Stepwise chemical separation methods and systems using resin packed columns

Abstract

A waste extraction system includes a column set located between and in fluid communication with an upstream section of a main waste path and a column effluent tank, the column set including uranium adsorption columns in fluid communication with supporting adsorption columns along the main waste path. The uranium adsorption column is upstream of the support adsorption column. The waste extraction system also includes a back extraction waste path, an anion exchange resin, and a cation exchange resin. A back extraction waste path extends from a uranium adsorption column to a column effluent tank bypassing a support adsorption column in which an anion exchange resin is contained, to a column effluent tank, an anion exchange resin is contained.

Description

[0001] STATEMENT REGARDING FEDERALLY SPONSORED RESEARCH AND DEVELOPMENT

[0002] This disclosure was developed with government support under Contract No. DE-NA0004010 awarded by the U.S. Department of Energy. The government has certain rights in this disclosure. Technical Field

[0003] The present disclosure generally relates to systems and methods for extracting waste radionuclides, such as those generated during the production of medical isotopes. Background Art

[0004] Current nuclear waste separation technologies include solvent-based extraction, often referred to as solvent extraction or liquid-liquid extraction. Solvent extraction is a separation technique that contacts an extractant containing an organic phase with an aqueous phase containing metal ions. After mixing, the metal ions are transferred from the aqueous phase to the organic phase. Despite industry reliance on solvent extraction, solvent extraction still has many disadvantages, including challenges related to phase separation, the formation of a heavy phase or "third phase", and the generation of large amounts of hazardous organic waste. From an environmental management perspective, the process's reliance on hazardous aromatic organic solvents poses particular challenges. Many commercial waste transporters have a low tolerance for the presence of benzene or other aromatic hydrocarbons in solidified waste.

[0005]

[0006] Therefore, there is a need for improved methods and systems for separation of nuclear waste, such as in the production of medical isotopes. Summary of the invention

[0006] According to a first aspect of the present disclosure, a waste extraction system comprises: a column group located between an upstream section of a main waste path and a column effluent tank and fluidly connected to the upstream section of the main waste path and the column effluent tank, the column group comprising a uranium adsorption column, the uranium adsorption column being fluidly connected to a support adsorption column along the main waste path, wherein the uranium adsorption column is upstream of the support adsorption column; a strip waste path, the strip waste path bypassing the support adsorption column and extending from the uranium adsorption column to the column effluent tank; an anion exchange resin, the anion exchange resin being contained in the uranium adsorption column; and a cation exchange resin, the cation exchange resin being contained in the support adsorption column.

[0007] The second aspect includes the waste extraction system described in the first aspect, wherein the support adsorption column is a first support adsorption column, and the column group also includes a second support adsorption column, which is located downstream of the first support adsorption column along the main waste path, and the second support adsorption column contains a cation exchange resin.

[0008] The third aspect includes the waste extraction system described in the first aspect or the second aspect, wherein the uranium adsorption column comprises: a waste flow inlet and an elution inlet, each of which is located at a first end of the uranium adsorption column, wherein the waste flow inlet is fluidically connected to an upstream section of the main waste path; and a waste flow outlet and an elution outlet, each of which is located at a second end of the uranium adsorption column, wherein the elution outlet is fluidically connected to the stripping waste path.

[0009] The fourth aspect includes the waste extraction system of the third aspect, further comprising: an elution path extending from the elution acid source to the elution inlet.

[0010] The fifth aspect includes the waste extraction system of the third aspect or the fourth aspect, wherein the first end of the uranium adsorption column is opposite to the second end of the uranium adsorption column, and the first end of the uranium adsorption column is above the second end of the uranium adsorption column.

[0011] A sixth aspect includes the waste extraction system described in any of the preceding aspects, wherein the support adsorption column includes a waste flow inlet and a waste flow outlet, the waste flow inlet is located at the first end of the support adsorption column, and the waste flow outlet is located at the second end of the uranium adsorption column.

[0012] The seventh aspect includes the waste extraction system of the sixth aspect, wherein the first end of the uranium adsorption column is opposite to the second end of the uranium adsorption column, and the first end of the uranium adsorption column is above the second end of the uranium adsorption column.

[0013] The eighth aspect includes the waste extraction system described in any of the preceding aspects, wherein: the cation exchange resin includes cation exchange resin beads; the anion exchange resin includes anion exchange resin beads; the cation exchange resin beads and the anion exchange resin beads are each polymer-based; and the average diameter of the cation exchange resin beads and the anion exchange resin beads is 400 μm to 800 μm.

[0014] According to a ninth aspect of the present disclosure, a method for extracting radioactive nuclide waste includes: directing a waste stream from an upstream section of a main waste path into a uranium adsorption column of a column group, wherein: the column group also includes a supporting adsorption column, the supporting adsorption column is fluidically connected to the uranium adsorption column along the main waste path and is located downstream of the uranium adsorption column, and the stripping waste path bypasses the supporting adsorption column and extends from the uranium adsorption column to a column effluent tank; adsorbing uranium from the waste stream onto an anion exchange resin, the anion exchange resin being contained in the uranium adsorption column; directing the waste stream from the uranium adsorption column into the supporting adsorption column; and adsorbing one or more target radionuclides onto a cation exchange resin, the cation exchange resin being contained in the supporting adsorption column.

