Treatment method of graphite waste containing radioactive elements

Abstract

The invention provides a treatment method of graphite waste containing radioactive elements. A treatment method of graphite waste containing radioactive elements comprises the following steps that (1) the graphite waste is added into fused salt, an oxidizing agent and an additive are added, the mixture fully reacts at the temperature of 500-1300 DEG C, the fused salt comprises alkali metal carbonate, and the additive is one or more of alkali metal nitrate, sulfate, copper oxide or lead oxide; (2) when the content of the precipitate containing radioactive elements in the reactant reaches 5-30 wt%, separating the precipitate; and (3) supplementing the molten salt to enable the reactants to continuously react until the content of the precipitate is less than 5wt%. The method has the advantages that the technological process is simple, convenient and flexible, environmental protection and safety are achieved, the generation amount of secondary waste is small, inclusiveness to radioactive elements is good, dispersion of the radioactive elements is reduced, and the content of the radioactive elements in tail gas is low.

Description

Technical Field

[0001] This application belongs to the technical field of radioactive waste treatment. Specifically, it relates to a method for treating graphite waste containing radioactive elements. Background Art

[0002] During the operation and decommissioning of nuclear facilities, graphite is widely used as a common neutron moderator and structural material. With the decommissioning of nuclear facilities, a large amount of radioactive graphite waste is generated. Graphite crucibles are also used in some operations in the post-treatment field, and these used graphite crucibles become radioactive waste after being contaminated by radioactive elements such as uranium and plutonium. Therefore, how to safely and effectively treat radioactive graphite waste has become an urgent problem to be solved in the nuclear industry and environmental protection fields.

[0003] These waste graphites are difficult to treat due to their stable structure and chemically indestructible properties, and the treatment process is complex. In recent years, extensive research has been conducted internationally on the treatment technologies, disposal strategies, and alternative solutions for radioactive graphite waste. The main technical means include fixed-bed incineration, fluidized-bed incineration, sealed-chamber incineration, laser incineration, steam pyrolysis technology, and solidification methods. Although the above technologies have made certain progress in treating radioactive graphite waste, these technologies have problems such as complex overall treatment processes, high technical difficulties, and easy secondary diffusion of radioactive elements. Therefore, developing a simple, low-treatment-difficulty, and efficient radioactive graphite treatment method that can recover elements such as uranium and plutonium is of great significance for promoting the sustainable development of the nuclear industry and environmental protection. Summary of the Invention

[0004] Aiming at the problems existing in this field, this application provides a method for treating graphite waste containing radioactive elements, which has the advantages of simple process, less secondary waste, radiation resistance, stability, reliability, economy, and high efficiency.

[0005] The technical solution of this application is as follows:

[0006] A method for treating graphite waste containing radioactive elements, comprising the following steps:

[0007] (1) Add the graphite waste into molten salt, add an oxidant and an additive, and react fully at a temperature of 500 - 1300 °C. The molten salt includes alkali metal carbonates, and the additive is one or more of alkali metal nitrates, sulfates, copper oxide, or lead oxide;

[0008] (2) Wait until the content of the precipitate containing radioactive elements in the reactants reaches 5 - 30 wt%, and separate the precipitate;

[0009] (3) Supplement the molten salt to make the reactants continue to react until the content of the precipitate is less than 5 wt%.

[0010] In the above method, the graphite waste can be graphite from the decommissioning of graphite reactors, graphite crucibles used in the nuclear industry, graphite fuel elements in high-temperature gas-cooled reactors, or graphite materials containing radioactive elements such as uranium and plutonium or having certain radioactivity itself.

[0011] In the above method, the radioactive elements include actinides, minor actinides, and major fission elements and their isotopes.

[0012] Preferably, the crushed graphite waste is added to the molten salt, and the size of the graphite waste is less than 200 mm.

[0013] Preferably, the molten salt further includes alkali metal hydroxides.

[0014] Preferably, the molten salt is in a molten state.

[0015] Preferably, the reaction temperature is 700 - 1000 °C.

[0016] Preferably, the alkali metal carbonate is one or more of Na2CO3, K2CO3, or Li2CO3, and the alkali metal hydroxide is NaOH and / or KOH. 。

[0017] Preferably, the content of the alkali metal carbonate in the molten salt during the treatment process is not less than 50 wt%.

