A method for extracting and recovering americium from radioactive waste resin containing americium-241

Abstract

The present invention relates to a method for extracting and recovering americium from radioactive waste resin containing americium-241, and discloses for the first time a method for extracting and recovering americium from radioactive waste resin containing americium-241. The radioactive waste resin is treated by direct leaching with TTA-xylene. After a single extraction, greater than 95% of the americium in the waste resin is extracted into the organic phase, and the total stripping rate of americium in the three-stage cross-current stripping is greater than 97%. Compared with the prior art, the present invention aims at the comprehensive utilization of americium in radioactive solid waste and proposes a method for extracting and recovering americium from radioactive waste resin containing americium-241. This method breaks the constraints of liquid-liquid extraction, simplifies the process for extracting and recovering americium from waste resin without requiring special pretreatment of the waste resin, and obtains a high-purity americium dioxide product, thereby reducing the difficulty of treating the waste resin.

Description

Technical Field

[0001] The invention belongs to the field of separation and extraction of radioactive nuclides, and in particular relates to a method for extracting and recovering americium from radioactive waste resin containing americium-241. Background Art

[0002] Nuclear energy has brought huge economic and social benefits to mankind, but with the widespread use of nuclear energy, a large amount of radioactive waste liquid will also be generated. The main treatment methods include precipitation, ion exchange, physical adsorption, extractant extraction, etc. Among them, the ion exchange method is widely used to treat radioactive waste liquid due to its low cost and good adsorption performance. However, when the adsorption resin is saturated, it must be replaced, which makes the amount of radioactive waste resin increase year by year. Radioactive waste resin belongs to a type of difficult-to-treat organic waste stream, especially those containing minor actinide radionuclides americium ( 214 Am), as plutonium ( 241 PuT 1 / 2 =14a) decay daughters, americium ( 214 Am) has low energy (59.6keV), high probability of gamma radiation and high specific activity (1.27×10 11 Bq·g -1 At present, most of the radioactive waste resins produced by the operation of nuclear facilities in China are still in temporary storage, which not only causes the surrounding environment to face greater safety risks, but also wastes 214 Am has great utilization value as a radionuclide.

[0003] 214 Am is prepared into α and γ sources, which are widely used in smoke alarm, surface thickness measurement and other fields; it is processed into neutron sources and used in oil detection and mining, reactor ignition, etc.; it is processed into heat sources and can be used in the manufacture of radioisotope batteries. With the increasing demand for it in recent years, 214 Am has once again received widespread attention. Summary of the Invention

[0004] The purpose of the present invention is to overcome the current situation in the prior art where radioactive waste resins generated by nuclear facility operation are in a temporary storage state and cannot realize their economic value, and to provide a method for extracting and recovering americium from radioactive waste resins containing americium-241.

[0005] The purpose of the present invention can be achieved by the following technical solutions:

[0006] The present invention provides a method for extracting and recovering americium from radioactive waste resin containing americium-241, comprising the following steps:

[0007] S1: Extraction: Weigh thenoyltrifluoroacetone (TTA) in xylene, stir and dissolve, and acid-wash to obtain a TTA-xylene solution; add the TTA-xylene solution to a reaction vessel containing radioactive waste resin containing americium-241 for leaching to produce a rosy red TTA-rich organic phase containing americium-241;

[0008] S2: Stripping: HNO3 is stripped with the TTA-rich organic phase containing americium-241, and concentrated by one evaporation to obtain a concentrated americium-containing stripping solution;

[0009] S3: Purification:

[0010] Adsorption method: a chelating resin is loaded into a resin column, and the americium-containing stripping solution described in step S2 flows through the chelating resin column to obtain a purified americium-containing solution;

[0011] or,

[0012] Extraction method: preparing a TBP dilute solution and mixing it with the americium-containing stripping solution in step S2 to perform extraction to obtain a purified americium-containing solution;

[0013] S4: Pre-transformation treatment: The purified americium-containing solution is directly evaporated and concentrated for a second time, and the nitric acid molecules are transferred from the americium nitrate solution to the gas phase. The solution volume is concentrated, and the concentrated solution is diluted and the acidity is adjusted to obtain a high-purity americium-containing solution that can be directly transformed.