[0015] The tenth aspect includes the method of the ninth aspect, wherein the one or more target radionuclides include strontium-90 and cesium-137.

[0016] The eleventh aspect includes the method described in the tenth aspect, further comprising: directing the waste flow from the supporting adsorption column to a column effluent tank, wherein the waste flow entering the column effluent tank contains less than 0.04 Curie / cubic meter of strontium-90 and less than 1 Curie / cubic meter of cesium-137.

[0017] A twelfth aspect includes the method of aspect ten or aspect eleven, wherein the waste stream in the upstream section of the main waste path contains a gram per liter level of uranium at least 500 times greater than the gram per liter levels of both strontium-90 and cesium-137.

[0018] Aspect 13 includes the method of any one of aspects 9 to 12, wherein the one or more target radionuclides include barium, cerium, cesium, lanthanum, molybdenum, sodium, neodymium, palladium, praseodymium, rubidium, rhodium, ruthenium, samarium, strontium, yttrium, zirconium, protactinium or a combination thereof.

[0019] A fourteenth aspect includes the method of any one of aspects nine to thirteen, wherein the anion exchange resin contained in the uranium adsorption column adsorbs 90% or more of uranium present in the waste stream.

[0020] A fifteenth aspect includes the method of any one of aspects nine to fourteen, wherein the waste stream in the upstream section of the main waste path contains 1 g / L or more uranium.

[0021] A sixteenth aspect includes the method of any one of aspects nine to fifteen, wherein the cation exchange resin contained in the supporting adsorption column adsorbs 90% or more of one or more target radionuclides present in the waste stream.

[0022] Aspect seventeen includes the method described in any one of aspects nine to sixteen, wherein the support adsorption column is a first support adsorption column, and the column group further includes a second support adsorption column, which is located downstream of the first support adsorption column along the main waste path, and the second support adsorption column contains a cation exchange resin.

[0023] An eighteenth aspect includes the method of aspect seventeen, wherein the first support sorbent column and the second support sorbent column together sorb 99% or more of one or more target radionuclides present in the waste stream.

[0024] A nineteenth aspect includes the method of any one of aspects nine to eighteen, further comprising: directing the elution acid to the uranium adsorption column to desorb uranium from the anion exchange resin; forming a uranium waste stream; and directing the uranium waste stream along a stripping waste path to a column effluent tank.

[0025] The twentieth aspect includes the method of any one of aspects nine to nineteen, wherein: the cation exchange resin comprises cation exchange resin beads; the anion exchange resin comprises anion exchange resin beads; the cation exchange resin beads and the anion exchange resin beads are each polymer-based; and the average diameter of the cation exchange resin beads and the anion exchange resin beads is 400 μm to 800 μm.

[0026] These and additional features provided by the embodiments described herein will be more fully understood in conjunction with the following detailed description and accompanying drawings. BRIEF DESCRIPTION OF THE DRAWINGS

[0027] The embodiments set forth in the accompanying drawings are illustrative and exemplary in nature and are not intended to limit the subject matter defined by the claims. The following detailed description of the illustrative embodiments may be understood when read in conjunction with the following drawings, in which like structures are represented by like reference numerals, wherein:

[0028] Figure 1 schematically illustrates a waste extraction system according to one or more embodiments shown and described in this disclosure; and

[0029] Figure 2 Schematically illustrates an ion exchange resin according to one or more embodiments shown and described in the present disclosure, the ion exchange resin being contained in Figure 1 In the adsorption column of the column group of the waste production system. Specific embodiments

[0030] Referring generally to the accompanying drawings, embodiments of the present disclosure are directed to waste extraction systems and methods for gradually removing target waste radionuclides from a waste stream, such as a waste stream formed during a medical isotope production process. The waste stream includes a variety of radionuclides, such as uranium-235 (U-235), cesium-137 (Cs-137), and strontium-90 (Sr-90). The waste extraction system includes a column group of adsorption columns containing ion exchange resins, which are used to selectively adsorb certain radionuclides. The column group is in fluid communication with a main waste path so that the waste stream passes through the column group and reaches a column effluent tank for final treatment.

[0031] The column group includes a uranium adsorption column and one or more supporting adsorption columns. The uranium adsorption column contains an anion exchange resin for selectively adsorbing uranium, and the one or more supporting adsorption columns are located downstream of the uranium adsorption column along the main waste path. One or more supporting adsorption columns contain cation exchange resins for selectively adsorbing one or more target radionuclides such as Cs-137 and Sr-90. The waste extraction system also includes a waste stripping path that bypasses one or more supporting adsorption columns to connect the uranium adsorption column fluid to the column effluent tank. The uranium adsorption column provides a method for removing uranium such as U-235 from the waste stream, so that the waste stream passing through one or more supporting adsorption columns contains minimal uranium. This allows one or more supporting adsorption columns to remove other target radionuclides such as Cs-137 and Sr-90 from the waste stream, which may be lower than uranium in the initial waste stream. An embodiment of the waste extraction system and a radionuclide waste extraction method using the waste extraction system will now be described, and whenever possible, the same reference numerals will be used throughout the drawings to refer to the same or similar components.