[0018] Preferably, when the additive is alkali metal nitrate and / or sulfate, the dosage of the additive is 0.5 - 20 wt% of the dosage of the molten salt; when the additive is copper oxide or lead oxide, the dosage of the additive is 5 - 40 wt% of the dosage of the molten salt.

[0019] Preferably, the treatment method further includes purifying the molten salt after the reaction by adding phosphate or recrystallization. The purification process of adding the phosphate includes: adding the phosphate as a precipitant to the molten salt in a molten state to purify the radioactive elements in the molten salt; the purification process of recrystallization includes: separating the upper molten salt after the precipitation in step (2), completely dissolving it with water, then performing evaporation and concentration. The crystallized alkali metal carbonate precipitated can be reused, and the remaining solution is subjected to solidification treatment.

[0020] This application has the following beneficial effects:

[0021] First, compared with the traditional method of graphite incineration, this application has the advantages of simple and flexible process, environmental protection and safety, less secondary waste generation, good inclusiveness for radioactive elements, reduction of radioactive element dispersion, and low content of radioactive elements in the tail gas, achieving effective volume reduction of solid waste. Through this application, most of the pollutants that may be contained in the graphite material can be retained in the molten salt through chemical reactions. In an alkaline molten salt environment, some volatile / semi-volatile fission products such as iodine, cesium, and technetium will dissolve in the molten salt, and only inert gases will be entrained in the treated tail gas, with a low overall content of radioactive elements in the tail gas. At the same time, actinides such as uranium and plutonium can be preliminarily separated from the fission products dissolved in the molten salt in the form of precipitation through chemical reactions, facilitating subsequent treatment.

[0022] Second, compared with the traditional method of graphite incineration, this application has the characteristics of strong adaptability and high economy. By optimizing the molten salt composition and operating conditions, this application can achieve the oxidation treatment of various radioactive graphite materials such as graphite waste from graphite reactor decommissioning, graphite crucibles used in the nuclear industry, and graphite fuel elements in the fourth-generation reactor while maintaining a high graphite oxidation efficiency.

[0023] Third, the reagents used in this application are mainly alkali metal hydroxides and / or alkali metal carbonates, which can not only efficiently oxidize radioactive graphite materials, but also remain stable under high-temperature and high-radiation conditions, effectively reducing secondary pollution. Moreover, the process flow is short, the cost is low, it is suitable for large-scale application, and has good industrial application prospects. BRIEF DESCRIPTION OF THE DRAWINGS

[0024] Figure 1 It is the XRD detection pattern of the orange precipitate in Example 2.

[0025] Figure 2 It is the SEM detection pattern of the orange precipitate in Example 2. DETAILED DESCRIPTION OF THE EMBODIMENTS

[0026] Next, the technical solutions of this application will be clearly and completely described in conjunction with the embodiments of this application. Obviously, the described embodiments are only a part of the embodiments of this application, rather than all the embodiments. Those not specified in the embodiments are carried out according to conventional conditions or conditions recommended by the manufacturer.

[0027] In this application, radioactive elements include actinides such as uranium and plutonium, minor actinides, and major fission elements such as zirconium, cerium, cesium, and strontium. The treatment method of this application can not only remove surface radioactive elements, but also achieve sufficient oxidation and volume reduction of graphite waste containing radioactive elements while highly "containing" radioactive elements and preventing their diffusion and spillage. During the treatment process, through the action of additives such as alkali metal nitrates, sulfates, copper oxide, and lead oxide, the rapid oxidation of graphite can be promoted, converting it into CO2. At the same time, elements such as uranium, plutonium, strontium, and cesium can be trapped by the alkaline molten salt through chemical reactions or physical effects during this process.

[0028] A treatment method for graphite waste containing radioactive elements, comprising the following steps:

[0029] (1) Add the crushed graphite waste with a size less than 200 mm to the molten salt, and add an oxidant and an additive. React fully at a temperature of 500 - 1300 °C, preferably at 700 - 1000 °C. The molten salt includes alkali metal carbonates and alkali metal hydroxides. During the treatment process, the content of alkali metal carbonates in the molten salt is not less than 50 wt%. The alkali metal hydroxide is NaOH and / or KOH, and the alkali metal carbonate is one or more of Na2CO3, K2CO3, or Li2CO3. The oxidant can be oxygen or air, or sodium peroxide, potassium superoxide, or a mixture of the above substances, etc. The additive is one or more of alkali metal nitrates, sulfates, copper oxide, and lead oxide. When the additive is alkali metal nitrate and / or sulfate, the dosage of the additive is 0.5 - 20 wt% of the dosage of the molten salt; when the additive is copper oxide or lead oxide, the dosage of the additive is 5 - 40 wt% of the dosage of the molten salt.