[0014] Furthermore, the feed ratio of TTA-xylene in step S1 is (0.5-0.6) mol: 1 L, and the acid washing solution is HNO3.

[0015] Furthermore, the solid-liquid ratio of the radioactive waste resin containing americium-241 to the TTA-xylene solution in step S1 is 1:5; and the extraction time is 10-15 minutes.

[0016] Furthermore, the reaction container in step S1 is a device carrying an ultrasonic reactor.

[0017] Furthermore, the concentration of HNO3 in step S2 is 0.1-0.3 mol / L; and the volume ratio of HNO3 to the TTA organic-rich phase containing americium-241 is 1:1.

[0018] Furthermore, the stripping time in step S2 is 10-15 minutes, the stripping times are 3-5 times, and the stripping temperature is 60-70°C.

[0019] Furthermore, in step S3, the acidity of the americium-containing stripping solution is adjusted to 5-7M by an adsorption method.

[0020] Furthermore, the chelate resin described in the adsorption method in step S3 is a chelate resin based on a styrene-divinylbenzene cross-linked polymer and containing a phosphoric acid functional group; the adsorption method adopts a dynamic adsorption mode, the resin column has a liquid flow-through capacity of ≤1.5 Bv / h, that is, the residence time of the resin column is not less than 40 minutes, and the resin column size satisfies a height-to-diameter ratio (i.e., the ratio of the resin column height to the internal diameter) of ≥5.0.

[0021] Furthermore, the specific preparation process of the TBP dilution solution described in the extraction method of step S3 is: TBP is mixed with a diluent, washed with HNO3 acid, and washed with Na2CO3 alkaline; the diluent is hydrogenated kerosene; TBP and kerosene are mixed at a concentration ratio of 3:7; the volume ratio of the TBP-kerosene to the americium-containing stripping solution in step S2 is 1:1.

[0022] Furthermore, the extraction time of the extraction method in step S3 is 10-15 minutes, and the number of extractions is 1-3 times.

[0023] Furthermore, in step S3, the acidity of the americium-containing stripping solution is adjusted to 5-7M using an extraction method.

[0024] Furthermore, deionized water (pH ≈ 6.60, κ ≤ 1.5 × 10 -4 ·S·m -1 ) Dilute the concentrated solution to adjust the acidity.

[0025] Furthermore, the acidity of the dilution in step S4 is 0.1M-0.5M, and the americium concentration is ensured to be no less than 1g·L -1 .

[0026] Compared with the prior art, the present invention has the following advantages:

[0027] (1) The present invention provides a method for extracting and recovering americium from radioactive waste resin containing americium-241. The method utilizes a one-step extraction process to extract the adsorbed nuclides from the radioactive waste resin, followed by a one-step stripping and a one-step purification process (1+1) to purify and recover the americium-241, thereby obtaining a high-purity americium dioxide product.

[0028] (2) The present invention aims at the comprehensive utilization of americium in radioactive solid waste and proposes a method for extracting and recovering americium from radioactive waste resin containing americium-241. This method breaks the constraints of liquid-liquid extraction and simplifies the process of extracting and recovering americium from waste resin without the need for special pretreatment of the waste resin, thereby obtaining a higher-purity americium dioxide product and reducing the difficulty of treating the waste resin. BRIEF DESCRIPTION OF THE DRAWINGS

[0029] Figure 1 The figure is a flow chart for extracting and recovering americium from radioactive waste resin containing americium-241 according to an embodiment of the present invention. DETAILED DESCRIPTION

[0030] The following description sets forth numerous specific details to facilitate a thorough understanding of the present invention. However, the present invention can be implemented in many other ways than those described herein, and those skilled in the art may make similar modifications without departing from the scope of the present invention. Therefore, the present invention is not limited to the specific implementations disclosed below.

[0031] Unless otherwise defined, all technical and scientific terms used herein have the same meaning as those commonly understood by those skilled in the art of the present invention. The terms used in this specification of the present invention are only for the purpose of describing specific embodiments and are not intended to limit the present invention.