[0032] Reference now Figure 1 and Figure 2 , shows a waste extraction system 100 according to an illustrative embodiment. The waste extraction system 100 includes a column group 110, which includes two or more adsorption columns, and the two or more adsorption columns are filled with ion exchange resins for adsorbing radionuclides. The column group 110 includes a uranium adsorption column 120 and one or more support adsorption columns 130. The adsorption column 120 contains an anion exchange resin 112, and the one or more support adsorption columns 130 contain a cation exchange resin 116. The column group 110 is in fluid communication with a main waste path 160, which includes one or more pipes, tubes or other fluid conveying mechanisms, which are used to facilitate the flow of waste from the production facility through the column group 110 to the column effluent tank 150 and ultimately to the waste tank 155. One or more pumps 180 are connected to the main waste path 160 to help create fluid flow within the main waste path 160. In one example operation, the upstream section 162 of the main waste path 160 fluidly connects the waste retrieval system 100 to a production area of ​​a medical isotope production facility, and the waste stream contains radionuclide waste generated by the medical isotope production process.

[0033] The column group 110 is located between the upstream section 162 of the main waste path 160 and the column effluent tank 150, and is in fluid communication with the upstream section 162 of the main waste path 160 and the column effluent tank 150. During operation, a waste stream containing radionuclide waste enters the waste extraction system 100 along the upstream section 162 of the main waste path 160 (e.g., an initial waste stream). The uranium adsorption column 120 is in fluid communication with the upstream section 162 of the main waste path 160 and is located upstream of one or more support adsorption columns 130 along the main waste path 160. During operation, the waste stream passes through the uranium adsorption column 120 and then passes through one or more support adsorption columns 130 along the main waste path 160. During operation, the anion exchange resin 112 contained in the uranium adsorption column 120 adsorbs uranium such as U-235 from the waste stream, thereby reducing the uranium in the waste stream entering the one or more support adsorption columns 130. The cation exchange resin 116 contained in one or more support adsorption columns 130 then adsorbs one or more target radionuclides, such as Cs-137 and Sr-90, from the waste stream.

[0034] By removing uranium from the waste stream at the uranium adsorption column 120, one or more support adsorption columns 130 can more efficiently and effectively adsorb other target radionuclides. Without intending to be limited by theory, the uranium present in the initial waste stream will be adsorbed by the cation exchange resin 116 in the support adsorption column 130 (along with one or more target radionuclides), and the relatively large amount of uranium may cause the cation exchange resin 116 to reach its adsorption limit before removing the desired amount of target radionuclides. For example, the initial waste stream may contain uranium at a gram / liter level that is at least 500 times higher than the gram / liter level of both strontium-90 and cesium-137, for example, at least 750 times higher, at least 1000 times higher, at least 1250 times higher, at least 1500 times higher, at least 2000 times higher, or multiples within a range of any two of these values ​​as endpoints. In some embodiments, the uranium in the initial waste stream may be at least 500 times higher than the gram / liter level of any of the following individual radionuclides: barium, cerium, lanthanum, molybdenum, neodymium, palladium, praseodymium, rubidium, rhodium, ruthenium, samarium, yttrium, and zirconium, for example, at least 750 times higher, at least 1000 times higher, at least 1250 times higher, at least 1500 times higher, at least 2000 times higher, or multiples within a range of any two of these values ​​as endpoints. In addition, uranium may account for 40% to 60% of the mass of the total radionuclides in the initial waste stream.

[0035] In some embodiments, the initial waste stream contains 1 g / L or more uranium, such as 1.5 g / L, 2 g / L or more, 2.5 g / L or more, 3 g / L or more, and values ​​within a range with two of these values ​​as endpoints. Although Cs-137 and Sr-90 are referred to as target radionuclides in the embodiments described herein, other target radionuclides may also be present in the waste stream and adsorbed by the cation exchange resin 116. For example, other target radionuclides that may be present in the waste stream and adsorbed by the cation exchange resin 116 include barium, cerium, cesium, lanthanum, molybdenum, sodium, neodymium, palladium, praseodymium, rubidium, rhodium, ruthenium, samarium, strontium, yttrium, zirconium, protactinium, or a combination thereof.

[0036] The uranium adsorption column 120 includes a waste inlet 123 and an elution inlet 124 and a waste outlet 125 and an elution outlet 126. The waste inlet 123 and the elution inlet 124 are each located at a first end 121 of the uranium adsorption column 120, and the waste outlet 125 and the elution outlet 126 are each located at a second end 122 of the uranium adsorption column 120. The waste inlet 123 is in fluid communication with an upstream section 162 of the main waste path 160, and the waste outlet 125 is in fluid communication with an inter-column section 165 of the main waste path 160. The main waste path 160 may include a plurality of inter-column sections 165, which are between adjacent adsorption columns of the column group 110 and between the final adsorption columns (e.g., Figure 1 The second support adsorption column 132b in the illustrated embodiment) and the section between the column effluent tank 150. It should be understood that in the embodiment of the treatment assembly included between the column group 110 and the column effluent tank 150, additional column inter-section 165 is considered. In addition, in some embodiments, the first end 121 of the uranium adsorption column 120 is opposite to the second end 122 of the uranium adsorption column 120, and the first end 121 of the uranium adsorption column 120 is above the second end 122 of the uranium adsorption column 120. This orientation facilitates the gravity-assisted flow of the waste stream through the uranium adsorption column 120. Gravity-assisted flow can reduce the pumping pressure and pumping power required to flow the waste stream through the uranium adsorption column 120. Gravity-assisted flow can also maximize the contact between the anion exchange resin 112 and the waste stream, thereby maximizing uranium adsorption.