[0030] (2) The radioactive element compounds contained in or contaminated by the radioactive graphite produce precipitates after being oxidized by the molten salt. When the content of the precipitates containing radioactive elements in the reactants reaches 5 - 30 wt%, separate the precipitates.

[0031] (3) Supplement the molten salt to make the reactants continue to react until the content of the precipitates is less than 5 wt%.

[0032] Purify the molten salt after the reaction by adding phosphate or recrystallization. The purification process of adding phosphate includes: adding phosphate as a precipitating agent to the molten salt in a molten state to purify the radioactive elements in the molten salt. The precipitates generated in this process can be directly subjected to vitrification treatment. The purification process of recrystallization includes: completely dissolve the upper-layer molten salt after separating the precipitates in step (2) with water, then perform evaporation and concentration. The crystallized alkali metal carbonate precipitated can be reused, and the remaining solution is subjected to solidification treatment.

[0033] Example 1

[0034] A method for treating plutonium-containing graphite waste, comprising the following steps:

[0035] (1) Take 3 g of Li2CO3, 5 g of Na2CO3 and 4 g of K2CO3 and mix them in a 30 mL corundum crucible. Place the 12 g of binary carbonate mixture in the corundum crucible and heat it to the molten state using an electric resistance furnace at 650 °C. Then add 0.3 g of NaNO3 and 0.5 g of Na2SO4. Place the corundum crucible in the electric resistance furnace and heat it to 600 °C to form a mixed suspension molten salt. Subsequently, add a mixture of 0.05 g of PuO2 and 0.2 g of graphite with a particle size of about 1 mm. Pass air through the corundum tube at a flow rate of 5 ml / min and raise the temperature to 800 °C, and fully react for 6 h.

[0036] (2) The content of the precipitate in the reactant is about 0.4 wt%.

[0037] (3) After taking out the sample, obvious stratification can be observed, and the graphite has completely reacted. Subsequently, the upper-layer sample is taken and dissolved in dilute nitric acid to test the plutonium concentration therein. It is calculated that the plutonium content in the upper-layer molten salt is 210 μg / g. It shows that plutonium dioxide can be separated and treated in the carbonate molten salt.

[0038] Example 2

[0039] A method for treating uranium-containing graphite waste, comprising the following steps:

[0040] (1) Take 3 g of Li2CO3, 5 g of Na2CO3 and 4 g of K2CO3 and mix them in a 30 mL corundum crucible. Place the 12 g of binary carbonate mixture in the corundum crucible and heat it to the molten state using an electric resistance furnace. Then add 0.4 g of NaNO3 and 0.4 g of Na2SO4. Place the corundum crucible in the electric resistance furnace and heat it to 700 °C to form a mixed suspension molten salt. Subsequently, add a mixture of 0.1 g of UO2 and 0.3 g of graphite with a particle size of about 2 mm. Pass air through the corundum tube at a flow rate of 5 ml / min and react fully for 6 h at 700 °C. The brown uranium dioxide solid powder in the molten salt is completely converted into an orange-yellow solid powder. A large number of fine bubbles are generated during the reaction of graphite, and finally it completely reacts and disappears.

[0041] (2) The content of the precipitate in the reactant is about 0.7 wt%.

[0042] (3) Take a sample of the upper-layer molten salt, dissolve it in dilute nitric acid, and test the uranium concentration therein by ICP-MS. It is calculated that the uranium content in the upper-layer molten salt is 72 μg / g. After taking out the sample, obvious stratification can be observed, and the graphite has completely reacted. Separate the lower-layer precipitate and characterize the mixture of orange-yellow precipitate by XRD and SEM, as Figure 1 andFigure 2 As shown, the orange precipitate is a mixture of diuranate and uranate. The complete oxidation of graphite and the oxidative separation treatment of uranium dioxide in carbonate molten salt are achieved.