[0032] Example 1

[0033] (1) Extraction

[0034] Preparation of the leaching solution: Accurately weigh 2.75 g of thenoyltrifluoroacetone (TTA) into a 100 mL beaker, dissolve it in xylene, and transfer it to a 25 mL volumetric flask to obtain 25 mL of 0.5 M TTA-xylene solution;

[0035] Direct leaching: Accurately weigh 5 mL of radioactive waste resin containing americium-241 into a centrifuge tube, measure its content using a gamma spectrometer and record it, then pour 25 mL of the prepared extract TTA-xylene solution directly into the centrifuge tube containing the radioactive waste resin for solid-liquid leaching at a solid-liquid ratio of 1:5. After oscillating for 10-15 minutes, a rose-red TTA-rich organic phase containing americium-241 is obtained. The metal content is determined using graphite crystal pre-diffraction X-ray fluorescence, and the americium-241 extraction rate is quantitatively measured using a gamma spectrometer. Am =96.28%, E Zr =99.36%(E Am is the extraction rate of americium-241 in the extraction stage, E Zr is the extraction rate of zirconium-95 in the extraction stage).

[0036] (2) Stripping

[0037] Prepare 75mL of 0.3mol / L HNO3, heat it to 60℃ as the stripping solution, and perform three-stage cross-current stripping with 25mL of TTA rich organic phase containing americium, plutonium, zirconium and uranium at an O / A ratio of 1:1. The shaking time is 10-15min. Collect 75mL of stripping solution and perform one evaporation concentration with a concentration factor of 16 to obtain 4.5mL of 5M nitric acid concentrate containing americium, i.e., the americium stripping solution. The americium metal content in it is determined by graphite crystal pre-diffraction X-ray fluorescence method, and E′ is obtained by detection. Am =97.6%(E′ Amis the stripping rate of americium-241 in the stripping stage).

[0038] (3) Purification

[0039] Preparation of the extract: Evenly mix 4.05 mL of tributyl phosphate (TBP) and 9.45 mL of hydrogenated kerosene, place in a 25 mL separatory funnel, acid wash once with 1 mol HNO3 at a volume ratio of 1:1, alkaline wash once with 5% sodium carbonate at a volume ratio of 1:1, and finally wash with deionized water at a volume ratio of 1:1 until neutral and set aside. Extraction of uranium and plutonium with TBP: Divide the prepared 13.5 mL of 30% TBP-kerosene into three parts and use them in the stripping process of step (2) to remove uranium and plutonium with an O / A ratio of 1:1. Each shaking time is 10-15 minutes. The metal content is determined by X-ray fluorescence graphite pre-diffraction, and E" is obtained by detection. U =98.96%, E″ Pu =98.38%(E″) U is the extraction rate of uranium-235 in the purification stage, E″ Pu is the extraction rate of plutonium-239 in the purification stage).

[0040] Example 2

[0041] Compared with Example 1, the difference is that the concentration of the TTA-xylene solution configured in the extraction section in (1) is 0.6 mol. The single leaching rate of zirconium E is 0.01%. Zr =99.32%, single leaching rate of americium E Am =95.17%, which is comparable to the extraction effect of 0.5M TTA-xylene solution.

[0042] In the stripping section, E' is stripped according to the same steps. Am =97.12%.

[0043] Example 3

[0044] Compared with Example 1, the difference is that in the stripping section (2), the HNO3 stripping solution is heated to 70°C and stripping is carried out according to the same operating steps. The single-stage stripping rate of americium is E′ at 60°C. Am =56.71% increased to E' Am =61.90%. This indicates that increasing the stripping temperature is beneficial to the stripping of trivalent americium in TTA-xylene solution.

[0045] Example 4

[0046] The remaining steps are the same as those in Example 1. This purification stage uses an adsorption method to remove uranium and trace plutonium, and this is used to amplify the entire process experiment in order to better illustrate the technical advantages of the present invention.

[0047] Take 500ml of adsorption saturated waste resin and record the type and content of adsorbed nuclides by the difference of metal ions between the adsorption original solution and the adsorption tail solution. The results are shown in Table 1 below:

[0048] Table 1: Amount of radionuclides adsorbed by 500 mL of resin

[0049]

[0050] 1 g of the mixed waste resin was taken and its Am content was measured by gamma spectrometer. The results are shown in the following table: The calculated error of the adsorption content was ≤2%, which met the authenticity of this example.