[0037] exist Figure 1In the illustrated embodiment, one or more support adsorption columns 130 include a first support adsorption column 132a and a second support adsorption column 132b, the first support adsorption column 132a and the second support adsorption column 132b being fluidly connected along the main waste path 160, for example, through the inter-column section 165 of the main waste path 160. The second support adsorption column 132b is located downstream of the first support adsorption column 132a, so that the waste flow passes through the first support adsorption column 132a before passing through the second support adsorption column 132b. Each support adsorption column 132a, 132b includes a waste flow inlet 134a, 134b and a waste flow outlet 135a, 135b. The waste flow inlet 134a, 134b is located at the first end 131a, 131b of each support adsorption column 132a, 132b, and the waste flow outlet 135a, 135b is located at the second end 133a, 133b of each support adsorption column 132a, 132b. In addition, in some embodiments, the first end 131a, 131b of each support adsorption column 132a, 132b is opposite to the second end 133a, 133b, and the first end 131a, 131b is above the second end 133a, 133b. This orientation facilitates the gravity-assisted flow of the waste stream through each support adsorption column 132a, 132b. Gravity-assisted flow can reduce the pumping pressure and pumping power required to flow the waste stream through one or more support adsorption columns 130. Gravity-assisted flow can also maximize the contact between the cation exchange resin 116 and the waste stream, thereby maximizing the adsorption of the target radionuclide.

[0038] Although two support adsorption columns 132a, 132b are shown, it should be understood that additional adsorption columns can be included to adsorb additional amounts of target radionuclides. Optionally, it is expected that the column group 110 only has the first support adsorption column 132a and does not include the second support adsorption column 132b. In fact, in this optional embodiment, the volume of the first support adsorption column 132a can be doubled (while the volume of the cation exchange resin 116 in the first support adsorption column 132a is doubled). Although this will help to remove similar levels of target radionuclides, in the waste extraction system 100 with size and design limitations, it may be difficult to double the volume of the first support adsorption column 132a. In addition, since the flow resistance increases with the increase in the size of the first support adsorption column 132a, by increasing the volume of the first support adsorption column 132a, the pumping pressure for the waste stream flowing through the first support adsorption column 132a is increased.

[0039] Still refer to Figure 1, the elution inlet 124 of the uranium adsorption column 120 is in fluid communication with an elution path 172, which extends from an elution acid source 175 to the elution inlet 124. The elution outlet 126 is in fluid communication with a stripping waste path 170, which extends from the uranium adsorption column 120 to the column effluent tank 150 and provides a fluid path from the uranium adsorption column 120 to the column effluent tank 150 while bypassing one or more support adsorption columns 130. One or more pumps 180 can be in fluid communication with the stripping waste path 170 to help facilitate fluid flow in the stripping waste path 170. The stripping waste path 170 provides a path for uranium removed from an initial waste stream (e.g., a uranium waste stream) by the uranium adsorption column 120 to reach the column effluent tank 150 without passing through one or more support adsorption columns 130.

[0040] During operation, elution acid may be directed to the uranium adsorption column 120 through the elution inlet 124. The elution acid flows through the uranium adsorption column 120 and desorbs uranium from the anion exchange resin 112, forming a uranium waste stream. The uranium waste stream may be directed from the uranium adsorption column 120 through the elution outlet 126 to the stripping waste path 170, where the uranium waste stream is directed to the column effluent tank 150. By flowing through the stripping waste path 170, the uranium waste stream bypasses one or more support adsorption columns 130, allowing the support adsorption columns 130 to adsorb other target radionuclides such as Cs-137 and Sr-90 that are present at levels below uranium in the initial waste stream.

[0041] like Figure 1 As shown, the column effluent tank 150 is in fluid communication with both the main waste path 160 and the stripping waste path 170. Therefore, the treated waste stream from the support adsorption column 130 is mixed with the uranium waste stream from the uranium adsorption column 120 in the column effluent tank 150 to form the generated waste. The waste tank 155 is in fluid communication with the column effluent tank 150 through the waste tank section 166 of the main waste path 160. The generated waste in the column effluent tank 150 can be directed to the waste tank 155 along the waste tank section 166 to be finally processed and transported to the off-site disposal (remove-off site). The final processing may include curing the generated waste with concrete to form a cured final waste, which may occur in the waste tank 155. In some embodiments, the column effluent tank 150 and the waste tank 155 have the same volume, for example, a volume of 25 gallons to 75 gallons, for example, 50 gallons. In other embodiments, column effluent tank 150 and waste tank 155 have different volumes and may include volumes ranging from 25 gallons to 75 gallons.

