in the reaction container 11 into cement. And a cement solidifying tank 12. The reaction vessel 11 and the cement solidification tank 12 are configured such that a pH adjusting agent, specifically, an oxidizing agent can be appropriately added according to the radioactive waste liquid treatment method of the present embodiment described below.

  A heater 111 for heating the radioactive waste liquid S in the reaction vessel 11 to an appropriate temperature is provided on the outer periphery of the reaction vessel 11, and a stirrer 112 is installed in the lower portion of the reaction vessel 11. .

  The reaction vessel 11, the cement solidification tank 12, and the pipe 21 constituting the processing apparatus 10 are made of a material having high corrosion resistance such as stainless steel. In addition, in the processing apparatus 10 of this embodiment, although the cement solidification tank 12 is comprised by the indrum system, it can also be set as the outdrum system which isolate | separated the kneading tank and the solidification tank.

  Next, the radioactive waste liquid processing method using the radioactive waste liquid processing apparatus 10 shown in FIG. 1 will be described.

First, a radioactive waste liquid S containing a radionuclide such as Co-60 stored in a buffer tank or the like at a high concentration of 1 × 10 ¹² Bq / ton or more in a reaction vessel 11 in the processing apparatus 10 of FIG. Introduce into the reaction vessel 11. Next, an oxidation-reduction potentiometer (not shown) (as a measurement electrode, a composite electrode in which a platinum electrode and a reference electrode are integrated is widely used, and in this case, it is different from the standard hydrogen electrode described above. The potential difference between the electrode and the standard hydrogen electrode is added to the measured potential difference), and the terminal is immersed in the reaction vessel 11 to measure the redox potential of the radioactive liquid waste S. Note that the oxidation-reduction potentiometer may be fixed to the wall surface of the reaction vessel 11 in advance.

  At this time, the oxidation-reduction potential of the radioactive liquid waste S is generally less than 0 mV, specifically in the range of minus several tens of mV to minus several hundred mV. Therefore, in this embodiment, the oxidation-reduction potential of the radioactive liquid waste S in the reaction vessel 11 is set to 0 mV or more. As a result, when the radioactive liquid waste S as described below is directly cemented, the high-concentration radionuclides in the radioactive liquid S do not cause radiolysis of pore water or crystal water present in the solidified cement. As a result, generation of hydrogen-based gas such as radioactive tritium gas can be suppressed.

  Such a relationship between the oxidation-reduction potential of the radioactive liquid waste S and the generation of hydrogen-based gas when producing the cement solidified body of the radioactive liquid waste S was obtained as a result of enormous and long-term research by the present inventors. Is.

  In order to set the oxidation-reduction potential of the radioactive liquid waste S to 0 mV or higher, for example, an oxidizing agent (pH adjusting agent) is used. As the oxidizing agent, strong acids such as sulfuric acid, nitric acid and hydrochloric acid can be used, and relatively weak acids such as molybdic acid, tungstic acid and perrhenic acid can also be used. Further, a salt that is easily soluble in the radioactive waste liquid S, such as sodium nitrate salt, and that generates nitric acid that is a strong acid by dissolving in the radioactive waste liquid S can also be used.

  However, when the radioactive waste liquid S contains a high concentration of oxo acid and its oxidation-reduction potential is 0 mV or more from the beginning, the operation for setting the oxidation-reduction potential of the radioactive waste liquid S to 0 mV or more is necessary. Not. Conversely, for the purpose of preventing corrosion of the reaction vessel 11 or the like, an alkali salt that is easily soluble in the radioactive waste liquid S such as sodium sulfide salt may be added.

  The mixing of the radioactive liquid waste S and the oxidizing agent or the like is performed using a stirrer 112 provided at the lower part of the reaction vessel 11.

  Moreover, in this embodiment, since the heater 111 is arrange | positioned in the outer peripheral part of the reaction container 11, the temperature of radioactive waste liquid S is changed by the heater 111, specifically, room temperature (25 degreeC to less than 100 degreeC). The oxidation-reduction potential of the radioactive waste liquid S can be changed (increased) by heating to a temperature of However, the oxidation-reduction potential of the radioactive liquid waste S primarily depends on the oxidant as described above.

  As described above, the upper limit of the oxidation-reduction potential of the radioactive liquid waste S is that the radioactive tritium is radioactively decomposed without radiolytic decomposition of pore water or crystal water, etc. present in the cement solidified body. Although it will not specifically limit if generation | occurrence | production of hydrogen gas, such as gas, can be suppressed, For example, it can be set as 0.5 mV-1.0 mV. Even if the oxidation-reduction potential of the radioactive liquid waste S becomes higher than this, it does not affect the suppression of the generation of the hydrogen-based gas as described above, and the reaction vessel 11 may be corroded.