[0051] Table 2: Am content in waste resin measured by gamma spectrometer

[0052]

[0053] The specific operations are as follows:

[0054] (1) Extraction

[0055] Accurately weigh 137.5g of TTA into a 2.5L beaker and add 2.5L of xylene to obtain 2.5L of 0.5M TTA-xylene solution. Unlike Example 1, the reactor uses an ultrasonic device for solid-liquid leaching, which shortens the leaching equilibrium time and reduces the operator's exposure to the gamma irradiation dose of Am-241. The leaching time is 10 minutes, and the metal content is determined by X-ray fluorescence graphite pre-diffraction. The leaching rate of americium-241 is quantitatively measured using a gamma spectrometer. Am =98.01%, E Zr =99.38%.

[0056] (2) Stripping

[0057] Prepare 7.5L of 0.3mol / L HNO3 and heat it to 60℃ as the stripping solution. Compared with the TTA loaded organic phase containing americium, plutonium, zirconium, and uranium, the O / A ratio is 1:1. After shaking for 10-15 minutes, 7.5L of 0.3M americium-containing stripping solution is obtained. The stripping solution is then concentrated 20 times to obtain 375mL of 6M americium-containing concentrated solution. The metal content in it is determined by X-ray fluorescence graphite pre-diffraction. The E′ is obtained by detection. Am =98.46%. The extract was subjected to ICP-AES impurity detection, and the results were as follows:

[0058] Table 3: Impurity content of stripping solution containing Am

[0059]

[0060] (3) Purification

[0061] In this example, a test column with an effective column volume of 46.1 mL was selected using the adsorption method. The residence time was similar to that of an engineering column with stable operation experience. The flow rate was set to 70 mL / min. The specific parameters are as follows:

[0062] Table 4: Technical parameters of the experimental adsorption column

[0063]

[0064] The test column was acidified with 6M nitric acid and swollen with deionized water. 375mL of 6M americium concentrate was passed through the resin column at a rate of 70mL / min. The adsorption tail liquid was collected and the metal content was determined by X-ray fluorescence graphite pre-diffraction. The E" was obtained by detection. U =99.96%, E″ Pu =98.38%. Finally, 375 mL of a high-purity americium solution containing 1.01 g of metal was obtained.

[0065] (4) Pre-transformation processing

[0066] A 6M 375mL high-purity americium solution with a metal content of 1.01g was evaporated and concentrated twice with a concentration factor of 20 to obtain a 14M 18.75mL americium-containing concentrate, which was then neutralized by adding deionized water without introducing impurities. Finally, a 0.5M 525mL precipitate stock solution was obtained for subsequent transformation treatment.

[0067] (5) Precipitation

[0068] With an excess of 0.1M oxalic acid, 80mL of the prepared 0.9M H2C2O4 precipitant was added. The mixture was stirred and allowed to settle at room temperature (25°C) for 4 hours. The mixture was then fed to a Buchner filtration apparatus at a rate of 100mL / min and a vacuum of 0.05MPa. After 6 minutes, the americium oxalate precipitate was filtered through a metal filter cloth, and the americium oxalate filter cake was washed with 100mL of deionized water. After washing, the americium oxalate filter cake was scraped into a quartz crucible and weighed to yield 3.68g of hydrated americium oxalate (approximately 10.65 units of bound water).

[0069] At this time, the mother liquor (precipitant + washing liquid) filtered out was 705 mL in total. The Am concentration was determined by X-ray fluorescence graphite pre-diffraction, which decreased from 1.66 g / L before precipitation to 18.0 mg / L in the end. The recovery rate of americium during the precipitation process was 98.74%.

[0070] (6) Calcination

[0071] A quartz crucible containing 3.71g of hydrated americium oxalate was placed in a calcining furnace. Three heating stages were set: drying at 120°C, nitrogen removal at 300°C, and thermal decomposition at 600°C. The holding times were 1.5h, 3.5h, and 4h, respectively, with a heating rate of 3°C / min. This thermal decomposition time was sufficient to completely decompose the excess oxalic acid and remove carbon impurities. Upon completion of the calcination, the pale yellow americium oxalate sample transformed into a black powder. A total of 1.13g of dioxide powder was collected, indicating that the recovered material was generally consistent with expectations.

[0072] The samples were taken and tested for impurities by spectroscopy. The total amount of impurity elements with γ radioactivity was ≤2%. After XRD and calorimetric data analysis, the purity of americium dioxide was greater than 98%.