[0042] Although the generated waste contains previously separated uranium, the target radionuclides are still removed. By removing the target radionuclides using the waste extraction system 100, the generated waste contains lower levels of radioactivity than the initial waste stream. In fact, the target radionuclides contribute a disproportionate amount of the total radioactivity in the initial waste stream. For example, Cs-137 is a gamma-ray emitting nuclide, and therefore, it is best to minimize the amount of Cs-137 in the generated waste. Adsorption of target radionuclides such as Cs-137 and Sr-90 can allow these target radionuclides to be handled separately from the generated waste, such as being sealed in concrete with a minimized volume.

[0043] Reference now Figure 2 , shows a schematic cross-sectional view of the column group 110. Figure 2 As shown, in some embodiments, the anion exchange resin 112 includes a plurality of anion exchange resin beads 114, and the cation exchange resin 116 includes cation exchange resin beads 118. The average diameter of the anion exchange resin beads 114 and the cation exchange resin beads 118 ranges from 200 μm to 1000 μm, for example, from 400 μm to 800 μm, for example, 400 μm, 425 μm, 450 μm, 475 μm, 500 μm, 525 μm, 550 μm, 575 μm, 600 μm, 625 μm, 650 μm, 675 μm, 700 μm, 725 μm, 750 μm, 775 μm, 800 μm, or any range having any two of these values ​​as endpoints.

[0044] In some embodiments, a plurality of anion exchange resin beads 114 and cation exchange resin beads 118 are dry poured into an adsorption column of the column set 110. After entering the support adsorption column 132, the anion exchange resin beads 114 and cation exchange resin beads 118 can be pretreated by passing two or more column volumes of a pretreatment acid, such as sulfuric acid, through the adsorption column of the column set 110 before introducing the waste stream into the column set 110. In other embodiments, the anion exchange resin beads 114 and cation exchange resin beads 118 can be pretreated with an acid, such as sulfuric acid, and added to the column set 110 as a resin slurry (e.g., resin in acid). The pH of the pretreatment acid is similar to that of the initial waste stream (e.g., within 1 unit). For example, the pretreatment acid can include sulfuric acid, which has a pH of 1. Pretreatment of the anion exchange resin beads 114 and cation exchange resin beads 118 can be performed at a location remote from the waste stream, such as a different location within the waste extraction system 100 or a different location within a medical isotope production facility. Therefore, personnel performing this preprocessing can be kept away from waste streams containing radionuclides.

[0045] In some embodiments, the anion exchange resin 112 and the cation exchange resin 116 are each polymer-based. The anion exchange resin 112 may include a weakly basic anion exchange resin or a strongly basic anion exchange resin. The cation exchange resin 116 may include a strongly acidic cation exchange resin. The anion exchange resin 112 and the cation exchange resin 116 have a combination of porosity that contributes to adsorption capacity and chemical functionality that contributes to selectivity. In addition, the cation exchange resin 116 tends to adsorb Cs-137 and Sr-90 more than uranium, so the cation exchange resin 116 is effective in supporting the adsorption column 130. Examples of the anion exchange resin 112 include Amberlite TM Resin and DIAON TM Examples of cation exchange resins 116 include SACMP (strong acid cationic macroporous polystyrene) resins (e.g., SACMP manufactured by ResinTech Corporation). SACMP) and AMP-PAN (ammonium molybdate phosphate polyacrylonitrile) resins.

[0046] In some embodiments, cation exchange resin 116 comprises an adsorption capacity of 100 mg of Cs-137 per gram of cation exchange resin 116 (i.e., 100 mg / g) to 200 mg / g of Cs-137, for example, 125 mg / g to 175 mg / g, for example, 105 mg / g, 110 mg / g, 115 mg / g, 120 mg / g, 125 mg / g, 130 mg / g, 135 mg / g, 140 mg / g, 145 mg / g, 150 mg / g, 153 mg / g, 155 mg / g, 157 mg / g, 160 mg / g, 165 mg / g, 170 mg / g, 175 mg / g, 180 mg / g, 185 mg / g, 190 mg / g, 195 mg / g, 200 mg / g, or a value within a range with any two of these numbers as endpoints. In some embodiments, the cation exchange resin 116 includes an adsorption capacity of 0.1 mg of Sr-90 (i.e., 0.1 mg / g) to 1 mg / g of Sr-90 per gram of cation exchange resin 116, for example, 0.15 mg / g to 0.5 mg / g, for example, 0.15 mg / g, 0.2 mg / g, 0.25 mg / g, 0.3 mg / g, 0.35 mg / g, 0.4 mg / g, 0.45 mg / g, 0.5 mg / g, 0.55 mg / g, 0.6 mg / g, 0.65 mg / g, 0.7 mg / g, 0.75 mg / g, 0.8 mg / g, 0.85 mg / g, 0.9 mg / g, 0.95 mg / g, 1 mg / g, or a value in a range with any two of these numbers as endpoints. In addition, in some embodiments, the anion exchange resin 112 includes an adsorption capacity of 85 mg uranium (i.e., 85 mg / g) to 165 mg / g of uranium, such as U-235, per gram of anion exchange resin 112, for example, 115 mg / g to 140 mg / g, for example, 85 mg / g, 90 mg / g, 95 mg / g, 100 mg / g, 105 mg / g, 110 mg / g, 115 mg / g, 120 mg / g, 125 mg / g, 130 mg / g, 135 mg / g, 140 mg / g, 145 mg / g, 150 mg / g, 155 mg / g, 160 mg / g, 165 mg / g, or a value in a range with any two of these numbers as endpoints.