[0043] Example 3

[0044] A method for treating uranium-containing graphite waste, comprising the following steps:

[0045] (1) Take 10 g of Na2CO3 and place it in a 30 mL corundum crucible. Heat it using an electric resistance furnace until it reaches the molten state. Then add 0.4 g of NaNO3 and 0.4 g of Na2SO4. Place the corundum crucible in the electric resistance furnace and heat it to 900 °C to form a mixed suspension molten salt. Then add a mixture of 0.2 g of UO2 and 0.3 g of graphite with a particle size of about 2 mm. Pass air through a corundum tube at a flow rate of 5 ml / min and react fully for 6 h at 900 °C. The brown uranium dioxide solid powder in the molten salt is completely converted into an orange solid powder. A large number of fine bubbles are generated during the reaction of graphite, and finally, the reaction is complete and the bubbles disappear.

[0046] (2) The content of the precipitate in the reactant is 1.8 wt%.

[0047] (3) Take a sample of the upper-layer molten salt, dissolve it with dilute nitric acid, and test the uranium concentration in it by ICP-MS. The calculated uranium content in the upper-layer molten salt is 143 μg / g. After taking out the sample, obvious stratification can be observed, and the graphite has completely reacted. The overall reaction phenomenon is the same as that in Example 2, and the complete oxidation of graphite and the oxidative separation treatment of uranium dioxide in carbonate molten salt are achieved.

[0048] Example 4

[0049] A method for treating graphite waste containing Cs, Sr, Mo, and Te, comprising the following steps:

[0050] (1) Take 2.79 g of Li2CO3, 5.21 g of K2CO3, and 2 g of NaOH and mix them in a 30 mL corundum crucible. Place the 10 g of the mixed molten salt in the corundum crucible and heat it using an electric resistance furnace to reach the molten state at 800 °C. Then add 0.4 g of KNO3 and 0.4 g of Na2SO4. Place the corundum crucible in the electric resistance furnace and heat it to 800 °C to form a mixed suspension molten salt. Subsequently, add a mixture of a simulated fission product powder containing 0.05 g of each of the oxides of Cs, Sr, Mo, and Te and 0.2 g of graphite with a particle size of about 2 mm. Use a peristaltic pump to blow air through a corundum tube, and the rotation speed of the peristaltic pump head is 20 rpm. React fully for 6 h at 800 °C.

[0051] (2) The graphite in the molten salt has completely reacted and the oxides have completely dissolved, and no precipitate is generated.

[0052] (3) Samples of the molten salt after the reaction were taken, dissolved in dilute nitric acid, and the element concentrations were measured by ICP-MS. The content of each element in the upper molten salt was calculated to be exactly the total amount of the added elements, realizing the dissolution of some simulated fission products and the oxidation treatment of the graphite material.

[0053] Example 5

[0054] A method for treating uranium-containing graphite waste, comprising the following steps:

[0055] (1) Take 2 kg of 50% Na2CO3 - 50% K2CO3 and place it in a 1.5 L crucible. Heat it in a reaction furnace to reach the molten state. Then add 50 g of NaNO3, 50 g of Na2SO4 as additives, and a mixture of 10 g of graphite particles, 5 g of silicon carbide particles and 10 g of UO2 powder. Heat the reaction furnace to 950 °C to form a mixed suspension molten salt. Pass oxygen into the molten salt through a corundum tube at a flow rate of 50 ml / min. Under the condition of 950 °C, react fully for 6 h. A large number of fine bubbles are generated by the graphite particles in the molten salt at the initial stage of the reaction, and the amount of bubbles gradually decreases and disappears as the reaction proceeds.

[0056] (2) During this process, the molten salt gradually becomes turbid and precipitates are formed. As the reaction proceeds, the color of the precipitate in the molten salt gradually turns orange-yellow.

[0057] (3) Samples of the upper molten salt were taken, dissolved in dilute nitric acid, and the uranium concentration was measured by ICP-MS. The uranium content in the upper molten salt was calculated to be 200 μg / g. After heat preservation and static settlement, obvious stratification can be observed. The lower precipitate is orange-yellow, and the graphite has disappeared. This shows that this example can realize the synchronous oxidation treatment of the mixture of graphite, silicon carbide and UO2 in the alkaline molten salt, and the precipitation separation of elements such as uranium in it.