[0073] The above examples are described in a relatively specific and detailed manner, but they cannot be understood as limiting the scope of the invention patent. For technicians in the same field, without departing from the concept of the present invention, they can also make process adjustments, deformations and improvements, which all fall within the scope of protection of the present invention.

Claims

1. A method for extracting and recovering americium from radioactive waste resin containing americium-241, characterized in that: The following steps are involved: S1: Extraction: Add the prepared TTA-xylene solution to a reaction vessel containing radioactive waste resin containing americium-241 for leaching to generate a TTA-rich organic phase containing americium-241; S2: Stripping: HNO3 is stripped with the TTA-rich organic phase containing americium-241, and concentrated by one evaporation to obtain a concentrated americium-containing stripping solution; S3: Purification: Adsorption method: a chelating resin is loaded into a resin column, and the americium-containing stripping solution described in step S2 flows through the chelating resin column to obtain a purified americium-containing solution; or, Extraction method: preparing a TBP dilute solution and mixing it with the americium-containing stripping solution concentrate in step S2 to perform extraction to obtain a purified americium-containing solution; S4: Pre-transformation treatment: The purified americium-containing solution is directly evaporated and concentrated for a second time, and the nitric acid molecules are transferred from the americium nitrate solution to the gas phase. The solution volume is concentrated, and the concentrated solution is diluted and the acidity is adjusted to obtain a high-purity americium-containing solution that can be directly transformed.

2. The method for extracting and recovering americium from radioactive waste resin containing americium-241 according to claim 1, characterized in that: The TTA-xylene solution preparation method described in step S1 is: weigh TTA, place it in xylene, stir to dissolve, and acid wash to obtain the TTA-xylene solution; the feed ratio of TTA to xylene is (0.5-0.6) mol: 1 L, and the acid wash solution is HNO3.

3. The method for extracting and recovering americium from radioactive waste resin containing americium-241 according to claim 1, characterized in that: The volume ratio of the radioactive waste resin containing americium-241 to the TTA-xylene solution in step S1 is 1:5; and the extraction time is 10-15 minutes.

4. The method for extracting and recovering americium from radioactive waste resin containing americium-241 according to claim 1, characterized in that: The reaction container described in step S1 is a device carrying an ultrasonic reactor.

5. The method for extracting and recovering americium from radioactive waste resin containing americium-241 according to claim 1, characterized in that: The concentration of HNO3 in step S2 is 0.1-0.3 mol / L; the volume ratio of HNO3 to the TTA organic-rich phase containing americium-241 is 1:

1.

6. The method for extracting and recovering americium from radioactive waste resin containing americium-241 according to claim 1, characterized in that: The stripping time in step S2 is 10-15 min, the stripping times are 3-5 times, and the stripping temperature is 60-70°C.

7. The method for extracting and recovering americium from radioactive waste resin containing americium-241 according to claim 1, characterized in that: The chelating resin described in the adsorption method in step S3 is a chelating resin based on a styrene-divinylbenzene cross-linked polymer and containing a phosphoric acid functional group; the adsorption method selects a dynamic adsorption mode, and the resin column has a liquid processing capacity of ≤1.5Bv / h, that is, the residence time of the liquid flowing through the resin column is not less than 40 min.

8. The method for extracting and recovering americium from radioactive waste resin containing americium-241 according to claim 1, characterized in that: The specific preparation process of the TBP dilution solution described in the extraction method in step S3 is: TBP is mixed with a diluent, washed with HNO3 acid, and washed with Na2CO3 alkaline; the diluent is hydrogenated kerosene; TBP and kerosene are mixed in a volume ratio of 3:7; the volume ratio of the TBP dilution solution to the americium-containing stripping solution in step S2 is 1:

1.

9. The method for extracting and recovering americium from radioactive waste resin containing americium-241 according to claim 1, characterized in that: The extraction time of step S3 is 10-15 min, and the number of extractions is 1-3 times.

10. The method for extracting and recovering americium from radioactive waste resin containing americium-241 according to claim 1, characterized in that: The acidity of the dilution in step S4 is 0.1 M-0.5 M, and the americium concentration is ensured to be no less than 1 g·L -1 .

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