[0047] Reference again Figure 1 and Figure 2, a method for performing radionuclide waste extraction using the waste extraction system 100 will now be described. First, the method includes directing a waste stream from an upstream section 162 of a main waste path 160 into a uranium adsorption column 120 of a column group 110. Next, the method includes adsorbing uranium from the waste stream onto an anion exchange resin 112 contained in the uranium adsorption column 120. During operation, the waste stream enters the uranium adsorption column 120 through a waste stream inlet 123, and the anion exchange resin 112 contained in the uranium adsorption column 120 adsorbs uranium present in the waste stream, thereby removing the uranium from the waste stream. The waste stream then leaves the uranium adsorption column 120 through a waste stream outlet 125 and flows to one or more support adsorption columns 130, the level of uranium present in its waste stream being lower than the level of uranium present in the waste stream when the waste stream enters the uranium adsorption column 120. For example, the anion exchange resin 112 contained in the uranium adsorption column 120 adsorbs 85% or more of the uranium present in the waste stream, for example, 86% or more, 87% or more, 88% or more, 89% or more, 90% or more, 91% or more, 92% or more, 93% or more, 94% or more, 95% or more, 96% or more, 97% or more, 98% or more, 99% or more, or a value within a range taking any two of these numbers as endpoints.

[0048] Next, the method includes directing the waste stream (e.g., uranium-reduced waste stream) from the uranium adsorption column 120 into the first support adsorption column 132a, for example, along the inter-column section 165. During operation, the waste stream (e.g., uranium-reduced waste stream) enters the first support adsorption column 132a through the waste stream inlet 134a, and the cation exchange resin 116 contained in the first support adsorption column 132a adsorbs target radionuclides such as Sr-90 and Cs-137 present in the uranium-reduced waste stream. During operation, the cation exchange resin 116 contained in the first support adsorption column 132a adsorbs 85% or more of one or more target radionuclides initially present in the waste stream (for example, present in the waste stream in the upstream section 162 of the main waste path 160), for example, 85% or more of the strontium-90 and cesium-137 present in the waste stream, for example, 86% or more, 87% or more, 88% or more, 89% or more, 90% or more, 91% or more, 92% or more, 93% or more, 94% or more, 95% or more, 96% or more, or a value within a range with any two of these numbers as endpoints.

[0049] The uranium-reduced waste stream then exits the first support sorbent column 132a through the waste stream outlet 135a and enters the inter-column section 165 of the main waste path 160. After exiting the first support sorbent column 132a, the waste stream may be directed to a second support sorbent column 132b (e.g., in embodiments including two or more support sorbent columns 130, e.g., Figure 1 ) or a column effluent tank 150 (e.g., in an embodiment including a single support adsorption column 130). In an embodiment including a second support adsorption column 132b, after the uranium-reduced waste stream enters the second support adsorption column 132b, the method includes adsorbing an additional amount of one or more target radionuclides present in the uranium-reduced waste stream onto a cation exchange resin 116 contained in the second support adsorption column 132b. By adsorbing the additional amount of the target radionuclides in the second support adsorption column 132b, the first support adsorption column 132a and the second support adsorption column 132b jointly adsorb 95% or more of one or more target radionuclides originally present in the waste stream (e.g., present in the waste stream in the upstream section 162 of the main waste path 160), for example, 95% or more of Sr-90 and Cs-137 present in the waste stream, such as 96% or more, 97% or more, 98% or more, 98.5% or more, 99% or more, 99.5% or more, 99.9% or more, or values ​​within a range with any two of these numbers as endpoints. In effect, the first support adsorption column 132a acts as a coarse filter for the target radionuclides, while the second support adsorption column 132b acts as a fine filter for the target radionuclides.

[0050] Still refer to Figure 1 and Figure 2 The method also includes removing uranium from the uranium adsorption column 120 and directing the uranium along the stripping waste path 170 to the column effluent tank 150. The removal process can include directing an elution acid such as sulfuric acid to the uranium adsorption column 120 to desorb uranium from the anion exchange resin 112 to form a uranium waste stream. The elution acid can include sulfuric acid. The elution acid can be directed from the elution acid source 175 along the elution path 172 and enter the uranium adsorption column 120 through the elution inlet 124. One or more pumps 180 can be in fluid communication with the elution path 172 to help promote the elution acid to flow in the elution path 172 and enter the uranium adsorption column 120. After desorption, the uranium waste stream can be directed from the uranium adsorption column 120 to the column effluent tank 150 along the stripping waste path 170. At the column effluent tank 150, the processed waste stream from the one or more support adsorption columns 130 is mixed with the uranium waste stream from the uranium adsorption column 120 to form a generated waste. In some embodiments, the method also includes directing the generated waste along the waste tank section 166 to the waste tank 155 for final treatment and transportation to off-site disposal.