  Next, the radioactive liquid waste S whose oxidation-reduction potential is set to 0 mV or more is transferred to the cement solidification tank 12 through the pipe 21 and mixed with cement in the cement solidification tank 12 to produce a cement solidified body.

  Examples of the cement used include alumina cement, blast furnace slag cement, fly ash cement, and Portland cement. Since these cement materials are easily available, are inexpensive, and are stable to seawater and chemical substances, they are suitable as cement materials for solidifying and stabilizing radioactive drainage as in this embodiment. Yes. In particular, alumina cement is made of bauxite and limestone, which are raw materials of aluminum, and is a cement containing aluminum oxide, and exhibits high strength immediately after kneading, and therefore has excellent ion confinement properties.

  On the other hand, the redox potential of the water derived from the radioactive liquid waste S and the like contained in the cement in the process of solidifying the cement is such that the high-concentration radionuclide does not radiolyze the water in the cement solidified body as described above. It is necessary to keep the voltage at 0 mV or higher.

  When producing a cement solidified body using Portland cement among the above-mentioned cements, if the redox potential of the radioactive liquid waste S is set to 0 mV or more in the reaction vessel 11, the redox potential of water contained in the cement solidified body is 0 mV. Although the above requirements are satisfied, when a cement solidified body is produced using other cements, the redox potential of the water becomes less than 0 mV. Therefore, it is preferable that the cement used when producing the cement solidified body is Portland cement.

In particular, when alumina cement is used, ettringite (3CaO • Al ₂ O ₃ • 3CaSO ₄ • 32H ₂ O), monosulfate (3CaO • Al ₂ O ₃ • CaSO ₄ ), which is a cement hydrate containing a large amount of crystal water, is used.・ 12H ₂ O) and calcium aluminate hydrate (3CaO ・ Al ₂ O ₃・ 13H ₂ O) are produced, so the generation of hydrogen-based gas becomes remarkable. For this reason, when a cement solidified body is obtained using a cement other than Portland cement, an additional operation is required to set the oxidation-reduction potential of the cement in the solidifying process to 0 mV or higher.

  In other words, it can be said that the method of the present embodiment is most effective when a cement solidified body is produced using alumina cement as the cement.

  As an additional operation, an operation similar to that performed on the radioactive waste liquid S in the reaction vessel 11, that is, an oxidizing agent (pH adjusting agent) is added. As the oxidizing agent, strong acids such as sulfuric acid, nitric acid and hydrochloric acid can be used, and relatively weak acids such as molybdic acid, tungstic acid and perrhenic acid can also be used. Further, a salt that is easily soluble in water, such as sodium nitrate, and that generates nitric acid that is a strong acid by dissolving in water in the cement solidified body can also be used.

  The cement solidified body obtained as described above is subjected to burying disposal at a burying disposal facility.

  As mentioned above, although several embodiment of this invention was described, these embodiment was posted as an example and is not intending limiting the range of 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 10 Treatment apparatus of radioactive liquid waste 11 Reaction container 111 Heater 112 Stirrer 12 Cement solidification tank

Claims (8)

Setting the oxidation-reduction potential of the radioactive liquid waste to 0 mV or higher;
Solidifying the radioactive liquid waste with cement;
A method for treating radioactive liquid waste, comprising:   The method for treating radioactive liquid waste according to claim 1, wherein the cement solidification is performed by setting an oxidation-reduction potential of water contained in the cement in the solidification process to 0 mV or more.   The method for treating a radioactive liquid waste according to claim 1, wherein the cement solidification is performed using Portland cement.   The method for treating a radioactive liquid waste according to claim 2, wherein the cement solidification is performed using alumina cement. A reaction vessel for setting the redox potential of the radioactive liquid waste to 0 mV or higher;
A cement solidification tank for mixing the radioactive liquid waste and cement to produce a solidified body;
An apparatus for treating radioactive liquid waste, comprising:   The apparatus for treating a radioactive liquid waste according to claim 4, wherein the oxidation-reduction potential in the cement solidification tank is maintained at 0 mV or more.   The radioactive processing apparatus according to claim 5, wherein the cement is Portland cement.   The apparatus for treating a radioactive liquid waste according to claim 6, wherein the cement is alumina cement.

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