[0058] Example 6

[0059] A method for treating uranium-containing graphite waste, comprising the following steps:

[0060] (1) Take 3 g of Li2CO3, 5 g of Na2CO3 and 4 g of K2CO3 and mix them in a 30 mL corundum crucible. Place the 12 g of binary carbonate mixture after mixing in the corundum crucible and heat it in a resistance furnace to reach the molten state. Then add 2 g of CuO powder. Place the corundum crucible in the resistance furnace and heat it to 800 °C to form a mixed suspension molten salt. Subsequently, add a mixture of 0.2 g of UO2 and 0.3 g of graphite with a particle size of about 2 mm. Pass air into the molten salt through a corundum tube at a flow rate of 10 ml / min. Under the condition of 800 °C, react fully for 6 h. A large number of fine bubbles are generated during the reaction of graphite, and finally it completely reacts and disappears.

[0061] (2) The content of uranium-containing precipitate in the molten salt is about 1.4 wt%.

[0062] (3) Samples of the upper molten salt were taken, dissolved in dilute nitric acid, and the uranium concentration was measured by ICP-MS. The content of uranium in the upper molten salt was calculated to be 140 μg / g. After heat preservation and static settlement, obvious stratification could be observed, and the graphite had completely reacted. The complete oxidation of graphite and the oxidation treatment of uranium dioxide in carbonate molten salt were achieved.

[0063] The above description of the disclosed embodiments enables those skilled in the art to implement or use the present application. Although the present application has been described in detail above with general descriptions and specific embodiments, on the basis of the present application, some modifications or improvements can be made, which are obvious to those skilled in the art.

Claims

1. A method for treating graphite waste containing radioactive elements, characterized in that, It includes the following steps: (1) Add graphite waste into molten salt, add an oxidant and an additive, and react fully at a temperature of 500 - 1300 °C. The molten salt includes alkali metal carbonate, and the additive is one or more of alkali metal nitrate, sulfate, copper oxide or lead oxide; (2) When the content of the precipitate containing radioactive elements in the reactants reaches 5 - 30 wt%, separate the precipitate; (3) Supplement the molten salt to make the reactants continue to react until the content of the precipitate is less than 5 wt%.

2. The treatment method of graphite waste containing radioactive elements according to claim 1, characterized in that, The radioactive elements include actinides, minor actinides, and major fission elements and their isotopes.

3. The treatment method of graphite waste containing radioactive elements according to claim 1, characterized in that, Add the crushed graphite waste into the molten salt, and the size of the graphite waste is less than 200 mm.

4. The treatment method of graphite waste containing radioactive elements according to claim 1, characterized in that, The molten salt further includes alkali metal hydroxide.

5. The treatment method of graphite waste containing radioactive elements according to claim 1, characterized in that, The molten salt is in a molten state.

6. The treatment method of graphite waste containing radioactive elements according to claim 1, characterized in that, The reaction temperature is 700 - 1000 °C.

7. The treatment method of graphite waste containing radioactive elements according to claim 1 or 4, characterized in that, The alkali metal carbonate is one or more of Na2CO3, K2CO3 or Li2CO3, and the alkali metal hydroxide is NaOH and / or KOH.

8. The method for treating graphite waste containing radioactive elements according to claim 1, characterized in that, The content of the alkali metal carbonate in the molten salt is not less than 50 wt%.

9. The method for treating graphite waste containing radioactive elements according to claim 1, characterized in that, When the additive is alkali metal nitrate and / or sulfate, the dosage of the additive is 0.5 - 20 wt% of the dosage of the molten salt; when the additive is copper oxide or lead oxide, the dosage of the additive is 5 - 40 wt% of the dosage of the molten salt.

10. The treatment method of graphite waste containing radioactive elements according to claim 1, characterized in that, It also includes purifying the reacted molten salt by adding phosphate or recrystallization. The purification process of adding the phosphate includes: adding the phosphate as a precipitant into the molten salt in a molten state to purify the radioactive elements in the molten salt; the purification process of recrystallization includes: for the upper molten salt after separating the precipitate in step (2), completely dissolve it with water, then perform evaporation and concentration. The crystallized alkali metal carbonate precipitated can be reused, and the remaining solution is subjected to solidification treatment.

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