[0051] As described above, the final treatment of the produced waste can be solidified with concrete to form a solidified final waste. This densifies the produced waste. Therefore, the final waste contains a higher Curie level per cubic meter than the produced waste used to form the final waste. However, the radionuclide waste extraction method using the waste extraction system 100 described herein effectively removes most of the target radionuclides, so that even after densification, the final waste formed from the produced waste still maintains a low level of radioactivity. For example, the waste generated can contain less than 0.25 Curie / m3 of Sr-90, for example, less than 0.2 Curie / m3, less than 0.15 Curie / m3, less than 0.1 Curie / m3, less than 0.08 Curie / m3, less than 0.06 Curie / m3, less than 0.05 Curie / m3, less than 0.04 Curie / m3, less than 0.03 Curie / m3, less than 0.02 Curie / m3, less than 0.01 Curie / m3, or any value in a range between any two of these values. The waste produced may also contain less than 10 Curies / m3 of Cs-137, for example, less than 8 Curies / m3, less than 6 Curies / m3, less than 5 Curies / m3, less than 4 Curies / m3, less than 2 Curies / m3, less than 1 Curie / m3, less than 0.75 Curies / m3, less than 0.5 Curies / m3, less than 0.25 Curies / m3, less than 0.1 Curies / m3, or any value in a range with any two of these values ​​as endpoints. In some embodiments, the waste produced contains less than 0.04 Curies / m3 of Sr-90 and less than 1 Curie / m3 of Cs-137. In some embodiments, the final densified waste contains less than 0.04 Curies / m3 of Sr-90 and less than 1 Curie / m3 of Cs-137.

[0052] Moreover, the above-described Curie / m3 values ​​in the generated waste can be achieved using the waste extraction system 100 described herein in embodiments where the initial waste stream (i.e., the waste stream passing through the upstream section 162 of the main waste path 160) contains greater than 150 Curie / m3 of Sr-90, e.g., greater than 200 Curie / m3, greater than 300 Curie / m3, greater than 300 Curie / m3, greater than 300 Curie / m3, greater than 500 Curie / m3, greater than 1000 Curie / m3, greater than 2500 Curie / m3, greater than 500 Curie / m3, greater than 1500 Curie / m3, greater than 2000 Curie / m3, greater than 3000 Curie / m3, greater than 5000 Curie / m3, greater than 10000 Curie / m3, greater than 2500 Curie / m3, greater than 50 ...0 Curie / m3, greater than 500000 Curie / m3, greater than 500000 Curie / m3, greater than 500000 Curie / m3, greater than 500000 Curie / m3, greater than 500000 Curie / m3, greater than 500000 Curie / m3, 00 Curie / m3, or any value in a range with any two of these values ​​as endpoints, and includes Cs-137 greater than 44 Curie / m3, for example, greater than 50 Curie / m3, greater than 100 Curie / m3, greater than 250 Curie / m3, greater than 500 Curie / m3, greater than 800 Curie / m3, greater than 1000 Curie / m3, greater than 1500 Curie / m3, greater than 2000 Curie / m3, greater than 3500 Curie / m3, or any value in a range with any two of these values ​​as endpoints.

[0053] As used herein, the terms "approximately," "about," "substantially," and similar terms are intended to have a broad meaning consistent with common and accepted usage by those of ordinary skill in the art to which the subject matter of the present disclosure belongs. Those skilled in the art who read this disclosure should understand that these terms are intended to allow description of certain features described and claimed without limiting the scope of these features to the precise numerical values ​​or ideal geometric forms provided. Therefore, these terms should be interpreted as indicating that insubstantial or insignificant modifications or changes to the subject matter described and claimed are considered to be within the scope of the disclosure as described in the appended claims.

[0054] The term "connected" and its variants used in this article refer to two elements being directly or indirectly connected to each other. This connection can be static (e.g., permanent or fixed) or removable (e.g., removable or releasable). This connection can be achieved by directly connecting the two elements together, or by connecting the two elements to each other using a separate intermediate element and any additional intermediate elements, or by connecting the two elements to each other using an intermediate element that is integrally formed as a single entity with one of the two elements. If "connected" or its variants are modified by additional terms (e.g., directly connected), the general definition of "connected" above will be modified by the general language meaning of the additional terms (e.g., "directly connected" refers to the connection of two elements without any separate intermediate elements), resulting in a narrower definition than the general definition of "connected" above. This connection can be mechanical, electrical, optical or fluid.

[0055] References to element positions herein (e.g., "top," "bottom," "above," "below") are only used to describe the orientation of the various elements in the figures. It should be noted that the orientation of the various elements may be different according to other exemplary embodiments, and such variations are intended to be encompassed in the present disclosure.

[0056] Although the figures and description may illustrate a particular order of method steps, the order of these steps may be different from that shown and described, unless otherwise specified above. In addition, two or more steps may be performed simultaneously or partially simultaneously, unless otherwise specified above. Such variations may depend on, for example, the selected software and hardware systems and the designer's choice. All of these variations are within the scope of the present disclosure. Similarly, software implementations of the method may be implemented by standard programming techniques, wherein rule-based logic and other logic are used to implement various connection steps, processing steps, comparison steps, and decision steps.

[0057] Although specific embodiments have been illustrated and described herein, it should be understood that various other changes and modifications may be made without departing from the spirit and scope of the claimed subject matter. In addition, although various aspects of the claimed subject matter have been described herein, these aspects do not have to be used in combination. Therefore, the appended claims are intended to cover all such changes and modifications within the scope of protection of the claimed subject matter.

Claims

1. A waste extraction system comprising: a column group, the column group being located between an upstream section of a main waste path and a column effluent tank and being in fluid communication with the upstream section of the main waste path and the column effluent tank, the column group comprising a uranium adsorption column, the uranium adsorption column being in fluid communication with a support adsorption column along the main waste path, wherein the uranium adsorption column is upstream of the support adsorption column; a stripping waste path, the stripping waste path bypassing the support adsorption column and extending from the uranium adsorption column to the column effluent tank; an anion exchange resin contained in the uranium adsorption column; and A cation exchange resin is contained in the supporting adsorption column.

2. The waste extraction system according to claim 1, wherein: The supporting adsorption column is a first supporting adsorption column and the column set further comprises a second supporting adsorption column, the second supporting adsorption column being located downstream of the first supporting adsorption column along the main waste path, the second supporting adsorption column containing a cation exchange resin.

3. The waste extraction system according to claim 1, wherein: The uranium adsorption column comprises: a waste flow inlet and an elution inlet, each of the waste flow inlet and the elution inlet being located at a first end of the uranium adsorption column, wherein the waste flow inlet is in fluid communication with the upstream section of the main waste path; and A waste flow outlet and an elution outlet are each located at the second end of the uranium adsorption column, wherein the elution outlet is in fluid communication with the stripping waste path.

4. The waste extraction system according to claim 3, wherein: The waste extraction system also includes an elution path extending from an elution acid source to the elution inlet.

5. The waste extraction system according to claim 3, wherein: The first end of the uranium adsorption column is opposite to the second end of the uranium adsorption column and the first end of the uranium adsorption column is above the second end of the uranium adsorption column.

6. The waste extraction system according to claim 1, wherein: The support adsorption column comprises a waste inlet and a waste outlet, wherein the waste inlet is located at a first end of the support adsorption column, and the waste outlet is located at a second end of the uranium adsorption column.

7. The waste extraction system according to claim 6, wherein: The first end of the support adsorption column is opposite to the second end of the support adsorption column and the first end of the support adsorption column is above the second end of the support adsorption column.

8. The waste extraction system of claim 1, wherein: The cation exchange resin comprises cation exchange resin beads; The anion exchange resin comprises anion exchange resin beads; The cation exchange resin beads and the anion exchange resin beads are each polymer-based; and The cation exchange resin beads and the anion exchange resin beads have an average diameter of 400 μm to 800 μm.

9. A method for extracting radioactive nuclide waste, the method comprising: The waste stream is directed from an upstream section of the main waste path to a uranium adsorption column of a column bank, wherein: The column set further includes a support sorption column in fluid communication with the uranium sorption column along the main waste path and located downstream of the uranium sorption column; and The stripping waste path bypasses the support adsorption column and extends from the uranium adsorption column to the column effluent tank; adsorbing uranium from the waste stream onto an anion exchange resin contained in the uranium adsorption column; directing the waste stream from the uranium adsorption column to the support adsorption column; and One or more target radionuclides are adsorbed onto a cation exchange resin contained within the supporting adsorption column.

10. The method according to claim 9, wherein: The one or more target radionuclides include strontium-90 and cesium-137.

11. The method according to claim 10, wherein: The method also includes directing the waste stream from the support adsorption column to the column effluent tank, wherein the waste stream entering the column effluent tank contains less than 0.04 Curie / m3 of Sr-90 and less than 1 Curie / m3 of Cs-137.

12. The method according to claim 10, wherein: The waste stream in the upstream section of the main waste path contains uranium at a level of grams per liter that is at least 500 times higher than the levels of both strontium-90 and cesium-137.

13. The method according to claim 9, wherein: The one or more target radionuclides include barium, cerium, cesium, lanthanum, molybdenum, sodium, neodymium, palladium, praseodymium, rubidium, rhodium, ruthenium, samarium, strontium, yttrium, zirconium, protactinium, or a combination thereof.

14. The method according to claim 9, wherein: The anion exchange resin contained in the uranium adsorption column adsorbs 90% or more of the uranium present in the waste stream.

15. The method according to claim 9, wherein: The waste stream in the upstream section of the main waste path contains 1 g / L or more uranium.

16. The method according to claim 9, wherein: The cation exchange resin contained in the supporting adsorption column adsorbs 90% or more of the one or more target radionuclides present in the waste stream.

17. The method according to claim 9, wherein: The supporting adsorption column is a first supporting adsorption column and the column set further comprises a second supporting adsorption column, the second supporting adsorption column being located downstream of the first supporting adsorption column along the main waste path, the second supporting adsorption column containing a cation exchange resin.

18. The method according to claim 17, wherein: The first support sorption column and the second support sorption column together sorb 99% or more of the one or more target radionuclides present in the waste stream.

19. The method according to claim 9, wherein: The method further comprises: directing elution acid into the uranium adsorption column to desorb uranium from the anion exchange resin to form a uranium waste stream; and The uranium waste stream is directed along the stripping waste path to the column effluent tank.

20. The method of claim 9, wherein: The cation exchange resin comprises cation exchange resin beads; The anion exchange resin comprises anion exchange resin beads; The cation exchange resin beads and the anion exchange resin beads are each polymer-based; and The cation exchange resin beads and the anion exchange resin beads have an average diameter of 400 μm to 800 μm.

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